ML20126D783
| ML20126D783 | |
| Person / Time | |
|---|---|
| Issue date: | 12/11/1992 |
| From: | NRC COMMISSION (OCM) |
| To: | |
| References | |
| REF-10CFR9.7 NUDOCS 9212280158 | |
| Download: ML20126D783 (155) | |
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NUCLEAR REGULATORY COMMIS SION:
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PERIODIC MEETING WITH THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS)
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ROCF'JILLE, MARYLAND b3(3 DECEMBER 11, 1992 23965 61 PAGES
-NEALR.GROSSANDC0.,INC.
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DISCLAIMER This is an unofficial transcript of a meeting of the United States Nuclear Regulatory Commission held on December 11, 1992 in the Commission's office at One White Flint North, Rockville, Maryland.
The meeting was open-to public attendance and observation.
This transcript has not been reviewed, corrected or edited, and it may contain inaccuracies.
l The transcript is intended solely for general informational purposes.
As provided by 10 CFR 9.103, it is not part of the formal or informal record of decision of the matters discussed.
Expressions of opinion in this transcript do not necessarily reflect final determination or beliefs.
No pleading or other paper may be filed with the Commission in any proceeding as the result of, or addressed to, any statement or argument contained herein, except as the Commission may authorize.
0 0
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1 1
1 UNITED STATES OF AMERICA 1
2
-NUCLEAR REGULATORY COMMISSION
.. i 3
4 5
PERIODIC MEETING WITH THE ADVISORY COMMITTEE 6
ON REACTOR SAFEGUARDS (ACRS)-
7 8
PUBLIC MEETING 9
10 Nuclear Regulatory Commission 11 One White Flint North 12 Rockville, Maryland 13 14 Friday
+
15 December 11, 1992 16 17 The Commission met -in - open session, pursuant to-18-notice, at 1:30 p.m.,
the Honorable IVAN SELIN, Chairman-19 of the Commission, presiding.
20-21 COMMISSIONERS-PRESENT:
t
'22 IVAN.SELIN, Chairman of the Commission 23 KENNETI
^
ROGERS,_ Member of-the. Commission.
i -
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L 24 JAMES R. CURTISS, Member of the Commission 25 FORREST J.: REMICK, Member of the Commission NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS-1323 RHOOE ISLAND AVENUE, N W -
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PROCEEDI N G:S 2
(1:31 p.m.)_
3 CHAIRMAN SELIN:
Bucking to the independent 4
opinion of our respected
- advisors, I'll say good 5
afternoon, ladies and gentlemen.
I'm pleased to welcome 6
the members of the Advisory Committee on Reactor 7
Safeguards and, in particular, to welcome Dr. Shewmon as 8
the new Chairman of the Committee, as well as Mr.' Peter 1
not here?
Welcome aboard, Mr.
9 Davis, who is inere 10 Davis.
11 This Committee provides a valuable service to.
12 the Commisoion by providing advice on the safety aspects 13 of proposed and existing nuclear facilities.
The 14 activities are therefor important as we try to-solve and 15 address technical concerns in. licensing and regulation.
16 In recent months, the Committee has discussed 17 various aspects of the advanced reactor reviews with the 18 Commission.
I'm pleased to know that there is an advanced 19 reactor item on the agenda-for today's-briefing.-
20 I personally would like to thank the Committee 21 for_its willingness _to focus on a relatively small-number 22 of significant long-range issues in meeting af ter meeting, 23 and allow us to have the follow-up and the ' continuity that-24 your experience and your expertise.can provide us.
25 The dialogue between the Committee and the NEAL R. GROSS COURT REPORTERS AND TRANSCRIAERS 1323 RHODE ISLAND AVENUE. N W (202) 734 4433 WASHINGTON. O C. 2tx105 (202) 234 4433
I 1
1 UNITED STATES OF AMERICA 2
NUCLEAR REGULATORY COMMISSION 3
4 5
PERIODIC MEETING WITH THE ADVISORY COMMITTEE 6
ON REACTOR SAFEGUARDS (ACRS) 7 8
PUBLIC MEETING 9
l 10 Nuclear Regulatory Commission 11 One White Flint North 12 Rockville, Maryland 13 14 Friday 15 December 11, 1992 16 17 The Commission met in open session, pursuant to 18 notice, at 1:30 p.m., the Honorable IVAN SELIN, Chairman 19 of the Commission, presidina.
20 21 COMMISSIONERS PRESENT:
22 IVAN SELIN, Chairman of the Commission 23 KENNETH C. ROGERS, Member of the Commission 24 JAMES R. CURTISS, Member'of the Commission 25 FORREST J. REMICK, Member of the Commission NEAL R. GROSS COURT REPORTEF 3 AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE. N W.
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STAFF AND PRESENTERS SEATED AT THE COMMISSION TABLE:
2 JOHN HOYLE, Assistant Secretary 3
WILLIAM C.
PARLER, General Counsel 4
PAUL G.
SHEWMON, Chairman, ACRS 5
CARLYLE MICHELSON, ACRS 6
JAMES C. CARROLL, ACRS 7
HAROLD W. LEWIS, ACRS 8
IVAN CATTON, ACRS 9
J. ERNEST WILKINS, ACRS 10 CHARLES J. WYLIE, ACRS 11 12 13 14 15 16 17 18 19
-20 s
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PROCEEDINOS 2
(1:31 p.m.)
3 CHAIRMAN SELIN:
-Bucking to the independent
^
4 opinion of our respected advisors,- 'I'll say good 5
afternoon, ladies and gentlemen.
I'm pleased to welcome 6
the members of-the Advisory Committee on Reactor 7
Safeguards-ande in particular, to welcome Dr. Shewmon as 8
the new Chairman of_the Comuittee, as well as Mr. Peter not here?
Welcome aboard, - Mr.
9 Davis, who is here 10 Davis.
11 This Committee provides a valuable service to-12 the Commission by providing advice on the safety aspects 13 of proposed and existing nuclear facilities.
The 14 activities are therefor important as we try to solve and 15 address technical concerns in licensing and regulation.
16 In recent months, the Committee has discussed-17 various aspects of the advanced reactor reviews with the 18 Commission.
I'm pleased to know that there is an advanced 19 reactor item on the agenda for today's briefing.
20 I personally would like to thank the Committet 21 for its willingness to focus on a relatively ;smallL number 22 of significant long-range issues in meeting af ter meeting, 23 and. allow us to have the follow-up and the continuity.that l24 your experience and your expertise can provide us.
25' The dialogue between the. Committee and the NEAL R. GROSS COURT REPOP7ERS AND TRANSCRIBERS 1323 RH000 ISLAND AVENUE, N.W (202) 2344433 WASHINGTON, O C. 20005 (202) 234 4433
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- l technical staff is also very useful--in the resolution of' 2_
technical issues, but the continuity and the willingness 3
to sort of have a triage between those things you. discuss
~
4 with the Commission, those things you discuss with the 5
- staff, those things you discuss, I don't know, with 6
nobody, or each other, has been really very, very helpful.
j 7
We'll also be briefed by the Committee on.
8 Digital Control and Protection Systems, Implementation of 9
the Plant Life Extension Rule, and Risk-Based Regulation.
10 My fellow Commissioners and I
are looking 11 forward to your briefing today.
I understand that copies 12 of your letters to the Commission on today's topics, are 13' available at the entrances to the room.
14 Any other comments?
l 15 (No response.)
16 Dr. Shewmon, the floor is yours, and welcome 17 again.
18
.DR. SHEWMON:
Thank you.
The first item has to 19 do with advanced and-evolutionary reactors.
.Since we
-20 aren't quite sure what we are calling which there, we'll-p 21 use both words. -The first and main statement on this will 22 be.Mr. Michelson, and then the other reactors comments
'23 will come from Carroll and Wilkins.
Go ahead.
24 MR. MICHELSON:
Thank you..
I have a short-25-statement I would like to read to you, af ter which you;can NEAL R. GROSS COURT REPOATERS AND TRANSCRIBERS l.
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-then ask questions.
This is'a' status-report on-the GE
'2 application for the' advanced boiling water. reactor.
3 Since July-of this year, we have been working 4
]
4 towards completing our safety review of the ABWR by 5
holding various meetings.
Our-ABWR Subcommittee met with 6
the staf f and GE on August 19th, to discuss our April 13th a
7 letter which identified several areas of concern related 8
to the ABWR d.) sign.
We also considered at that time, 9
eight chapters from the staf f's then existing drsf t safety 10 Evaluation Report.
11 Additional meetings were held in October ' and 12-November, when the staff's Draft Final Safety Evaluation 13 became available, but the staff was still-not ready to 14 discuss three major chapters Chapter 7,
on 15 Instrumentation and Ccatrol; Chapter 18, on Human Factors 16 Engineering, and Chapter 19, on Severe Accident 17 Considerations, including the PRA, 18 All of these chapters are undergoing extensive 19 rework by the staff, and the staff has not yet given us a-l _
20 date as to when they will be ready to discuss-them.
21 We-stand. prepared to. complete our review:
22-whenever the _FSER and the_ associated design descriptions, l
j'
-23 DACs, -and ITAACs are reasonably complete, up-to-date, and l
l 24 they have been subjected to an appropriate quality check, 25 and the staff and GE are ready to support a final meeting ~.
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'l The quality checks are essential
&o tihe 2
resolution of. the numerous technical conflicts that 3
presently exist in the SSAR, the FSER, and the various
~4-design descriptions, DACs and ITAACs.
5 Certain other ACRS review work remains.
We are 6
proposing a schedule to meet with the staff when they are 7
ready to discuss Appendix 9-A, which is the-Fire Hazards 8
Analysis, and Chapter 16, which contains the Proposed 9
Technical Specifications.
10 Still remaining for our selective review will be 11 the final closure on over 300 items in the draft FSER, 12 consideration of certain design description DACs--'and 13 ITAACs as they reach final development, and s number of 14 requests for additional information from previous ABWR 15 Subcommittee ~ meetings.
16 Also pending is our review of the Proposed
_17 Technical Resolutions of the unresolved safety issues and 18 the medium and high-priority generic safety issues which 19 are technically relevant to the ABWR-design.
In addition, 20 there are a number of miscellaneous SSAR sections.the' ABWR l
21 Subcommittee has not seen for the first time.
22 It-is our view that the remaining FSER chapters 23 and _ other items which we have-asked - to see, must be 24 completed in reasonably-final-form, quality checked, and' 25 provided to us during. the March, -'1993, full Committee NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHOOE ISLAND AVENUE. N W.
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7-1' meeting, if our: report is to be ready'to support:a fina1' 2
design approval following the June full Committee meeting..-
3 This would be close to what GE is proposing, which is June 4-1st for the FDA.
5 The key documents which we need' to receive 6
before we can prepare our final report include GE's final 7
raendment to-the SSAR, the staff's final FSER, and.the 8
associated final design descriptions, DACs and ITAACs.
9 Thus far, it is not clear to us that the staff and GE can 10 prepare these documents in an acceptable form by our March 11 full Committee meeting.
Questions?-
12 CHAIRMAN SELIN:
Commissioner Rocers?
13 COMMISSIONER ROGERS:
Well, I just have some 14 questions on some of the comments in some of your letters.
15 With respect to the assumptions made in PRAs and the' lack-16 of some way of determining those -reliabilities, of-17 verifying those reliabilities, have you any thoughts on 11 how to do that?
Any ways-that'you feel that might'be 19 carried out prior to the work of-a COL applicant?
20 MR. MICHELSON: -Well, we have not yet reviewed
-21 the Chapter 19,.which is severe accidents and the PRA. We 22 have a proposed schedule to work on_that'in February of 23-next year.
If-anybody would wish to venture a. reply to 24-your question of any way, I think that would be welcome..
25 But we really haven't come to grips with the PRA ~ since NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHOOE ISLAND AVENUE, N W.
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-8 'it's:not yet available to us in reasonable form.
-2 MR.-CARROLL:
I'm not sure-I understand your-4 3
concern, Commissioner.-
4 COMMISSIONER ROGERS:
Well,.my concern deals 5
with the point that was made in your August 12th letter --
6 MR. CARROLL:
August 12th.
7 COMMISSIONER ROGERS: -- in which you say "A PRA_
8 has been performed for the ABWR-.... certain conclusions 9
about the safety of the design can be drawn from this.
In 10 performing the PRA, many assumptions were necessary about-11 the performance reliability of components and systems.
12
'"here appears to be no means by which Tier 1 requirements 13 will ensure that components and systems in the plant can 14 be expected to have reliabilities that are consistent with 15 those assumed in the PRA".
And then you further go_on to 16 say that you "were told that appropriate ' reliability.
l 17 values for components and systems will be ensured'-through.
18 a reliability assurance program developed bys a-COL
-19 applicant.
We believe this matter deserves-mmre study".
20 And I _was just wondering if you had any thoughts 21 about what that study might entail, any-ways'of dealing-22 with those reliability questions.
23 (No response.)
2 4..
- CHAIRMAN SELIN:
Let's_ start with a yes or a no.
25 (Laughter.)
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- 1 DR. SHEWMON:
I don't see any volunteers coming-2 forward to explain the position.
l 3
_MR.'WILKINS:
I'm not.sure that the Committee, A
4 as such, has a position.
I think Carl stated - that -
5 correctly.
I might comment on this to the fo] lowing i
6 extent, that the things that are left over for the COL 7
applicant to do are a very hazy, gray area.
And it would 8
appear at least to me -- this is a personal opinion.now, 9
not necessarily a Committee opinion -- that the staff and 10 the Commission needs to pay attention to.the boundary 11 between the applicant and the COL.
The applicant for the but the design-my jargon is no good now 12 13 certification and the COL applicant, that boundary needs 14 to be made a little bit more precise.
15 COMMISSIONER ROGERS: Well, the concern,-I take 16 it, has to do with the f act that the PRA is necessary for
=
17 the design certification.
Certain assumptions have to be 18 m a d e i n p e r f o r m i n g t h a t P R A,: and to what extent can one 19 develop some limited degree of assessment of what the 20 range is of the overall PRA - that might come about?
21 Because yc.1_ don't.really know the exact systems that are--
22 going to be in the plant when it's filled, you have to 23 use, I suppose, some average -performances, or something of 24 that sort.
25 MR. CARROLL:
Generic data, yes.
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COMMISSIONER ROGERS' Yes, and whether there's 2
.anything that-.could be done in._ addition to thet, to give 3
some' credibility to how -- what this PRA finally --'the 4
bottom line result finally leads to.
It's.the old problem 5
of point values with no error bars on them.
i 6
MR. LEWIS:
Let me take a crack at this, since 1
7 nobody else seems eager to try. -There's certainly that.
8 You don't know what -- there's a lack of definition of the 9
actual structure finally, as you go to look at it, not.in 10 the sense that you normally would have it at a later -
11 stage, or under an earlier type of development. But on top 12 of
- that, there is a
pre-occupation with assigning 13 reliability values as point estimates, as you've said, and 14 for things which are extremely reliable,- there is 15 essentially no way in which one can require anyone to 16 maintain a given-reliability, or even to assess whether he 17 has maintained it.
18 I think I mentioned to you in Florida, a case in 19
=which the staff has sent a message to GE requiring that 20 certain circuit breakers be maintained at a reliability of L
21
-.9976.
These are things that are sprung a couple of times 22 a year.
And there's -just no way_ on earth in which anyone 23 can, you know, - in: an intellectually honest way, certify 24-that a thing has a reliability better than that.
25 So, I think there_is hidden in this -both the NEAL R. GROSS COURT HEPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE N W.-
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uncertainty in' what -the actual ~ gadget will be whenl-it 2
comes out and the dif ficulty of using_ point values, 'again,
~3 to use your term, to assign the reliabjlity the we look 4
at.
5 DR. SHEWMON:
We state here "We believe this 6
matter deserves more study".
Do we have any suggestions 7
on how that might be done?
8 COMMISSIONER ROGERS:
That's really --
9 MR. LEWIS:
That's what he's asking.
10 COMMISSIONER ROGERS:
That's my question.
11 DR. SHEWMON:
I guess.the answer is no.
12 MR. LEWIS:
Oh, no, no, no, no, no.
13 DR. SHEWMON:- We do have an answer?
14 MR. LEWIS:
Well, some of us have an answer.
15 DR. SHEWMON:
Good, let's hear it then.
'All 16 we've heard is concern, so far.
17 MR. LEWIS:
Well, you know, it's this class of-18 questions about how-you use reliability numbers.
I'll 19 make a speech-I've mado many, many times.
A reliability-20 consists of two things.
It consists of some kind of point' 21-estimate, of which you ought to kn'ow for yourself:what the-22-
- point' estimate is supposed to be a point estimate of,
'23
.whether.it's the middle of a distribution, or the mean of-24 a.. distribution, whatever,- but it 's - a _ point estimate, _ but-25 it - also includes - some kind - of spread around _ the point-NEAL R. GROSS COURT REPORTERS AND TRANSCR10ERS 1323 RHODE ISLAND AVENUF. N.W.
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estilmate, _and the uncertainty is as much a part.of the 2
reliability as is the central _ estimate.
Two numbers are 3
not.suf ficient to characterize a curve, but they are twice 4
as good as one number.
5 And if you have those two things we'have 6
actually written a letter in which we suggested that there 7
might be a way, based on confidence levels, in which you-8 could actually regulate using these two-numbers, to assure 9
the reliability within a certain confidence of-things.
?
10 And we have not developed such a thir.g, although I could 11 --
do.it on the spot and you wouldn't.like-it, but we think 12 it can be done.
And certainly my interpretation of-_this 13 is that the staff ought to be moving in the direction --
14 or putting it dif ferently, that the Commission ought to be :
15 directing the staf f to explore the feasibility of the kind-16 of regulation that would go with both reliability numbers, 17 point estimates of =
some
- kind, and-.
uncertainties 18:
characterized in some way-with it.
-19 On the-other hand, -we'll later be talking = about
- 20 the success of a similar ef fort on risk-based regulations, 21
- -
it will-come up again there.
But, - you - know,- I1 could -
22-deJeribe-a research program.
23
'DR._
SHEWMON:
Do you want to pursue this 24 further?
25 MR.-CARROLL:
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=1 aspect of this that we were expressing here,.and that is,-
2 this is one example, in my mind, of where the staf f and GE' 3
seem very satisfied to pull some, numbers out of the air 4
and say-"this is going to be the basis for final design j
5 approval, and we'll let the COL applicant justify the-6 numbers at a later stage in the game".
7 COMMISSIONER ROGERS: " ell, that's the position 8
that I have a little trouble with, - and. that's why I'm 9
asking for perhaps a little help from you folks, on what 10 we might do that's a little better than that.
11 MR. CARROLL:
Well, I think.he stafI should --
12 or one thing they certainly can do is be more questioning 13 of the assumed reliability values that GE is using in the 14 PRA.
15 MR. MICHELSON:
Well, ' I think, though, the I
the reason for this particular paragraph, 16 problem 17 though, Jay, was that the Tier 1-requirements, which is 18 what we were looking at, do not contain any of this.
They:
19 don't even specifyEnumbers.
And we wondered-how--do you 20-somehow convart Tier i requirements into a system that 21 then has-a reliability that you're predicting through the 22' PRA.
Where-is the connect, and there isn't ono 23 You know, you could' procure to-a _ reliability 24 number,,but that's not' necessarily a good approach either.
25 COMMISSIONER ROGERS:. That would probably be a NEAL R. GROSS COURT BEPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE, N W.
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I n'ightmare to try it that way.
-2 MR. MICHELSON:
It would be a nightmare to try.
3-'
COMMISSIONER CURTISS:
Can I suggest what I 4
think the staff has in mind here?
I don't make it a 5
practice to come to their defense, but let mc just 6
describe the approach as I understand it, if it's not 7
imminent at this point or, if this isn't the approach, 8
this is obviously an issue to pursue.
9 I do think an appropriate question has been 10 raised about the dividing line between what a design 11 certificate ought to address and what the COL applicant 12 ought to have the responsibility for addressing.
13 In this area, I think the staff's plans are a A4 little bit clear, as I understand them.
Let me describe 15 them and see if it's a reasonable approach.
16 We do require, as part of Part 52, that a PRA be 17 submitted.
That's in Part 52, and1.I think it's a sound Kii requirement.
The PRA that has been submitted-by GE for 19-the-ABWR is something that the staff hasitaken a good, 20 hard-look at, with the help and assistance of Brookehem,-
21-and they've gone back and forth on what the numbers are -in
.22 the. PRA and how the PRA is-being used to info 2.m the 23' design, at the-staff's most recent briefing as well'as at 24
- the GE briefing, and there was. considerable discussion.of 25
.how the PRA-is being used in the design process.
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15 1
1 There's.an important policy question, which is,5
-2 do you wantlthe PRA to be a_living document.that's kept 3
.up-to-date, and the. staf f indicateci that - they will be:
4 coming back to us with a recommendation on the question of 5
whether a PRA ought to be kept living,-at least for the 6
advanced reactors.
7 With respect to the question of-what you do with 8
the
- PRA, and setting aside from - that whether -it's 9
appropriate for the COL applicant to have to do this, my 10 understanding is that the use of the PRA by O c_ COLL 11 applicant would be very similar to the use of the PRA by 12 a licensee today, in complying with the maintenance ru 'e.
13 In fact, when I asked the staff what is the difference 14 between
- ORAP, the Operational Reliability-Assurance 15 Program, and the maintenance regime, the answer is, very 16 little.
17 Now, somebody at one of your meetings, in one of :
18 your transcripts that I read, suggested that we ought to i
19 take a look at making ORAP, or at least coordinating'our 20 thoughts on ORAP and the maintenance rule.
I don't know 21 who it was who suggested that,.but it was in-one-o'f your l
22 transcripts.
I think that is going--on.
And in my. view, 23 what we are doing with the PRA in the context of the COL 24 applicant is going to wind up to-be very similar to how a 25 PRA is used in the maintenance context. -Recognizing that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE, H W.-
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16 you -haven't had your final say on the maintenance 2
guidance, at. some point it would be helpful _ to see if you
-3
-think that's an appropriate use of the PRA.
But I think 4
that maybe the discussion could best be informed by 5
focusing on this issue as the-details emerge, some of I
6 which I think are.on the table now.
But that's my 7
understanding of the general framework of the issue.
8 MR. CARROLL:
I don't think I disagree with 9
that, but I think what we were really -- we were talking 10 about a much narrower issue, and that is, you've got to 11 assume something about the reliability of components in 12 performing a PRA, and I think what we were suggesting here 13 is that the staff ought to make sure that whatever is 14 assumed is achievable, or is a reasonable number, so that 15 the poor COL holder-doesn't end up with requirements for 16 reliability of components that just can't be accomplished.
17 COMMISSIONER ROGERS: Yeah.
I. guess my questic,n 18 really was whether you had some particular thoughts _that 19 might be useful to-the staff in carrying that out, that 20 from your-point of view would be a credible and useful way _
21 to proceed.
That's really what my question was.
22 DR. SHEWMON:
I--think you've' heard-all.we have.
-23 COMMISSIONER ROGERS:
Yeah.
24 MR. CARROLL:
But we.could'go back.and think-25 some more about it..
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COMMISSIONER ROGERS: Well, you've heard of some -
2 of what we have, too.
3 MR. CARROLL:
Yes.
4 COMMISSIONER ROGERS:
Let me just touch on a 5
couple of other things that you may not want to get into 6
because you've said that you still have more work to do on 7
8 DR. SHEWMON:
In some areas.
9 COMMISSIONER ROGERS:
-- in some areas.
But in 10 this question of the ABWR plant cavity area for quenching 11 molten core, you stated that - you thought that more 12 attention needed to be given to that problem, and outlined 13 some of your concerns.
14 I guess my question is, do you believe that 15 really there's a fundamental scientific basis at this 16 point, to really consider some of these questions of 17-debris spreading, debris quenching,.that are useful in 18 this context?
Are the'e enough open-issues there that 19 it's really still diff.lcult to establish some kind of' 20 numbers for this? I know we've heard some numbers, but --
21 DR. SHEWMON:
I think Ivan has been looking at--
22 that.
He could perhaps --
23 MR. CATTON:
I' think I can: address that a little 24 bit.
First, the word " quench" was used early on,_like in 25 90-016.- Now, the word "coolable" is used.- I think the
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18
-1 first thing you'veLgot to decide is, which do you want?
2
-If you're talking about quench, we can do -- or people can-3 do sort of limiting calculations to find you an area.that 4
it certainly_ will' quench.
And it turns out ifiyou do-5' that, it's about twice the.02 where you're sure.
So, 6
it's probably somewhere between the-two.
And my personal:
7 view is that if you get the right group of people 8
together, they probably could have talked that and come up-9 with something sensible.
10 The analysis that led to the.02, in my view, is 11 - j not credible.
I don't say that it's-wrong, but it just 12 doesn't have enough substance in.it that you can believe 13 it.
That's sort of where it's at, and what we're saying 14' is, why don't_ you get on with it and address the question,.
15 first, what do you want?
If you want it quenchable,-then 16 ask what you have to do-to quench it.
'17 The word "coolable" is too open. No matter What-18 it is, it will eventually cool.
So, do you want-it'to 19 quench? :Do you want to keep it-in there?- Do you want to 20 let the base mat.be eroded?. Justowhere are you coming-21 from?
To me, it's-'not cle'ar..=And-I think the staff has 22 shifted in their position._ 90-016, the; wording, to me at I east,.was. clear.
.I know what " quench" means.-
I don't l
23 24-
.know what "coolable" means, which is the-wording-that's:-
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90-016, 2
MR. CARROLL:
Well, we were talking yesterday y
3 about what appears to be a change in the GE' approach.
I'm 4
at least getting the impression that they are now going to 5
rely on external cooling of the reactor vessel.as their 6
first line of defense, but I'm not positive of_that.
7 MR. CATTON:
Well, to me,-that makes sense.
I 8
think, at least analysis-that I've seen, seems to. point to 9
the boiling water reactor as being coolable, by keeping 10 the bottom of the vessel in water.
The PWR is a different 11 story, I don't think it's coolable.
But-the one is.
And 12 they certainly ought to be looking.at that.
That makes a 13 lot of sense.
Use the bottom of your vessel as a core-14 catcher.
15 MR. MICHELSON:
We do intend to talk to GE and.
16 the staff in February, on this issue. We're trying to set 17 up a. meeting now to go in on.it.
18 COMMISSIONER ROGERS:
Well, your statementiin 19-your April 13th letter was, "There's little evidence _that-20 the plant. cavity area will lead to-_ quenching following 21 flooding or-that the ABWR flooding plans 4will'not lead to
-22 ex-vessel-steam explosions", and so on.
23 MR. CARROLL:- That's right.
COMMISSI'ONER-ROGERS:
So, there are some 24' 25 important questions there.
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20 1-MR. CARROLL: This is not the first time we have 2
said this..This goes back to 90-016 and, you know, this 3
is not just an ABWR concern.
If, in fact, we decide we 4-need more area under the vessel, it blows System.80-plus S
out of the water because, with that spherical containment, 6
they --
7 COMMISSIONER ROGERS:
They can't get it.
8-MR. CATTON:
You_need a bigger sphere.
9.
MR. CARROLL:- -- they need a bigger sphere.
So, 10 we've always felt the staff has been slow in getting to-a 11 position, getting the research done that's needed to 12 establish a pouition here.
13 MR. CATTON: The other approach that we've heard.
14 is, you can decide that you just need to keep it bottled-15 up for some period of time.
If that's the decision that's 16 being made, we know how to address that, too.
But, to me, 17 it's not clear what we're shooting for.
-18 DR. SHEWMON:
Let me;just say on the other two-19 items then -- are you through with that?
20 COMMISSIONER ROGERS:- Yes.
l 21
-DR.
SHEWMON:
.That we will be _ starting our-22
-review of - the System ' 80 plus in February,_ and we are 23 starting a
review of the what-I-
guess you'd call p
24.
'" evolutionary plants" --
25 MR.-CARROLL:
No, advance plants.
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DR. SHEWMON:
-- advance plants, okay, more the 2
non-LWR'--
9 3
MR. CARROLL: You're talking about pious prism -
4 5
DR. SHEWMON: Yeah, and can-do, yes. We haven't 6
got a good phrase to summarize or describe all of those.
7 So, if there's questions on those --
8 COMMISSIONER ROGERS:
Well, no, not.those, but 9
you did raise some questions about multiple steam 10 generator. tube breaks in your letter of Septembbr-16th, 11 and I guess the question is really, do you think' a 12 concurrent steam 'line break with steam generator tube 13 ruptures, multiple tube ruptures, is a realistic scenario *t 14 I mean, the question is, - those two together, is that 15 something that you really feel is an important scenario to 16 be considered?
i 17 DR. SHEWMON:
The steam line break comes first?
)
18 COMMISSIONER ROGERS:
Well, together.
I'm not q
19 sure just what the order of business would be, but' --
20 DR. SHEWMON: Well, there are, as you know, very 21 stringent plugging requirements where, 'if. you get even pin'-
1 22-holes that.go-through a fraction.of the tube-wall,. they-23 have to plug-the.' tubes.
24-
-COMMISSIONER ROGERS:
Yes.
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so that'it can, indeed,.take full-system pressure if you.
2 have a steam line break. And th ?t's;why m trying. to say 9
3 if that comes first, then does it increase the probability 4
of breaking the tubes, or rupturing them, and the answer 5
is yes, but I think the staff hav ; insisted on a very 6
conservative limit in that regard, so I person 0.ly would 7
not find the scenario particularly credible.
8 COMMISSIONER ROGERS:
Well, I think you roem to 9
say that you thought that there should be a better 10 technical basis for estimating the frequency of occurrence 11 of multiple tube breaks --
12 MR. CARROLL:
We're talking about the AP600.
13 COMMISSIONER ROGERS:
The AP600, yrs.
14 MR. CATTON:
I think that came about because in 15 the Japanese reactor, where they had the.;taam generrJ.or 16 tube rupture, when they took a look, they f und that there 17 were multiple tubes-that were ruptured.
18 COMMISSIONER ROGERS:- Yes..
19-MR. CATTON:
And the-multiple tube rupturas, it 20-turns out, all of the other tubes that rupt:? red, wi' the 21" exception of the one,-had been plugged, h7 feeling is 22-that somehow'some of that experience 9eeda ta be-brought.
.23
-into:the basis.
We'didn't disagree that-they thould look 24
--into the multiple-steam: generator tube rupture because it i
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basis for doing so could be atrengthened.
2 The' Japanese have gotten very excited about.that 3
- particular incident, to the point that tney are making all 4
sorts of demands on their' industry,'yet nothing seems to l-5 be happening here, and I can't quite figure out why.
And 6
it was sort of to push them a little bit. to come to grips 7
with it, that that statement was written.
And it was.my i
l 8
fault that it's in there.
9 MR. CARROLI - Nell, but the point about AP600 is 10 that multiple steam generator tube ruptures is a much more 2
11 serious situation on a passive plant than it is c,n _ the.
12 population.of BWRs that exist today.
13 MRr CATTON: And at preser t, it'a very dif ficult 14 to come-to grips with the cliff cz3sociated with fluid 15 elastic instability that causes_it.
It's very difficult t
16 to calculate it.
So, you really don't know whers a' steam i
17 generator is relative to che clif f, all you know is that, I
^
13,
well, it hasn't happened very of ten, but -it has happened.
D F
19-ER. SHEWMON: -Multiple?-
20
.MR. CARROLL:
Well, in that sense.--
21
- MR. CATTON: Well, you take a-look, and you find 22 that for whate'ze <- reason, the tubes around it were 23 plugged.
You don't know"if they all.went loose in the 24-saine incident,
- Maybe some of the others had < happened 25-earlier.
You don't know.
t-1-
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COMMISSIO!iER ROGERS:
All right.
I think I'm 2
finished.
i 3
CilAIRMA11 SELIll Commissioner Remick?
4 COMMISSIO!1ER REMICK:
Are we on ABWR, or are we 5
on ABWR System 80-plus?
6 DR. SilEWM0ll We're anyplace you want to put us.
7 MR. CARROLL:
All of them.
8 (Laughter.)
9 COMMISSIO!iER REMICK:
All of them.
Well, like 10 Commissioner
- Rogers, I
had a
number of indiv idual 11 questions on some of your letters from August and April 12 and September, but I think, in general, from what-I'm 13 hearing, that a lot of those concerns ;,u expressed are 14 still open and are not resolved.
So, I',u just try to-15 pick out a couple that may or may not be in that category.
16 How about in the case this was your 16 17 September letter -- you were talking about a prototype 18 control room for passive light w' ter reac. tors such as the 19 AP600 and SBWR.
Have you come to any kind of resolution 20 with the staff on that issue?
21 MR. CARROLL:
I guess I believe that the staff 22 anJ the industry have pretty much come to an agreement.
2?
Initially, when I read some of the staf f stuf f, I was very 24 concerned that we were going to force people to build a 25 full simulator-driven control room very early in the NEAL R. GROSS.
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design.
And from the dialogue we've had with the staff e
and the applicants, it sounds like it's a much more i
3 reasonable approach where you 'might build pieces of it 4
very early but, you know, it would follow the deeign 5
evolution in a fairly sensible manner.
6 So, there does not seem to be a great dif ference 7
of opinion between Westinghouse and-or GE or 8
Westinghouse and the staff, on that issue, which surprised 9
me to some degree.
10 COMMISSIONER REMICK:
I see.
But are you still 11 continuing to look at that? Will we hear more from you on 12 that subject?
13 MR. CARROLL Yeah.
Yeah.
14 COMMISSIONER REMICK:
Okay.
15 MR. CAhROLL:
We'll certainly hear more about 16 it.
You'll certainly hear more about it when we provide 17 our final status report on DAC.
18 COMMISSIONER REMICK:. Okay.
All right.
19 MR. CARROLL:
Because that is a DAC issue on 20 ABWR.
you had'-
21 COMMISSIONER - REMICK:
How about 22 concern over. the question of safety and nonsafety 23 equipment, how that would be handled -- and, once again,.
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than were expressed -- I think that was your Septmnbor 2-letter,' September of '92.
3 DR.
SilEWMoti What was the subject of the 4
September letter?
5 HR. WILKINS: Design Certification and Licensing 6
Policy Issues Pertaining to Passive and Evolutionary 7
Advanced Light Water Reactor Designs.
8 (Simultaneous discussion.)
9 MR.
MICIIELSON:
- Okay, let me got it into 10 context.
11 MR. WILKINS:
Page 58, in the lower right-hand 12 corner.
And this specific item is item II, which is on-13 page 61.
14 MR. MICl!ELSON:
Well, I will cite one examplo 15 where we are continuing our dialogue, of course, and that 16 is in the case of the advanced boiling water reactor.
GE 17 is using a common nonsafety ventilation system which f ans 18 out to all three divisions.
And, so, we are pursuing in 19 great depth, how they are going to isolato one division-20 from ancther, if they break pipes or whatever, in any one
-21 of the divisions. This is still being pursued, and we are 22 waiting for answers.
23 COMMISSIONER REMICKt That_ was one of the issues __
24-in one of the other letters that I-just assumed that.you.
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thinking of was the one specifically in the case of the 2
passive light water reactors, the question of safety-3 nonsafety, as pointed out by Mr. Wilkins.
That was the 4
item 11 in the September letter.
I assume you have no 5
further information on that.
6 MR. CARROLL:
11 0, we really haven't started our 7
review of the passive plants in any detail.
I mean, we've 0
looked at some specific issues like heat transfer test 9
f acilities and things like that, but in terms of detailed 10
-- and been briefed on the general design, but we haven't 11 started a review yet.
12 COMMISSIO!1ER REMICK:
I'm currently trying to i
13 better understand, and I must admit I'm a little confused 14 on things like generic
- ITAAC, discipline
- ITAAC, 15 programmatic ITAACs, and hoping the staf f will provide me 16 with some samples so I have a little better understanding 17 what it is.
But do you happen to know, will security plan 18 be a ITAAC of some kind and, if it is an ITAAC, is that a 19 certification ITAAC, or is it a COL ITAAC7 And then the 20 last question is, would that be something the Committee 21 eventually would be looking at, wherever it might be?
22 MR.
CARROLL:
My impression is that some 23 features of the security design would be covered by an 24 ITAAC, but that the security plan would be a COL holder 25 activity that would be needed.
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COMMISSIO!1ER REMICK So, the separation and so 2
forth, that type of thing, would be in the design, I 3
assume?
4 MR. CARROLL And I'm pretty sure I've seen an 5
ITAAC that covers that.
6 COMMISSIO!1ER REMICK Do you know if the plan I don't knv.
as I say, I'm 7
itself would be a 8
confused between discipline, programmatic, and generic.
9 Would the plan itself fall in that category?
It's a 10 question perhaps I'd better address to the staff, but I 11 thought perhaps you might have thought about it.
12 DR. SHEWM0!1:
You're referring to the security 13 plan?
14 COMMISSIO!1ER REMICK:
The plan, yes.
15 MR. CARROLL:
11 0, that would be something you'd 16 need to -- the COL holdar would need to get to get hinself 17 a combined operating license.
18 COMMISSIO!1ER RFMICKt So, it would be a ITAAC 19 and COL, do you know, or not?
20 MR. CARROLL:
llo, the security plan would be a 21 security plan that needs to be submitted under --
22 COMMISSIO!1ER REMICK:
Part of the problem I'm 23 having, I understand training presumably would be a COL 24 ITAAC. 11ow, what's the dif ference between a training plan 25 and a security plan?
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[
1 MR. CARROLL:
Okay.
Well, I guess I didn't
^
ip 2
think training was.
3 COMMISSIONER REMICK:
Well, I think it's one of 4
the ones propesed but, as I say, I really haven't seen a 5
list of these various things so I
nave a
better 6
understanding but, okay, that answers my question.
7 One other observation -- I'm curious to see if 8
you agree -- I certainly originally viewed ITAACs-to be 9
something that once the plant is built you are able to-10 test whether the plant was built according to the design 11 and meets the acceptance criteria but, in hindsight, as we 12 go through this process, I believe that ITAACs are forcing the vendor, the staff, and ACRS. -- to 13 all of us 14 probably do -- and the Commission -- to probably do a more 15 thorough review of the details of the design and thought 16 process of the relationship between design and operation,-
17 than perhaps might have originally been foreseen.
And as 18 a result, I think we're probably ending up with a far more 19 detailed review at the design certification stage, than we 20 perhaps even imagined a year or two ago.
Is that your 21 observation also?
3 22 MR. CARROLL ' I would certainly agree with that, 23 Forrest'.
L-24 COMMISSIONER -REMICKt It's certainly make us 25 think about things in greater detail.
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'l MR. CARROLL:
It's tough to, you know, take a 2
big description of a system and boil it down to the 3
essentials, of course, as you really think about the 4
fundamentals.
5 MR. MICHELSON:
The drafting of the ITAACs is 6
certainly forcing us to go back and realize the depth of 7
the design and, where it is not suf ficient, you have to do 8
some more thinking and some more designing in order to 9
know that you have an acceptable finally -- you make a 10 final safety determination on it.
11 COMMISSIONER REMICK:
Well, as I-say,-in your 12 letters there were a number of issues that you raised.
I 13 assume that you haven't spoken finally on those, so I'll 14 look with interest to your future letters on those 15 subjects, without going into detail now.
16 MR. MICHELSON:
In general, experience on the 17 ABWR at least, has been that we're getting over a lot of 18 easy ground, but the real sticky issues-have got a long 19 way to go, and those are the ones that will hang you up at 20 the end.
21 CHAIRMAN SELIN: But, on the other hand, I think 22 we owe you a vote of thanks for basically riding out what 4
23 is a difficult storm.
You're seeing through the ITAAC 24 concept with us.
I talked to your predecessor _a little 25 bit earlier, about trying to get geared up so that we NEAL R. GROSS CoVRT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE, N W.
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don't end up with the staff and the vendor going on a 2
bone-crunching schedule and then have a very long ACRS i
3 review.
You're doing all that could be asked, and then i
4 some, to keep on top of this and be prepared so that as 5
soon as you get a document, you're ready to do your own 6
review. We will ask the staf f to get back to you in early 7
January, when they have a more concrete schedule for the 8
next six months, so that there aren't any surprises on 9
this.
10 Personally, I think the full Commission is 11 really very cognizant of the flexibility that you - are -
12 showing, and have shown, as we all try to figure out this 13 process that we've created along the way.
So, thank you 14 very much on that.
15 MR. MICilELSON:
Fine.
16 CilAIRMAN SELIN:
Should we goon to the digital 17 control and protection --
18 DR. SilEWMON :
Yes.
19 CilAIRMAN SELIN:
Before you start, I'd like to 20
.just say I found your letter very, very useful on this.
21 I've talked- -to-the staff-at -some
- length, so -has 22 Commissioner Remick.
I think they've responded to the 23 particular questions we -brought up, in particular.this t
l
-24 question of not mixing up analog and digital. versus -
25 hardware and sof tware, et cetera.
So, we're interested in.
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your report, but I'd like you to keep following up on this 2
so that we don't always end up one cycle, "Oh, we answered
-j 3
those questions, but now we have new questions" kind of 4
thing.
5 DR. SHEWMON:
Yes.
6 CHAIRMAN SELIN:
So, Dr. Lewis?
7 MR. LEWIS: Well,-I'm delighted that you started 8
chat way because I told Paul in an earlier meeting, " Gosh, 9
they've read our letters, so we don't have to go through 10 f t " '.
He said, "Go through it, they have short memories".
11-(Laughter.)
12 CHAIRMAN SELIN: I've read your letter recently, 13 so I don't have to make up any decision.
14 MR. LEWIS:.He's tryig to be quantitative about 15 the duration of your memory.
16 CHAIRMAN SELIN:
I thought the letter was very 17 good.
We talked to the staff on.--
I think it was this 18 week, sometime.
They have a concrete approach to -- and 19 I just want to make sure you keep close to them as we all 20 go through this.
This is all rather new.
I.was actually 21-quite impressed with --
22 HR. LEWIS:
Thank you..
23 CHAIRMAN SELIN:
-- with your letter and with 24 their response to it, but let's catch it on each cycle, to 25 make sure that we continue to converge on it.
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1 MR. LEWIS:
I can only say I read it again this 2
morning, so my memory is even more --
3 CHAIRMAN SELIll Did you like.it?
4 MR. LEWIS:
I thought it was pretty good.this 5
morning --
6 CHAIRMAN SELIN:
Same letter, right?
7 MR. LEWIS:
-- but I'd forgotten what was in it.
8 No.
This is a peculiar subject to deal with right now.
9
- because, as you
- know, we're running a
series of 10 subcommittee meetings on the general subject in which we 11 are trying to do things comprehensively. We are-trying to 12 contact the staff, the vendors, the outside experts, the 13 foreigners, the natives, and so forth, and try to put 14-together a whole packr.ge.
We're not quite all there=yet.
15 We have another meeting scheduled.
So, after that, there 16-will be, hopefully, some' serious and.long-winded, like 17 this one, committee letter that w3 '.1 lay it all out.
18 There may not be.
But, in any case, that's what we're 19 going to.
20
,Therefore, the things that will appear in-that 21 are not yet firm.
So, all I can'.do is reallyLeope'with 22 the things that are in this letter, maybe repeat them in 23 case-both our memories are not quite that. good, and then 24 embellish them slightly for you.
25 The only difference'I have with what you said,
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34 1
Mr. Chairman, was that in some places the staff response 2
-ascribed to us positions which I was not able to find in 3
our letter. And, so, it will probably be well to fix that-4 up.
5 First of
- all, I
think that as a
semi-6 philosophical point, one of our members made the point 7
very well a couple of days ago, when he said that one has 8
the feeling that the NRC approach to digital technology is 9
to treat it as a disease rather than an opportunity --
10-that is, to look at the down side -- treat it sort of like 11 arthritis, something you can't prevent but you can-learn 12 to live with, if you keep it under control.
And I-think 13 that encapsulates part of the problem.
14 Our letter, we tried very hard to tell you, 15 although you already ; know, what the advantages were of 16 digital technology.
It really is more reliable than 17-analog technology.
It really is further advanced.
You 18 really can-have more functions. You can build in'self-19
. testing out of your kazoo.. Every time you turn.on your 20-prin.ter or your home computer, it tests itself.
Probably 21 even your microwave oven tests itself.-
So, that these 22 capabilities that are endemic to-the technology are real-
.23 opportunities-and not'something to be sort of kept under 24 control. And 'I think it's important to bear that in mind..
25 CIIAIRMAN SELIN:
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staff either has come to believe that, or they have found 2
briefers who believe that, that they now send over to see-3 me --
4 MR. LEWIS:
I see. Well, I'm glad to know that.
5 Cl! AIRMAN SELIN:
-~ but we've been getting very 6
much that same appreciation, more recently, from the staf f 7
especially.
8 MR. LEWIS:
Okay.
Well, I think there has 9
actually been some forward progress since we started our 10 series of meetings.
It does pay to exchange views from 11 time to time.
12 This morning, during the very -- one of the very-13 few dull moments in an ACRS-meeting --
14 CHAIRMAN SELIN: Somebody else was speaking, Dr.
15 Lewis.
16 (Laughter.)
17 MR. LEWIS:
Except for the sign of the effect, 18 you've got it right.
The dullest moments are when I'm 19 speaking.
But I realize that my computer at-home is 20 churning away at this very moment, doing some miserable 21 calculation, and it will be for four days.
So, I
22 multiplied numbers together, and I found that during these 23-four days it will do 4 x 10" o n e r a t i o n s,-- a n d I will 24 expect it to do them without a mistake.
And, you know, 25-that's really' pretty high reliability.
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I'll throw away five orders of magnitude if you-2 start to quibble, but that's very high reliability, and we 3
expect it.
And the reason is that it's easy to build in 4
error correction on these things, not just error detection-5 but error correction, and you can do it from now 'til next 6
- Thursday, and it's
- very, very easy.
That's high-7 reliability, and it's built into the system, and it's an
[
8 opportunity which is not available with the analog system.
9 So, one really shouid look at these things with a forward 10 looking view.
11 I' don't know if I I've got to throw in-one 12 anecdote.
A friend of mine who is-a lawyer -- and that 16 13 not a contradiction in terms -- told me about a Supreme-14 Court decision of about 1975 which is extremely 15 interesting because it's one in which a case was decided 16 because the probability that something could have happened-17 by accident was 101 , and the Court decided that 10'' was 2
18 really very unlikely, and it opined-that way, and the-19 Chief Justice at that time filed a dissent--saying that 20-that didn't convince him.
-21 So, you know, the understanding of these low 22 probabilities and how to -- I'll give you-the reference 4
23-
-1:.ter -.to understand these things, is not as widespread.
24 as one wish:it were.
'25 So, within that context that we all understand NEAL R. GROSS COURT REPORTERS AND TRANSCRIDERS 1323 7 HOOE ISLAND AVENUE, N W..
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i the capabilities, we looked at what the staff is doing 2
about that, and the staff will tell you that they are, 3
- indeed, in contact with many experts.
We think not 4
enough, and not as broad a group.
We can list much larger 5
groups of experts.
I've spoken to some of my experts 6
about their experts, I've no doubt that they've spoken to 7
their experts about our experts, and so forth.
But there 8
is a fairly large community of people out there, who know 9
a great deal about the subject and have been dealing with 10 it in connection with the telephone company, the money-11 handlers were very concerned about these matters, and so 22 forth, and there's a lot known, and it is not true that 13 all the people who are concerned about computer 14 reliability are high-in-the-sky computer scientists, there 15 are some real people out there.
16 I heard a talk a couple of weeks ago, in a 17 completely d i f f e r e n t c o n t e x t,- by a really first-rate 18 computer scientist, on the question of whether there is 19 such a subject as software engineering.
And it was an 20 extremely interesting talk because her conclusion was that 21 there is not, that the sof tware _ business has not yet 22 reached the point at which it can be called in.any.way 23 engineering.
And there's an. interesting definitionLthat-24 she used, namely,.that engineering is the development of 25 a craft to the point at which even dunderhea'ds can do it NEAL R. GROSS COURT REPORTERS AND TRANSCHiBERS 1323 RHODE ISLAND AVENUE, N W (202) 2344433 -
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and get the right answer.
It's a way of describing.
- And, 2
therefore, you need handbooks.
And as she correctly 3
pointed out, there are no hundbooks about how to do it.
4 That's why we have many of these conflicts.
But, in any l
5 case, I think most people are agreed that the advantages 6
are there.
The staff then has narrowed the field of 7
discourse fairly dramatically.
8 If you look at the Federal Register 11otice,
9 which I just got out this morning because I have a short 10 attention span, the Federal Register Notice of August of 11
'92, lists concerns as the use of software,.the effect of 12 electromagnetic interference, the use and control ~ of 13 configuration equipment, the effect that some digital 14 designs have on diverse trip functions failure-specific to 15 digital' hardware, the effect of system integration, and 16 the commercial dedication of digital electronics -- I'm 17
'not sure what that means, but commercial use, I assume it 18 means -- of digital electronics.
19 Then it says the most notable of these concerns 20 is the use of software.
But then the staff has gone out 21 pretty narrowly to the question of sof tware, which I would 22-agree is-a major point.
It's gone a step further.
It's 23-gone to the: question of - what is: called " common mode 24 failures" in software.
They-made the point that if you-25 use a specific suite - of software in? several: redundant NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE, N W.
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1 channels, you're going to get the same mistake in these i
several redundant channels, that's a form of common mode 3
failure.
4 I've thought through the anecdotes that I know -
5
- and-I don't think there are any good statistics -
.I 6
thought through the anecdotes that I know of software 7
failures, and I don't know any of them that are actually 8
common mode fallures due to the same mistake in different 9
channels.
They are usually strange failures. -They are 10 unique, always unique, and they have to do with an error 11 in coding in a channel, which usually doesn't prevent a'
+
12 function which one wanted to occur, but usually generates 13 a function that one didn't want to occur, make something-P 14 happen.
The accident at Bruce was a single mistake in a 15 line of code.
It was actually miswriting -of - a return 16 instruction, and it produced a.
motion that wasn't 17 intended.
Actually, what it did was to recall a return 18 address on a subroutine from the day before, which 19 happened to be stored in memory at the location which was.
20
. referred to at a later time.
.I know many such things like 21 that.
22 There is a lot known about how to validate'and 23 verify software.- Much of the anecdotal information-that 24 we all-hear, - things that have come from the telephone 25 company and so forth, have te do with extremely.large NEAL R. GROSS
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systems, although lines of code are not a measure of the 2
complexity of a program, but you're dealing often with 3
things that involve millions of lines of code.
4 Dave Parness, who was the principal advisor to 5
the Canadians when they were trying to do the Darlington 6
thing, is a guy whose popular fame -- not among computer 7
scientists -- whose popular fame comes from his concerns 8
about the software that goes with the Strategic Defense 9-Initiative, which is an order, many orders of magnitude 10 different from the kind of thing one has in the nuclear
-11 business.
12 So, we tend to be skewed in our feeling about 13 what can go wrong with the software, by anecdotes which 14 come from more complicated systems than we-have in the 15 nuclear business.
So, I guess.I'm not convinced that the 16 common modo failure issues are the most important ones.
17
-Now, what we said in our letter which I was told 18 to tell you about, is to emphasize also that-software 19 doesn't wear out,-ano there are no worn out electrons in 20 the system, that when it works, it works, and it'll go on 21 working for a long time.
That's true of the circuits, 22 too.
If they are not over-heated, over-stressed, over-23 voltaged, and so forth,= they work extremely well.
But.
24 they are more complicated, and-so forth, and there:-- a 25-viewpoint I forgot to mention -- and there are no explicit-NEAL R. GROSS COURT REPOR1ERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUES N W.-
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safet y standards. That's the real issue because the staff 2
may be improving, but in the things I've heard, they_
3 continue to speak of probability of failure per million 4
!.ours of operation, that sort of thing, the kind of S
measute that is appropriate for valves and switches and 6
mechanical devices, but is really not useful for sof tware.
7 Software, if it works, will continue to work 8
and, if there is a defect in it, the defect will not be i
9 one that will be found by testing it once.
It will be 10 found only because some other malfunction exercises some-11 subunit in the software that has not boot, tested under 12 normal conditions.
And that's why, of course, full-scale 13 validation and verification can only be done by mapping 14 the entire input space to the entire output space, on a 15 theory that something else will exercise any given part of 16 the input space.- You cannot count on.just the expected 17 normal inputs, in the same sense that if you take a homo-18 computer and hit 17 keys at once, nobody really knows what 19 will happen.
Nobody has broached and did the response _to 20 17 keys at once, in general, something-like that is_the 21 case.
22-So, the fact that the software in computers is 23 simpler means-that,- in-f act -- one c.amputer sci entist -I've -
24 consulted, who is _ a realistic person, who is an expert in 25 V&
V,_ believes that it is just on the verge of being NEAL R. GROSS COURY HEPORTERS AND TRANSCRIBERS L
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I suf ficiently sirnple to do full-scale V &
V.
In the 2
subcommittee meeting we had on this subject, at which a 7
3 number of-nembers of the staff were present, there were 4
differences of opinion.
We had some people standing up 5
and saying you could, and should, do e formal V & V cn 6
nuclear software, others said it was impossible; mostly 7
they were calculating lines of code.
8 So, I think that the bottom line on this, you 9
know, we wrote you a letter saying that the specific item 10 that the staf f had concluded, which was to require analog 11 backup for digital sof tware, was inappropriate because one 12 could have just as reliable digital stuff.
You know, you 4
13 really don't have to tow a horse behind your car.
The car 14 may break down, but you don't need a horse.
We think that 15 was inappropriate.
16 The staff response is to weaken that very 17 slightly, but I'm a suspicious person because the way in 18 which they weakened it is by saying that the staf f will be 19
-flexible and will consider alternate solutions-involving 20 digital systems, if they are suf ficiently simple.
And
'21 that is a very subjective judgment, presumably to be made 22 by the same people who originally wanted only analog 23-systems.
So, we'll wait and see.
-But-it is true that 24 they've weakened that insistence-on analog backups.
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about whether one ought to -- how one ought to validate 2
software, and how one ought to assess the reliability of 3
aof tware, in what terms, the sort of thing, I think you're 4
going to have to wait until we finish our subcommittee 5
sequencs and write you a real letter.
6 CilAIRl4All EELIlit A certain amount of skepticism 7/
[9 always called for, but I really do ask that you 8
continue to work with tha staff, and watch what-they're 9
doing because I'd be very interested myself, since I've 10 had sema acquaintance with this field.
Il 14R. LEWIS We are trying to work with the 12 staff.
13 C il A I R M A 14 S E L I lit And you are continuing to do 14 that, to do a letter comparable to the one you did with-15 the --
16 MR. LEWISt I think I can say that we find-it ns 17
-easy to work with them as-they find it to work with us.
18 MR. CARROLL:
I would add just one point.to 19 this.
One of the hats I.'m wearing on this is,_I want to 20 continue to challenge Hal on his bad-mouthing of the 21 staf f's capability, if you will, because I think there are 22 some very good people on the staff that are trying very-23 hard to do a good job in this area.
So, I'm going to be 24 a check-and-balance on Hal here.
25 MR'. LEWIS:
llave I said otherwise?
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MR. CARROLL:
No.
2 MR. LEWIS:
Okay.
3 CilAIRMAN SELIlls Could I ask -you a favor, 4
Doctor?
5 DR. S!!EWMON:
Yos?
6 CilAIRMAN SELIN:
Would you mind going to the 7
risk-based regulation next, and then como back to the F
8 plant life extension rulo?
9 COMMISSIONER ROGERS: Just before wo leave this, 10 with your meeting, llal, will you como out with something i
11 helpful in the way of identifying ~ standards for -I&C 12 systems?
13 MR. LEWIS:
llow can I know?
I circulated among 14 the members of ACRS a cartoon I took out of the paper the-15 other day, in which Senator So-and-So was ' asked - his-16 opinion about a given tax bill, and he said, "That will be 17 a committoo decision". - And the reporter says, "Is that.a 18 cop-out?"
And he-says, "No,.it's an oxymoron".
So, I 19 don't know --
20 COMMISSIONER ROGERS Well, lot mo just put-It 21 this way. -Is-that on your agenda?
22 MR. LEWIS:
Yes.
Tho question is-whether I can-23 got it through the committee.
You really want all the 24
-facts.
.2T COMMISSIONER ROGERS:
Well, I was thinking from
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your own -- in your own committee.
2 MR. LEWIS:
Oh, I have no problem -- no, that's 3
a subcommittee. We have to go through the full Committee.
4 I think we can try to do that.
On the other hand, there's-5 a
delicate balance between the ACRS advising the 6
Commission and the ACRS functioning as an alternate staf f.
7 So, we have to be careful about that.
But I think we will 8
have some -- I hope we will have some ideas.
If we don't 9
have any ideas, don't pay us.
10 COMMISSIONER ROGERS:
Okay.
9 11 COMMISSIONER REMICK: I have one question in the 12 I&C area.
It's not directly related to what you're 13 saying, but indirectly.
14 Recently, the-Nuclear Safety Research Review-15 Committee wrote a letter also, on the I&C area, and it was '
16 very good letter.
And in a number of areas they agree 17 with the type of things that you've been putting in your 18 letter, and I think that's good, and some _ overlap is 19 important, but just a question went through my mind when 20 I read that.
21 Han the ACRS and the NSRRC, whatever it is,- have 22 you ever talked about what is a logical-area of overlap, 23 or areas in which you concentrate on?
~*
24 MR. LEWIS: No, we have not.
I didn't see their-25 letter until yesterday..
I found it interesting.
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Especially, I found the fact that they recommend work on 2
genetic systems, which I think meant generic, but they did 3
say many of the things we said.
4 There were two differences.
One is that they 5
are happier with the questions of expert systems, fuzzy 6
logic, and things like that, than I am.
And they also 7
were concerned about the human interf ace with the computer 8
systems, which we've not been dealing with, but I thought 9
it was a pretty good letter.
10 COMMISSIOllER REMICK:
Oh, yes, it definitely 11 was.
But my point is, have the committees thought _about 12 what are logical areas to cover and not cover, and so 13 forth, something like you've done with AC&W, in general.
14 MR. LEWIS:
Ho, we have not communicated.
15 COMMISSIO!IER REMICK:
Okay.
16 MR.
WILKINS:
What has happened in that 17 occasionally one of us will sit in on one of their 18 meetings, as an observer, not a participant, of course, 19-but as an observer.
I've done that, Carl has done that on 20
- occasion, and
-I believe that Mr.
Fraley hasl open 21
. communications between them so that we get copies of' thei'r 22
' letters and they get copies of our letters, but there's no 23 MOU, so to speak.
24 MR. CARROLL:
I would also point out, Forrest, 25 that most of our interaction with the' staff in this area NEAL Ri GROSS.
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is with NHR, it's not with Research.
2 COMMISSIONER REMICK:
Oh, yes.
I see.
Thank 3
you.
4 MR. LEWIS:
Well, they say RES has to become 5
aware of artificial intelligence expert systems, neuro-6 networks, fuzzy systems, that sort of-thing.
I've never 7
seen anything goad come out of neuro-networks, but a lot 8
of fun.
9 Cl! AIRMAN SELIN:
Dut they did it with brains, 10 Dr. Lewis.
11 MR. LEWIS:
On a special neuro-network.
12 CllAIRMAN SELIN:
Oh.
13 COMMISSIONER REMICK:
The certainly did stress 14 the.Importance of knowing what others are doing in this.
15 area.
They suggested that the NRC should:not be a lead because there's other 16 researcher in this area, but 17 research going on, but it should be familiar with what's 18 going on in this area.
19 MR. LEWIS:
Well, in fact, that's a general 20 pattern that goes through a number of things.. There are 21
-a number of places where NRC probably shouldn't be doing 22 research, but shoul'd in fact be-taking advantage of the 23 rest'of the-world.-
24
. COMMISSIONER REMICK:
I agree.
25' CilAIRMAN SELIN:
Dr. Shewmon?
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[
f 1
DR. SHEWMON:
Okay.
Unfortunately, risk-based 2
regulation comes back to our friend Hal again, who is 4
3 almost becoming speechless.
4 MR. CARROLL:
Which may not be a bad thing.
5 DR. SHEWMON:
What we have here --
6 MR. LEWIS:
I'll do it.
7 DR. SHEWMON:
I know you will, you'll get it in 8
a minute -- that we think there's promise here, as you did 9'
in your original letter. And we were quite unimpressed --
i 10 while we thought, you know, the draft document which we 11 got from the staff wac preliminary, and look forward to 12 seeing another version of thnm, or an improved version 13 later on, you --
14 CHAIRMAN SELIN:
I should just say something'in 15
. response to that.
I've gone into some depth with the 16 staff about exactly what they're doing, and some is good 17 and some is'less good.
I do believe that there's an f
18 attempt to have a true coordination 'at a -- I wouldn't say 19 a low level, but at a working level -- to see who is 20 applying what methods, what's working, what's not working, 21 not from a management' initiative, but more of'an after-22 the-fact what's going on, who's using which methods, what 23
-- are' people doing them well or not.
I still think that1
-24 what's missing is more of a. top-down equivalent of an EO.
25' coordinator or some coordinator says -- you, ideally, PRA NEAL R, GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE, N W.
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is not a field by itself, it's something that everybody 2
should be doing, but there should be objectives at the 3
rate in which PRA kind of thinking is introduced into all 4
areas of regulation, and I don't believe we're to the 5
point of having such objectives or such goals at the top.
j I mean, coordinator is a very 6
So, your point about 7
vague phrase, but your point about having somebody who is 8
seeing through the implementation of PRA methods and could -
9 report to the - EDO and the Commission about how we're 10 doing, that's not been set up yet.
The lower level point 11 about are people using the right tools, are they using 12 them well, are they talking to each other, I think there's 13 much more progress in that area than in the first area.
14 DR. SilEWMON:
1.et me bring up two items.
- One, 15 at this meeting, we had a presentation from people at 16 River Bend, who have an active PRA for their reactor and 17 were using it to drive or ' influence their maintenance 18 program to increase reliability.
And what was motivating 19 it and making it pay off was improved plant reliability, 20 not what they were getting in terms of relief from the 21 NRC.
22-The other thing, when we heard about this 23 sometime back -- Murley was down talking.to-us Bill 24 Windblatt, who was over on that side of the room, leaned 25 over to him and said, Do you think we can also-get this NEAL R. GROSS -
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mwrem?-
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FC 1
far enough to where we can get relief from some of the 2
pieces of paper which don't do anybody eny good and we 0
3 might even agree on that they don't do any good", and i
4 Murley was not total.1y discouraging, but promised nothing.
5 So, I hope that that will be another part of it that it l
6 will come to view.
7 MR. LEWIS:
Sure. Well, in f act, most of it has 8
been covered.
In fact, one other thing, I think, and I 9
think the Committee thinks, that the idea of risk-based 10 regulation in the initiative --
11 VOICE:
You need the microphone, !!al.
12 MR. LEWIS:
-- think that the initiative that 13 the Commission took in March was very sensible, very 14 important.
It's appropriate not only for the direction of
'15 Commission resources, but also for industry resources.
16 I think it would be inappropriate to respond in 17 any detail to the paper we saw.
It was just very bad, or l
18-let me put it differently, it was newborn, and it needed 19 to grow up a little bit.
But there is a point that goes 20 through all of these things.
One is one there are 21 several points -- one is one that I mentioned-earlier, 22 that. I need to mention again, the other '.s the question of' 23 whether, in fact, one can push it to the-point of getting 24 relief from regulations also, which is the point that Paul 25 mentioned.
- NEAL R. GROSS-CoVRT REPORTERS AND TRANSCR!BERS 1323 RHODE ISt.AND AVENUE, N W.-
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The ono I want to repeat -- and then I'll shut-2 up Lecause I obviously have to -- ia that the thing we see 3
l that is unfortunately coherent within the staff, is the 4
use of PRA through bottom-line numbers without-5 uncertainties.
Thera 16 just a constant search for 1
6 l
thresholds, for thresholds for action, or thresholds for-
?
7 inaction, throughout the staff.
F.very small group doing 8
it independently has at least that in common.
9 When they speak of unenrtainty, they speak of it 10 as -- to use the word again -- as a " disease", as aone t
11 defect in the calculation.
We heard only this morning of 12 an effort to reduce the seismic hazard probability by 13 bringing the two cordenders together and forcing them to 14 a consensus, which may be wrong becsuse, in fact, the 15 diversity of views is part of the -information you have 16 about the system.
So, there's this quest for certainty.
17 The question of whether
.you can foldL because everyone.is 18 probabilities through-the PRAs 19 going to have PRAs whether you can -fold the 20 probabilities and the uncertainties together into a.
21 regulatory procesa, is a very deep and complicated one on 22 which very-little progress has been made.
23 The. committee wrote a letter six months' ago, in 24 which we sort of skimmed by the idea that-one might be 4
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when Ernest and I wrote additional r 'mments to spelling it 2
out inta little more detail.
But I-think some serious-3-
attention to ways in which one can incorporate bottom-line 4
numbers and uncertainties into the system, 11nto 'a 5
decisionmaking system, on t. : part of the staff, would be 6
well worthwhile.
7 For example, everyone knowe that if you have one 8
risk which is, in terms of bottom-line, less severe'or 9
le ss threatening than
- another, but it has larger 10 uncertainty, all your effort will go onto that,.
even 11 though your bottom-line number is lower, because, people 12 won't be sure. They will draw error bands, and then their 13-hearts and minds will go to the top of the error bands 14 and, if the top of this error band is higher than the top 15 of that error band, that's where it will be.
16 Now, there's a quantitative way to do this, and 17 one needs to devote some attention to trying to find some 18 decisionmaking mechanisms and get them into the business
+
- 19 that take advantage.of the cases which are increasing-in 20 number, in which we have information both about-bottom 21 lines, about uncertainties, and I. don't see that happening 22 within the staff.
I think you-can't get to risk-based
_ regulation-until you begin-to understand how to do that.
23 24 That's all I have te say.
l' 25 CHAIRMAN SELIN:
Commissioner Remick?
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COMMISSIONER REMICK:
I don't-think I-have-a<
2
' question on this subject.
3-CHAIRMAN SELIN:
Commissioner Curtiss?
4 COMMISSIONER CURTISS: I do think you're correct-5 that much more detailed work needs to go.on on the risk--
-6 based regulation initiative.
We've made'some progrese,.
7 but in terms of_what I think the Commission expected and 8
hoped for, and I do think it was a pretty sensible request 9
to make, from looking at the draf t paper and in listening 10 to your comments, I'm hopeful that that kind of attention.
11 at a higher level within the agency, will help bring this 12 together in a coherent way with some of the_ other_
f 13 activities thct you've alluded to.
14 There is some very interesting-work going on I
15 that I would commend to you.
I've actually personally met 16 with two people, Bill Vesley and' Ernie Lof tgren, from 17 SAIC, and in addition to the briefing that the Commission 18 had with Herschel Specter, I-actually think some of the
. things that they're talking about"-- I'm'not sure-they're 20 exactly what you're suggesting, Hal, but they might - be s
as 21
' worth-inviting in for a presentation, 1f you-haven't, 22
'a group, met with them.
They had some vety, I think, 23
- stimulating. thoughts.
. 24-
- CHAIRMAN SELIN:. I would just say that the staf f 25 is less -oriented towards how to manipulate compound NEAL Rc GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLANO AVENUE, N W.
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probabilities that have uncertainties around them, than 2
they are to using the concept of risk, whether it's a 3
point to focus their incremental ef forts, and with all the 4'
methodological questions that you bring up, I think your 5
calling attention to the lack of coherence, et cetera, has 6
been very helpful.
And this sounds like a funny request 7
to make of a scientific advisory group, but you could, I 8
think, be of more help to us and to the staf f, by focusing 9
on the management progress and instilling PRAs into the 10 way we do busineso, than the scientific way of how we do, 11 you know, probabilities of compound events.
12 I shouldn't have used EO as an example, prose it 13 a better example.
I mean, we expect everybody to use good 14 prose, but having an assistant director for prose isn't 15 the way to get it done.
It's to keep functioning on just 16 good, analytic work.
And I think that the help you've 17 given and the pressure you've given to apply these methods 18 to get some coherence has been very helpful, and I hope 19 you'll keep up -- not in PRA interactions with the staff, 20 but in all interactions with the staff since, ideally, 21 risk-based priority setting and analysis should suffuse 22 everything that they do and everything that you do.
23 Mr. Carroll?
24 MR. CARROLL:
One thing that we talked about 25 this morning that you didn't mention, Hal, was, we were --
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE. N W.
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\\
1 I guess at least Hal and I both were -- at the prospectus 2
for the academy study on risk-based decisionmaking.
If 3
you haven't seen it, it's very worth reading.
It looks 4
like a
very worthwhile project because risk-based 5
regulation is going to have a
problem with public 6
acceptance, and that's the issue the academy study is 7
dealing with.
8 CHAIRMAN SELIN: Okay.
Dr. Shewmon, shall we go 9
on to the plant life extension, since that's -- you know, 10 we've had a number of meetings on that with the staf f, and 11 it's very, very important.
12 DR. SHEWF'N:
Well, yes, this may have been 13 overtaken by events. Our concern has been that though the 14 rule sounds reasonable, that we will continue licensing 15 basis and protect the public health and safety, when it 16 has gotten down to the implementation, it seems to sort of 17
-- the staff seems to feel that there's this great cliff 18 that we b.ve to climb over, or fall off of, we may fall 19 off of, when you come to the end of 39 and a half years.
20 And, so, what they have come out with, their 21 position, is sometimes -- which we feel f ar exceed, or do 22
- exceed, the current licensing
- basis, and also are 23 appreciably more than can be justified by any sort of a 24 health and safety argument.
25 I think we saw -- well, we did see the recent NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVENUE. N W (202) 2344433 WASHINGTON. O 0. 20005 (202) 234 4433
56 1
memo of Commissioner Curtiss with regard to plant life 2
extension in the maintenance rule, and I think that's an 3
attractive avenue, and I think that something of that sort 4
would certainly be more consistent with what we think 5
should be the branch here than what is ending up with the 6
implementation.
I think it's more of a statement of 7
concern about where the implementation has been going.
8 CHAIRMAN SELIN:
We thought, when this meeting 9
was scheduled, that we would have a chance to discuss your 10 views on metal fatigue and environmental qualifications, 11 but the staf f has decided they want to do another cycle on 12 their own technical position.
So, I'm not sure w.
ther 13 that would be as fruitful to get into at this point, as we 14 had thought when we had scheduled the --
15 MR. CARROLL I'm glad to hear that.
16 CHAIRMAN SELIN:
Did your colleagues want to go 17 into some other of the points on license extension?
You 18 know how important we take this.
We think this is just a 19 terribly important issue not just for license extension, 20 but for continued operation of existing plants and provide 21 a framework for decisions on capital investment throughout 22 the life of the process.
23 MR. WILKINS:
Am I correct that there are no 24 present utilities that have committed themselves to file 25 an application for --
NEAL R. GROSS COURT REPORTERS AND THANSCRIBERS 1323 RHODE ISLAND AVENUE. N W (202) 234 4433 WASHINGTON, O C 20005 (202) 234 4433
57.
-1 CHAIRMAN SELIN: Well, you know,- even af ter they 2
file the-applications, they're not' going to get to them, 3
so that's a -- there's an enormous amount of interest and -
4 a-lot of waiting to see what we will do and what other 5
utilities will do.
So, the closest thing to-an 6
application is a statement of intent by the Babcock and 7
Wilcox Ownerc G :oup, to try to do a generic application, 8
which will at _ *st apply to their seven plants.
l, 9
Commissioner Curtiss' position, which I hopo.
I 10 will show a lot of sympathy -- I'll put it-in the, broadest l
11 way, is that they should not be allowed to go too far i
12 forward until the staff and the Commission have settled a j
13 number of policy issues which could have a severe impact 14 on how they spend their resources.
1 15 So, the answer to your question is, there are no 16 applications close to being on the table.
But I think a-17 realistic answer is that we are-not yet in a position to 18 have appl'ication so close to being on the table, so the 19 homework evolves on the agency, probably.
The first 20 expiration is 2012, I think it is, but -- I'm sorry, 2006, 3
21 the first one but,'in any event, there's a lot of work-to-22 be done between now and then, and it-would be timely to 23 get this work done.
24 COMMISSIONER CURTISS:
Could I
-make a
25 prospective request of you in this area?- You'll-have.an NEAL R. GROSS' CC#JAT REPORTERS AND TRANSCRtBERS 1323 RHOOE ISLAND AVENUE. N W.
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opportunity, I think, in the next several months, to take 2
a more careful look at the maintenance guidance, -which is 3
now winding its way through the V & V program, and the 4
results of which will be finalized, I think, in a report 5
at the end of December, that should be available.
I 6
encourage you to take a look at that as well as the two 7
interim reports.
8 The particular issue that I guess I'd be anxious 9
to hear your views on as you look at the maintenance rule 10 and consider in more detail the license renewal rule, as 11 the staff's process works its way forward, is one that I 12-raised in my memo so you may be familiar with it, but let 13 me just sharpen it up a little bit.
14 I'm interested in knowing whether it makes sense 15 from the standpoint of taking a
look at.important 16 structures and components within the plant, to place 17 greater reliance on the maintenance rule and the 18 activities that we will be undertaking in that context, to -
19 the point where, if we did that, if'we are comfortable 20 with those activities and all of the attendant machinery 21 that goes along with that, surveillances and tests and so-22 forth, we could truncate the need to do:a. component-by -
23 component analysis-of the aging effects as currently 24-required,:LI think, under the staff's view of what an IPA.
25 has to include.
NEAL R. GROSS COURT REPORTERS AND TRANSCRtBERS 1323 RHOOC ISLAND AVENUE, N W, (202) 234 4433 WASHINGTON, D C 20005 (202) 2344433 :
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g..
59-1 The staf f is going to take a look at that issue.
2 It was the subject of discussion at the meetings that the-3 Chairman referred to on December 7th and 8th.'
What I'm 4
suggesting there, to put-it in a different context,_is a-5 concept that in some respects would be similar to the 6
Japanese approach to the maintenance of their f acilities, 7
which is to say, as you know, no set limit on the license 8
in Japan but, instead, a maintenance process that provides 9
for a continuing, evolving and rolling approach to the 10 maintenance' of structures and components important in 11 those facilities.
12
- Now, recognizing that there are significant 13 dif ferences between the details of what may be required in 14 the Japanese maintenance program vice what we have here in 15 our maintenance approach, the concept at least is that you 16 would have the sort of approach that they take to the 17 maintenance of their f acilicies, without a drop-dead date, 18 if you will, 40-year license in our case, required, of 19 course, by the Atomic Energy Act for a limit but, instead, 20 view it as an opportunity to address those aging 21 mechanisms that are not unique to the _ license renewal-22 period through a process like that. -And as you get into a
23 the evaluation of the maintenance guidance, if you'd bear 24 in mind, as I'm sure you will,-its-potential application 25 in that context in the licenae renewal, at some point I'd NEAL R. GROSS COURT REPORTERS AND TRANSCTBERS 1323 RHODE ISLAND AVENUE, N W.
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60 1
appreciate your thoughts on that.
2 MR. CARROLL:
And one should add to that also o
3 the reliability assurance programs that-are going to be 4
required
. of.
the advance plants because that is 5
maintenance, too.
6 COMMISSIONER CURTISS:
That's right.
7 CHAIRMAN SELIN:
This has been a fun::y meeting 8
because, on the one hand, you've shown, as I said earlier, 9
extraordinary flexibility and willingness to ride with the 10 punches on this overall review of the ABWR and, by 11 implication, of the System-80, 12 On the other, there's been more motion faster by 13 the staff both in a management and in.a technical. sense, 14 in all three of-the other areas -- the specific technical 15 areas'on fatigue-and equipment ~ qualification to go at 16 license
- renewal, the questions of I&C and how one 17 approaches valid instrumentation and control of..not.only 18 digital systems, but sof tware,- and this whole-question ~of 19-probabilistic work which-makes it hard. to. exactly 20 synchronize their appearances and your appearances, but'I-21 think that's a very good sign.
22
-When I first came on here, as you know, I was-23 dismayed by - how of ten - the staf f and -- the Committee. just -
24-seemed to be talking past each other. Land the-fact that 25 you're in head-to-head discussions, confrontations, useful NEAL-R. GROSS COURT REPORTERS AND TRANSCRIBERS -
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61-1 discussions, et cetera, meetings, that it's less easy to 2
predict exactly what will be topical when'one'of these 3
meetings comes up.
Some of your letters are overtaken by 4
them because they just do everything you say and,-et 5
cetera.
6 MR. CARROLL:
Well, I hadn't noticed-that.
7 CHAIRMAN SELIN:
But, you know, the bottom line-8 of all this is a thank you for what you've done so far, 9
and encourage you or ask you to keep, if possible, even in 10 closer touch with the management of-the staff.
There's 11 always been close level between the " working level" and 12 the ACRS, but there seems to me now to be much - more 13 willingness at the top levels of staff management to 14 interact with the Committee, learn from each other,-and 15 keep it going.
16 The meetings may not be quite as exciting as 17 otherwise, but I think the progress is much greater.
So, 18
-we thank you for the progress and for the willingness to 19 take on what_I think is a much more productive working 20
-interaction.
Thank you very much.
21 DR. SHEWMON:
Thank you.
22 CHAIRMAN SELIN:
Merry Christmas to you.
e-23
- DR. SHEWMON:
Merry Christmas to you all.-
24 (Whereupon, at-2:52 p.m.,
the meeting was 25 adjourned.)
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE If, LAND AVENUE, N W (202) 234 4413
~ WASHINGTON. D.C. 20005 (202) 234-4433
CERTIFICATE OF TRANSCRIBER This is to certify that the attached events of a meeting
?.s of the United States Nuclear Regulatory Commission entitled:
f,-
TITLE OF MEETING:
PERIODIC MEETING WITH THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (ACRS)
PLACE OF MEETING:
ROCKVILLE, MARYLAND DATE OF MEETING:
DECEMBER 11, 1992 were transcribed by me. I further certify that said transcription is accurate and complete, to the best of my ability, and that the transcript is a true and accurate record of the foregoing events.
l Cl W1 1
V
(
/
~
Reporter's name:
PHYLLIS YOUNG a.
-s
' 18' HEAL R; GROSS COURT RIPCHtTIR3 AND TRANSCRIRER$
1323 RMODE ISLAND AVENUI, N.W.
(202) 234 4433 WASHINGTON. O.C.
20005 (202) 232 4 600
n t,
na asa UNITED STATES og.
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.- NUCLEAR REGULATORY COMMISSION n
y
g-WASHINGTON, D. C. 20555 i
' December 3,;1992
-MEMORANDUM-FOR:
Samuel J. Chilk,-Secretary of the Commission.
FROM:
a nd F.
Fra ey,.Execu ve Director, ACRS
SUBJECT:
-ACRS: MEETING WITH THE NRC--COMMISSIONERS ON DECEMBER 11, 1992 - BACKGROUND INFORMATION The ACRS is scheduled to meet with the NRC Commissioners on Friday, December-11, 1992, between 1:00 and'3:00 P.M. to discuss items of mutual. interest, - including the following.
Background material related to these matters is attached:
- 21. Status of ACRS Review of Advanced Reactor Desions - C. Michelson I
and J. Carroll (PP. 2-52) 2'. Dioital Control and Protection Systems - H.-Lewis (PP. 53-64) 3.,Imolementation of-the Plant' Life Extension Rule -
P. Shewmon
- (PP. 65-83)
- 4. Risk-Based Reculation - H.' Lewis (PP. 84-87)
Ittachments: As' Stated cc:-ACRS-Members
'ACRS Technical Staff 4
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NUCLEAR REGULATORY COMMISSION UNITED STATES <
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iI ADVISORY COMMITTEE ON REACTOR SAFEGUARDS -
WASHtNGT ON, D, C. 20555.
- (...../
1 December-3,.1992-J.
i 1
MEMORANDUM FOR:
Samuel J._Chilk, Secretary of the-Commission-lj FROM:
a) nd F. Fra ey,;Execu. ve Director, ACRS
- {
SUBJECT:
ACRS MEETING WITH THE NRC COMMISSIONERS ON DECEMBER 11, 1992 -~ BACKGROUND INFORMATION.
l t
The ACRS is scheduled to meet with the NRC Commissioners on Friday,
~
December 11, 1992, between 1:00 and 3:00 P.M.-to discuss items of-mutual
- interest, including the following.
Background material.
related to these matters is attached:
- 1. Status of ACRS-Review of Advanced Reactor Desions - C. Michelson -
and J. Carroll (PP. 2-52) q
- 2. Dioital-Control and Protection Systems - H. Lewis (PP. 53-64)
- 3. Imo2 ementation of the Plant LifA Extension Rule-
-P.
Shewmon (PP. 65-83) j 4.
Risk-Based Regulation H. Lewis (PP. 84-87)
.i 2
Attachments: As Stated cc: ACRS Members-ACRS Technical Staff
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ITEM 1: STATUS OF ACRS REVIEW OF ADVANCED REACTOR DESIGNS The Committee previously discussed the status of the advanced reactor reviews with the Commission on. March 5 and September 11, 1992. Since then, a number of Committee and Subcommittee meetings have been held to discuss various aspects of the advanced reactor reviews, mainly, the GE ABWR, EPRI Requ.irements Document, Testing for the H AP600 and the GE SWBR, Design Acceptance Criteria (DAC),
and Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC).
The following is a brief summary of the reviews for individual plant reviews and the EPRI submittals.
EVOLUTIONARY PLANTS
- General Electric Advanced Boilina Wa_t.er Reactor (GE ABWR)
The ACRS Subcommittee on Advanced Boiling Water Reactors and other subcommittees riave held 22 meetings beginning in October,1989, to discuss the NRC staff's Draft Safety Evaluation Reports (DSERs),
the GE Standard Safety Analysis Report for the ABWR, and related matters. The Committee has provided four letter reports to the EDO on matters related to this review.
The Committee anticipates issuing its final report on the final design approval (FDA) for the ABWR, shortly after the FSER is provided for review.
The following documents are attached:
- ACRS report to the Commission dated August 12, 1992.
Subject:
Inspections, Tests, Analyses, and Acceptance Criteria Program' for the GE ABWR Design (PP. 6-9)
- ACRS letter report to James M. Taylor (EDO) dated April 13, 1992.
Subject:
Review of the Draft Safety Evaluation Reports on the GE Advanced Boiling Water Reactor Design (PP. 10-18)
- Westinghouse RESAR SP/90 Desian The ACRS review of the Westinghouse's application for the Preliminary Design Approval (PDA) for the RESAR SP/90 design has been completed. The Committee provided a report to the Commission dated December 12, 1990 on this matter.
<c w
The following document is attached:
ACRS Report to the Commission dated December 12, 1990.
1
Subject:
Westinghouse's Application for Preliminary Design Approval for the RESAR SP/90 Design (PP. 19-24) y
- ABB-CE System 80+ Desion The ACRS Subcommittee on Advanced Pressurized Water Reactors has held six meetings beginning in April 1990 to discuss the ABB-CE Systems 80+
design features and related issues such as the Licensing Review Basis (LRB) document. The Committee provided 'a report to the Commission dated November 14, 1990-in regard to the LRB. The staff's Draft SER on the Systems 80+ design was provided to' the ACRS on October 1, 1992, and a Subcommittee meeting has been scheduled for February 10, 1993 to discuss this document.
The following document is attached:
- ACRS report to the Commission dated November 14, 1990.
Subject:
SECY-90-353, Licensing Review Basis Document for the Combustion Engineering, Inc. System 80+ Evolutionary Light Water Reactor (PP. 25-26)
- EPRI Utility Recuirements Document fc-Evolutionary Plants The ACRS Subcommittee on Improved Light Water Reactors has held seven meetings to discuss the staff's draf t SERs related to various chapters of the EPRI Requirements Document for Evolutionary Light-Water Reactor Designs. The Committee provided reports to the Commission dated April 23, 1991 and August-18i 1992 on this matter.
The staff plans to issue a supplement to the FSER af ter ' all:
I evolutionary policy issues have reached final resolution. The ACRS expects to review the supplement to the FSER.
i The following document is attached:
--ACRS Report to the Commission dated August 18, 1992.
Subject:
Electric Power Research Institute Advanced Light Water Reactor -
Utility Requirements Document -- Volume II, Evolutionary' Plants (PP. 27-30) 1
RAESTVE PLANTS
- Westinchouse AP600 The Committee and the Subcommittees on. Advanced Pressurized Water Reactors / Thermal Hydraulic Phenomena, have heard presentations regarding design details for the Westinghouse AP600 passive plant and the test programs proposed by both, Westinghouse and the staff in support of the AP600 passive plant design certification. The Committee provided reports dated November 14, 1991 and March 10, April 6, and July 17, 1992 to the Commission in regard to the test programs. The Standard Safety Analysis Report for the AP600 was issued on June 26, 1992. The Committee will continue its discussion of this matter on a schedule consistent with the development of the staff's SER.
The following document is attached:
- ACRS report to the Commission dated July 17, 1992.
Subject:
Integral System and Separate Effects Testing in Support of the Westipghouse AP600 Plant Design Certification (PP. 31-35)
- General Electric SBWR The Committee and the Subcommittees on Advanced Boiling Water Reactors / Thermal Hydraulic Phenomena have heard presentations regarding design details and test programs for the General Electric SBWR passive plant.
The Committee provided a report to the Commission dated June 10, 1992 regarding the proposed test programs in support of the SBWR design certification.
The Standard Safety Analyses Report for the SBWR was received on August 26, 1992. The Committee will continue its discussion of this matter on a schedule consistent with the development of the staff's SER.
The following document is attached:
ACRS report to the Commission dated June 10, 1992.
Subject:
Testing and Analysis Programs in Support of the Simplified Boiling Water Reactor Design Certification (PP. 36-39)
- EPRI Reouirements Document for Passive Plants The Committee and the Subcommittee on Improved Light Water Reactors have been briefed on the EPRI Requirements Document for Passive Plant Designs. The ACRS plans to continue its review of this matter after-receiving the staff's final SER.
POLICY ISSUES FOR EVOLUTIONARY AND PASSIVE PLANTS The Committee has discussed the resolution of technical and policy issues at various meetings beginning in early 1990. The Committee has provided five reports dated April 26, 1990, December 18, 1991, May 13, 1992, August 17, 1992 and S.eptember 16, 1992 to the Commission or the EDO on this matter. Subject to the availability of all necessary information from the st.af f and industry groups and satisf actory completion of the reviews, the Committee expects to continue providing reports to the Commission on these issues. The Committee will continue its review of additional issues as they are identified by the staff and/or the Committee.
The following documents are attached:
- ACRS letter report to James M. Taylor (EDO) dated September 16, 1992.
Subject:
Draft Commission Paper, " Design Certification and Licensing Policy Issues Pertaining to Passive and Evolutionary Advanced Light Water Reactor Designs" (PP. 40-43)
- ACRS letter report to James M. Taylor (EDO) dated August 17, 1992.
Subject:
Issues Pertaining to Evolutionary and Passive Light Water Reactors and Their Relationship to Current Regulatory Requirements (PP. 44-48)
- ACRS letter report to James M. Taylor (EDO) dated May 13, 1992.
Subject:
Issues Pertaining to Evolutionary and Passive Light Water Reactors and Their Relationship to Current Regulatory Requirements (PP. 49-52)
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UNITED STATES -
NUCLEAR REGULATORY COMMISSION m
S3
-I
- ADVISORY COMMITTEE ON REACTOR SAFEGUARDS -
WASHIFWGTON, D. C 20655
/
August 12, 1992' 4
.c The Honorable.Ivan Selin Chairman U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Chairman Selin:
SUBJECT:
INSPECTIONS, TESTS, ANALYSES,-AND ACCEPTANCE CRITERIA-PROGRAM FOR THE GE ABWR DESIGN During the 388th meeting of the Advisory Committee on Reactor--
Safeguards, August ' 6-8,
- 1992, we. reviewed a sample of -the' Inspections, Tests, Analyses, and Acceptance criteria. (ITAAC) which are being prepared. by GE. Nuclear Energy - (GE) as a - part _ of. :its application fbr cartification of the ABWR design.
This topiciwas-also reviewed at a joint meeting of our Subcommittees on Decay Heat-Removal'- Systems = and Advanced Boiling - Water Reactors on August. 5, _
1992._ During these meetings, we~had-the. benefit of presentations by members of the. NRC staff and L by. representatives of GE.
Our ~
review has.been-in response to a request by.the Commission made at-our meeting with them on March 5,'1992,.and confirmed.in~a-Staff:
. Requirements Memorandum datedi April 11,71992. :
We also had - thel benefit of the documents referenced.
ITAAC-are an important part of Tier 1 submittals.which ' the' NRC
_ requires - of applicantsL for _ design certificationx under Part 52.
They_ are intended to abstract from the more voluminous: source, the:
Standard Safety' Analysis Report (SSAR), the'information needed by the-NRC: staff.toLaake its-final safety _ determination and to ensure.
that-this.information is agreed
- to : at : the time -_of-. design certification and verified in~the completed. plant.
- The_ form andi content. of - individual ITAAC:-
are still !beingL developed _ by an iterative-process between GE'and the~NRC staff.
There are:several-_ types-of ITAAC, as described:by the staff:-
Systems Generic Interface.
Design Acceptance _ Criteria-(DAC)
Combined " Operating - License - (COL) w y
e a
- a
a e
The Honorable Ivan Selin 2
August 12, 1992 our present review has been confined to the general program and to the first type, which includes the largest number of individual ITAAC.
We were told that the entire plant design can be described in terms of about 140 systems.
of these, GE has proposed that about 85 have sufficient safety significance. to be covered by_
individual ITAAC.
These comprise the' " Systems ITAAC."
We have reviewed 5 of these.85 ir some detail', as a means for evalaating the ITAAC process.
We intend to continue our review by investigating examples ef the Generic and Interface ITAAC.
We were told there are nine Generic ITAAC for the ABWR, covering subjects which apply to many or all systems, such as walding and equipment qualification requirements.
We have commented 'on DAC in an interim report of June 16, 1992.
The COL ITAAC, which will be concerned with such matters as operator training, will be developed by a COL applicant-after the design certification.
We would expect to review these in the future when appropriate.
We conclude from our review that the ITAAC process appears to be generally well founded and can be made to work as the staff and GE visualize.
The general form and scope of the individual ITAAC we studied were satisfactory.
There is, however, a problem with content of the ITAAC.
Although the examples we examined were a part of what was described as the final Stage 3 GE uibmittal, there was a significant lack of consistency, accuracy, and completeness.
We were informed by both the staff and GE that this is a problem beyond the five examples we selected for our review.
Both are individually committed to major efforts to improve the quality of the content of all ITAAC.
We were t'old by the Director of NRR that he plans an extensive and in-depth review of the submitted ITAAC and will not recommend approval of a Final Design Approval (FDA) until the results of the review are fully satisfactory.
This could mean a delay in the presently projected date for the FDA issuance.
For its part, GE expressed its commitment to respond'to prot.lems indicated by the staff review and to conduct its own quality review in parallel. GE l
intends to ensure consistency among ITAAC and other Tier 1 and Tier 2 documents.
In addition, we were told that }WMARC intends to carry out an independent review of the ABWR ITAAC.
GE already has comments from utilities on the Stage 3 ITAAC.
These will be incorporated into the continuing iterations between the staff and GE.
We are concerned.with the structural adequacy of walls and associated penetrations within buildings housing critical systems outside of primary containment during possible fires, floods, or i
pipe breaks.
It was not clear from the material presented to us how structurcl requirements for these will be verified through the 7
l The Honorable Ivan Selin 3
August 12, 1992 ITAAC process.
We expect to pursue this matter at a future meeting.
A PRA has been performed for the ABWR design and certain conclusions about the safety of the design can be drawn from this.
In performing the PRA, many assumptiohs were necessary about the performance reliability of components"and systems.
There appears t.; be no means by which Tier 1 requirements (e.g., ITAAC) will ensure that components and systems in the plant can be expected to have reliabilities which are consistent with those assumed in the PRA.
The SSAR provides some information on this, but does-not close the loop.
We were told that appropriate reliability values for components and systems will be ensured through a reliability assurance program developed by a COL applicant.
We believe this matter deserves more study.
In our report to you of September 10, 1991 on ITAAC, we expressed a preference for Option 3 in SECY-91-210 which would allow for completing the ITAAC after issuance of the FDA for ABWR.
The staff position is that completion of the ITAAC before the FDA is essential.
Given our evaluation of the current status of ABWR documentation,' we agree.
We trust the above discussion and comments have been helpful.
We expect to complete our review in the near future.
Sincerely, O
a' David A. Ward Chairman
References:
1.
SECY-91-210, dated July 16,
- 1991, from James M.
- Taylor, Executive Director for Operations, for the Commissioners,
Subject:
Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Requirements for Design Review and Issuance of a Final Design Approval (FDA).
2.
Staff Requirements Memorandum dated April 1,1992, from Samuel J.
- Chilk, Secretary, for David A.
- Ward, ACRS,
Subject:
Periodic Meeting with the Advisory Committee on Reactor Safeguards on March 5, 1992.
3.
Excerpts of Inspections,
- Tests, Analyses, and Acceptance Criteria from. GE Nuclear Energy Report:
" Tier 1 Design Certification Material for the GE ABWR," dated June 1992, as follows:
4 The Honorable ivan selin 4
August 12, 1992 Standby Liquid Control System (2.2.4) e Residual Heat Removal System (2.4.1) e Reactor Building Cooling Water System (2.11.3) e e Emergency Diesel Generator System (Standby ac Power. Supply -
2.12.13)
Control building (2.15.12) e 4.
Report dated September 10, 1991, from David A. Ward, Chairman, ACRS, to Ivan Selin, Chairman, NRC,
Subject:
Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) for Design Certifications.
as e
4, p
'o, UNITED STATES P
NUCLEAR REGUL ATORY COMMISSION U
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April 13, 1992 Mr. James M. Taylor Executive Director for Operations
~
U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Taylor:
SUBJECT:
REVIEW OF THE DRAFT SAFETY EVALUATION REPORTS ON THE GE ADVANCED BOILING WATER REACTOR DESIGN During the 383rd and 384th meetings of the Advisory Committee on Reactor Safeguards, March 5-7 and April 2-4, 1992, we discussed the Draft Safety Evaluation Reports (DSERs) on the Advanced Boiling Water Reactor (ABWR) design which is described by GE Nuclear Energy (GE) in its Standard Safety Analysis Report (SSAR), as amended, and for which GE tias applied for design certification in accordance with 10 CFR Part 50, Appendix 0.
The DSERs which are the basis for this report were sent to the commissioners for information as six SECY papers (SECY-91-153, 235, 294, 309, 320, and 355).
These generally cover the SSAR and its first eighteen amendments.
Our Subcommittee on Advanced Boiling Water Reactors discussed these papers with representatives of GE and the NRC staff during its meetings on September 18 and October 23, 1991 and January 23-24 and February 20-21, 1992.
We also had the benefit of the documents referenced.
Our first report to you concerning the DSER for this project was dated November 24, 1989.
That report conveyed our comments on Module 1 of the design (former GE designation).
We also sent a report to you on July 18, 1991, outlining several ABWR design concerns that developed during subsequent review.
l We note a marked improvement in the quality of the staff's DSER evaluations since our November 24, 1989 report.
The staff reviewers appear to be following the guidance outlined in the applicable Standard Review Plans (SRPs) to the extent possible, and they are asking good in-depth questions in most areas.
The SECY-91-161 schedule indicates that the Final Design Approval (FDA) is to be issued before the end of Calendar Year 1992.
If we are to provide our final report on this subject in December 1992, it will be necessary that we receive a complcce and final SER no later than early September 1992.
There are now more than three i
hundred open items in the DSERs, many of which are major.
In l
M
Mr.' James M. Taylor 2
April 13, 1992 addition, there is a number of important policy issues which are
-unresolved.
With the staff programs in place, it is probable that these issues can be resolved.
- However, this is a
large undertaking, and we have concerns about whether it can be accomplished on the schedule now indicated.
In the course of our review, we have identified technical issues for which resolutions should be achieved before we write our final report.
These are listed and discussed as follows:
1.
Control Buildinc Floodina The proposed ABWR plant design locates the Reactor Building Cooling Water (RBCW) System at the lowest elevation in the control building, with the essential 250 V dc battery rooms and the main control room at a higher elevation, but still below ground.
Our concern with this arrangement is the potential for control building flooding due to an unisolated break in the Reactor Service Water (RSW) System which provides cooling water from the Ultimate Heat Sink (UHS) to the RBCW System.
The proposed UHS is a' ground-level spray pond which we assume to be at building grade and likely to contain sufficient water to flood the control building.
i The staff should obtain sufficient information on the i
interface and conceptual design of the RSW System and UHS to support an adequate evaluation of the flooding potential. The staff's evaluation should include consideration of isolation valve arrangements, the feasibility of and time available for response, and the assumption of a single active component failure during the response.
The design information and flooding analysis should be included in the SSAR.
2.
Adecuacy of Physical Seoaration Pipe breaks, internal plant flooding, and external events such j
as fire are of major concern if their effects -cannot be-l confined in order to protect required safe-shutdown equipment.
We believe that the key to confinement is the provision of appropriate separation barriers. However, a classical barrier such as the 3-hour-rated fire barrier wall and its penetrations (e.g., doors and dampers) may not, of itself, be-L sufficient to ensure separation under (a) the combined effects of-pressure, heat, and smoke _ from a - fire, and the flooding which results from fire mitigation, (b) the effects'of pipe whip, jet impingement, or compartment pressurization due to pipe breaks, or (c) the influx of water and hydrostatic pressure buildup due to internal floods.
/
Mr. James M. Taylor 3
April 13, 1992 We believe that the SSAR should describb and the staff should i
evaluate the adequacy of proposed separation barriers for the full range of events and conditions for which separation must be ensured.
We continue to recommend that systems required for safe shutdown not share a common Heating, Ventilating and Air conditioning (HVAC) System during normal plant opeiation.
The secondary containment HVAC System for the~ABWR is such a shared system.
i 3.
Protection of Environmentally Senriitive Eauinment The ABWR makes extensive use of environmentally sensitive equipment (including solid-state electronic components) for essential protection,
- control, and data transmission functions.
Such components are known to be susceptible to adverse environmental
- changes, particularly temperature extremes.
We are concerned that a number of these componento may be located in plant areas where postulated events such as pipe breaks, fire, internal flooding, or loss of room cooling may create an adverse environment.
Such environments need to be identified in the SSAR to ensure appropriate environmental qualification of the equipment.
4.
Beview of Chilled-Water Systems The ABWR uses large chilled-water systems to provide essential environmental cooling, which in turn includes cooling of the solid-state electronic components.
Because there was no SRP for chilled-water systems, the staff used other guidance such as SRP Section 9.2.2 (Reactor Auxiliary Cooling Water Systems) when the safety evaluation was performed.
However, this guidance is not appropriate for the evaluation of refrigeration systems.
The NRC staff needs to evaluate the performance of chilled-water systems under varying accident heat loads and during loss-of-offsite-power events, and to consider their ability to restart and function after a prolonged station blackout.
The DSER sections which should evaluate the performance of large chiller packages do not address these issues. We believe they should.
5.
Use of Leak-Before-Break Methodoloav It is our understanding that GE will not propose the use of leak-before-break methodology for the ABWR standard plant.
Thus, the DSER,should be revised to ensure that consideration is given to pipe break effects for all systems and locations.
This may introduce additional structural protection and environmental qualification requirements in the SSAR.
'n Mr. James'M.-Taylor 4
April 13, 1992 6.
Use of Intectral Low-Pressure Turbine Rotors In our July 18, 1991 report to you, we recommended that the staff review the issues -involved with the use of integral low-pressure (LP) turbine rotors.
It is our understanding that this new design for LP rotors will be used for the ABWR.
(Rotors of this type are being issed in rot'or replacement programs at currently operating, plants.)
The practice of turbine manufacturers has been to bore the centerline of this type of rotor to remove impurity inclusions.
We. were concerned that the use of unbored rotors-was being contemplated.
The Electric Power Research Institute (EPRI) has recently added a-requirement in its Advanced Light Water Reactor Utility Requirements Document (URD) that LP rotors be center-bored.
7.
Cavity Floor Area Beneath Reactor Vessel The cavity area beneath the reactor vessel is sized to meet 2
the EPRI URD specification of 0.02 m /Mwt.
The ABWR design includes flooding of the cavity.
Little consideration has been given to.how this should be accomplished.
There is little evidence that ' the planned cavity area vill lead to-quenching following flooding or that the ABWR flooding plans will not. lead to ex-vessel steam explosions.
Further attention needs to be given in the SSAR as to when and how fast the cavity should be-flooded in order to avoid exacerbating a core-melt accident if it should occur.
8.
Adecuacy of the ABWR PRA It is impossible to determine whether the PRA submitted by the applicant will be adequate for a safety determination absent information on. how it is-to be used by the staff.
In our February 14, 1992 report'to the-Commission on the Use of Design Acceptance Criteria During 10 CFR Part 52 ' Design Certification Reviews, we commented on the need for guidance'.
on the use of PRA in the review of'new plant designs.. At this point.the applicant has submitted a PRA, a contractor. has performed an extensive review, and the. staff has-prepared al DSER._ However, the use of the PRA in the design certification process is still. undefined.
Presumably, the results of the PRA will be.used in the course-of the staff's determination that-the design-is expected to produce a nuclear power plant that has an appropriate response-
' to severe accidents.
In the Severe Accident Policy Statement, the Commission ' indicated that a PRA would-be required for each '
new design,- and that the results of this PRA would be part of -
the information which would -guide the staff in -
its determination.that a-design is adequate to deal with severe B
.c Mr. James M. Taylor 5
April 13,.1992 accidents.
The policy statement published in the Federal Reaister of August 8,
1985, also states that "Accordingly, within 18 months of the publication of this Severe Accident Policy Statement, the staff will issue guidance on the form, purpose and role that PRAs are to play in severe accident analysis and decision making for,. both existing, and future plant designs...."
The Statement says further, "The PRA guidance will deceribe the appropriate combination of deterministic and probabilistic considerations as a basis for severe accident decisions."
The staff has yet to produce the promised guidance.
We urge that the staff formulate a set.of criteria that it plans to use in making severe accident decisions.
This should include the way in which the results of a PRA are to be used in the process (not just whether the PRA has been done properly).
9.
Containment Hydrodynamic Loads Air-clearing loads on containment structures are the result of a complex process resulting from the drywell air being forced into the yetwell by the primary system blowdown. The water in the vent system is pushed down and out until the horizontal vents are cleared.
The water-clearing process produces a jet of water into the suppression pool which causes a load on the outer part of the wetwell wall.
This water clearing is followed by an air-steam mixturo which creates a large bubble as it exits into the pool.
The steam condenses but the air expands forcing the water above it up into the wetwell air space.
The wetwell air space is compressed due to the momentum of the water in the layer above the bubble.
The wetwell air space will be subjected to an energetic two-phase eruption as a result of the air-clearing-process.
The vacuum breakers which are in the vicinity will be exposed to this environment unless protected.
The SSAR should describe what the environment will be and what protective measures, if any, are needed to ensure survival of the vacuum breakers.
If a vacuum breaker does not close, the suppression pool is bypassed and the wetwell/drywell pressures will rise at a rate dictated by the capability of some means other than-the suppression process (e.g., containment sprays) to remove heat and condense steam.
The SSAR should contain an analysis of such a situation.
The early work to address problems arising from analyses of the Mark I,
II,, and III containments is not sufficient to address similar processes that will occur following a LOCA in an ABWR containment.
The ABWR is different for two reasons:
(a) the volume of the wetwell air space in the ABWR is approximately that of a Mark II, and (b) the impact of the
/
Mr. James _M.
Taylor 6
April 13, 1992 air-clearing loads will be alleviatedsomewhat because the expected blowdown flows are much smaller than those expected in a Mark I or Mark II.
Nevertheless, the combination of a much smaller wetwell and the lower mass flow from the break have not received sufficient attention to be written off by the staff or GE without further analysis or experlmental investigation.
We are not aware 6f'any testing of the ABWR type geometry. We believe there are sufficient differences in both geometry and LOCA characteristics to require further eve.luation of the air-clearing phase of the LOCA by more extensive analysis and/or experimental inveatigation.
10.
Adecuacy of SSAR Treatment of the Reactor Water Cleanuo System We performed a review of the Reactor Water Cleanup (RWCU)
System using our own staff.
This system was chosen because it is a non-safety system located outside of primary containment, but inside the building which houses engineered safety features.
It uses pipes up to 8-in, nominal diameter whose rupture would result in a LOCA and a source of serious environmental disruption in the building.
This system is not seismically qualified or built to quality assurance standards.
Our review identified a number of deficiencies in the SSAR, some of which are listed below:
There is little useful information presented in the SSAR that describes how the Japanese codes and standards used for the RWCU System design can be converted to domestic design standards.
The Quality Group classifications for certain portions of the RWCU System are inconsistent with the Japanese code-related classifications shown on the Piping and Instrumentation Diagrams.
The Safety Class / Quality Group transition between the piping inside primary containment and that outside primary containment is not in accordance with ANSI /ANS safety class standards for BWR fluid systems.
The questionable ability of system isolation valves to e
close under large-break-LOCA conditions has been the subject of extensive NRC testing and a Generic Letter (GL 89-10).
- However, the SSAR. specifies no special performance requirements for these valves.
e The safety-grade leak detection _ and isolation system which actuates the system isolation valves was not described.in detail sufficient to support an assessment of its adequacy.
The ABWR PRA did not evaluate as initiating events RWCU System line breaks (or other LOCAs) outside the primaryW
a.
Mr. James M. Taylor 7
April 13, 1992 containment.
The exclusion of t'h'ese breaks was based erroneously on an analysis of.the effects of suppression pool bypass events on overall risk.
- However, the analysis failed to take into account that the bypass path (e.g., RWCU System pipe break) could be the initiator for the core-damage event.
The PRA analysts took credit for the RWCU System as a e
heat removal system in all sequences where reactor pressure is assumed to remain 'high. The analysts assumed that the capacity of the non-regenerative heat exchanger (NRHX) is adequate to remove the decay heat.
The capacity appears to be adequate;
- however, our calculations indicate that the outlet temperatures on the RWCU System side and cooling water side of the NRHX would -
exceed the design limits for the piping.
Furthermore, a temperature sensor between the NRHX and the RWC0 System pumps in the present design would automatically isolate the NRHX on high temperature, making it unavailable.
The items mentioned above are among a number of issues that were identified.
It is important for the staff to ensure -that the shortcomings of the RWCU System and PRA related portions of the SSAR are not indicative of problems in the remainder of that report.
11.
Plant Desion Life and Acina Manacement We recommend that the SSAR clearly define the scope of the 60-year design life for the ABWR and describe a program plan for achieving it.
This program should include those aging management measures which are necesaary to uaintain the plant within its design basis throughout its design life.
This program should specify the original design and application criteria and, where lequired, the projected refurbishment or replacement requirements with appropriate rationale.
To the extent applicable, the lessons learned from the NRC's Nuclear Plant Aging Research Program as 'well as other ogin projects should be incorporated into this program.g research We note that the EPRI URD (Volume II, Chapter 1, Paragraph 3.3) includes a requirement for a plant design life of "60 years without necessity for an extended refurbishment outage,"
and discusses the requirements-for its achievement in Paragraph 11.3.
Hr. James H. Taylor 8
April 13, 1992 12.
Station Groundino and Surce Protection Chapter 8 of the ABWR SSAR defines the scope of and specifies the requirements for the electrical power systems.
The scope is limited to the onsite electrical power systems and to the interface requirements with the offsite electrical power systems.
Notably absent are lightning protection, _ station grounding systems, and surge protection measures whic:h are necessary to protect plant personnel and equipment during normal and abnormal conditions. These measures are required to eliminate or reduce electrical shock hazards to personnel, and to protect systems and equipment against damage or misoperation as the result of lightning strikes, switching operations, electrical arcs, short circuits, _ static electricity, etc.
These protective measures and their interf ace requirements should be included in the SSAR.
The ABWR makes extensive use of sensitive rolid-state elec-tronic compcnonts for essential protection, control, and data transmission functions.
These components should be protected from extraneous electrical impulses that will damage them or cause improper performance.
To the extent practical, these components should be isolated from potential adverse signals that may be transmitted over control or data links from remote locations, meteorological stations, switchyards, etc.
We note that the EPRI URD (Volume II, Ch @ter 11, Item 9,
" Electrical Protective Systems") addrusses aquirements for these systems.
We recommend that these grounding, surge protection, and isolation features be i.ncluded in the SSAR.
13.
Corrosion Control for Structures The SSAR should include an interface requit at. f or a corrosion control program to identify the potencial for the corrosion of structures and components and to determine the corrective measures to be taken The program should commence prior to the completion of the detailed design of building substructures and underground installations.
The program should consider the potential for corrosion from galvanic direct currents which may flow as the result of copper ground mats on site, including the electrical switching stations' ground - mats.
The potential for corrosion of containment building substructures and liners should be considered.
The mitigation measures may include coatings, wrappings, cathodic protection, electrical
- bonding, elimination of galvanic currents, or other mitigation means.
/
a Mr. James M. Taylor 9
April 13, 1992 4
We do not expect to receive a separate replyto the above items if they are covered appropriately in the final SER.
We will keep you informed of any additional concerns as our review proceeds.
Sincerely O
Davi'd A. Ward Chairman
References:
1.
GE Huclear Energy, Standard Safety Analysis Report, " Advanced Boiling Water Reactor," Chapters 1 through 20 (Amendments 1 through 18) 2.
BECY-91-153, dated May 24, 1991, for-the Commissioners from James M.
Taylor, Executive Director for Operations, NRC,
Subject:
Draf t Saf ety Evaluation Report (DSER) on the General Electric Company Advanced Boiling Water Reactor Design Covering Chapters 1,
2, 3,
4, 5,
6, and 17 of the Standard Safety Analysis Report (SSAR) 3.
SECY-91-235, dated August 2, 1991, for the Commissioners from JameF M. Taylor, EDO, NRC,
Subject:
DSER on the GE Boiling Water Reactor Design Covering Chapters 1, 3, 9, 10, 11, and 13 of the SSAR 4.
SECY-91-294, dated September 18, 1991, for the Commissioners from James M.
Taylor, EDO, NRC,
Subject:
DSER on the GE Boiling Water Reactor Design Covering Chapter 7 of the SSAR 5.
SECY-91-309, dated October 1,1991, for the Commissioners from James M. Taylor, EDO, NRC,
Subject:
DSER on the GE Boiling Water Reactor Design Covering Chapter 19 of the
- SSAR,
" Response to Severe Accident Policy Statement" 6.
SECY-91-320, dated October 15, 1991, for the Commissioners from James M.
Taylor, EDO, NRC, Subject OSER on the GE Advanced Boiling Water Reactor Design Covering Chapter 18 of the SSAR 7.
SECY-91-355, dated October 31,~1991, for the Commissioners from James M.
Taylor, EDO, NRC,
Subject:
DSER on the GE Boiling Water Reactor Design Covering Chapters 1, 2, 3, 5, 6,
8, 9,
10, 12, 13, 14, and 15 of the SSAR 8.
Electric Power Research Institute,
" Advanced Light Water Reactor Utility Requirements Document" (Volume II)/ALWR Evolutionary Plant, Revision 3, Issued Novemb6c 1991 e
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December 12, 1990 The Honorable Kenneth M. Carr chairman U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Chairman Carr:
SUBJECT:
WESTINGHOUSE'S APPLICATION FOR PRELIMINARY DESIGN APPROVAL FOR THE RESAR SP/90 DESIGN During the 367th meeting of the Advisory Committee on Reactor Safeguards, November 8-10,
- 1990, we completed our review of Westinghouse's spplication for Preliminary Design Approval (PDA) for the Westinghouse Reference Safety Analysis Report (RESAR SP/90) nuclear power; block (NPB).
We heard presentations from the NRC staff and the applicant concerning thn staff's draft Safety Evaluatio!. Report (SER)
(NUREG-1413) for this PDA during our meeting.
Representatives of the staff and of the Office of the General Counsel (OGC) discussed the related draft PDA document.
i Our Subcommittee on the Advanced Pressurized Water Reactors has held a series of meetings with the staff and reprecentatives of the applicant regarding this matter over the past two and a half years.
He also had the benefit of the documents referenced.
1.0 Scope and History of RESAR SP/90 ADDlicatiQD The RESAR SP/90 is an evolutionary (as contrasted with passive)
Advanced Light-Water Reactor (ALWR) design for a single-unit NPB, I
rated at a reactor pcwer of 3800 MWt.
Although many basic design decisions were made by Westinghouse prior to completion of the EPRI ALWR Utility Requirements Document, the design of this four-loop pressurized water reactor generally conforms to the EPRI require-ments for such designs.
RESAR SP/90 NPB contains preliminary design information for the portion of the design that encompasses NPD buildings, structures, systems, and components.
Specifically excluded from the scope are the turbine building, the waste disposal building, the servici building, the administratior building, the service water / cooling water structure, and the ultimate heat sink.
These features will be the design responsibility of an applicant proposing to build a facility referencing the RESAR SP/90 design. Interface information addressing the pertinent safety-related design requirements necessary to ensure the compatibility of the referenced system with
~
4 C>=
O e The Honorable Kenneth M. Carr
.1 December 12, 1990
'o the plant-specific portion of the f acility has been included in the RESAR SP/90 application.
On October 24, 1983, Westinghouse submitted an application for a PDA for RESAR SP/90 HPB design in accordance with 10 CFR Part 50, Appendix 0,
" Standardization of Design:
Staff Review of Standard Designs,"
which was the then existing regulatory basis for this type of application.
The application was docketed on November 30, 1983 (Docket No. 50-601).
The RESAR SP/90 application describing the design of the HPB was submitted in modular form during the period from October 23, 1983 to March 9, 1987.
In addition, the information in RESAR SP/90 has been supplemanted by 47 amendments to these modules.
2.0 Reculatory Backaround Before the promulgation of 10 CFR Part 52 in May of 1989, the review of RESAR SP/90 had been performed by the staff pursuant to Appendix 0 to 10 CFR Part 50, using a procedure similar to that used for custom plant reviews for which guidance to staff reviewers is provided 11 the Standard Review Plan.
This evaluation was 3
analogous to a construction permit (CP) licensing review for a specific facility and conducted with the intent that, following satisfactory completion of the reviews performed by the staff and the ACRS, a PDA could be issued by the staff.
The promulgation of 10 CFR Part 52 resulted in the transfer of Appendix 0 to 10 CFR Part 52; hence a PDA can now be issued for this application pursuant to 10 CFR Part 52.
A PDA is optional for a Final Design Approval (FDA) and/or Design Certification under the provisions of 10 CFR Part 52.
3.0 The Staff's SER and the PDA The SER and PDA represent the first stage of the staff's review of the design, construction, and operation of the RESAR SP/90 design.
During our meetings, we learned that there is no prospective CP applicant nor does Westinghouse intend to apply for an FDA and/or Design Certification of the RESAR SP/90 design until there is a proven interest on tho' part of a domestic or foreign utiliti.
The staf f 's SER summarizes the results of the staff's radiological safety review of the RESAR SP/90 NPB design and delineates the scope of the technical details considered in evaluating the proposed design.
This review took place over the period of October 1983 to October 1989 (the date on which the staff decided to close its review).
Environmental aspects were not considered in the staff review of RESAR SP/90, but would be addressed in a utility's plant-specific application.
The lionorable Kenneth H.
Carr 3
December 12, 1990 3.1 Comments on the Staff's SER There are 170 open items that will require resolution during the review of a plant-specific application for an Operating ~ License (OL).
Most of those appear to be the kind of open issues expected at this stage of the design.
of the 170 open items, 17 are site specific, 110 involve information in the scope of an OL or FDA and/or Design Certification application, and 43 had not been resolved by the staff when it closed its review in October 1989.
(Westinghouse submittals on many of these 43 open items, including its proposed resolution of Generic Safety Issues, Unresolved Safety
- Issues, issues a: post-THI regulatory requirements, and outstanding PRA e yet to be reviewed by the staff.)
In view of these open items and our concerns regarding the SER and the many unresolved 1
covere accident
- issues, we indicated to the staff that its conclusions on page 25-1 of the draf t SER were stated too strongly.
The staff agrood to revise this language.
The Committee is not of one mind regarding the issuance of a PDA for the RESAR SP/90.
On the one hand, there is merit to the argument that ', Westinghouse's application for the RESAR SP/90 PDA was made in good f aith in 1983 under a dif ferent set of regulations and that it is now appropriate to document the reviews that have taken place to date and issue the PDA for potential future use as a reference design for an individual plant CP application or as the starting point for an FDA and/or Design Certification application.
Both Westinghouse and the staff advocate this approach; neither believes that it can devote further resources to this effort.
On the other hand, we view the RESAR SP/90 SER as a mixed bag of staf f evaluations that were performed over the seven-year period since the application was filed.
Some are current and well done; others are poorly done and/or were performed years ago and do not meet the standards that we believe should be applied to a current SER. A ma jor contributor to this problem appears to be the staf f's reliance on the July 1981 Standard Review Plan (SRP) (NUREG-0800) in performing this review.
This SRP needs updating to reflect the current situation for the licensing of ALWRs.
Some examples of our concerns with the staff's SER are:
3.1.1 SER Chapter 7, Instrumentation and Controls, references a staff review that was performed in 1979 for the Westinghouse RESAR 414 design.
The staff concluded that the computer based integrated reactor protection system design for RESAR SP/90 is acceptable for a PDA on the basis of the " similarity" of the RESAR 414 design to that proposed for RESAR SP/90.
It is our view that the staff should have developed improved standards for the review of such systems during this 11-year period.
We are J/
The Honorable Kenneth H. Carr 4
December 12, 1990 particularly concerned about the verification and validation of the software employed with computer based reactor protection systems.
It appears that there is a need to augment existing staff resources with expertise in the computer science area so that appropriate standards can be developed *for the review of computer based reactor protection systems.
All of the proposed evolutionary and passive ALWRs employ such systems.
' 1.2 For materials used in the fabrication of pressure boundary components, Westinghouse has committed to follow applicable codes, standards, and regulatory guides. Many of these are not representative of current industry practice for such materials.
We learned that Westing-house has developed internal specifications for pressure boundary materials that presumably do reflect current industry practice.
These were not submitted for the staff's review.
3.1.3 The proposed design employs water displacer control rods and; associated control rod drive mechanisms, which is a new feature for Westinghouse plants.
The SER describes the function of and strategy for use of these control rods.
The SER, however, does not discuss the pressure boundary integrity of those new control rad drive mechanisms or the potential for reactivity insertion accidents that could result from misoperation of these control rods.
Although Westinghouse submitted informa-tion on these subjects, the staff has not completed its review of this information.
In general, we believe that new features of this kind should be thoroughly reviewed at an early stage of review.
3.1.4 Our review, which represents only a sampling effort, revealed a number of factual errors and inconsistencies in the SER; the staff has agreed to correct these errors.
We believe that a review of the draf t SER by Westing-house, which has not yet had access to this predecisional document, would reveal additional errors that should be corrected.
We recommend that this be done.
3.2 Comments on the PDA Document The PDA states that the preliminary design information contained in RESAR SP/90 " complies with the requirements of 10 CFR Part 52, Appendix o... arid is acceptable for incorporation by reference in applications for individual construction permits.
The PDA does not describe how this preliminary design information would be used in a future FDA and/or Design Certification application.
J-
The lionorable )tonneth H.
Carr 5
December 12, 1990 We were told by OGC that this results from the fact that Westing-house has not made an application under 10 CFR Part 52.
Given the quality of the SER for this PDA, we are concerned with the language of the PDA that requires the staf f and ACRS to utilize and rely on the " approved preliminary' design" in their reviews of any individual f acility construction permit application ".
unless significant information which~ substantially affects the determination set forth in this PDA, or other good cause, is present."
OGC advised un that thin requirement would apply only to the staff and ACRS reviews of a CP application and that both entities would be able to revisit any issue in their review of any type of application that would Icad to an OL.
This is satisfac-tory to us but could present problems for the staff in dealing with a contested CP application.
4.0 Commentr on the SP/90 Desian u
We have two concerns regarding SP/90 design features:
4.1 Our revi6w of the NPD layout indicates that Westinghouse has provided many desirable features from the standpoint of separation of equipment trains for protection against fires and industrial sabotage.
However, we are concerned about the location of the emergency diesel generators (EDGs) on the same floor and corridor from the control room.
We believe that another location for the EDG room should be specified in view of the potential for fire and/or explosions associated with the operation of largo diesel generators.
4.2 The proposed RESAR SP/90 design employs a spherical contain-ment.
To deal with core / concrete interaction, the layout of the containment employs a cavity floor area beneath the rgactor vessel that is based on the EPRI requirement of 0.02 m per MWt.
If a larger area is required, major changes to the containment sizing and layout may be needed.
Timely development of a
Commission position on this issue is important not only to this design Lut also to the design of all of the ALWRs.
5.0 ACRC Recommendations on the Issuance of a PDA We believe, subject to the above comments, that the proposed design of the RESAR SP/90 NPD can be successfully completed and used in an application for an individual plant CP.
Accordingly, we recommend that a PDA be issued for the proposed Westinghouse RESAR SP/90 HPB.
r The !!cnorable Kenneth M.
Carr 6
December 12, 1990 6.0 concludino Remarks Finally, we wish to commend the Westinghouse Electric Corporation, the Japanese APWR program participants, tne EPRI ALWR Utility Steering Committee, and the EPRI staff for the-effort they have expanded in the development of this evolutionary design. The RESAR SP/90 design represents an important stcp forward in providing improved LWR designs that incorporate many of the lessons related to safety, performance, and reliability that have been learned by the nuclear power industry over the past 30 years.
Sincerely, M
L Carlyle Michelson Chairman
References:
1.
U.S. Nuclear Regulatory Commission, Draf t HUREG-1413, " Safety Evaluation Report Related to the Preliminary Design of the Standard Nuclear Steam Supply Reference System, RESAR SP/90" (Predecisional) 2.
Draft Westinghouse Electric Corporation, Docket No. 50-601, Reference Safety Analysis Report (RESAR SP/90 Nuclear Power Block Standard Design), Preliminary Design Approval (PDA)
(Predecisional) (Discussed during the November 8-10,1990 ACRS full Committee meeting) 3.
Letter NS-EPR-2675 dated November 1, 1982 from E.
P.
- Rahe, Jr., Westinghouse Electric Corporation, to F. Miraglia, U.S.
Nuclear Regulatory Commission,
Subject:
Westinghouse Advanced Pressurized Water Recctor Licensing Control Document
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November 14, 1990 The lionorable Kenneth M. Carr Chairman U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Chairman Carr:
SUBJECT SECY-90-353, LICENSING REVIEW BASIS DOCUMENT FOR THE COMBUSTION ENGINEERING, INC. SYSTEM 80+ EVOLUTIONARY LIGHT WATER REACTOR During the 367th meeting of the Advisory Committee on Reactor Safeguards, Nover' tr 8-10,*1990, we reviewed the staff's SECY 353, " Licensing E iew Basis Document for the Combustion Engineer-ing, Inc.
System 80+ Evolutionary Light Water Reactor," dated October 12, 1990.
Our Subcommittee-on Advanced Pressurized Water Reactors also considered this matter during a subcommittee meeting on November 1,: 1990.
During this review, we had the benefit of discussions with representatives of the NRC staff and of Asea Brown Boveri combustion Engineering.
We also had the benefit,of the documents referenced.
The staff has recommended that the Licensing Review Basis (LRB) ef fort for the combustion Engineering (CE) System 80+ design, which is well advanced, be continued to completion.
There does not apperr to be any substantive disagreement between the staff and CE on issues addressed in the LRB document.
The only approved LRB docuunt was proposed by the General Electric Company (GE) as a way of obtaining early agreement with the staff on major process and technical issues for the review of its advanced boiling water reactor design certification application.
It was approved by' the Director of HRR in a letter to Mr.
R.
Artigas, GE, on August 7,1987.
This letter contains the qualifi-cation that the LRB represented the approach in "certain key areas" that GE was committed to follow "
until final Commission positions and staff requirements are defined and implemented." At that time, neither 10 CFR Part 52 nor Commission-approved staf f positions relating to the certification of advanced light water reactors such as SECY-90-016 (referenced) were available.
We note that 10 CFR Part 52 does not discuss the use of LRB documents as a part of the final design approval or certification process.
These regulatory requirements and others under development have preempted the need for and diminished the usefulness of an LRB document for the CE System 80+ design.
We recommend that no further effort be devoted to the proposed LRB document for the CE System 80+ design.
The Honorable Kenneth H. Carr 2
November 14, 1990 Additional comments by ACRS members Ivan Catton, Paul.G. Shevmon, and J.
Ernest Wilkins, Jr.,
are presented below.
Sincerely,
,gf?
E
/
Carlyle Michelson Chairman Additional Comments by ACRS Members Ivan Catton. Paul G. Shevmon.
and J.
Ernest Wilk(np. Jr.
We understand that this LRB document can be completed and issued with relatively little additional effort.
If so, we would prefer to see an orderly disposition of this LRB document in accordance with the staff recommendation in SECY-90-362 (referenced).
We would agree with our colleagues that the CE System 80+ LRB effort be terminated now if the Commission, the staff, and the ACRS'need to invest any significant additional effort.
Esferences:
1.
" Licensing Review Basis Document for. the Combustion Engineering, Inc. System 80+ Evolutionary Light Water Reactor," dated October 12, 1990.
2.
SECY-90-362, " Staff Comments on the continuing Heed for a License Review Basis Document for Each Passive Design," dated October 24, 1990.
3.
" Evolutionary Light Water Reactor (LWR)
Certification Issues and their Relationship to Current Regulatory Requirements," dated January 12, 1990.
4.
Letter LD-90-005 dated January 22, 1990 from A.
E.
- Scherer, Combustion Engineering, to R..Singh, subject:
System 80+
Licensing Review Basis Document.
5.
Letter LD-90-060 dated August 28, 1990, from E.
H. Kennedy, Combustion Engineering, to Thomas V.
Wambach, NRC,
Subject:
Licensing Review Basis for the System 80+ Standard Design.
e
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UNITED STATES y*
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August 18, 1992 The Honorable Ivan Selin Chairman f
U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Chairman Selin:
SUBJECT:
ELECTRIC POWER RESEARCH INSTITUTE ADVANCED LIGHT WATER REACTOR UTILITY REQUIREMENTS DOCUMENT VOLUME II, EVOLUTIONARY PLANTS During the 387th and 388th meetings of the Advisory Committee on Reactor Safeguards, July 9-11 and August 6-8, 1992, we reviewed the NRC staff's Final Safety Evaluation Report (FSER) for Volume II of the Electric Power Research Institute's (EPRI) Advanced Light Water Reactor (ALWR) Utility Requirements Document (URD) for Evolutionary Plants.
Our Subcommittee on Improved Light Water Reactors held meetings on June 17-18 and July 27, 1992, to review this subject.
During these meetings, we had the benefit of discussions with representatives of the NRC staff and EPRI.
We also had the benefit of the documents referenced.
In the early 1980s, EPRI established the ALWR program to, support the United States utility industry efforts to ensure a viable nuclear power generation option for the 1990s and beyond.
The objective of the program was to ensure that future nuclear power plants would be safer, simpler, more robust with greater margins, more easily operated and maintained. and more certain of being constructed and licensed without delays.
This was accomplished using utility experience, by establishing design philosophy, producing design criteria and guidance to achieve the objective, and addressing the policies and regulations of the NRC.
The EPRI ALWR URD is a compendium of technical requirements for design,, construction, and performance of ALWR nuclear power plants for the 1990s and beyond.
The URD consists of three volumes:
Volume I, "ALWR Policy and Summary of Top-Tier Requirements,"
e is a management-level synopsis of the URD, including the design objectives and philosophy, the o/erall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant.
The Honorable Ivan Selin 2
August 18, 1992 Volume II, "ALWR Evolutionary Plant," c nsists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1350 Hwe).
Volume III, "ALWR Passive Plant," contains utility-design requirements for passive nuclear p,ower plants (approximately 600 Hwe).
/
We have followed the develop' ment of the.EPRI ALWR program from its inception and offered suggestions regarding safety improvements on several occasions.
We also held numerous subcommittee and Committee meetings to consider and discuss the development of the EPRI URD program and the NRC staff's reviews.
The staff's review of the URD was conducted as described in NUREG-1197.
As noted therein, the staff used NUREG-0800,
" Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants," for review guidance.
In addition, the staff's review reflects the requirements of 10 CFR 52, the Commission's policy statements on severe accidents, and the safety goals.
Although the SRP was used by the staff as guidance, the level of detail in the EPRI submittal did not permit a review of its completeness.
(The SRP was written to support the review of safety analysis reports on specific plant designs for which a significant amount of design and construction information was available.)
The staff conducted its review with the understanding that EPRI design criteria would mee".
all current regulations, except where deviations were identified.
The staff's review of the URD focused primarily on determining whether the EPRI criteria did or did not conflict with current regulatory requirements.
In its review of Volume II of the URD, the staff identified a number of issues that will require additional information before the staff can reach a final conclusion.
Initially, the staff divided the outstanding issues into three categories:
(1) open policy issues on which the staff has proposed a position, but for which the Commission has not yet provided guidance, (2) open issues that rust be satisfactorily resolved oefore the staff can complete its review of the URD, or (3) confirmatory issues for which the staff will ensure follow up of commitments in the URD.
At this date the staff has identified 21 open policy issues that are included in a draft Commission paper, " Issues Pertaining to Evolutionary and Passive Light Water Reactors and Their Relationship to Current Regulatory Requirements" that was issued on February 27, 1992.
We provided our recommendations on'tha open policy issues pertaining to evolutionary plants in our letters which addressed SECY-90-016, SECY-91-078, and the draf t SECY paper of February 7, 1992.
i.
4 The Honorable Ivan Selin 3
August 18, 1992 The staff has handled the remaining 410 open issues which were identified in the FSER for Volume II by classifying them as " Vendor or Utility Specific Items" which must be satisfactorily addressed during the staff's review of a
vendor-or utility-specific application.
The staff plans to issue a supplement to the FSER af ter all evolutionary policy issues have reached final resolution.
The staff indicated that they plan to interact with EPRI in an attempt to resolve significantMipen issues which may be resolved generically, and to include in a supplement any which are resolved.
We recommend generic resolution of as many of these issues as possible.
We commend EPRI for developing a comprehensive set of requirements.
These will aid in the design of nuclear plants which will be safer, simpler, more robust, and more easily operated and maintained.
We commend the NRC staff for a very thorough review of the EPRI ALWR Evolutionary URD, and its work with EPRI to identify a7d resolve many issues relevant to licensing future LWRs.
We recognize the NRC staff's position that its review necessarily is incomplete.
Sincerely, David A. Ward Chairman
References:
1.
SECY-92-172, dated May 12,
- 1992, from James M.
- Taylor, Executive Director for Operations, for the Commissioners,
Subject:
Final Safety Evaluation Report for Volume II of the Electric Power Research Institute's Advanced Light Water Reactor Requirements
- Document, including the following enclosures:
Draft Safety Evaluation Report for Volume I,
" Program Summary of the NRC Review of the Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document," prepared by the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, dated May 1992 Safety Evaluation Report for Volume II, "NRC Review of e
the Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document for Evolutionary Plant Designs," prepared by the Office of Nuclear Reactor Regulation, U.S.
Nuclear Regulatory Commission, dated May 1992
1 The Honorable Ivan selin 4
August 18, 1992 2.
Advanced Light Water Reactor Utility Requirements Document, Volume II, "ALWR Evolutionary Plant," Chapters 1-13, through Revision 4,
dated April 1992, Prepared for Electric Power Research Institute 4
4 l
30
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%...l.. f July 17, 1992 The Honorable Ivan Selin Chairman U.S. Nuclear Regulatory Commission S
Washington, D.C.
20555
Dear Chairman Selin:
SUBJECT:
INTEGRAL SYSTEM AND SEPARATE EFFECTS TESTING IN SUPPORT OF THE WESTINGHOUSE AP600 PLANT DESIGN CERTIFICATION During the 387th meeting of the Advisory Committee on Reactor Safeguards, July 9-11, 1992, we discussed the programs of integral system and separate effects testing being planned by both Westinghouse and NRC to support the certification effort for the Westinghouse Electric Corporation's AP600 passive plant design. We held discussions on this matter during our 381st.through 384th (January-April 1992) meetings, inclusive.
Our Subcommittee on Thermal Hydraulic Phenomena held meetings on December 17, 1991, March 3, 1992, and June 23-24, 1992 to review this issue.
During these
- meetings, we had the benefit of discussions with representatives of the Westinghouse Electric Corporation and the NRC staff.
We also had benefit of the referenced documents.
We have previously reported to you on this matter in our letters of March 10 and April 6, 1992.
B CKGROUND Appropriately validated thermal hydraulic computer models must be relied on to support the safety assessments required for certification of the AP600.
Westinghouse has indicated that it plans to use its more mechanistic assessment code, HCOBRA/ TRAC, only for large-break LOCA analyses, and will rely on its evaluation model, NOTRUMP, for analyses of all other design-basis events. The NRC plans to use RELAP5/ MOD 3 to support its assessments.
The NOTRUMP code is an evaluation model code that is based on 10 CFR Part 50, Appendix K, requirements.
The other two codes, HCOBRA/ TRAC and RELAPS/ MOD 3, are more mechanistic codes that have been qualified re, best-estimate tools only for large-break LOCAs.
All of these analysis tools will be required to simulate the AP600 behavior in regimes'where the codes are known to be weak.
These regimes include phenomena such as horizontal (perhaps countercurrent stratified)
- flows, interface movements, thermal w/
The Honorable Ivan Selin 2
July 17, 1992 stratification, rapid " shock" condensation, b*oron mixing, and low-pressure gravity-driven flows.
To develop the necessary data for improvement and validation of
)
these models for AP600 assessment, Westinghouse now has plans for conducting a number of separate effects tests at several different facilities, and integral system tests.The integral system test programs are to be conducted in a low-pr' essure facility now nearing final design at the ' Oregon State University (OSU) and in an existing high-pressure facility, SPES (in Italy), to be modified to better simulate AP600.
The NRC has proposed to conduct high-pressure confirmatory testing by modifying and using the existing ROSA-IV_ facility at JAERI in Japan.
The modified facility will be referred to as ROSA-V.
The NRC has no specific plans for_ additional separate effects testing.
The staf f does plan to conduct low-pressure integral system testing in the OSU facility after the Westinghouse program has been completed.
At this time, we have the following comments and recommendations regarding various aspects of these planned and proposed efforts.
WESTINGHOUSE PROGRAM We believe that, with certain enhancements, the Westinghouse program will be adequate for the certification process.
We have the following specific comments and recommendations:
We are concerned that Westinghouse plans to rely primarily on e
its NOTRUMP evaluation model (EM) code.
It is a step backwards to use computer codes of only EM sophistication and capabilities to evaluate the thermal hydraulic behavior of new nuclear power plants.
The Westinghouse separate effects tests of most importance to the certification of-AP600 are the Core Make-up Tank (CMT) tests and the Autontic Depressurization System (ADS) tests.
The test matrices for these do not cover ranges of conditions that are broad enough to yield an adequate data base-for the required model development.
We _ recommend that pressure disturbances of the types that would-be caused by either ADS-valve actuation _or by rapid steam condensation when cold CMT fluid is injected into the downconer region be part of the test program.
e An additional separate - ef f ects test facility is needed to investigate the asymmetric effects associated with _the-downcomer and with the cold-side plenum of-the ' steam.
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The lionorable Ivan selin 3
July 17, 1992 SPES is generally a good choice for conducting full-height, full-pressure integral system tests, llowever, in addition to the scaling problems associated with a high ratio of surface area to fluid volume that plague small-scale simulations of this kind (and must be dealt with), the proposed modified version, SPES-II, has two important scaling defects that should be eliminated:
(a) the
- aspect ratio (height to diameter) of the simulated pressur'izer is different from that of the AP600 and (b) the cold leg configuration is not geometrically similar to that of AP600.
We recommend that Westinghouse be required to preserve the scaling of the pressurizer and the geometrical configuration of the cold legs, to. better simulate AP600 behavior (this would include simulation of a reactor coolant pump in each leg).
The method proposed for simulating steam generator tube ruptures in SPES-II is flawed in that it does not appear to allow the break flow from the primary system to be from both the hot and cold sides of the tube.
We recommend that Westinghouse develop a better simulation method.
The OSU low-pressure integral system testing facility is well conceived.
We commend Westinghouse for its efforts with respect to this facility.
Our evaluhtion of the scaling rationale for the facility design (discussed during the subcommittee meeting of June 23-24, 1992) is that it is soundly based. Further, the 400 psia design capability should allow considerable simulation of high-pressure effects, while providing the more important low-pressure behavior.
Imc PROGRAM our undustanding of the justification provided by the NRC staff for its proposed confirmetory high-pressure integral system testing in the ROSA-V facility is as follows:
o Because ROSA-V is considerably larger than SPES-II, such confirmatory testing would provide an additional check on the adequacy of the scaling capabilities of the codes, and would help confirm that important effects have not been overlooked.
The confirmatory test program would provide the opportunity to e
maintain the staf f's - thermal hydraulic expertise and up-to-date knowledge in this field.
While we agree that the above considerations have some merit, we have not been persuaded that confirmatory high-pressure testing by the staff is needed before the AP600 design certification and, even-if this were the case, we have significant reservations about the 33
The Honorable Ivan Selin 4
July 17, 1992 adequacy of the ROSA-V facility for this pu'rpose.
These positions are based on the following observations:
e The NRC staff has not presented convincing arguments supporting its needs for confirmatory testing, particularly at high pressures.
The SPES-II facility appears to b sufficient to meet all tho high-pressure integral system testing needs.
The NRC will be able to use the SPES-II facility for its confirmatory testing needs just as it plans to use the OSU facility.
The desired staff experience will come from pre-test and post-test evaluations of the various tests using the RELAPS/ MOD 3 code.
This experience can just as easily be obtained by evaluating the SPES-II and OSU tests and results.
The ROSA-V facility contains several atypicalities that will manifest themselves in difficult-to-explain behavior relative to that expected for AP600 (the sensitivity of the ROSA-V thermal hydraulic behavior is well documented in the INEL report, NUREG/CR-5853).
The tests would be in a distant location.
There would be a very limited number of tests, because of the expense involved.
In
- addition, we are concerned that the adequacy of instrumentation (for example) might have to be compromised in order to reduce overall program costs.
For the above reasons, we believe that NRC resources would be bette" used by focusing on three areast (a) possible additional separute effects testing to support the modeling needs for REIAPS/ MOD 3, (b) participation in the pre-test and post-test analyses efforts associated with the SPES-II and the OSU test programs, and (c) consideration of utilizing the SPES-II facility for high-pressure confirmatory testing needs in the same way the staff plans to use the OSU facility for its confirmatory low-pressure testing needs.
To accomplish the above objectives, we believe that the staff should consider the establishment of a task force of experts in related fields to assist it in the development of the analytical and experimental programs necessary for timely certification of the AP600 passive plant design.
Sincerely,
+
Paul Shewmon Acting Chairman
The Honorable Ivan Solin 5
July 17, 1992 4
11e f erengag:
1.
U.S.
Nuclear Regulatory Commission, HUREG/CR-5053,
" Investigation of the Applicability and Limitations of the ROSA-IV Largo Scale Test Facility for AP600 Safety Assessment (Draft)," dated May 1992 2.
T. Doucher, Idaho National Engine'ering Laboratory, et al.,
" Scaling Issues for a
Thermal-Hydraulic Integral Test Facility," Paper transmitted via a memorandum from L. Shotkin, NRC-RES, for P. Boehnert, ACRS, dated June 29, 1992 3.
Oregon State University Report, OSU-NE-9204 (Draf t), " Scaling Analysis for the OSU AP600 Integral System and Long Term Cooling Test Facility,"
J.
- Reyes, Jr., dated June 1992 (H Proprietary Report) 4.
Letter dated January 22, 1992, from G. Saporano, ENEA, Italy, to E.
S. Dockjord, NRC, transmitting documentation on SPES test facility S.
Homorandum dated June 13, 1991 from S.
Hodro, INEL, for L.
Shotkin, NRC-RES, transmitting draft report, " Evaluation of Scaled Integral Tout Facility Concepts for the AP600" by Hodre, et al.
6.
U.S.
Nuclear Regulatory Commission, SECY-92-219, "NRC-Sponsored ' Confirmatory Teating of the Westinghouse AP600 Design," dated Juno 16, 1992 (Predecisional) 7.
U.S. Nuclear Regulatory Commission, SECY-92-219A, " Addendum to SECY-92-219 - Providing Additional int'rmation to Justify Sole Sourco Procurement," dated July 9, 1992 (Pradeoisional) 8.
Homorandum dated April 21, 1992, from S. Chilk, Secretary, for J.
H. Taylor, EDO, and W. Parlor, General Counsel,
Subject:
SECY-92-037 Need for NRC-Sponsored Confirmatory Integral System Testing of the Westinghouse AP600 Design 9.
Westinghouse Topical Report, WCAP-13277, " Scaling, Design and Verification of the SPES-2, the Italian ExperimentaJ Facility Simulator of the AP600 Plant," dated April 1992 (E Proprf stary Report) 3f
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The lionorable Ivan Selin Chairman U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Chairman Solin:
SUBJECT:
TESTING AND ANALYSIS PROGRAMS IN SUPPORT OF TiiE SIMPLIFIED BOILING WATER REACTOR DESIGH CERTIFICATION During the 385th and 386th meetings of the Advisory committee on Reactor Safeguarde, May 6-9 and June 4-5, 1992, we reviewed the testing and analysis programs in progress and proposed by GE Nuclear Energy (GE) in support of the certification effort for the Simplified Boiling Water Reactor (SBWR) passive plant design.
Our Subcommittee on Thermal liydraulic Phenomena held meetings to discuss this topic on April 23 and June 2,
1992.
During these mootings, we had the benefit of discussions with representatives of GE and the NRC staff.
We also had the benefit of the documents referenced.
GE will use its best-estimate code, TRACG, to evaluate the SBWR thermal hydraulic behavtor under accident conditions ranging from ATWS with instabilities to long-term behavior of the Passive containment Cooling System (PCCS).
GE representatives presented a very good analysis of processes and phenomena important to accident scenarios postulated for the SBWR.
The results were summarized in tables which are to be used by GE to validate the TRACG computer
- code, flowever, these same tables appear not to have been used to guide the design and operation of the experimental f acilities that are to support the code validation process.
The GE experimental program consists of three elements:
1)
Laboratory scale experiments to obtain fundamental heat transfer data, 2)
Separate effects tests to obtain data for parts of the total
. system and full-scale components where necessary, and 3)
Integral system tests to obtain system data.
_ - _ - - _ _. = _
The lionorable Ivan Selin 2
June 10, 1992 Although we were shown some comparisons of T'RACG predictions with data f rom GE's integral system tests (GIST and GIRAFFE f acilities),
the question of whether or not the facilities can scale the important phenomena was not addressed in either GE's presentation or in the documents supplied to the ACRS by GE.
A rigorotto scaling analysis is needed if integral system test data alone are'to be used to demonstrate that a TRACG calculation is meaningful.
We have some comments about the elements of the GE test plan.
The initial conditions for the integral system tests are based on conditions assumed to exist some time after vessel depressurization.
These conditions include an initial drywell and pCCS nitrogen mass fraction of 15 percent.
The nitrogen concentration could.be much higher.
GE should develop a basis for its choices of initial conditions or broaden its test matrix to include some tests at much higher values of the nitrogen concentration, both in the drywell and in the PCCS.
Separate effects tests to be conducted in the PANTHERS facility will yield the data needed to characterize heat exchanger behavior under a variety of expected conditions.
In particular, GE has agreed to add instr.umentation to the individual heat exchanger tubes to obtain local heat transfer data.
This will make the GIRAFFE integral system experiments more meaningful. We believe GE has been very responsive to issues raised by both the ACRS and the NRC staff in this regard.
The oscillatory behavior observed in the GIRAFFE integral system tests needs more detailed study to ensure that the suppression pool does not overheat due to steam bypass of the PCCS through the suppression pool top horizontal vents.
The steam flow rate will be low which could lead to a stratified condition.
The suppression pool is not a very effective heet sink when this process occurs.
This may well require a separate effects study to obtain data for development of a low steam flow model for the horizontal vent.
Further, review of the GIRAFFE facility instrumentation is needed to ensure that the resulting data will support TRACG model validation.
The SBWR has full pressure isolation condensers (IC) capable of removing 4.5 percent of full power decay heat at full system pressure.
The behavior of isolation condensers is well understood and introduces no new processes.
GE has indicated that it will collect relevant IC operating data for staff review.
The SBWR is automatically depressurized when the vessel water level drops to some prescribed value by a staged opening of squib-type valves.
Further, GE has had,a great deal of experience with automatic depressurization and only the squib-type valve itself is of a new design.
As a result, we do not believe that full-height, full-37
The Honorable Ivan Solin 3
June 10, 1992 pressure integral system testing is required'for certification of the SBWR design.
The GE program includes conduct of integral system testing at the PANDA facility located in Switzerland.
The NRC staff would like GE to obtain data from this facility in time to support its' design certification review of the SBWR.
To do so, GE would have to accelerate its schedule by six months.
We agree with the NRC staff that further integral system testing of the PCCS is needed prior to the final design approval.
It has not been demonstrated by GE that existing data obtained from GIRAFFE or GIST testing are sufficient for validation of the TRACG code, nor that the PANDA test facility will yield the needed data.
A more definitive assessment by GE is needed; this assessment should include both the scaling rationale for the GIRAFFE, GIST, and PANDA f acilities, and a demonstration of how the effects of test facility scaling distortion impact the important processes and phenomena outlined by GE in its evaluation
-l of TRACG.
As a part of such an effort, it may be possible to show that one can obtain the needed data by some combination of additional separate effects tests and judicious use of the GIRAFFE and GIST facilities.
To summarize, we agr'oe with the NRC staff views that full-height, full-pressure integral system testing is not needed to support the SBWR design certification.
Further, we agree that early integral system testing of the PCCS is essential to meet the present design certification schedule.
We have not, however, seen evidence that-the PANDA facility is adequate to obtain the needed data.
Sincerely, W.
David-A. Ward Chairman References 1.
Memorandum dated February 26, 1992, for the Commissioners from James M.
- Taylor, Executive Director for operations, transmitting Advance Copy of proposed Commission paper,
" Evaluation of the General Electric Company's (GE's) Test Program to Support Design Certification for the Simplified Boiling Water Reactor --(SBWR)"
2.
Letter dated February 3, 1992, from R. C. Mitchell, GE Nuclear Energy, to U.S. Nuclear Regulatory Commission,
Subject:
GE Response to Request for Information on SBWR Testing program 37
The lionorable Ivan Selin 4
June 10, 1992 3.
Joint Study Report,
" Feature Technology of Simplified BWR (Phase I) GIRAFFE (Final Report)," dated November 1990, The Japan Atomic Power
- Company, et al.
(GE Proprietary Information) 4.
GE Huclear Energy, GEFR-00850,
" Simplified Boiling Water Reactor (SBWR) Program Gravity-Driyen Cooling _ System (GDCS)
Integrated Systems Test - Final Report," A.F. Billig, dated October 1969 (Applied Technology Restriotion) 5.
" ALPHA - The Long Term Passive Decay Heat Removal and Aerosol Retention Program at the Paul Scherrer Institute, Switzerland,"
by P.
Coddington, et al.,
Paul Scherrer Institute, undated 6.
Paper from the Proceedings of The International Conference on Multiphase Flows
'91
- Tsukuba, Japan, September 24-27,
" Condensation in a
Natural Circulaticn Loop with Noncondensable Gases Part 1 - lleat Transfer," K. M. VieroW, GE Nuclear Energy, and V.
Schrock, University of California 7.
GE Draft Report: " Test Specification for IC & PCC Tests,"
undated (GE Proprietary Information) 8.
Paper submitted to the Department of Energy, "The Effect of Noncondensable Gases on Steam Condensation Under Forced convection conditions,"
M.
- Siddique, Ph.D.
Thesis Massachusetts Institute of Technology, dated January 1992
[# ts:g'o UNITE D STATES NUCLEAR REGULATORY COMMISSION g
e ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
/
WASHINGTON,0. C. 20665 o
September 16, 1992 Mr. James H. Taylor Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Dear Mr. Taylors
SUBJECT:
DRAFT COMMISSION
- PAPER,
" DESIGN CERTIFICATION AND LICENSING POLICY ISSUES PERTAINING TO PASSIVE AND EVOLUTIONARY ADVANCED LIGHT WATER REACTOR DESIGNS" During the 389th meeting of the Advisory committee on Reactor Safeguards, September 10-12, 1992, we reviewed the NRC staff's positions and recommendations concerning the certification issues for evolutionary and passive light water reactor designs contained in the draft Commission paper, which was forwarded to the Commis-sion on June 25, 1992.
Our Subcommittee on Improved Light Water Reactors met on September 9, 1992, to review this subject.
During these meetings we had the benefit of discussions with representa-tives of the NRC staff and EPRI.
We also had the benefit of the document referenced.
We previously provided comments to you on other policy issues related to design certification in our letters of May 13, 1992 and August 17, 1992.
Our comments and recommendations on the proposed policy issues contained in the draft Commission paper are given below.
Issues A, B, C, D, E, and G apply to evolutionary and passive plant designs-and Issues F and H apply only to passive plant designs.
The issue titles and letter designations correspond to those of the draf t Commission paper.
A.
Defense Acainst Common-Mode Failures in Dioital Instrumenta-tion and Control (I&C) Syst, mag It is our view that the thrust of the staff recommendations concerning defense _ against common-mode failures in digital ~ I&C systems as u.nderlined in Issue A of the' draft Commission paper is appropriate.
We agree with the staff-that the applicant should be required to ' assess the defense in depth and diversity of the proposed designs for the events postulated in the-Safety Analysis Report, and demonstrate an acceptable plant response for each. The staff proposes that the instruments,
- controls, and equipment required to demonstrate an acceptable response be independent of any common-mode failure mechanisms associated with'the event.
We view this requirement to be essential, but remain open as to the 9a
Mr. James H. Taylor 2
September 16, 1992 best approach.
The staff proposes an indep'6ndent set of safety-grade displays and controls in the main control room.
We believe that other arrangements might be shown to be acceptable.
In a separate letter to Chairman Selin dated September 16, 1992, we have provided additional comments and advice regarding the general approach being taken by the staff in its review of digital instrumentation and control systems.
D.
Analvnes of External Eventn Devond the Desian Basis To assist in the closure of severe accident issues, the staff recommends that (1) analyses submitted in accordance with the requirements of 10 CPR 52.47 (concerning the contents of applica-tions for standard design certification) include an assessment of internal and external events and (2) during the design certifica-tion review, the staff should evaluate those external events that are not site dependent (e.g., fires, internal floods) and certain bounding analyses.
We agree with this staff recommendation.
c.
nipination of the onoratina Basis Parthouake from seismic Desian The staff is s ill reviewing this issue and has expressed only an interim position._ We believe the staff in taking an appropriate approach in its. interim position.
D.
Multinle Steam Generator Tube Ruotures (MSGTRs)
The staf f is recommending that the applicant for design certifica-tion perform additional analyses to determine the AP600 response to multiple breaks of up to 5 steam generator tubes.
We agree with the staff's recommendation, but believe the staff should have a better technical basis for estimating the frequency of occurrence of such multi-tube breaks.
The staff is also recommending that the applicant for design certification of a passive or evolutionary PWR assess design features necessary to mitigate the amount of containment bypass leakage that could result from MSGTRs.
We agree with the staff's recommendation.
E.
Probabilistic Risk Assessment (PRA) Beyond Desicm Certificq-tion The staff is recommending that, throughout the duration of the combined or operating license, the PRA be revised to address significant plant m6difications, operating experience, and other developments that may affect previous PRA insights.
/
l Mr. James H. Taylor 3
September 16, 1992 We are convinced that it is worthwhile for a blant operator to have an up-to-date PRA ard are, therefore, reluctant to recommend against this position.
However, if this is to be required, the staff should more clearly specify how it intends to use the up-dated PRA and what is meant by keeping it current.
We think such guidance is part of the overall issue of appropriato use of PRAs in regulation and would be helpful to licensees and to the statf.
F.
Role of the Ooerator in a Passive _ Plant Control Room We agree with the first part of the staff's position "that suf ficient man-in-the-loop testing and evaluation be performed...
to demonstrate that functions and tasks are integrated properly into the man / machine interface design" of passive ALWR control rooms.
The second part of the staff's underlined position states "that a fully functional integrated control room prototype is necessary for passive plant control room designs to demonstrate that functions and tasks are integrated properly into the man / machine interface design."
We pointed out to the staff that the non-underlined last sentence of this paragraph is inconsistent with this language in that it would ermit an applicant to " demonstrate that a control room prototype of reduced scope is sufficient."
We also pointed out that the non-underlined paragraph preceding the underlined paragraph states that such a prototype "would likely" be required (not would be required) to demonstrate that functions and tasks are integrated properly into the man / machine interface design.
We believe that the staff should clarify its intent by reconciling these various statements.
The staff believes that operators of passive plants will be confronted with a new operating philosophy.
The staff argues that "the operators of passive plants must understand the operation of
' investment protection' systems and their interfaces with the safety-related passive systems" and that they will be confronted with "new functionc and tasks unlike those required for evolution-ary plants" (or current plants)
"due to the not approach in operational philosophy" and "the increase in automstion, and the greater use of advanced technology in the passive plant designs."
As a result of our discussions with the staff and EPRI, we believe that the staff may be overreacting to the " newness" of these issues.
It appears to us that additional discussion of this issue among the staff and EPRI and the vendors is needed.
G.
Control Roon Annunciator (Alarm) Reliability We agree with the st'aff's position that the alarm system for ALWRs should neet the requirements of the EPRI Utility Requirements Document.
Mr. James H. Taylor 4
September 16, 1992 Reculatory Treatinent of Nonsaf ety Systefns We were told that the staff is still engaged in significant on-going discussions and review of this issue and that the associated position and recommendations are subject to modification.
We believe the issue is substantial and has broad implications with respect to such items as use of PRAs in_ regulation, safety goal implementation, and reduction of regulatory burdens, and we expect to have additional future interactions with the staff and the industry.
Consequently, we are not prepared to express a position on this issue at this time.
Sincerely, David A. Ward Chairman Poference:
1.
Draft Commission Paper dated June 25, 1992, from James M.
- Taylor, Executive Director for Operations,
- liRC, for the Commissioners,
Subject:
Review of the Draf t Commission Paper,
" Design Certification and Licensing Policy Issues Portaining to Passive and Evolutionary Advanced Light Water Reactor Designs"
e
[g* *8 cwo UNITED STATES y
NUCLEAR REGULATORY COMMISSION o
.I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
/
WASHINGTON, D C. 20666 o
%...../
August 17, 1992 Mr. James M. Taylor Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dear Mr. Taylort
SUBJECT:
ISSUES PERTAINING To EVOLUTIONARY AND PASSIVE LIGHT WATER REACTORS AND THEIR RELATIONSHIP TO CURRENT REGULATORY REQUIREMENTS a
During the 386th,
- 387th, and 388th meetings of the Advisory Committee on Reactor Safeguards, June 4-5, July 9-11, and August 6-8, 1992, We discussed with representatives of the NRC staff the st'aff's positions, recommendations, and resolution schedules concerning the certification issues for evolutionary and passive light water reactors contained in the draf t SECY paper dated February 7, 1992.
This supplements our -letter of May 13, 1992, and provides our comments and recommendations on some of the staff's positions for the passive light water reactors.
The section titles and letter designations correspond to those in the draft SECY paper.
I.
SECY-90-016 Issues (For Passive Plants)
E.
Fire Protection The NRC staff is seeking Commission approval to use the enhanced fire protection criteria previously approved for evolutionary Advanced Light Water Reactor. (ALWR) plants by. the Commission's Staff Requirements Memorandum (SRM) of June 26, 1990.
This SRM approved the staff's position on fire protection as presented in SECY-90-016 and supplemented by the staff's April 27, 1990 response to our report on the SECY.
We recommended separate
- Heating, Ventilating, and Air conditioning (HVAC) systems for each division as an important step toward ensuring - adequate environmantal separation of safety systems. The staff agreed that consideration of smoke, heat, and fire suppressant migration may result in separate HVAC systems, but other options may be available to the designer.
Our report to the Commission of April 13, 1992, on the Draf t Safety Evaluation Report for the ABWR identified the adequacy of physical separation as a continuing issue for. the 3
a
Mr. James M. Taylor 2
August 17, 1992 ABWR, due in part to the use of a shared HVAC system for multiple trains of redundant safety systems during normal plant operation.
our concern with shared HVAC systems is relat ' to the need for adequate isolation of such systems during certain disruptive events (e.g.,
fires, floods, or pipe breaks).
If the isolation is.not adequate, the HVAC arrangement may become a pathway whereby effluents from the even; are conducted to locations where required safe shutdown equipoent is located.
This is not a concern if either (1) the HVAC isolation provisions are able to withstand the event consequences (e.g.,
pipe whip, jet impingement, static and dynamic pressure, and elevated temperature) during and after closure with consideration of single active component failures and acceptable leakage, or (2) the safe shutdown eclipment is qualified for the environmental exposure resulting from a release of the adverse environment at any credible location along the HVAC pathway such as duct openings or blowout locations.
Except for. the concern with shared HVAC, we support the staff recommendation that the passivo plants should be reviewed against the enhanced fire protection criteria approved in the Commission's SRM.
F.
Intersystem Loss-of-Coolant-Accident The staff's pos. tion is that designing these low-pressure fluid systems that interface the reactor coolant system (RCS) to withstand full RCS pressure (to the extent practicable) is an acceptable means for resolvi:ng this issue.
For those systems t.% t have not been designed to withstand full RCS
- pressure, ua staff indicates tnat other measures will be required.
O, fecommend approval of the proposed staff resolution, provided consideration is given to all elements of the low. pressure-piping system (e.g., instrument lines, pump seals, heat exchanger tubes, and valve bonnets).
G.
Hydrocen Control The staff recommends that the evolutionary LWR designs provide a system for hydrogen control that can safely accommodate hydrogen gent wted by the reaction of steam with 100 percent of the fuel
'. adding surrounding the active fuel.
(Note:
This is not 1 percent of the reactive metal in the core.)
We support the taff's recommendation.
The staff also recommends that the system be capable of precluding uniform containment concentrations of hydrogen greater than 10 percent.
We are aware of analytical work in i
Mr. James M. Taylor 3
August 17, 1992 support of the resolution of Generic Issue 106, " Piping and the Use of Highly Combustible Gases in Vital Areas," that suggests the possibility of transition to detonation at average concentrations as low as 12 percent.
We recommend that the staf f do a similar analysis of the impact of hydrogen combustion, and possible detonation including stratification, before establishing a
limit for the average hydrogen concentration.
This is of particular importance to steel-shell containments.
I.
Hich Pressure Core Melt Eiection To cope with the possible effects of direct containment heating (DCH), the staff concludes, that ALWR design should include a depressurization system and cavity design features to contain ejected core debris."
DCH is an extremely improbable event, and we see no need to require two modes of coping with the possibility.
Either depressurization or cavity design provisions alone should be adequate.
Because of possible safety benefits for other events, reliable depressurization is the preferred approach.
J.
Containment Performance The str*f has not yet developed an adequate technical position celats.r ti to requirements for containment performance in cassive LWRs.
We agree that the proposed value of 0.1 for a
- onditional containment-failure probability (CCFP) is reasonable but, as we stated in our letter of April 26, 1990, regarding " Evolutionary Light Water Reactor Certification Issues and Their Relationshi;p to Current Regulatory Requirements," this value is defined only within the context
,f a family of initiating events.
It should be used by the staff in the development. of its requirements and not merely passed on to applicants.
The dettrainistic criterion proposed by the staff is not a
simple alternative to the CCFP.
It could be used more logically as a complement.
Using ASME Code Service Level C stress limits is not unreasonable given a known loading for which the containment is to be designed.
Howeve.r, determination of the appropriate loading is the hard part of the problem and the suggested deterministic criterion is essentially meaningless without it.
The staff states that
" applicants using the deterministic approach will be required to define the challenges considered in this evaluation."
The staff takes no position on what those challenges should be or how they are to be quantified.
Apparently the intent is to default to a " design specific review."
This approach leaves
Mr. James M. Taylor 4
August 17, 1992 the applicant without any real guidance from the' Commission on this important topic.
We acknowledge that it is a very difficult task to establish containment performance criteria but is importarit.
We suggested what we believe to be the best approach in our letter of May 17, 1991, " Proposed Criteria to Accommodate Severe Accidents in Containment Design."
K.
Dedicated Containment Vent Penetration The staff proposes that the decision on the need for a containment vent for passive designs should not be made at this time but should wait until specific plant designs are evaluated.
We believe that the Commission should make a generic judgment about the acceptability of containment vents for LWRs.
This should be a part - of establishing general criteria for containment design as proposed in our letter of May 17, 1991.
L.
Ecuinment Survivability We agree with the staff's recommendation that features provided only for severe-accident mitigation for the passive plant designs not be subject to the environmental qualification requirements of 10 CFR 50.49,' quality assurance requirement of 10 CFR 50, Appendix B,
and redundancy /
diversity requirements of 10 CFR 50, Appendix A.
N.
In-Service Testina of Pumns and Valves We support the staff recommendation that the special pump and valve design, testing, and inspection provisions be imposed on all safety-related pumps and valves for the passive ALWRs.
III.E - Control Room Habitability There vere several significant differences between the staff and EPRI at the-time the staff drafted this policy issue.
EPRI has subsequently made a proposal to modify its Utility Requirements Document to include a requirement for a passive, safety grade, control room pressurization system that would use a bottled air supply to maintain-operator doses within regulatory limits for the first 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an accident.
(The regulations require that operator doses be so limited for the, duration of the accident.) The pressurization system proposed by EPRI would be designed to be replenished by off-site portable supplies after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if needed.
Accordingly, EPRI has recommended that the staff close this issue.
Mr. James M. Taylor August 17, 1992 5
no We discussed this matter with the staff and EPRI during our June 4-5, 1992 meeting.
The staff told us that it is currently evaluating the EPRI proposal and is not-prepared to close this issue.
ACRS had several comments regarding design -
features of the passive control room pressurization system proposed by EPRI.
We believe that the staff should take these comments into account in its evaluation.-
We may provide additional recommendations after the staff has completed-its evaluation.
Sincerely, O
r-
-Y David A. Ward Chairman
References:
1.
Draft SECY Paper dated February 7,1992, from James M. Taylor, Executive Director for Operations, NRC, for the Commissioners,
Subject:
. Issues Pertaining to Evolutionary and Passive Light Water Reactors and Their Relationship to Current Regulatory Requirements 2.
SECY-90-016 dated January 12, 1990, from James M.
- Taylor, Executive ' Director for Operations, for the Commissioners, subject: Evolutionary Light Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements 3.
Memorandum dated April 27,
- 1990, from James M.-Taylor, Executive Director for Operations, NRC, for NRC Commission,
Subject:
Staff.. Response to ACRS Conclusions Regarding Evolutionary Light Water Reactor Certification Issues
~
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'n NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 0,
W ASHINGTON, D. C. 20$$5
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- May 13, 1992 Mr. James M. Taylor Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Taylor:
SUBJECT:
ISSUES PERTAINING TO EVOLUTIONARY AND PASSIVE LIGHT WATER REACTORS AND THEIR RELATIONSHIP TO CURRENT REGULATORY REQUIREMENTS During the 383rd,
- 384th, and 385th meetings of the Advisory Committee on Reactor Safeguards, March 5-7, April 2-4, and May 6-9, 1992, we discussed with representatives of the NRC staff the staff's positions, recommendations, and resolution schedules concerning the certification issues for evolutionary and passive light water reactors contained in the draft SECY paper dated February 7,
1992.
We also had the benefit of the documents referenced.
The staff requested ACRS comments on the draft SECY paper.
Our comments and recommendations on some of the staff's positions are given below.
I.
SECY-90-016 Issues Item M.
Elimination of Operatina Basis Earthauake Appendix A to 10 CFR Part 100 currently establishes the Operating Basis Earthquake (OBE) at a level one-half of the Safe Shutdown Earthquake (SSE).
With this specification, the OBE exerts undue influence over the seismic design and requires a
full spectrum analysis in addition to that of the SSE.
The staff's proposal-is to effectively decou-ple the OBE from design.
We agree with the staff's recommendation.
II.
Other Evolutionarv and Passive Desian Issues Item A.
Industry Codes and Standards We agree with the staff's recommendation to use the newest codes and standards that have been endorsed by the NRC in its reviews of both the evolutionary and passive plant design applications, and its 9
Mr. James M. Taylor 2
May 13, 1992 recommendation that unappro d revisions to codes and standards be reviewed on a case-by-case basis.
Item D.
IAak Before Break We agree with the staff (s recommendation to extend the application of the. leak-before-break approach for both evolutionary and passive advanced light water reactors.
Item E.
Classification of Main Steamlines of Boilina Water Reactors (BWRs)
We agree with the staff's recommendation for reso-lution of the main steamline classification for both evolutionary and passive BWRs.
Item F.
Tornado Desian Basis Based on a study (NUREG/CR-4661) that compiled a considerable quantity of tornado data, the staff
. recommends that the maximum tornado wind speed of
'300 mph (compared with the present 360 mph) be used for the design-basis tornado.
We agree that the best available data should be used, but caution that design-basis specifications have sometimes been established conservatively to provide margins to deal with events not specifically addressed in the design basis.
We recommend that the staff's position be approved with a qualification that the staff require assurance that other potential loads that may have been previously subsumed within the tornado design basis be taken into account if necessary.
Item H.
Containment Leakaae Rate Testina
~
The staff recommends that the maximum interval between Type C leakage rate tests for both evolu-tionary and passive designs be increased to a 30-month interval from the 24-month interval now required in 10 CFR Part 50, Appendix J.
No signif-icant safety penalty caused by this change has been identified.
We agree with the proposed staff position.
Item I.
PostdAccident Samolina System (PASS)
The staff is requesting approval of changes in requirements for the PASS currently four.d in 10 CFR i
50,35 (f) (2) (viii).
These requirements, and the
l l --
Mr. James M. Taylor 3
May 13, 1992 guidance contained in Regulatory Guide 1.79 and in NUREG-0737, resulted from consideration of the TMI-2 accident.
We agree with the staff's proposal but have the following comments:
1.
The requirements as contained in the above referenced regulation refer to "the reactor coolant system and containment that may con-tain TID-14844 source term radioactive materi-als" and to measurement of these and other materials.
In light of source terms now considered in severe accident analysis, it is advisable to revise this obsolete description.
2.
The proposal for " Elimination of the Hydrogen Analysis of Containment Atmosphere Samples" is appropriate, given that safety grade hydrogen monitoring instrumentation will be installed.
3.
The Electric Power Research Institute (EPRI) proposed elimination of an existing require-ment for the capability to sample the reactor coolant at operating pressure in order to measure the dissolved gas and chloride in the coolant.
EPRI claims that maintaining the systems on existing plants produces signifi-cant exposure of operating personnel, and that given a severe accident, no useful informa-tion, not otherwise available, is provided by this capability.
The staff proposes to retain the requirement, but to change the time after accident onset at which the capability must be available from 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
During our discussion with the staff, we were unable to elicit any reason for this requirement other than that it was established following the TMI-2 accident.
We cannot endorse continua-tion of the requirement for high pressure sampling on the basis of information available to us.
4.
The staff proposes approval of a position that "would require the capability to take samples for boron and for activity measurements 8
' hours and'24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, respectively, after the end of power operation."
.The intent appears appropriate, however, we suggest that it might be better to specify a time at which the information from measurements becomes avail-I f/
m.;
Hay [13,-1992
~
Mr. James M. Taylor 4-o, able~to the operator.rather-than'the: time-at' which samples ; can be_ taken.
- Further, we assume that what is~ required is boron concen-tration rather than the presence or absence of--
Finally, wel suggest that ' the; phrase _-
"a.f ter the and of cpower operation" be made-more specific.
1 Item N.
Site-Soecific Probabilistic Risk Assessment If, as concluded by the staff, enveloping analyses are practical for both seismic - events and torna -
does, itois appropriate that these?be part of the:
a submittal at the time-of certification.
- However, enveloping analyses are not as-practical for other-external-events such as river
- flooding, storm-surge, tsunamis, hurricanes, and volcanism..There-fore, the.staif re m Nonds that these other types of site-specific. PRA information be submitted at:
the combined operating license (COL) stage.'
We agree with thisLrecommendation but would--like to-
' hear more about how the staff proposes to deal with.
any' unacceptable findings--at the.. COL. stage.
Sincerely OY' David A. Ward-Chairman
References:
1.
Draft SECY paper dated: February 7, 1992, for the Commis'sion.-
ers,,from James M.. Taylor,-NRC Executive Director..for Opera-tions,
Subject:
Issues Pertaining to Evolutionary and Passive Light Water Reactors and Their -Relationship.to: Current
- Regulatory Requirements (Draft Predecisional).
~ 2.
SECY-90-016 dated January 12,_1990 forfthe Commissioners from James. M.
- Taylor, NRC Executive Director' : for Operations,
Subject:
.' Evolutionary Light Water Reactor (LWR)- Certification Issues and=their Relationship to Current' Regulatory Require-e F
ments 3.
U.S.. Nuclear Regulatory commission, NUREG/CR-4661,
Subject:
Tornado Climatology-ofc the: Contiguous United States, ' dated May -
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1 ITEM 2: DIGITAL CONTROL AND PROTEGTION SYSTEM The Committee provided two reports in September 1992 regarding the staf f's recommended approach toward the design certification policy issue associated with digital instrumentation and control systems.
In a letter to the EDO dated Septembe,r.16, 1992,,the Committee generally agreed with the staff's recommendations, with the exception that independent " hardwired" analog safety-grade displays and controls might n'ot be the_ only solution to concerns over common-mode failures of digital systems. In a separate report to the Commission, the Committee expanded upon its concern that the staff's recommended position may reject a digital-based I&C backup system as a means of overcoming common-mode failure, without having assessed the reliability of such a system. In response, the'EDO provided a letter dated October 23, 1992, in which he indicated that, based on comments received from the ACRS, EPRI and industry, the staff will allow flexibility in implementing the required independent set of displays and controls, dependent upon-an evaluation of the specific design features provided by the vendor.
The final ACRS Subcommittee meeting of the two year series on these issues is scheduled for February 1993. Topics.for discussion will include quantitative software assessment methods, and the staff's progress on its proposed analog-to-digital generic letter.
The following documents are attached:
- ACRS report to the Commission dated September 16, 1992.
Subject:
Digital Instrumentation and Control System Reliability (PP. 54-57)
- ACRS letter report to James M. Taylor (EDO) dated September 16, 1992..
Subject:
Draft Commission Paper, " Design Certification and Licensing Policy Issues Pertaining to Passive and Evolutionary Advanced Light Water Reactor Designs" (PP. 58-61)-
- Letter for D. Ward (ACRS) from the EDO dated October 23, 1992.
Subject:
Defense Against Common Mode Failures in Digital Instrumentation and Control (I&C) Systems (PP. 62-64) 53
i eMar(#o, UNITED STATES 1
7, NUCLEAR REGULATORY COMMISSION-1 e,E ADVISORY COMMITTEE oN REACTOR SAFEGUARDS S
g WASHINGTON. D. C. 20$$$
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September 16, 1992 w
The Honorable Ivan Selin Chairman
,U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Chairman Selin:
SUBJECT:
DIGITAL INSTRUMENTATION AND CONTROL SYSTEM RELIABILITY During the 389th meeting of the Advisory Committee - on Reactor Safeguards, September 10-12, 1992, we reviewed the staff's proposed approach with respect to defense against common-mode failure of digital I&C systems, as discussed in policy issue "A" of the draft Commission paper
- entitled,
" Design Certification and Licensing Policy Issues Pertaining to Passive and Evolutionary Advanced Light Water Reactor I)esigns," forwarded to the Commission on June 25, 1992.
Specific comments on policy issue "A"
are contained in a letter to Mr. Taylor dated September 16, 1992.
The concerns we raise here are, however, more generally applicable, e.g.,
in connection with tue staff's proposed generic letter on analog-to-digital replacements.
The trend in most induttries over the last few decades has been toward the replacement of analog instrumentation: and control-systems with digital alternatives, and-the nuclear industry has been no exception.
This has been true -for both functional replacements within existing nuclear facilities and. for new
- designs, so it has been necessary for the staff to develop regulatory practices to deal with both the novel opportunities and the novel threats posed by these ' systems.
Experience, both military and. industrial, has generally shown the-digital systems to be more reliable and versatile than their analog =
counterparts. There are, however, some caveats and some regulatory conundrums.
An advantage is that the digital: systems are capable of --more - complex functions, - so it is possible,to build 'in self-testing capabilities that provide continuous assurance of operabil -
ity with - negligible system stress.
In addition, the digital-systems don't wear out; a, billion ~ activations of a CMOS gate are no more'damagina-than a thousand.-
.While much has been made of:the vulnerabilities of multiplexed data transmission systems, some of
-which-are doubtless real, such systems generally provide greater fidelity and reliability of data transfer, along with greater fault-f-
The lionorable Ivan Selin.
2 September 16, 1992 tolerance through error-correcting coding.
('if an analog signal is corrupted, it is often not possible to know it has happened.)
Indeed, error detection and error correction can be carried to arbitrary lengths for digitized data.
There are many other advantages, and the future clearly belongs to digital systems, where they can be used.
On the negative side, the available comp 3exity of function afforded by digital systems invites the creation of complex software, which can be difficult to validate and can be subject to surprising error (modes.
Such systems are also hard to regulate, because only the simplest programs are amenable to formal validation and verifica-tion (V&V), in the sense of a complete analysis of the mapping of
=
the input space to the output space.
For more complex programs (relevant to nuclear control systems, but not necessarily to-instrumentation or safety actuation systems),
there are many analytical techniques in use, none perfect.
That is also true of analog systems.
Solid-state systems, whether digital or analog, are also peculiarly vulnerable to environmental damage, e.g., from overheating.
Finally, programmable digital systems have their own special vulnerabilities to human error.
The staff has, concentrated its ' attention on one of these many issues, the vulnerability of digital systems to certain kinds of common-mode
- failures, principally through programming errors introduced into the software, and therefore common to all channels.
To deal with this supposedly special susceptibility to common-mode failure, the staff has proposed a set of regulatory requirements.
The set includes some unarguable items, like the provision of adequate diversity to cope with common-mode failures that can affect safety systems, and analysis of the appropriate accident sequences.
The set also includes some items whose desirability is less clear, and we now turn to these.
Since each of these would require an extensive discussion to-develop the point completely, and since our recommendation is that the staff revisit all these-points, we will be brief.
There is no special order.
The lack of explicit and quantifiable safety standards for instrumentation and control systems is particularly troublesome -
here.
The staff speaks of reliability for digital systems in the same terms (failures per demand) that it uses for items which do wear out, like relays and switches. The entirely different f ailure mechanisms make this an inappropriate transfe/ of terminology.
Indeed, a-simple software-based system, in which the hardware is-kept within its environmental constraints, and whose software is simple enough to have been subjected to a full validation and verification (in the sense used above) can be expected to never fail.
(Never is only a slight exaggeration.)
The failure anecdotes we all know are typically in systems that are too complex for formal VAV, leaving the door open to software errors, or have
!.1 1.
The lionorable Ivan Selin 3
September 16, 1992-been mistreated, opening the door to hardware" failures. The latter problem is not unique to digital systems.
In view of the lack of explicit standards for the-reliability of the digital systems, the staff seems to have drifted to what has been called the " bring me a rock" posture, in which the industry is
~
asked to analyze its own vulnerabilities, after"which-the' staff will make its ruling about the adequacy of the design.
The spirit of the safety-goal initiative was presumably to help make regula-
, tion more predictable, and this approach is clearly in the other direction.
The focus on common-mode failures is troublesome.
Software' errors in single systems can Icad to accidents just as serious as those due to common-mode failures in redundant systems, and the entire-question of software reliability greatly transcends the issues raised here.
We have been conducting a coordinated series of meetings on the safety issues inv.olved in the inevitable computer-ization of the industry,_already in progress.
When we. report on these, we will doubtless raise the question of whether sufficient talent, both in quantity and in experience, is being' directed ~at these issues by Imc.
That question is also an underlying issue here.
For the specific issue of protection against common-mode failures, whether for digital systems or such devices as_ diesel generators, there is a set of standard prophylaxes like diversity and defense _
in depth, which are useful when applied sensibly.
(Slogans-can be overplayed. It makes ne sense to insist that multi-engine aircraft-have a suitable mix of turbine and piston engines.)
The most controversial specific position taken by the staff is that there must be a safety-grade set of. displays and controls located in _ the control room, -independent-of the _ computer systems, and
" conventionally hardwired" to the ' lowest level practicable. Though the intent of the_words in quotations is. unclear, we were assured that it was to require analog backup' systems.
We do not concur in this proposed requirement. We think that the staff is unnecessari-ly mixing up the issues of digital / analog, hard wire / multiplex, and software / hardware.
Each instrumentation %
+rol system that is-important to the-s..
safety of a plant <%
set some identifiable standard - of reliability and fa t
.snce, regardless of the. hard-ware / software basis u :
igning and fabricating the system, It i s -. n o t necessary s
f_given element of the system be m
perfect, but that the,_,.mem as a whole meet some _ recognized standard, presumably in.the form of a relevant surrogate for the Commission's safety goals.
Both the identification of that standard and the evaluation of conformance for the. system: in question pose problems, but each should somehow be - completed
A 4
The lionorable Ivan Selin 4
September 16, 1992 before, not after, a regulatory position is established.
For example, the staf f proposes to require that a' backup system provide protection equivalent to that of the primary system, whereas the need is for suf ficient protection-to assure the adequate safety of the plant.
It is not at all uncommon for backup systems to be designed to lower standards than the primaries; taking into account the fact that they will be called upon le,ss of ten.
(Consider spare tires.)
It is entirely possible that a digital system may turn out to be a
',better backup than an analog system.
(The proposed position does accommodate thic idea, but the staff briefings did not.)
For some situations a light beam is a more reliable means of communication than a hard wire.
A general-purpose microprocessor that is in widespread commercial use may be more reliable (and more thoroughly tested) than a special-purpose analog switch.
And so forth.
In each case it is necessary to make a specific reliability analysis, measured against a reasonable standard, and the staff gave no evidence of having done so for any case.
Instead, it has adopted a general requirement for an analog backup for all cases, and we were not convinced by the justification provided.
We recommend thAt the staff revisit these issues, augment its own capabilities, and broaden its interaction with those elements of the outside world who have previously dealt with such problems.
It would be unwise', however, to read too literally into the nuclear arena the considerations that are relevant to far more complex systems.
We are dealing here with the relatively simple safety-centered parts of the computerized instrumentation and control system, and an architecture that exploits this fact may be more robust.
Sincerely, O
David A. Ward Chairman
References:
1.
Memorandum dated June 25, 1992, from James M. Taylor, Execu-tive Director for Operations, NRC, for The Commissioners,
Subject:
Review of the Draft Commission Paper,
" Design Certification and Licensing Policy Issues Pertaining to Passive and Evolutionary Advanced Light Water Reactor Designs" 2,
57 Federal Recistel, 36680, August 14, 1992, Proposed Generic Communication; Analog-to-Digital Replacements Under the 10 CFR 50.59 Rule 57
/p ratg#e UNITED STATES NUCLEAR REGULATORY COMMISSION n
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- ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g
j
- WA$HWGTON, D. C. 20656 September 16, 1992 Mr. James:M. Taylor Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Taylor:
SUBJECT:
DRAFT COMMISSION
- PAPER,
" DESIGN CERTIFICATION AND LICENSING-POLICY ISSUES PERTAINING TO PASSIVE AND EVOLUTIONARY ADVANCED LIGHT WATER REACTOR DESIGNS" During the 389th meeting of the Advisory Committee on Reactor Safeguards, Septe.aber 10-12, 1992, we reviewed the NRC staff's positions and recommendations concerning the certification issues for evolutionary and passive light water reactor designs contained in the draft commission paper, which was forwarded to the commis-sion on June 25, 1992.
Our Subcommittee on Improved Light-Water Reactors met on September 9, 1992, to review this subject.
During-these meetings we had the benefit of discussions with representa-tives of the NRC staff and EPRI.
We also had the benefit of the document referenced.
We previously provided _ comments to you _ on other policy-issues related to design certification in our letters of May 13, 1992 and August 17, 1992.
Our comments and recommendations on the proposed policy - issues contained in the draft Commission paper are given below.
Issues'A, B, C, D, E, and G apply to evolutionary and passive plant designs and Issues F and H apply only to passive plant designs.
The issue titles and letter designations correspond to those of the draft Commission paper.
A.
Defense Acainst Common-Mode Failures in Dicital-Instrumenta-tion and Control (IGC) Systegg It is our view that the thrust of - the staff recommendations concerning defense against - common-mode fallures in digital I&C systems as underlined in Issue A of the draft commission paper is appropriate.
We agree with the staff that the applicant should.be required.to assess the defense in depth and diversity of the proposed designs for the events postulated in.the Safety Analysis Report, and demonst' rate an acceptable plant response for each.
The staff proposes that the instruments,
- controls, and equipment required to demonstrate an acceptable response be independent of any common-mode failure mechanisms associated with the event.
We:
view this requirement to be essential, but remain open as to the
4 Mr. James M. Taylor 2
September 16, 1992 best approach.
The staff proposes an independent set of safety-grade displays and controls in the main control room.
We believe that other arrangements might be shown to be acceptable.
In a separate letter to chairman selin. dated September 16, 1992, we have provided additional comments and. advice regarding the general approach being taken by the staff in its review of digital instrumentation and control systems.
B.
Analyses of External Events Beyond the Desian Bqgja To assist in the closure of severe accident issues, the staff recommends that (1) analyses submitted in accordance with the requirements of 10 CFR 52.47 (concerning the contents of applica-tions for standard design certification) include an assessment of internal and external events and (2) during the design certifica-tion review, the staff should evaluate those external events that are not site dependent (e.g.,
fires, internal floods) and certain bounding analyses.
We agree with this staff recommendation.
C.
Elimination of the Ooeratina Basis Earthcuake from Seismic Desien The staff is still reviewing this issue and has expressed only an interim position.
We believe the staff is taking an appropriate approach in its interim position.
D.
Multiple Steam Generator Tube Ruotures (MSGTRs)
The staff is recommending that the applicant for design certifica-tion perform additional analyses to determina the AP600 response to multiple breaks of up to 5 steam generator tubes.
We agree with the staff's recommendation, but believe the staff should have a better technical basis for estimating the frequency of occurrence of such multi-tube breaks.
The staff is also recommending that the applicant for design certification of a passive or evolutionary PWR assess design features necessary to mitigate the amount of containment bypass leakage that could result from MSGTRs.
We agree with the statf's recommendation.
E.
Probabilistic disk Assessment (PRA) Beyond Desien Certifica-l tion The staff is recommending that, throughout the duration of the combined or operating license, the PRA be revised to address l
significant plant modifications, operating experience, and c:her developments that may affect previous PRA insights.
q Mr. James M. Taylor 3
September 16, 1992 We are convinced that it is worthwhile for a plant operator to have an up-to-date PRA and are, therefore, reluctant to recommend against this position.
However, if this is to be required, the staff should more clearly specify how it intends to use the up-dated PRA and what is meant by keeping it current.
We think such guidance is part of the overall issue 6f ~ appropriate use of PRAs in regulation and would be helpful to li'censees and to the staff.
F.
Role of the ooerator in a Pagsive Plant Control Room agree with the first part of the staff's position "that We sufficient man-in-the-loop testing and evaluation be performed...
to demonstrate that functions and tasks are integrated properly into the man / machine interface design" of passive ALWR control rcoms.
The second part of the staff's underlined position states "that a fully functional integrated control room prototype is necessary for passive plant control room designs to demonstrate that functions and tasks are integrated properly into the man / machine interface design." We pointed out to the staf f that the non-underlined last sentence of this paragraph is inconsistent with this language in that it would pernit an applicant to " demonstrate that a control room prototype of reduced scope is suf ficient."
We also pointed out that the non-underlined paragraph preceding the underlined paragraph states that such a prototype "would likely" be required (not would be required) to demonstrate that functions and tasks are integrated properly into the man / machine interface design.
We believe that the staff should clarify its intent by reconciling these various statements.
The staff believes that operators of passive plants will be confronted with a new operating philosophy-The staf f argues that "the operators of passive plants must-. understand the operation of
' investment protection' systems and their interfaces with the safety-related passive systems" and that they will be confronted with "new functions and tasks unlike those required for evolution-ary plants" (or current plants)
"due to the new approach in operational philosophy" and "the increase in automation, and the greater use of advanced technology in the passive plant designs."
As a result of our discussions with the staff and EPRI, we believe that the staff may be overreacting to the " newness" of these issues.
It appears to us that additional discussion of this issue among the staff and EPRI and the vendors is needed.
G.
Control Room Annunciator (Alarm) Reliability We agree with the staff's position that the alarm system f or ALWRs should meet the requirements of the EPRI Utility Requirements Document.
I' 4
i Mr. James M. Taylor 4
September 16,- 1992-4 H. --
Reculatory Treatment of NonsafetV Systems We were told that the staff is still engaged in significant on-going discussions and review of this issue and that the-associated position and recommendations are subject to ' modification.
We believe the issue is substantial-and has broad-: implications with respect to such items as use of PRAs in regulation, safety goal-implementation, and reduction of regulatory burdens, and we expect to have additional future interactions with the staff and the
- industry, consequently, we are not prepared to express a position on this issue at this time.
Sincerely, David A. Ward Chairman
Reference:
1.
Draft Commission Paper dated June 25, 1992, from James M.
- Taylor, Executive Director for Operations, NRC,.for the-Commissioners,
Subject:
Review of the Draf t Commission Paper,
" Design certification and--Licensing Policy Issues _ Pertaining to Passive and Evolutionary Advanced Light Water Reactor Designs" i
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'e UNRED STATES 3
~I NUCLEAR REGULATORY COMMISSION I
WASHINGTON, D.C. 20655 o,
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,o October 23, 1992 Mr. David A. Ward, Chairman Advisory Comittee on Reactor Safeguards U.S. Nuclear Regulatory Comission Washington, D.C.
20555
Dear Chairman Ward:
NA
SUBJECT:
DEFENSE AGAINST COMMON HODE FAILURES IN DIGITAL. INSTRUMENTATION AND CONTRGL (I&C) SYSTEMS I am responding to your letter to the Chairman df September 16, 1992, in which you comented on the staff's proposed approach with respect to defense against the common-mode failure of digital I&C systems. This approach was discussed in policy issue "A" of the draft Comission paper, " Design Certification and Licensing Policy Issues Pertaining to Passive and Evolutionary Advanced Light Water Reactor Designs," June 25, 1992.
I also received your letter of September 16, 1992, in which you provided specific coments on policy issue "A."
In the letter to the Chairman, you also raised concerns that apply generally to the staff's proposed generic letter on analog-to-digital replacements. The staff will be presenting this proposed generic letter to the ACRS for its review and consideration, and our dialogue on this topic can continue.
In the interim, this response is intended to address the specific ACRS coments raised on palicy issue "A" by the September 16, 1992, letters.
This' issue of diversity and requisite level of independent backup capability for shutdown purposes is an issue that I feel strongly needs to be addressed in the use of digital protection systems.
In the introduction to the proposed policy issue "A," the staff stated that the two principal factors for defense against comon-mode failures in digital computer systems are quality and diversity. These factors and segregation (or separation), which are needed to provide and maintain independence between redundant ce diverse equipment, were discussed more fully in SECY-91-292 September 16,1991, " Digital Computer Systems for Advanced Light Water Reactors."
As stated in SECY-91-292 and documented in the reviews of advanced light water reactor designs and analog-to-digital conversions at operating reactors, the staff reviewed and accepted several Institute of Electrical and Electronic Engineers (IEEE) standards that govern quality in the development of software.
The staff is also assessing means to improve software quality by conducting research, obtaining assistance of expert consultants, exchanging technical information with other nations, and participating in national and
e-
'e_
4.
v Mr. David A. Ward, Chairman-international software engineering standards comitteos. These activities will provide the information needed to develop regulatory guidance and-acceptance criteria to improve the quality of. computer-based I&C systems.
The staff agrees with the ACRS that quality of. digital computer systems.is of principal importance and that-improving the quality of software will-reduce the potential for a software caused common mode failure. However, based upc:
staff interaction with the international comunity, discussion with experts';
the field of software engineering and our experience with digital control systems, the staff concludes that software quality by itself is not sufficient to reduce the potential for a common mode failure to an acceptable level. As a result the staff concludes that overall control system designs must include diversity in addition to high quality to provide reasonable assurance t1at a common mode software error cannot disable required controls.
The staff recommended four points in assessing diversity and ensuring its adequacy for digital I&C system applications.
The first three points address
- requirements for the applicant to assess the defense in depth and diversity of the proposed designs against comon mode failure vulnerabilities'for events postulated in the -safety analysis report (SAR) and demonstrate an: acceptable-plant response to each event. The staff proposed asia feurth point a set of-safety-grade displays and controls in the main control room. independent of
'the computer systems, for system-level actuation and monitoring:of critical safety functions and parameters for shutdown purposess The. staff did not intend that the independent backup provisions be designed to a reliability equivalent to that of the first-line digital system.
Your specific comments on this issue indicate th'at the ACRS basically agrees with the first three points-of the staff's-proposal; that.is, you consider the recommendations for defense against comon-mode failures-in digital: I&C -
systems to be appropriate and the requirement to' assess _the defense in _ depth and diversity of the systems-to be essential. -However, the ACRS believes that other arrangements might be 'shown' to be. acceptable for.' the fourth point that proposed an independent set of safety-grade displays and controls. Upon considering your specific coments, the ::taff does not believe that the entire -
policy issue "A" needs to be revisited, only its. fourth point.
The staff position on the fourth point has changed as.a result-of the~ comments received from the:ACRS., EPRI, and industry o'n the draft position ~. The. staff would consider allowing more flexibility in implementing the independent set of displays and controls. The flexibility necessary depends on the speci_fic equipment and design features of the I&C _ system and will be evaluated-individually with each vendor.
The intent is to permit the 'use of digital G3
~1.,
l c.-
Mr. David A. Ward, Chairman-equipment that is not affected by the identified common-mode failures and to
^
reduce complexity in the design.
The staff.will not be so inflexible as to-require only analog equipment and will consider allowing simple digital
-equipment.
Safety parameter displays may include dedicated-digital components.
The system-level actuation controls that are " hardwired" to.the lowest level practicable in the I&C architecture may use dedicated and diverse digital equipment.
I believe that this revised approach is consistent with-the comments received from the ACRS.
Sincerely.
/
T or ecutive irector for Operations CC' The Chairman Commissioner Rogers Commissioner Curtiss Commissioner Remick Commissioner de Planque SECY
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ITEM 3: IMPLEMENTATION OF THE PLANT LIFE EXTENSION RULE The Committee has provided five reports to the Commission or the EDO regarding the staff's approach toward the proposed rule on plant license renewal and its -implementation, and the proposed Branch Technical Positions (BTP) on fatigue evaluation procedures and environmental qualification of electrical equipment. In these reports, the Committee has expressed a number of concerns with the major ones based upon the following:
o the definition of " current licensing basis" o coordination with the Maintenance Rule o BTP on metal fatigue evaluation o BTP on environmental qualification of electrical equipment.
The following documents are attached:
- ACRS letter to James M. Taylor (EDO) dated October 22, 1992.
Subject:
Proposed Branch Technical Position on Environmental Qualification of Electrical Equipment for License Renewal (PP. 66-6S)
- Letter for David A. Ward (ACRS) from the EDO dated October 1, 1992.
Subject:
Response to the Augast 17, 1992 ACRS Report (PP. 69-72)
- ACRS letter to James M. Taylor (EDO) dated August 17, 1992.
Subject:
Proposed Regulatory Guide and Interim Standard Review Plan for License Renewal and a Related Branch Technical Position on Fatigue Evaluation Procedures (PP. 73-76)
- ACRS report to the Commission dated April 17, 1991.
Subject:
Draft Final Rule on Nuclear Power Plant License Renewal (PP. 77-78)
- ACRS letter to James M. Taylor (EDO) dated October 11, 1990.
Subject:
Draft Implementation Documents for the Proposed License Renewal Rule (PP. 79-81)
- ACRS report to the Commission dated April 11, 1990.
Subject:
Proposed Rule on Nuclear Power Plant License Renewa3 (PP. 82-83)
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NUCLEAR REGULATORY COMMISSION o-F E
ADVISORY COMMITTEE ON REACTOR SAFdGUARDS h
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WASHINGTON. D. C. 20$$$
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October 22, 1992 Mr. James M. Taylor Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Taylor:
SUBJECT:
PROPOSED BRANCH TECHNICAL POSITION ON ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT FOR LICENSE RENEWAL During the 390th meeting of the Advisory Committee on-Reactor Safeguards, October 8-10, 1992,- we reviewed a proposed Branch-Technical Position (BTP) on Environmental Qualification of Electrical Equipment for License Renewal.
Our Subcommittees on Plant License Renewal and Reliability and Quality reviewed this matter during a joint meeting on September 16, 1992. The staff proposes that the BTP be issued for public comment.
During these meetings, we had the benefit of discussions with members of the NRC staff, its consultants, and representatives of industry.
We also had the benefit of the documents referenced.
Under the License Renewal Rule, 10 CFR Part 54, applicants will be required to develop a comprehensive prograr to identify lin their plants all structures, systems, and components (SSCs) which may be subject to age-related degradation -unique - to the license renewal period.
A' further program to manage these components to ensure-continued safe operation of the plant is also required.
The'. staff is now proposing an additional program, by means~of-a BTP,-which singles out environmental qualification of electrical equipment for:
special_ treatment in the_ license renewal--period.. The particular-concern of the staff seems to be that the. qualification standardsi for
' insulation used on electrical cables prior-to 1984-(representing 87 of 111: licensed nuclear power plant units); may not ensure adequate performance of cables for extended plant life.
That, of course, is-the issue for all SSCs in_a plant, and it is not clear-to us why the more general treatment of SSCs called.for under 10 CFR Part 54 is not adequate' for electrical cables as well.
Industry representatives expressed objection to the staff proposal for a BTP.
They believe ' that while older plant cables. were qualified to a _ lesser standard than has been in use since 1984,
//
L these' cables have been approved for continued use in the plants (as gg
Mr. James H. Taylor 2
October 22, 1992 has much other equipment where standards have evolved) and are part of the Current Licensing Basis (CLB) for each of these plants.
Their interpretation of 10 CFR Part 54 is that the CLB is to be preserved with the exception that those SSCs subject-to age-related degradation unique to the license renewal period should be subjected to specific management programs.
They see no need for the BTP and believe it will result in unnecessary cable replacements and add significantly to plant costs for license renewal.
We are not convinced that the proposed BTP has been shown to be necessary or appropriate.
It should not be issued for public comment until the matters discussed below have been addressed.
Neither the staff nor the industry presented any risk perspective on this issue.
In simple terms, the risk is as follows:
During the license renewal period the electrical cable in a key system might degrade in a way that the degradation would remain undetected during normal operation and by normal maintenance, testing, and surveillance practices.
Then, during an accident, i.e.,
a LOCA, the insulation would fail and the key system would not perform its design function to mitigate effects of the accident.
Present licensing practice assumes, and experience seems to confirm, that the probability of this sequence during the initial license period is acceptably low.
At issue is whether the probability during the license renewal. period is significantly greater.
No evidence has been presented either way.
Analysis of the risk importance of this issue should be made before the BTP is finally accepted or rejected.
Such an analysis should include estimates of downside risks inherent in major projects intended to improve nuclear power plant safety.
Many electrical cables are covered with fire retardant materials.
These coatings could have important effects on the aging of the cable insulation.
Apparently, these effects have not been considered by the staff in development of this BTP.
We do not know whether they have yet been explicitly considered in the selection and evaluation of important SSCs in license renewal programs.
They should be.
Dr.
Thomas Kress did not participate in the Committee's deliberations regarding this matter.
Sincerely, David A. Ward Chairman
Mr. James M. Taylor 3
October 22, 1992 Reference.ji:
1.
Memorandum dated July 10, 1992, from John W. Craig, Office of Nuclear Reactor Regulation,
- NRC, for Raymond F.
Subject:
Request for Review of Branch Technical
- Position on Environmental Qualification of Electrical Equipment for License Renewal, with enclosures 2.
Letter dated October 7, 1992, from H. H. Philips, Jr., and W.
A. Horin, Counsel to the Nuclear Utility Group on Equipment Qualification, to D.
A.
Ward, Advisory Committee on Reactor Safeguards,
Subject:
NRC Staff Proposed License Renewal BTP Regarding Environmental Qualification of Electric Equipment, with enclosures t
47 1
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' ~ %
UNITED STATES.-..
I i-NUCLEAR REGULATORY COMMISSION
'o WASHINGTON, D.C. 20566 y
j October 1, 1992-1 Mr. David A. Ward, Chairman Advisory Committee on Reactor Safeguards OM' S"'
U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Ward:
J
SUBJECT:
PROPOSEDREGULATORYGUIDEANDINTERIN.STANDARDREVIEWPLANFOR-4 LICENSE RENEWAL-AND~A RELATED BRANCH TECHNICAL POSITION ON FATIGUE-EVALUATION PROCEDURES
)
I am responding to your August-17, 1992, letter about the 387th and 388th meetings of the Advisory Committee on Reactor Safeguards (ACRS).
In your
?
letter..you-indicated that the-ACRS did not object to the staff publishing the documents so'pporting license renewal for public comment, but gave three-comments. The following paragraphs below provide the staff's response to the-ACR$' concerns.
7 The ACRS commented that the process for selecting systems, structures,.and components (SSCs) important to license renewal should have a better control mechanism.
In response to public comments that_the draft version of'10 CFR Part 54 required too much judgment in determi_ning-whether SSCs were important -
to license-renewal, the Commission-revised the definition of what. constitutes SSCs important.to license renewal in the final Part 54 rule. The revised' definition effectively places more controls on the process of identifying ~ SSCs
.important to license renewal.- -Additionally,.the staff expects. to gain.
experience with the. implementation of the rule during the lead plant applica-tion review. Until then, the staff will review applications for license-renewal using the definition in the rules ascit was explained in the Statementi of Consideration.- If changes are necessary, the staff will seek the ACRS views on.all proposed changes to the rule.
The ACRS commented that somel consideration should be given to combining the-
- maintenance and license renewal rules and recommended completing a compre-
.hensive. study before. implementing the maintenance and license renewal rules to increase regulatory coherence and reduce -regulatory burden. The staff _does CONTACT:
Francis M. Akstulewicz,.NRR 504-1136
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Mr. David A. Ward recognize that the license renewal rule and the maintenance rule share, in a general sense, the similar purpose of managing age-related degradation.
However, the staff does not believe a compreh,ensive study is necessary at this time. The license renewal rule can be implemented in a manner that is compatible with the maintenance rule, if the' licensee chooses that option.
The staff has stated that the aging management actions. carried out to meet the maintenance rule could develop the type of in' formation that could be used to assess the technical feasibility and resource requirernents for renewing an operating license.
For example, if a licensee identifies age-related degrada-tion mechanisms as part of the implementation of the maintenance rule and, if necessary, improves the effectiveness of its maintenance activith i to control the identified degradation, that licensee will be in a much stronger position to demonstrate the effectiveness of its aging management practices in a license renewal application.
Therefore, the staff believes that implementa-tion of the maintenance rule presents an opportunity for licensees to lay r.
substantial portion of the ground work for a license renewal application in a fashion that minimizes the potential for duplication of effort.
As Commis-sioner Curtiss noted in his presentation at the recent Regulatory Information Conference, the. key is to take advantage of the flexibility afforded by the maintenance rule and to adapt or implement maintenance activities to the more prescriptive license renewal framework.
The ACRS commented on how PRA would be used in the implementation of the two rules. The scope of each rule is determined by specific definitions contained within each rule. These definitions do not depend upon nor include the use of PRA. However, licensees can use PRA insights as one factor in their judgments on the scope of SSC for both the maintenance rule and the license renewal rule.
The ACRS commented that the branch technical position (BTP) on metal fatigue appears to require more of licensees than is justified. Specifically, the ACRS was concerned that calculating the cumulative usage factor (CUF) would be-time-consuming and that using it in the way the BTP suggests would require replacement of components that would exceed only one-third of the American Society of Mechanical Engineers (ASME)Section III design life. The ACRS proposed using the procedures of Section XI of the ASME Code to inspect and repair SSCs instead of implementing the design requirements in ASME Section III during the renewal term.
The staff acknowledges that the calculation of the CUF will be time-consuming for some older plants, however, the staff believes that use of the CUF is necessary to determine whether or not components initially designed for a 40-year plant life will retain sufficient fatigue resistance during 60 years of plant operation. As the staff noted during its presentation, the licensing basis for metal fatigue consists of the calculated design life, transient and cycle monitoring, and inspection.
The staff does not believe that the 70
1 1
Mr. David A. Ward licensing basis in the area of fatigue can be maintained by relying solely on ASME Section XI inspections, as proposed by the industry.- An approach which relies exclusively on limited inspections, which may not detect a flaw before it reaches critical size, would not assure that the licensing bases is maintained during the renewal period. Maintaining the licensing bases is one of the two fundamental principles established by the Commission as the bases for license renewal.
The staff notes the ACRS concern that the BTP requires component replacement at one-third of a component's design life.
This is not a correct interpreta-tion of the BTP because component monitoring, inspection, and repair are included as options under the BTP under specific conditions. The staff notes that figure F.1.0-1 of the BTP shows that augmented inspection and transient tracking are acceptable options for monitoring and managing fatigue resistance when the CUF exceeds 0.4 for identified components.
The staff BTP also mentions component repair as an acceptable means of managing fatigue degrada-tion in lieu of component replacement.
Therefore, the staff BTP proposes no unreasonable action that would require replacement of satisfactory piping or components. The staff notes that the BTP value of 0.4 comes from ASME Section XI for inspecting pipe welds.
However, this BTP applies to both welds and..
base metal. As part of the public comment process, the staff will be request' ing specific comments on the use of the CUF value for both welds and base metals.
At the briefing, the ACRS expressed a concern about the use of the environ-mental factor as part of the process for establishing the design life of a component for license renewal.
The staff has reviewed existing data and is aware of ongoing confirmatory research which suggests that the environmental factor should be included in calculations that estimate the remaining fatigue resistance of components for a renewal term.
The staff's intent in publishing the BTP is to obtain review and comment on the technical merits of the BTP and to stimulate the technical debate on the process for estimating remaining design life of components important to license renewal. Additionally, the staff is participating in various Code groups to keep abreast of their activities in this area. The staff will review the public comments received and any new test or research data to determine what changes to the BTP are necessary before developing a final techni_ cal position on fatigue. However, should the staff receive a renewal application before a final. position is published, the staff believes that the approach presented in the BTP is reasonable and will provide assurance that fatigue degradation will be adequately managed for any period of extended operation.
The ACRS also suggested that the staff request clarification from the ASME on what should be done when the CUF of a component approaches 1.0.
The staff understands that ASME is aware of the issue and will be providing comments.
The staff will consider comments from the ASME along with other public comments received during the public comment period when it is developing the final version of the BTP.
Mr. David A. Ward The staff plans to issue the regulatory guide and standard review plan for license renewal for public comment and will reconsider these ACRS coments along with the public coments received during the comment period.
Upon establishing final positions, the staff will again present them to the ACRS for eview.
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. Td, or mes xecutive Director for Operations cc:
The Chairman g
Commissioner Rogers Commissioner Curtiss Commissioner Remick Commissioner de' Planque SECY Mike Case 7A
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August 17, 1992 g
Mr. James M. Taylor Executive Director for Operation U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Ta"'Jr:
4
SUBJECT:
PROPOSED REGULATORY GUIDE AND INTERIM STANDARD REVIEW PLA:1 FOR LICCNSE RENEWAL AND A RELATED BRANCH TECIUiICAL "91 TION ON FATIGUE EVALUATION PROCEDURES During the 387th and 388th meetings of the Advisory Committee on Reactor Safeguards, July 9-11 and August 6-8, 1992, we reviewed a proposed Regul'atory Guide (DG-1024) and an interim Standard Review Plan (SRP) (NUREG-1299) to be used in plant license renewal.
We also considered a proposed Branch Technical Position (BTP)
(an appendix to the SRP) on fatigue evaluation procedures which would provide a basis for license renewal reviews.
These matters were also considered udring a joint meeting of our Subcommittees on Plant License Renewal and Materials and Metallurgy on July 7, 1992.
During these meetings, we had the benefit of presentations by the NRC sta.*f, its consultants, and representatives of industry.
We-also had the benefit of the documents referenced.
We commented on the earlier version of the Regulatory Guide and on the interin SRP in our report of October 11, 1990, and on an early version of the License Renewal Rule in our report of April 11, 1990.
Since these reports were issued, the NRC staff has issued a final rule, which incorporates some significant changes from the earlier version, and has also received and evaluated public comments on the 1990 Regulatory Guide and the SRP proposals.
The now-proposed Regulatory Guide and SRP are intended to reflect the fina.1 rule, the public comments, and ACRS comments.
Because there
?
have been significant changes in these documents, the statf proposes to publish them for another round of public comments.
In addition, in its development of the license renewal process, the staff has identified a concern about the adequacy for extended service of some existing plant componentr. in accommodating metal fatigue.
Generally, plants were designed with a 40-year 11fe expectation.
Operation for 60 years could mean that some fatique limits would be exceeded.
A BTP has been proposed as a basis for staf f evaluation of f atigue status in the license renewal process.
73
Mr. James M. Taylor 2
August 17, 1992 In general, the new versions of the Regulatory Guide and SRP seem to be improvements and appropriately reflect changes made in the rule.
W. ' have no objection to their being published for public comment.
We have, however, three comments at this time:
1.
The major concern expressed in our October 11, 1990 report was about control of the process for selection of Structures, Systems and Components (SSCs) "important to license renewal."
Without adequate constraintr., reviewers are likely to expand the list of SSCs beyond that needed to provide reasonable assurance that aging of important plant systems will be adequately controlled.
This could be carried to a point of being unnecessarily burdensome on licensees, thereby discouraging plant owners from seeking license renewals.
We were told that this concern was being addressed by improving the definition of SScs.
We have reservations about whether this is adequate and retain our concern.
We were told that estimates by management of the lead plant involved in the license renewal effort indicate that 65 percent of the components in the plant will be on the SSC list and that the cost of developing the required program could be as much as $25 million.
The length of the SSC list is obviously a substantive issue.
We believe some better mechanism for control should be established.
Creation of a review function for each plant's SSC list at a senior level within the agency, perhaps something similar to the CRGR, should be considered.
2.
Requirements imposed on licensees with the Maintenance Rule have much commonality with requirements under the License Renewal Rule.
We asked the staff presenters whether consideration had been given to combining or at least coordinating the two rules.
Apparently none has been.
We believe this is a mistake.
Requirements for the two rules have been developed in different branches of the Office of Nuclear Reactor Regulation. The two sets of requirements have somewhat, but not greatly, different scopes and purposes.
We were told that a licensee who decides to apply for license renewal and moots the requirements of the rule will not automatically meet the Maintenance Rule because of some scope differences.
Nor will the opposite be true.
Thus, a licensee will have to meet both sets of requirements over the life of a plant.
One interesting difference in the proposed implementation of these two rules is in how PRA would be used.
With the Maintenance Rule, risk arguments will be accepted by the staf f for either excluding or including specific items.
In the case of che License Renewal Rule, we were told that risk arguments 7
Mr. James H. Taylor 3
August 17, 1992 would be accepted by the staff for inclusion but not for exclusion of specific items in the plant from the SSC list.
We note that a
reliability assurance program is being developed for the ABWR and understand that such programs will be required for all ALWRs.
These programs should be coordinated with the requirements of the License Renewal and Maintenance Rules.
We have often commented on the need for greater enherence among the many parts of the overall fabric of NRC regulatory policios and practices. The Commission has recently spoken on the need to reduce unnecessary regulatory burden on licensees.
We recommend that before the Maintenance Rule and License Renewal Rule are implemented, a comprehensive study be carried out to determine if combining the rules would foster the aims of increased regulatory coherence and reduced regulatory burden.
3.
The BTP on metal fatigue appears to require more of licensees than is justified.
The BTP would require evaluation of the fatigue life (cumulative usage
- factor, CUF) in certain components of the reactor coolant pressure boundary based on 60 years of service.
The actual transient history could be used, resulting in a lower CUF than that assumed in the original design, but it would require the replacement of any component which exceeded one-third of the ASME Section III design life.
Calculating the CUF can be time-consuming, and using it in the way suggested by the BTp will usually require the replacement of components which would otherwise perform satisfactorily for the remaining life of the plant.
The industry position is that the calculated CUF should not be regarded as an absolute service limit.
Industry spokesmen suggested that equally appropriate and more economical approaches are available and should be used.
We believe the staff proposal is unreasonable.
A better approach would be to use the procedures of ASME Section XI for inspection and repair.
Consideration might be given to requesting a
clarification from the appropriate ASME Code Committee on what it believes should be done if CUF approaches 1.0 for a component.
This would be time-consuming, but time does seem to be available in this instance.
1
t Mr. James M. Taylor 4
August 17, 1992 Additional comments by ACRS Members William Kerr, Thomas S. Kress, Harold W.
Lewis, and Charles J. Wylic are presented below.
Sincerely, W
y
-yv David A. Ward chairman Additional comments by ACRS Members William Kerr. Thomas S. Kress, Harold W.
Lewis, and Charles J.
Wylie.
We do not wish to address the details of the elaborate regulatory structure the staff proposes to erect to support its review of licent.e extension applications for currently operating nuclear plants, but only to provide some perspective.
We are impressed by the contrast between the licensing of nuclear plants and of other complex systems, like aircraft.
Both nuclear plants and aircraft are complex systems, each synthesized from componenta of a wide variety of effective lifetimes, rates and modes of degradation, and importance to the safety of the system.
In_the nuclear case, a term license is issued for the system-in the aviation case the aircraft airvorthiness certificate is permanent, provided the maintenance and replacement of the aging components are managed in a timely and effective manner.
That seems to us to be far more effective, and is consistent with the Committee's recommendation to coordinate this plan with the Maintenance Rule.
The purpose of-licensing is to ensure and maintain the protection of the public health and safety-it is not an end in itself.
Of course we recognize that initiatives in
'.his matter are constrained by the tarms of the Atomic Energy Act, but'it is not unthinkable that laws can be adjusted if it is in the public interest to do so.
This was not an important matter forty years ago-it is now.
Refere'Agg:
i.
'acwrandum dated June 10, 1992, from John W. Craig, Office of Kurlear Reactor Regulation, NRC, -for Raymond F.
- Fraley, A l'Hory Committee on Reactor Safeguards,
Subject:
Request ft.,1 Review of Branch Technical Position on Fatigue for License.
Renewal, with enclosures 2.
Memorandum dated June 5, 1992, from John W. Craig,-Office of Nuclear - Reactor Regulation, NPr for Raymond F.
- Fraley, Advisory Committee on Reactor ~ s' Squards,
Subject:
Request for Review of Interim Regulatory Guide and Standard Review Plan for License' Renewal, with enclosures
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April 17, 1991 The Honorable Kenneth H. Carr Chairman U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Chairman Carr:
SUBJECT:
DRAFT TI!U1 RULE Oli NUCLEAR POWER PLANT LICENSE RENEWAL During the 372nd meeting of the Advisory Committee on Reactor Safeguards, April 11-13, 1991, we reviewed the draft of the final rule on nuclear power plant license renewal (10 CFR Part 54).
Our Subcommittee on Plant License Renewal discussed this matter during its April 8,
1991 meeting.
During our consideration of this matter, we hpd the benefit of discussions with representatives of the NRC staff, NUMARC, and Northern States Power Company.
The latter is the licensee for the Monticello Nuclear Gene.cating Plant, which is a lead plant in the license renewal program.
We also had the benefit of the document referenced.
The ACRS reported to you on the proposed license renewal rule in its report of April 11, 1990.
Since that time, the proposed rule was published for public comment.
The staff received 197 comments.
It has assimilated information from these comments and information received in a number of interactions with industry and has prepared a draft final rule.
The schedule calls for the final rule to be published by June 28, 1991, and for other parts of the rulemaking package, a regulatory guide and a standard review plan, to be oublished about one year later.
As stated in our April 11, 1990 report, we concur with the approach being taken by the staff in this rulemaking.
However, there are two areas of disagreement between the staff and NUMARC that we would like to bring to your attention.
The first might require a modification in the draft final rule.
The second is related to implementation of the rule.
The first matter is an issue on which we do not have a recommenda-tion except that it should receive your consideration.
The draft final rule requires that each applicant for license renewal develop a " compilation" of its current Licensing Basis.
Although it is not precisely clear what this means, it was agreed that it would, at a minimum, include a list of all licensing commitments agreed-to-by the applicant over the history of its plant.
Industry represen-tatives believe this is unnecessary.
77
The lionorable Kenneth H. Carr 2
April 17, 1991 The second issue in how implementation of the rule will be limited in scope to concentrate resources for aging management where needed.
The rule would require that each applicant develop a list of Systems, Structures, and Components Important to License Renewal (SSCITLR) and then implement an aging management program ap-propriate for items on that list.
The staff's position is that the original SSCITLP list should include all those items in the plant that play a role in meeting any docketed commitment the licensee has made.
This would include the original license; commitments related to itew rules as they came into being; and commitments made in response to Safety Evaluation
- Reports, Information Notices, Dult MS
%.9ric Letters, and Orders.
The industry representat.c;D 4+
vt. hat such a definition of SSCITLR would result in a ( ( vit Att includes 65 to 90 percent of all equipment in the plant.
may believo that application of a special aging program to all of th3se items would be unnecessary and onerous.
The process of reducing the initial SSCITLR list to just those items to be covered by a special aging program is critical.
Items important to impicment other commitments would not thereby be ignored.
They would be maintained through the new license period just as they are now.
We believe that selection of those items to be subjected to a special aging program should be based on technical rather than legal argument, our understanding is that a program of this nature can be developed with the rule as presently draf ted.
- Ilowever, implementation will require careful crafting of the regulatory guido and the standard review plan.
We would like the opportunity to-review these documents before they are issued.
Sincerely, David A. Ward Chairman
Reference:
Memorandum dated March 6,
1991 from Warren Minners, Office of Nuclear Regulatory Research, to Raymond F.
Fraley, ACRS,
Subject:
Final Rule on Nuclec Power Plant License Renewal, with enclosures (Predecisional) 78
p ** en o,
UNITED STATES 3,
NUCLEAR REGULATORY COMMISSION e
2-ADVISORY COMMITTEE ON RE ACTOR SAFEGUARDS W ASHINGTON, D. C. 205$5
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October 11, 1990 Mr. James M. Taylor Executive Director for Operations U.S. Nuclear Regulatory Commission '
Washington, D.C.
20555 Dear Mr. Taylor
SUBJECT:
DRAFT IMPLEMEffTATION DOCUMENTS FOR THE PROPOSED LICENSE RENEWAL RULE During the 366th meeting of the Advisory Committee on Reactor Safeguards, October 4-6, 1990, we reviewed draft Regulatory Guide, Task DG-1009, " Standard Format and Content of Technical Information for Applications to Renew Nuclear Power Plant Operating Licenses,"
and associated draft NUREG-1299, " Standard Review plan - License Renewal."
Our Subcommittee on Plant License Renewal also reviewed this matter during its meeting on October 2, 1990.
During this review, we had the benefit of discussions with representatives of the NRC staff and of the documents referenced.
These documents are an important part of the program to implement the proposed license renewal rule, 10 CTR Part 54, that was published for public comment on July 17, 1990.
We commented to the Commission on this proposed rule in our report of April 11, 1990.
We believe that the general approach proposed by the staff for implementation of the license renewal process is reasonable, and we agree that both of the subject documents should be published at this time for public comment,
- llowever, we have a concern, discussed below, about control of the process for selecting structures and components important to license renewals (SCITLRs).
We believe that this matter should be considered further as public comments on the rulemaking are evaluated.
We also offer several comuents on the implementing documents.
'lhere is justification for the general philosophy of the proposed license renewal rule.
Aging-degradation issues should be dealt with by more explicit programs as the plant age passes beyond the general target age for which it was designed, our understanding.
is that a 40-year operating life _ has been used for most structures and components'in nuclear power plants.
However, that target age and the design were not so precisely defined that there should be a step increase in licensing requirements as the plant passes its 40th anniversary of operation.
As we said in our April 11, 1990 report, "no specific form of plant aging lwcomes magically decisive at forty."
We have a concern that th<>.
license renewal process under the proposed 10 CFR Part 54 wi il permit or encourage a 77
l Mr. James H. Taylor 2
October 11, 1990 significant expansion of regulatory requirements as a plant phases into operation under a renewed license.
We had hoped and expected that the implementing documents would provide some clear indica-tions of how such regulatory expansion would be constrained.
They do not.
Introductory material in the proposed 10 CFR Part 54 indicated that the backfit rule would somehow be used in control-ling the extent to which regulatory requirements would be expanded.
However, the rule itself does not make it clear how this is to be done, nor do the draft implementing documents.
We recommend that the rule or the implementing documents be revised to ensure that the process for selecting SCITLRs and developing new requirements is sufficiently disciplined.
In addition, we have several specific comments on the proposed implementing documents:
(1)
In the proposed process for evaluating age-related degrada-tion, the draft Regulatory Guide indicates that a decision about classification of a given structure or component should be made on the basis of whether the structure or component is routinely replaced or refurbished (see Block 12 of Figure 1B in the draft Regulatory Guide).
We recommend that satisfac-tory results of inspection or monitoring should also be credited at this decision point.
(2)
Many of the unresolved safety issues and generic safety issues that have been analyzed over the past several years have had assumptions about expected plant life factored into their resolution.
The staff has indicated that, in general, an expected life of 60 years instead of 40 years would make little difference in cost-benefit analyses, given the large uncertainty inherent in the calculated results.
However, the staff also indicated that a review of all such resolutiens will be made, in the linhr of new expectations about plant lifetimes, given the chan,es of 10 CFR Part 54.
We would like to be kept informed about the results of this review.
(3)
Certain industry topical reports on the _ subject of aging degradation are being developed by NUMARC, and are expected to be approved by the staff as acceptable references in license renewal applications.
We encourage the development of these industry reports as a means of providing a comprehen-nive technical base for license renewal reviews.
Because the license renewal process can be expected to extend over many years, much technical information about aging will be in need of revision, and some means for formally updating these industry reports and their approval by the NRC should be provided.
(4)
Perspectives gained from applicable risk assessment should be used in the selection of SCITLRs.
Mr. James H. Taylor 3
October 11, 1990 (5)
Consideration should be given to lacluding physical security systems in the SCITLR program.
We plan to continue our review of this important subject after public comments on this proposed rul.c, the Regulatory Guide, and the proposed Standard Review Plan are received and assimilated.
Sincerely, Carlyle Michelson y,
Chairman
References:
1.
U.S.
11uclear Regulatory Commission, Office of fluclear Regulatory Research, Draft Regulatory Guide, Task DG-1009,
" Standard Format and Content of Technical Information for Applications to Renew fluclear Power Plant Operating Licenses,"
Revision 5A dated August 1990, and U.S.
11uclear Regulatory Commission, Of fice of fluclear Reactor Regulation, Draf t NUREG-1299, " Standard Review Plan, License Renewal," dated August 1990, transmitted by memorandum dated August 31, 1990, from Eric S.
Beckjord, RES, and Thomas E. Hurley, NRR, to Raymond F.
Fraley, ACRS 2.
U.S.11uclear Regulatory Commission, Rules and Regulations,10 CFR Part 54, " Requirements for Renewal of Operating Licenses for liuclear Power Plants," Proposed Rule Making, Published July 17, 1990 8
[pa ren UNITED SYATES
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NUCLEAR REGULATORY COMMISSION n
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS I
O WASHINGTON D, C,70666 s*..*/
g April 11, 1990 The Honorable Kenneth M. Carr Chairman U.S. Nuclear Regulatory Commission i
Washington, D.C.
20555 Dear Chairman Carrt
SUBJECT:
PROPOSED RULE ON NUCLEAR POWER PIANT LICENSE RENEWAL During its 360th meeting, April 5-7, 1990, the Advisory Committee on Reactor Safeguards reviewed the staff's proposed rule on nuclear plant license renewal.
This matter was also discussed during a meeting of the Regu'* tory Policies and Practices Subcommittee on March 26, 199Q.
I,.
Ang these meetings we had the benefit of discussions wit re; -esentatives of the NRC staf f, NUMARC, EPRI, Northern States er Company, and Yankee Atomic Electric Company.
He also had th benefit of the referenced document and its enclosures.
The decisive issues for license renewal and assOiated plant aging, and the potential for further aging during the. proposed license extension, should be addressed throughout the life of a plant.
Attention to aging phenomena, and the criteria for safe operation (adequate protection of the health and sat'ety of the _ public),
should be the same just after as just before license renewal.
There may be components or systems which are not aging issues during the first forty years, but become so later, and which therefore may require special attention.,
At the time that the forty year period for a license was chosen, there was no special technical rationale for its choice, and no specific form of plant aging becomes magically decisive at forty.
The regulatory job for license renewal-is to identify the aging elements of the plant, and ensure that they receive timely attention during the extended' license period.
In that context, we were surprised by the lack of emphasis on pressure vessel integrity during our - viefings. This is surely one of the driving technical-issues for extended life, and we assume that it will move to a more central position as the plans develop.
The Honorable Kenneth H. carr 2
April 11, 1990 The staff proposes to use the " current licensing basis" of a plant as the basis for license renewal, but there seems to be some ambiguity about the interpretation of the tern. The industry seems concerned that this may provide an opportunity to impose arbitrary new requirements.
It is important 'that this ' terminology be clarified, so that any future conflicts of interpretation are minimized.
With these observations, we concur in the approach being proposed by the staff, which emphasizes attention to aging phenomena, avoids the temptation to treat license extensions as relicensing, and makes a timely start toward providing an integrated policy for dealing with aging phenomena.
I Sincerely, Carlyle Michelson chairman
Reference:
Memorandum dated March 6,
1990 from Warren Minners, Office of Nuclear Regulatory Resoarch, NRC, to Raymond F.
Fraley, ACRS,
Subject:
Proposed Rule on Nuclear Power Plant License Renewal, w/ enclosures Draft Commission Paper, " Proposed Rule on Nuclear Power Plant License Renewal" 83
ITEM 4: RISK-BASED REGUIATION During the November 1992 Meeting, the C'ommittee discussed various aspects of risk-based regulations with representatives of the staf f and industry. Both efforts by the staff and MIMARC had just gotten underwry. The staff provided the following definition for the approach to risk-based regulation:
"an approach to regulation where quantitative insights derived from a probabilistic risk assessment are used to focus utility and regulatory attention on design and operational iseues commensurate with their impact on risk to the public".
Following a metating with industry on March 10, 1992, the Commission issued a SRM dated March 26, 1992, in which they requested that the staff provide their views on the practicality of risk based regulations and the feasibility of developing a transition strategy from deterministic _ based regulations. The Committee discussed a draft Commission paper with the staff during the November meeting but chose not to comment on the details in the paper as chey were in need of much further development.
The following documents are attached:
- ACRS report to the Commission dated November 16, 1992.
Subject:
Risk-Based Regulation (PP. 85-86)
- SAM dated March 26, 1992.
Subject:
Briefing on Risk-Based Regulations Transition Strategy, March 10, 1992 (PP. 87)
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NUCLEAR REGULATORY COMMISSION n
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W ASHINGT ON, D. C. 20666 November 16,T992 i
The Honorable Ivan Solin Chairman I
U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Chairman selin:
SUBJECTt RISK-BASED REGULATION i
During the 391st meeting of the Advisory committee on Reactor Safeguards, November 5-7, 1992, we reviewed a draf t Commission paper on Risk-Based Regulation..
The paper responds to the Staff Requirements Memorandum (SRM) dated March 26, 1992.
During this meeting, we had the benefit of discussions with representatives of the NRC staff, and of the document referenced.
We interpret the Commission's chargo to the staff as reflecting a recognition of the increasingly sophisticated and widespread use of analytical risk assessment techniques in the nuclear enterprise, a natural evolution of a process that began with the 1975 publication' of the Reactor Safety Study, WASH-1400.
Since it is now possible to make informed and quantitative statements about' many (but not all) of the contributors to nuclear risk, it is correspondingly ~
possible to optimize the deployment and 'use of the regulatory resources available to the Commission.
The SRM directed the staff to both examine the feasibility of such a risk-based approach to regulation and to suggest means by_which it could_.be-implemented.
The draft paper _on which we were briefed is - the preliminary response to that charge.
We would prefer not to comment in rdetail on the paper itself,.
except'to note that it needs a' great deal of work before it can be
' t considered responsive.to the Commission's charge at the level-of sophistication demanded by the importance of-the question.-
The--
staff-is still working on the paper, and we expect to see a later-and improved version.
It is simply-not 'yet ready for public.
. 1 comment.
-Far more important to us-is_the issue of coherence of the variousi efforts now-_in progress _in various parts of the staff to develop and implement activities that could be collected under the name of risk-based regulation.
We have commented earlier about the Maintenance Rule, Regulations Marginal to Safety, and other.
25~
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The Honorable Ivan Selin 2
November 16, 1992 initiatives involving the use of risk ana' lysis, and ve at this meeting heard about Risk-Based Regulation, revision of the Regulatory Analysis Guidelines, and the Prioritization of Generic f
Safety Issues.
Each of these requires informed use of quantitative risk information and appropriate attention to the Commission's safety goals, yet each is being analyzed by an independent group, with an independent perspective on the'NRC's needs.
In addition to this, there is the PRA Working Group,' whose progress we_have been following closely.
We are unable to, find any focal point for all these efforts, except at the level of the EDO.
We continue to call for increased coherence in the treatment of all tha% matters, bound to each other by the common need to weave the threads of the safety goals (the expression of the ultimate objective of regulation) and quantitative risk assessment (the tool that makes more directed risk management possible) into the NRC fabric.
If it is not done at the level of the EDO it will not be done, and resources that could be devoted to assuring nuclear safety will be squandered.
In the past we have suggested strong measures to address this problem.
While not-pushing any particular solution, we still believe that the collection of issues discussed here is important to the future performance of the agency.
The coherence problemo will not be solved by an incoherent effort.
Sincerely, kb+
Paul Shewson Chairman R e f e r e q q fq :
Memorandum dated October 16, 1992,'from Warren Minners,' Office of Nuclear Regulatory Research, NRC, for Raymond F.
Fraley, ACRS, transmitting Draft SECY Paper (undated) from James M.-Taylor, Executive Director for Operations, for The Commissioners, subject:
Risk-Based Regulation (Predecisional) 74 1
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March 26, 1992 IN RESPONSE, PLEASE REFER TO:
M9203100 of teC8 0F THE llCA8TARY MEMORANDUM FOR:
James M. Taylor.'
ExecutiveDirectorforof4 rations FRON:
Samuel J. Chilk,' Secreta {
SUBJECT STAFF REQUIREMENTS = BRI CI HG ON RISK-BASED REGULATIONS TRANSITION S' 6TEGY, 2:00 P.M.,
TUESDAY, MARCH 10, 1992, COMMISSIONERS' CONFERENCE ROOM, ONE WHITE FLINT NORTH, ROCKVILLE, MARYLAND (OPEN To PUBLIC ATTENDANCE)
The commission was briefed on a transition strategy to the use or risk-based regulations.
The information was presented by:
Mr. John C.
Brons President and Chief operating of ficer How York Power Authority (NYPA)
Mr. Herschel Spector New York Power Authority Mr. William H. hasin Director, Technical Division Nuclear Management and Resources Council (NUMARC)
The Commission requested that the staff provide their views on the practicality of risk-based requistions and the feasibility of developing a transition strategy from deterministic based regulations.
Consideration should be given to the need for a threshold which ensures some attention to low-priority items.
(EDO)
(SECY Suspense:
12/18/92) cc:
The Chairman Commissioner Rogers Commissioner-Curtiss Commissioner Remick Commissioner de Planque OGC OCAA O!G ACRS fDR - Advanco VDCS - P1-24 i-9203300L;3 920326 0
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