ML042010350
| ML042010350 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 07/14/2004 |
| From: | Silberman R Neal R. Gross & Co. |
| To: | Office of Nuclear Reactor Regulation |
| Byrdsong A T | |
| References | |
| 50-413-0LA, 50-414-OLA, ASLBP 03-815-03-OLA, NRC-1588, RAS 8157 | |
| Download: ML042010350 (261) | |
Text
RAS: _F1571 Official Transcript of Proceedings; NUCLEAR- REGULATORY COMMISSION
&6 er'. .. opr ion
Title:
Docket Number: 50-412/414-OLA; ASLBP No.: 03-815-03-OLA Location: Rockville, Maryland DOCKETED USNRC July 16, 2004 (3:50PM)
Date: Wednesday, July 14, 2004 OFFICE OF SECRETARY RULEMAKINGS AND ADJUDICATIONS STAFF Work Order No.: NRC-1 588 Pages 2072-2330 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.
Washington, D.C. 20005 (202) 234-4433 ji enpIAtc - Se CY-O3,L :5ecy-:Oa_
2072 I UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
+ . . . .
ATOMIC SAFETY AND LICENSING BOARD (ASLB)
HEARING
.1 1l In the Matter of: Docket Nos. 50-412-OLA 50-414-OLA DUKE ENERGY CORPORATION ASLBP No.03-815-03-OLA Catawba Nuclear Station, Units 1 and 2 II II
-1 Wednesday, July 14, 2004 The above-entitled matter came on for hearing, pursuant to notice, at 1:00 p.m.
Before Administrative Judges:
( ANN MARSHALL YOUNG, Chair ANTHONY J. BARATTA NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE.. N.W.
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2073 APPEARANCES:
On Behalf of the Licensee:
DAVID A. REPKA, ESQ.;
ANNE W. COTTINGHAM, ESQ.; and TAMIKA SHAFEEK-HORTON, ESQ.
of: Winston & Strawn, L.L.P.
1400 L Street, N.W.
Washington, D.C. 20005-23502 (202) 371-5726 DR (202) 371-5950 fax On Behalf of the Petitioner:
DIANE CURRAN, ESQ.
I/
of: Harmon, Curran, Spielberg & Eisenberg, LLP Suite 600 1726 M Street, N.W.
Washington, D.C. 20036 (202) 328-3500 STAFF ATTORNEYS PRESENT:
MARGARET BUPP ANTONIO FERNANDEZ MARVIN ITZKOWITZ SUSAN UTTAL NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2074 I-N-D-E-X WITNESS DIRECT EXAMINATION STEVE NESBIT, 25 ROBERT HARVEY, BERT DUNN, and J. KEVIN McCOY EDWIN S. LYMAN 45 RALPH MEYER, 57 UNDINE SHOOP, and RALPH LANDRY EXHIBIT MARKED RECEIVED Numbers 1 through 24 38 38 Numbers 25 through 36 56 56 Numbers 37 through 46 73 73 Numbers 47 through 50 73 73 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2075 1 P-R-O-C-E-E-D-I-N-G-S 2 (1:05 p.m.)
3 CHAIRPERSON YOUNG: As I was saying, I am 4 Ann Marshall Young. I am the Chair of this Licensing 5 Board. To my left is Judge Anthony Baratta. As I 6 indicated in an e-mail to the parties this morning, 7 Judge Elleman has been delayed because of airline 8 delays and will not be with us today. He will be with 9 us tomorrow morning.
10 And unless there's any strong reason not 11 to, we're going to start tomorrow morning at 8:00 12 o'clock so that we can get as much done tomorrow as 13 possible.
14 Before we go into what we're going to do 15 today, let me ask. I will start with the staff. If 16 you could introduce yourself, Ms. Uttal, and then 17 everyone who is with you? And then we'll move to Duke 18 and then to BREDL.
19 MS. UTTAL: I'm Susan Uttal, staff 20 counsel. To my immediate right is Antonio Fernandez, 21 staff counsel; Margaret Bupp, staff counsel. And 22 Marvin Itzkowitz is our supervisor. And behind me is 23 Dr. Ralph Meyer, who is one of the staff witnesses.
24 CHAIRPERSON YOUNG: Mr. Repka?
25 MR. REPKA: I'm David Repka, counsel for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE.. N.W.
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2076 1 Duke Energy. And on my immediate left is Mr. Steve 2 Nesbit of Duke Energy. He is the Duke Energy MOX fuel 3 project manager.
4 Behind me, behind the partition, are my 5 colleagues in this case: Ms. Anne Cottingham of 6 Winston and Strawn and Ms. Tamika Shafeek-Horton of 7 Duke Energy.
8 The rest of the Duke witness panel is 9 behind me. I could introduce them now or wait until 10 they take the stand, whatever your --
11 CHAIRPERSON YOUNG: Why don' t you do both?
12 MR. REPKA: Okay. Mr. Nesbit, of course, 13 is on my left. But, in addition to that, perhaps just 14 raise your hand when I call your name: Mr. Robert 15 Harvey, Mr. Bert Dunn, and Dr. Kevin McCoy. And also 16 with us today is Ms. Rose Cummings, who is a public 17 affairs person for Duke Energy.
18 CHAIRPERSON YOUNG: Ms. Curran?
19 MS. CURRAN: I'm Diane Curran, counsel for 20 Blue Ridge Environmental Defense League. With me 21 today is BREDL's expert witness, Dr. Edwin Lyman of 22 the Union of Concerned Scientists.
23 I would also like to mention that Janet 24 and Lou Zeller of BREDL are here. Janet is president 25 of BREDL, and Lou is the nuclear issues coordinator.
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2077 1 They're sitting right behind me.
2 CHAIRPERSON YOUNG: I also noticed that we 3 have other people in the audience. I don't know 4 whether any of you are non-NRC people or non-party 5 people who are members of the public.
6 Everyone is aware, I'm sure, that we do 7 have the security screening downstairs. If you want 8 to go off this floor, you will need an escort. Just 9 let us know, and we will arrange for escorts. That 10 applies to the parties as well to any non-NRC person.
11 If anyone has a cell phone, please turn it 12 off or on vibrate. I am saying that with the proviso 13 that mine is on solely to take a call from Judge 14 Elleman, in which case we may need to take a short 15 break. We have had some difficulty reaching him. And 16 so for that reason, we have e-mailed him that we will 17 if necessary take a break to take with him between his 18 trying to make various planes.
19 There is a cafeteria downstairs. There 20 are restrooms on this floor. If there is anything 21 else that anyone needs to know about these kinds of 22 logistical issues, we have staff people here who can 23 assist you with those.
24 JUDGE BARATTA: Karen, do you want to 25 stand up?
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2078 1 CHAIRPERSON YOUNG: Thank you. All right.
2 As I said earlier, when we found out that 3 Judge Elleman was not going to be able to be here 4 today, I did send an e-mail out to the parties and 5 also filed with SECY, the Office of the Secretary, 6 indicating that because of that, we would take only 7 entering into the record the prefiled testimony 8 exhibits and any preliminary or procedural matters 9 that we could sort of get out of the way today.
10 As part of that, we envisioned that, to 11 the degree possible, if the parties can ascertain 12 together that there are certain areas that are not 13 disputed so that we can sort of get those out of the 14 way, so to speak, and focus more on the disputed 15 areas, that would be another useful thing in our minds 16 in how we envisioned using this afternoon.
17 Subsequent to our e-mail, Ms. Uttal by 18 e-mail sent a request to me asking that we start with 19 the testimony because we do have a quorum of the 20 board. Judge Baratta and I have discussed that. And 21 while we're glad to hear what the parties have to say, 22 I will tell you that our inclination is that we would 23 not want to deprive Judge Elleman of the opportunity 24 to hear all of the testimony given that he will be 25 taking part in the decision.
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2079 1 If we were able to get a transcript of 2 today's testimony done by 7:00 o'clock in the morning 3 -- and I suppose we can check on that -- then there 4 might be some possibility that he could read that 5 testimony or a portion of it before we started at 6 8:00.
7 Our expectation is that we won't get the 8 testimony that early. So that it would be a 9 significant limitation on his ability to participate 10 as meaningfully if he were not present to hear all of 11 the testimony, all of the cross-examination and verbal 12 testimony.
13 Do you want to add anything to that before 14 we hear --
15 JUDGE BARATTA: Yes. I guess it is to be 16 understood that we are working on an accelerated 17 schedule and because of prior commitments, we are 18 having to only have a limited number of dates 19 available to be able to do this.
20 It is unfortunate that the weather didn't 21 cooperate and that the gentleman's plane was delayed.
22 However, we are still committed to try to support an 23 early decision in this process, even if it meant that 24 we might to have another day or so worth of hearings 25 at some time in the future.
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2080 1 So I think that we would like to be fair 2 to the panel and make sure that everybody has an 3 opportunity to participate, as opposed to proceeding, 4 but we are willing to listen to what you might have to 5 say with regards to that.
6 CHAIRPERSON YOUNG: Also, we have 7 discussed -- and I know we discussed this with the 8 parties earlier -- we are going to be here for our 9 closed hearing on Friday. And I know the staff has 10 one expert that you were concerned who couldn't be 11 here Friday. You were concerned about going ahead, 12 even with other experts' testimony without him 13 present.
14 I guess let me start with the staff and 15 ask you first what you have to say about your request 16 in light of the concerns that we have expressed and 17 also about running over onto Friday. We could 18 reorganize the order of the testimony to accommodate 19 you if we did need to do that.
20 MS. UTTAL: Judge, we appreciate the fact 21 that the Board has accommodated us by having an 22 extended schedule for these two days, but the fact of 23 the matter is that Dr. Meyer is going out of the 24 country on Friday and will not be available at all 25 that day.
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2081 1 He is a very important witness and very 2 important source of information. And I cannot proceed 3 without having him here for the entire proceeding.
4 That is the reason why we have requested that we 5 proceed today, so that we can be assured that this 6 matter will be over by 9:00 o'clock tomorrow night.
7 So we would object to the hearing 8 proceeding on Friday because Dr. Meyer will not be 9 here.
10 JUDGE BARATTA: In light of that, if we 11 were to have to continue on another day, other than 12 Friday, when will Dr. Meyer be returning?
13 MS. UTTAL: Anytime after the 27th of 14 July.
15 JUDGE BARATTA: So Tuesday, the 27th would 16 be. As we hear from the other parties, would you 17 indicate what your availability might be after the 18 27th?
19 MS. UTTAL: Judge, one more thing.
20 CHAIRPERSON YOUNG: Go ahead.
21 MS. UTTAL: I become unavailable on August 22 14th.
23 JUDGE BARATTA: Fourteenth, did you say?
24 MS. UTTAL: August 14th or 13th, whatever 25 that Friday is.
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2082 1 JUDGE BARATTA: All right. So there is 2 about a three-week window, then?
3 CHAIRPERSON YOUNG: And I'm unavailable 4 for the period early in August as well.
5 MS. UTTAL: I would like to check with the 6 rest of my witnesses, too.
7 CHAIRPERSON YOUNG: We will all need to 8 check calendars if that is what we need to do because 9 I know several people have schedules. And I can't 10 recall what Judge Elleman's schedule during that time 11 is either.
12 Anything more from you before we move on?
13 MS. UTTAL: No, Your Honor.
14 CHAIRPERSON YOUNG: Okay. I was going to 15 go to Duke next, but you look like you want to say 16 something. Do you want to go on next, Ms. Curran?
17 MS. CURRAN: I look more enthusiastic than 18 he.
19 MR. REPKA: I'm happy to for Duke Energy.
20 We're here. We're ready to proceed. We're 21 disappointed that Judge Elleman is not here and think 22 it important that he hear the evidence. But 23 notwithstanding that, we're here. We're ready to 24 proceed. We're willing to proceed.
25 We support the staff's request. Unlike NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2083 1 the staff, we are here on Friday. We would be 2 available to continue on Friday. But we understand 3 the importance the staff places on Dr. Meyer. And so 4 we are willing to proceed today to make that 5 accommodation.
6 Beyond that, if we don't proceed today, 7 any trade-off that would result in a hearing, an extra 8 day of hearings, two weeks from now, three weeks from 9 now, whatever it might be, we would oppose that. We 10 much prefer to go forward today than to delay by a 11 number of weeks.
12 CHAIRPERSON YOUNG: Okay. Ms. Curran?
13 MS. CURRAN: We think it's important to 14 have the three judges present for the hearing. And so 15 we do not wish to go forward today. And, frankly, I 16 understood the Board as having made a decision and 17 proceeded accordingly thi's morning in a manner that I 18 prepared for today. I assumed that a decision had 19 been made and that this is the way things were going 20 to go forward. And so I relied on that.
21 But, more importantly, I think it is very 22 important that all three judges be present to hear the 23 testimony, to be able to question the witnesses while 24 a particular subject area is in discussion. I just 25 think it is so much more helpful to the judges' NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2084 1 understanding of the complicated issues here for all 2 judges to be present.
3 We, Dr. Lyman and I, are both available 4 for almost all the rest of July and August after the 5 27th except that I am going to be out of town the week 6 of August 15th. And we would be very glad to try to 7 accommodate whatever schedule the Board wants to make.
8 And, finally, we will do our best tomorrow 9 to be efficient in the way we handle our presentation 10 and try to conclude it tomorrow.
11 CHAIRPERSON YOUNG: Let me just ask, are 12 all the parties' experts going to be available all day 13 today and all day tomorrow?
14 MS. UTTAL: Yes.
15 MR. REPKA: Yes.
16 MS. CURRAN: Yes.
17 CHAIRPERSON YOUNG: Okay. Why don't we go 18 ahead and deal with the housekeeping matters? Well, 19 that may be overstating it, but let's go ahead and 20 have the prefiled testimony entered into the record, 21 have the exhibits admitted to the degree possible.
22 Have the parties had any communication 23 with each other about the exhibits? Do you anticipate 24 that there is going to be any dispute? It would be I 25 think helpful and efficient to go ahead and have NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2085 1 admitted all of those exhibits on which there is no 2 dispute? And if there are any disputes, we may be 3 able to take those up today as well.
4 In any event, if we can go ahead and get 5 all of this done, Judge Baratta and I can then confer 6 with each other and consider what to do with regard to 7 the staff's request.
8 Any objection to that?
9 MR. REPKA: We have no objection to that.
10 And we have no objection to any exhibits that have 11 been proposed by either the staff or BREDL and no know 12 of no objections to any Duke exhibits.
13 MS. UTTAL: We have no objections to 14 Duke's exhibits or to BREDL's exhibits.
15 CHAIRPERSON YOUNG: Okay.
16 MS. CURRAN: We don't have any objection 17 to the other parties' exhibits.
18 CHAIRPERSON YOUNG: Great. Then at least 19 we can go ahead with some of this activity that tends 20 to be a bit tedious but to take some time. So why 21 don't we start on that?
22 As I said in previous communications, we 23 will start with Duke and then BREDL and then the 24 staff. So I am about to ask you, Mr. Repka, if you 25 are ready to present your prefiled testimony.
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2086 1 Now, normally, after you did that, we 2 would allow you to question your witnesses as to 3 whether there are any changes. Let's hold off on any 4 verbal questioning of witnesses about changes. But 5 apart from that, do you know of any? And have you 6 prepared any revised version of your originally 7 submitted prefiled testimony?
8 MR. REPKA: We have two minor pen and ink 9 changes to the testimony, which we have identified and 10 we were going to explain.
11 CHAIRPERSON YOUNG: Okay.
12 MR. REPKA: I don't think that would take 13 more than a few minutes. And we have marked those 14 changes on the copies of the testimony that will be 15 provided to the court reporter. Obviously we would 16 expect the parties and the Board to make those changes 17 as we introduce the witnesses.
18 CHAIRPERSON YOUNG: Great. All right.
19 MS. UTTAL: Your Honor?
20 MR. REPKA: In addition to that, we would 21 offer the five exhibits we included with our proposed 22 testimony. And also in response to your, Judge Young, 23 e-mail yesterday regarding the figures, we have 24 prepared single pages with each of the figures from 25 the testimony, which we would then offer into evidence NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2087 1 if that were your intent as additional proposed 2 exhibits. They're simply duplicates of the figures 3 from the testimony.
4 CHAIRPERSON YOUNG: Okay.
5 MS. UTTAL: Judge?
6 CHAIRPERSON YOUNG: Yes?
7 MS. UTTAL: We have one issue on one of 8 our exhibits, exhibit number 5.
9 JUDGE BARATTA: Your exhibit number 5, 10 staff exhibit?
11 MS. UTTAL: Our exhibit number 5, yes.
12 JUDGE BARATTA: Is it --
13 MS. UTTAL: There was a notation on one of 14 the attachments to it that is some proprietary 15 information. We are checking to see if that 16 information has been released because the document is 17 in ADAMS. That particular page --
18 CHAIRPERSON YOUNG: It is in ADAMS?
19 MS. UTTAL: Yes. So we are checking to 20 see why. If you noted, in our rebuttal testimony, we 21 are only going to use a portion of each of those three 22 exhibits.
23 So I am asking to withdraw exhibits 3, 4, 24 and 5 and in their place put the limited pages, as 25 indicated by my footnote in the rebuttal testimony.
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2088 1 CHAIRPERSON YOUNG: It sounds like we are 2 in great shape to proceed on all of this. Did you 3 have something to add? Okay. Here is the way we are 4 going to do this as just a procedural, mechanical, 5 logistical approach. We will start with Duke.
6 And at the point at which we enter or 7 accept the prefiled testimony and enter it into the 8 record, I will indicate to the court reporter, at this 9 point, we are going to stop the transcript that is 10 being made by the court reporter. -
11 And when the transcript is created, 12 whichever prefiled testimony, Duke's first, will be 13 actually bound into the final transcript with the 14 changes, as Mr. Repka indicated. We will do that for 15 each of the parties.
16 It might be simpler while we are with each 17 party to have your exhibits entered as well. The way 18 we are going to do that is, in addition to whatever 19 markings you have on them, we are going to be creating 20 a new official exhibit number that will be a series of 21 numbers so as to make it easier to keep track of all 22 of the exhibits. And we will be keeping actually a 23 chart of that as well and will be glad to provide that 24 to the party when it's typed up to assist you in your 25 proposed findings and keeping track of the exhibits as NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2089 1 well.
2 Before we start with that, are there any 3 other preliminary matters or procedural or 4 housekeeping matters that we need to address?
5 MR. REPKA: I will just raise one 6 procedural issue. I think I know the answer, but I 7 may as well get consensus. When we put our witnesses 8 on the stand, they will introduce both their direct 9 testimony and their rebuttal testimony and the 10 associated exhibits.
11 At that point, my intent is to turn them 12 over for cross-examination by the Board and the 13 parties. I would point, though, that our witness 14 panel has not had the opportunity for surrebuttal 15 testimony with respect to the rebuttal testimony by 16 Dr. Lyman.
17 And I want to reserve the right at some 18 point -- and I think the appropriate point would be my 19 own redirect examination -- to ask some questions by 20 way of surrebuttal if that has not already been 21 addressed through questions by the parties or the 22 Board during their cross-examination.
23 So as the party with the burden of 24 persuasion, I think that I did want to make it clear 25 that we do want to have that opportunity at some NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2U9U I 1 point.
2 CHAIRPERSON YOUNG: Sure. I don't see any 3 problem with that.
4 JUDGE BARATTA: I don't either.
5 CHAIRPERSON YOUNG: One of the things when 6 we do finish dealing with the prefiled testimony and 7 the exhibits and we break for Judge Baratta and myself 8 to confer with each other, to the extent that you can 9 talk with each other and you know of any issues that 10 you can resolve or you know of any parts of another 11 party's testimony that you have no dispute with and so 12 there is no need to go into that, any way that we can 13 make the taking of the actual testimony more efficient 14 and go more smoothly is fine.
15 We want all parties to have opportunity to 16 get out all that you wish to present on the issues 17 that have been raised in the prefiled testimony and 18 would just encourage that that be done in the most 19 efficient manner possible.
20 So, for example, if there were something 21 very brief that you wanted to point out at the very 22 beginning, I don't see any problem with your doing 23 that very briefly and then still being able to reserve 24 surrebuttal for your redirect as well.
25 MR. REPKA: That would be fine.
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2091 1 JUDGE BARATTA: I have a question on one 2 of the exhibits. I don't know whether to raise it now 3 as a general issue or to discuss it specifically with 4 respect to that exhibit. It deals with proprietary 5 information in that --
6 CHAIRPERSON YOUNG: Go ahead because it 7 may be one of the ones that falls within what Ms.
8 Uttal was talking about.
9 JUDGE BARATTA: Yes. One of the charts 10 that you have for I assume proprietary reasons has the 11 scale omitted on it.
12 MR. REPKA: The numbers. I think you're 13 referring to one of the figures --
14 JUDGE BARATTA: Right.
15 MR. REPKA: -- has no numbers.
16 JUDGE BARATTA: Right.
17 MR. REPKA: And yes, those were 18 proprietary, number one. And, number two, the actual 19 numbers themselves we didn't consider to be relevant 20 to the point we were trying to make. We were trying 21 to show a correlation.
22 JUDGE BARATTA: It did appear in the text.
23 I mean, I could follow this up with a question during 24 the cross-examination that there was one number that 25 was germane to that figure which was raised in the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2092 1 prefiled testimony. So you might think about how you 2 want to handle -- it's not a question of needing the 3 numbers for all of the points but for one or two 4 points on there, it might be important.
5 MR. REPKA: Perhaps you could direct us to 6 the specific figure you are referring to because there 7 are a couple where there is -- not necessarily 8 numbers.
9 JUDGE BARATTA: Let me see if I can --
10 give me a couple of minutes to look for it.
11 MR. REPKA: Sure.
12 CHAIRPERSON YOUNG: Do you want my copy?
13 JUDGE BARATTA: I can probably do this.
14 In the meantime, if you want to go ahead and continue 15 while I check this?
16 MS. CURRAN: Judge, I just want to clarify 17 that all of the parties will have an opportunity for 18 surrebuttal in the course of redirect.
19 CHAIRPERSON YOUNG: We want to make sure 20 that all of the parties have the opportunity to bring 21 out all of the issues that need to be brought out 22 without going into any duplicative testimony, of 23 course.
24 JUDGE BARATTA: Yes. Go ahead. Please go 25 ahead.
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2093 1 CHAIRPERSON YOUNG: All right. Then let's 2 start with Duke. And you indicated that there were a 3 couple of minor changes. Do you want to tell us what 4 they are? And then we can formally place the --
5 MR. REPKA: Why don't I have the witnesses 6 do that? When I put them on the stand, they will 7 swear or affirm to the truth of the testimony and 8 explain the changes that they wanted to make.
9 CHAIRPERSON YOUNG: Okay. Then why don't 10 we have your witnesses come over here to the witness 11 box? And we'll swear them in.
12 JUDGE BARATTA: As a procedural issue, do 13 you want to use the Elmo camera for any of this?
14 MR. REPKA: Not at the outset, no, 15 certainly not for introducing the testimony at this 16 point.
17 JUDGE BARATTA: Or for explaining the 18 changes, possibly?
19 CHAIRPERSON YOUNG: Were they longer than 20 just a couple of words?
21 MR. REPKA: It's really one clause.
22 JUDGE BARATTA: Okay.
23 CHAIRPERSON YOUNG: Okay.
24 JUDGE BARATTA: We wanted to see whether 25 it was lawyer-proof or not. That's why we offered it.
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2094 1 MR. REPKA: It's not my idea. The witness 2 will explain it. I will call to the stand the panel 3 on contention 1: Mr. Nesbit, Mr. Harvey, Mr. Dunn, 4 and Dr. McCoy.
5 JUDGE BARATTA: While they're getting 6 situated, figure 14 was the one I was .referring to in 7 the testimony from Duke. Maybe I misinterpreted 8 something, but I thought there was some discussion in 9 the paragraph 1.11 following that that suggested that 10 there was a particular point that was discussed on 11 that graph.
12 It's not a question of knowing exactly 13 where that point is but within the general area of 14 that point or something.
15 MR. REPKA: We could explain. There is a 16 reference to the typical strains, --
17 JUDGE BARATTA: Right.
18 MR. REPKA: -- between 50 and 75 percent.
19 JUDGE BARATTA: Yes.
20 MR. REPKA: The question is, where does 21 that appear on the --
22 JUDGE BARATTA: Right.
23 MR. REPKA: The witness is going to answer 24 that.
25 JUDGE BARATTA: Okay. That was the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2095 1 question.
2 CHAIRPERSON YOUNG: I will put this down 3 so I can see you. Let me ask all of you to raise your 4 right hands.
5 Whereupon, 6 STEPHEN P. NESBIT, ROBERT C. HARVEY, 7 BERT M. DUNN, and J. KEVIN McCOY 8 were collectively called as witnesses by counsel for 9 the licensee and, having been first duly sworn, were 10 examined and testified as follows:
11 CHAIRPERSON YOUNG: Thank you. Go ahead, 12 Mr. Repka.
13 DIRECT EXAMINATION 14 MR. REPKA: Okay. I'm going to ask you 15 each to introduce yourself in turn. And I will start 16 to my left with Dr. McCoy.
17 DR. McCOY: I'm Dr. J. Kevin McCoy.
18 MR. DUNN: I'm Bert Dunn.
19 MR. NESBIT: I'm Steve Nesbit.
20 MR. HARVEY: Robert Harvey.
21 MR. REPKA: And your titles and 22 responsibilities are described in the testimony, is it 23 not?
24 DR. McCOY: Yes, they are.
25 MR. REPKA: Do you have a document in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2096 1 front of you that's titled "Testimony of Stephen P.
2 Nesbit, Robert C. Harvey, Bert M. Dunn, and J. Kevin 3 McCoy on Behalf of Duke Energy Corporation on 4 Contention 1 (MOX Fuel Lead Assembly Program, MOX Fuel 5 Characteristics and Behavior and Design Basis Accident 6 LOCA Analysis"?
7 DR. McCOY: We do, yes.
8 MR. REPKA: Let me ask you each to respond 9 in turn. Was this document or at least those portions 10 of it to which your name is assigned prepared by you 11 or under your supervision?
12 DR. McCOY: It was.
13 MR. DUNN: Yes.
14 MR. NESBIT: Yes.
15 MR. HARVEY: Yes, it was.
16 MR. REPKA: In this document, which is 17 dated July 1, 2004, do any of you have any changes or 18 corrections or clarifications you would like to make?
19 MR. NESBIT: I have one clarification to 20 make.
21 MR. REPKA: Yes. Do you want to go ahead 22 and describe that?
23 MR. NESBIT: In paragraph 154, we would 24 like to insert the following words at the beginning as 25 follows, "For the cases establishing the LOCA limits" NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2097 2 CHAIRPERSON YOUNG: I'm sorry. You're 3 adding this at the beginning of it?
4 MR. NESBIT: At the very beginning.
5 CHAIRPERSON YOUNG: Okay.
6 MR. NESBIT: Prior to the "The."
7 CHAIRPERSON YOUNG: Okay.
8 MR. NESBIT: "For the cases establishing 9 the LOCA limits or the MOX fuel lead assemblies,"
10 comma. That's it. The rebuttal testimony of Dr.
11 Lyman indicated that there was some confusion as to 12 which LOCA analyses and which peak cladding 13 temperatures we were referring to.
14 We thought it was clear from the context 15 of the testimony in paragraph 154 that we were 16 referring to the analyses that performed the bases for 17 the LOCA limits. However, there appeared to be some 18 confusion on that point. So we just wanted to clarify 19 that.
20 MR. REPKA: You've identified an insertion 21 at the beginning of the paragraph. Was there not a 22 deletion in it later in that sentence? I believe you 23 wanted to delete the words so it should delete "For 24 Catawba described in the MOX fuel lead assembly 25 license amendment request."
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2098 1 We have a copy marked up, and I'm looking 2 for it.
3 MR. NESBIT: It was -- have we got the 4 marked-up copy?
5 MR. REPKA: Give me one minute.
6 MR. NESBIT: I'll get that for 7 clarification.
8 MR. REPKA: If I can find Ms. Cottingham, 9 she -- I'm going to hand a copy of this to Mr. Nesbit.
10 MR. NESBIT: Yes, we had a deletion as 11 well. We would delete -- in that same first sentence 12 of paragraph 154, we would delete the following, "In 13 the LOCA calculations for Catawba described in the MOX 14 fuel lead assembly license amendment request."
15 MR. REPKA: So that sentence reads, "For 16 the cases establishing the LOCA limits or the MOX fuel 17 lead assemblies, the highest PCT at the ruptured 18 location was approximately 1,750 degrees Fahrenheit, 19 and the local oxidation on that fuel pin is 3 20 percent." Is that correct?
21 MR. NESBIT: That's correct.
22 MR. REPKA: Dr. McCoy, did you have a 23 clarification you wanted to make?
24 DR. McCOY: I had one clarification also 25 to make. In paragraph 4, the third and fourth lines NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2099 1 refer to 20 years of experience in the nuclear 2 industry. I'd like to insert comma, "primarily" after 3 "experience."
4 MS. CURRAN: I'm sorry. I couldn't hear 5 that.
6 DR. McCOY: I would like to insert comma, 7 "primarily" after "experience." So the sentence would 8 read, "I have more than 20 years of experience, 9 primarily in the nuclear industry."
10 MR. REPKA: Okay. With those changes, is 11 this testimony true and correct, to the best of your 12 knowledge and belief?
13 DR. McCOY: It is.
14 MR. DUNN: It is.
15 MR. NESBIT: It is.
16 MR. HARVEY: It is.
17 MR. REPKA: And do you sponsor this 18 testimony as your direct testimony in this proceeding?
19 DR. McCOY: I do.
20 MR. DUNN: I do.
21 MR. NESBIT: I do.
22 MR. HARVEY: I do.
23 MR. REPKA: Okay. I have another document 24 in front of me, which is the exhibits related to that 25 testimony. And it identifies five proposed exhibits, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE.. N.W.
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2100 1 which have been pre-served on the Board and the 2 parties.
3 If the Board wants, I can identify the 4 exhibits for the record, but --
5 CHAIRPERSON YOUNG: Why don't you do that?
6 I think that would be helpful.
7 MR. REPKA: Okay. Exhibit 1 is a 8 selection of relevant portions of Duke Energy's 9 license amendment request submitted to the NRC on 10 February 27th, 2003, specifically attachment 3 to that 11 submittal, section 3.7.1.
12 CHAIRPERSON YOUNG: 3.71?
13 MR. REPKA: 3.7.1.
14 CHAIRPERSON YOUNG: 7.1.
15 MR. REPKA: Exhibit 2 is a selection of 16 relevant portion of Duke Energy's response to an NRC 17 request for additional information. The Duke response 18 was submitted to the NRC on November 3rd, 2003 and 19 addresses LOCA analysis.
20 Exhibit 3 is a paper by M. Lambert and 21 others titled "Synthesis of an EDF and Framatome ANP 22 Analysis on Fuel Relocation Impact and Large-Break 23 LOCA Presented at Aix-en-Provence in March of 2001."
24 Exhibit 4 is another paper presented at 25 Aix-en-Provence in March 2001 by a C. Grandjean and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2101 1 others entitled "High Burnup U02 Fuel LOCA 2 Calculations to Evaluate the Possible Impact of Fuel 3 Relocation After Burst."
4 Exhibit 5 is a presentation from Argonne 5 in May of 2004 at a topical meeting on LOCA issues 6 presented by V. Guillard and others titled "Use.of 7 CATHARE2" -- that's C-A-T-H-A-R-E 2 -- "Reactor 8 Calculations to Anticipate Research Needs."
9 So do you gentlemen sponsor these as 10 proposed exhibits in this proceeding?
11 DR. McCOY: We do.
12 MR. DUNN: I do.
13 MR. NESBIT: I do.
14 MR. HARVEY: I do.
15 MR. REPKA: Okay. I have another document 16 I'm going to have Ms. Cottingham hand out, which is a 17 compilation of the 19 figures from your direct 18 testimony of July 1st, each on a single page.
19 Rather than identifying each and every one 20 for the record, I would just state that they are 21 labeled as figures 1 through 19. They have the same 22 numbers as they do in the testimony. And we'll pass 23 those out to the Board and parties.
24 CHAIRPERSON YOUNG: Okay.
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2102 1 shortly, but those could be marked as the next 19 Duke 2 exhibits. I would simply say, do you gentlemen 3 sponsor those as exhibits of Duke Energy in this 4 proceeding?
5 DR. McCOY: We do.
6 MR. DUNN: We do.
7 MR. NESBIT: We do.
8 MR. HARVEY: We do.
9 MR. REPKA: Next you have a document in 10 front of you dated July 8th, 2004 titled "Rebuttal 11 Testimony of Stephen P. Nesbit, Robert C. Harvey, Bert 12 M. Dunn, and J. Kevin McCoy on Behalf of Duke Energy 13 Corporation on Contention 1."
14 DR. McCOY: Yes, we do.
15 MR. REPKA: At least with respect to the 16 relevant parts that each of you sponsored in this 17 rebuttal testimony, was this document prepared by you 18 or under your supervision?
19 DR. McCOY: Yes, it was.
20 MR. DUNN: Yes, it was.
21 MR. NESBIT: Yes, it was.
22 MR. HARVEY: Yes, it was.
23 MR. REPKA: Do any of you have any 24 corrections or clarifications you want to make to this 25 testimony?
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2103 1 DR. McCOY: No.
2 MR. DUNN: No.
3 MR. NESBIT: No.
4 MR. HARVEY: No.
5 MR. REPKA: Is the testimony true and 6 correct, to the best of your knowledge and belief?
7 DR. McCOY: Yes, it is.
8 MR. DUNN: Yes.
9 MR. NESBIT: It is.
10 MR. HARVEY: Yes.
11 MR. REPKA: And do you sponsor it as your 12 testimony in this proceeding?
13 DR. McCOY: Yes.
14 MR. DUNN: Yes.
15 MR. NESBIT: Yes.
16 MR. HARVEY: Yes.
17 MR. REPKA: Okay. With that, then, we 18 will pass out the figures 1 through 19 labeled Duke 19 exhibits 6 through 24. And, if appropriate to do so, 20 I would simply move that Duke exhibits 1 through 24 be 21 admitted into evidence in connection with the direct 22 testimony of Duke Energy.
23 CHAIRPERSON YOUNG: Okay. Why don't we 24 first in a moment I'm going to ask the court reporter 25 to stop. And at that point in the transcript, the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2104 1 direct prefiled testimony of Duke Energy and the 2 prefiled rebuttal testimony of Duke Energy will be 3 bound into the transcript.
4 Did that make sense to you how I've 5 described it? Okay. Then let's just stop right here.
6- And at this point, those two documents will be bound 7 into the transcript.
8 9
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salo5 July 1, 2004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of: )
)
DUKE ENERGY CORPORATION )
) Docket Nos. 50-413-OLA (Catawba Nuclear Station, ) 50-414-OLA Units I and 2) )
)
)
TESTIMONY OF STEVEN P. NESBIT, ROBERT C. HARVEY, BERT M. DUNN, AND J. KEVIN McCOY ON BEHALF OF DUKE ENERGY CORPORATION ON CONTENTION I (MIOX FUEL LEAD ASSEMBLY PROGRAM, MIOX FUEL CHARACTERISTICS AND BEHAVIOR, AND DESIGN BASIS ACCIDENT (LOCA) ANALYSIS)
zIQb TABLE OF CONTENTS
- 1. INTRODUCTION ........................................................ 1 II. OVERVIEW ........................................................ 5 A. The MOX Fuel Project....................................................... S5 B. MOXFutel Experience........................................................ 7 C. LAR Overview ....................................................... 10 III. MOX FUEL APPENDIX K LOCA ANALYSIS ....................................................... 11 A. Summnnary ofRequirements ....................................................... 11 B. Overview of a PressurizedIMater Reactor (PIVA) Design Basis LOCA ................... 13 C. Overview ofMOXFutel LOC.A Analysis ................................. ...................... 18 D. Conservatisms in the LOCA Analysis ....................................................... 24 IV. DIFFERENCES IN LEU AND MOX FUEL BEHAVIORS ........................................... 26 A. Fuel-Related "Differences....................................................... 27 B. Cladding-Related "Differences....................................................... 37 V. FUEL RELOCATION AND RELATED ISSUES ..................................... 40 A. Futel Relocation- Description and RegulatoryIlistory........................................... 41 B. V'ERCORS Tests ....................................................... 47 C. AMS Ballooning....................................................... 51 D. FillingRatio of Relocated Fuel....................................................... 54 E. Fuel-CladdingInteraction....................................................... 59 F. MOXFuel Relative Power at IHigh Burnip....................................................... 61 G. Conclusions on Fu el Relocation ................. ...................................... 66 VI. UNCERTAINTIES ............ 70 VI. CONCLUSIONS ............. 74 i
2407 July 1,2004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of: )
DUKE ENERGY CORPORATION )
) Docket Nos. 50-413-OLA (Catawba Nuclear Station, ) 50-414-OLA Units 1 and 2) )
)
TESTIMONY OF STEVEN P. NESBIT, ROBERT C. IIARVEY, BERT M. DUNN, AND J. KEVIN McCOY ON BEHALF OF DUKE ENERGY CORPORATION ON CONTENTION I (MOX FUEL LEAD ASSEMBLY PROGRAM, MIOX FUEL CHARACTERISTICS AND BEHAVIOR, AND DESIGN BASIS ACCIDENT (LOCA) ANALYSIS)
I. INTRODUCTION
- 1. (Nesbit) 1,Steven P. Nesbit, am an Engineering Supervisor II employed by Duke Energy Corporation (Duke). I currently serve as the Duke Mixed Oxide (MOX) Fuel Project Manager. I have 22 years experience with Duke and 24 years experience overall in nuclear engineering and management, including both the commercial sector and U.S. Department of Energy (DOE) nuclear projects. I have particular experience in the area of design engineering, including nuclear safety analysis and nuclear reactor safety reviews. In my current position I direct the technical, Nuclear Regulatory Commission (NRC) licensing, and business activities associated with the MOX Fuel Project. A full statement of my Professional Qualifications is included as Attachment I to this testimony.
1
- 2. (Hanrey) 1, Robert C. Harvey, am a Senior Engineer employed by Duke, responsible for the Loss of Coolant Accident (LOCA) analyses supporting the Oconee Nuclear Station (Oconee), McGuire Nuclear Station (McGuire), and Catawba Nuclear Station (Catawba).
Prior to my employment by Duke, I was employed by Yankee Atomic Electric Company to perform LOCA and severe accident analyses to support various nuclear reactor licensing initiatives and core reloads. Overall, I have over 25 years of experience in nuclear thermal hydraulic and safety analyses. With respect to the Duke MOX Fuel Project, I provided oversight to AREVA (formerly Framatome ANP) in the performance of the supporting LOCA analyses discussed further below. A full statement of my Professional Qualifications is included as to this testimony.
- 3. (Dunn) I, Bert Dunn, am an Advisory Engineer employed by AREVA Framatome ANP Inc., located in Lynchburg, Virginia. I have 34 years experience in the nuclear engineering field - primarily in the area of LOCA analyses and safety analyses to support nuclear fuel design and licensing activities. For the Duke MOX Fuel Project, I am responsible for the LOCA analyses supporting Duke's proposal to utilize four MOX fuel lead assemblies at Catawba. A full statement of my Professional Qualifications is included as Attachment 3 to this testimony.
- 4. (AMcCo) I, J. Kevin McCoy, am an engineer in the fields of metallurgy and materials engineering, employed by Framatome ANP, Inc. I have a Doctorate in Materials Engineering and a Masters Degree in Metallurgical Engineering. I have more than 20 years of experience/in the nuclear industry, with my most recent work in the area of nuclear fuel including recent work on MOX fuel performance. A full statement of my Professional Qualifications is Attachment 4 to this testimony.
2
- 5. (All) Based on our specific job responsibilities, we are all very familiar with Duke's License Amendment Request (LAR), dated February 27, 2003. The LAR seeks NRC approval for Duke's proposal to use four MOX fuel lead assemblies at Catawba. The MOX fuel lead assemblies will be included in cores that will be predominantly comprised of Low Enriched Uranium (LEU) fuel assemblies.
- 6. (All) The purpose of this testimony is to specifically address Contention I of the Blue Ridge Environmental Defense League (BREDL), an intervenor in this NRC license amendment proceeding. That contention, as admitted by the Atomic Safety and Licensing Board (Licensing Board) in its Memorandum and Order of March 5, 2004, asserts that:
The LAR is inadequate because Duke has failed to account for differences in MOX and LEU fuel behavior (both known differences and recent information on possible differences) and for the impact of such differences on LOCAs and on the [design basis accident (DBA)] analysis for Catawba.1
- 7. (All) As discussed in this testimony, LOCA is the only DBA analysis at issue in this proceeding. In our testimony we show that, contrary to Contention I, Duke has fully accounted for "known" differences in MOX and LEU fuel behavior in its LAR; that Duke has performed conservative LOCA analyses using well-known and approved methodologies; and that the LAR meets applicable NRC requirements. Duke has also considered issues raised in a recent research proposal concerning "possible" differences in behavior between MOX fuel and LEU fuel, and has concluded that these issues do not affect the safety or compliance of the MOX fuel lead assembly proposal.
X "Memorandum and Order (Ruling on Standing and Contentions)," Duke Energy Corp.
(Catawba Nuclear Station, Units I and 2), LBP-04-04 _ NRC _ (March 5, 2004, slip op. at 63).
3
21W
- 8. (All) This testimony focuses in particular on Duke's LOCA analyses, which were completed in accordance with the methodology and acceptance criteria in NRC regulations in 10 C.F.R. § 50.46 and 10 C.F.R. Part 50, Appendix K. Consistent with the direction of the Atomic Safety and Licensing Board,2 this testimony on Contention I does not include issues and analyses related to LOCA dose consequences. Duke has in fact prepared assessments of doses following a design basis LOCA in accordance with applicable regulatory requirements and guidance.
Those analyses are addressed in the LAR, Attachment 3, Section 3.7.3, and in responses to NRC Staff Requests for Additional Information (RAls). However, these matters are beyond the scope of Contention I and the scope of the testimony of this panel.
- 9. (All) Section 11 of this testimony provides an overview of the MOX Fuel Project and the requested regulatory approval.Section III provides an overview of LOCA analyses in general, and reviews the LOCA analyses that were performed for the MOX fuel lead assemblies.
Section IV addresses the potential differences in LEU and MOX fuel behavior that were postulated by BREDL.Section V reviews the fuel relocation issue raised by BREDL in more detail, and addresses potential fuel relocation and related phenomena in the context of the MOX fuel application.Section VI addresses the issue of uncertainties as they relate to lead test assembly programs and the LAR.Section VII provides the conclusion of this testimony, which is that the MOX fuel lead assembly program does not pose an undue risk to the health and safety of the public due to the LOCA issues raised in Contention 1. In particular, MOX fuel does not pose a risk in the area of fuel relocation that is significantly different than for LEU fuel. That 2 Order (Confirming Matters Addressed at April 6 Telephone Conference), April 8, 2004 (at p. 2) ("With respect to Contention 1, this contention encompasses those calculations involved in the determination of events up to and including LOCAs and DBAs, but does not include analyses related to any releases either in containment or offsite."); see also Tr. 1726-36.
4
2'lI risk is well-bounded by conservatisms in the regulatory acceptance criteria and in the LOCA analysis.
- 10. (All) Our testimony will show that the LOCA calculations performed by AREVA and submitted by Duke (i) meet the applicable regulatory requirements and (ii) adequately and conservatively characterize MOX/LEU differences. We will demonstrate that the MOX fuel LOCA analyses are inherently conservative, providing a large margin of safety for the public.
We will address purported MOX/LEU differences in detail, particularly those related to fuel relocation and related issues. We will address the BREDL concern over uncertainty in the context of the proposed Catawba lead assembly program. We will conclude that Contention I is not correct, and that the Catawba MOX fuel lead assembly program can proceed without adverse impacts on the health and safety of the public.
II. OVERVIEW A. The AOX Fuel Project
- 11. (Nesbit) On February 27, 2003, Duke submitted the LAR to the NRC to allow for the receipt and use of four MOX fuel lead assemblies at one of the McGuire or Catawba units.
On September 23, 2003, Duke amended the LAR to apply to Catawba only.
- 12. (Nesbit) Duke's proposal to use four MOX fuel lead assemblies is part of an important and ongoing nuclear non-proliferation program of the United States and the Russian Federation. The goal of this program is to dispose of surplus plutonium from nuclear weapons by converting the material into MOX fuel and using that fuel in nuclear power reactors. The current proposal for four MOX fuel lead assemblies supports the potential future use of larger quantities of MOX fuel at either Catawba or McGuire. Should such future use occur, however, it will be the subject of a separate NRC licensing action.
5
- 13. (Nesbit) The MOX fuel lead assemblies will be manufactured in France under the direction of AREVA. The MOX fuel assemblies will be based on the AREVA Advanced Mark-BW fuel assembly, a fully qualified LEU fuel assembly design that will be adapted for MOX application through changes to the fuel rod design. The Advanced Mark-BW fuel assembly is a standard lattice 17 x 17 fuel assembly specifically designed for use in Westinghouse reactors (such as Catawba). The Advanced Mark-BW adaptation for MOX applications, the Mark-BW/MOXI, is dimensionally and structurally identical to the Advanced Mark-BW with the only change appearing in the fuel rod internal design. The Advanced Mark-BW and the Mark-BW/MOXI share as a design feature M5TMalloy fuel rod cladding, guide thimbles, and spacer grids. (M5 is a trademark of AREVA.)
- 14. cNesbit) Duke is currently targeting Catawba Unit 1, Cycle 16 (CIC16), with a Spring 2005 startup, for initial insertion of MOX fuel lead assemblies. Plans call for the lead assemblies to be irradiated for a minimum of two cycles to confirm acceptability of the fuel assembly design, verify the ability of Duke's and AREVA's models to predict fuel assembly performance, and confirm the applicability of the European database on MOX fuel performance to Duke's use of MOX fuel.
- 15. (Nesbit) The CIC16 core will be predominantly comprised of Westinghouse Robust Fuel Assembly (RFA) fuel assemblies. In addition to the RFA fuel and the four MOX fuel assemblies, there will be eight Westinghouse Next Generation Fuel (NGF) lead test assemblies in the core. The eight NGF test assemblies will be loaded into core locations that are not adjacent to the MOX fuel.
6
2113
- 16. (Nesbit) On April 5, 2004, the NRC Staff issued a Safety Evaluation Report (SER), concluding that the health and safety of the public will not be endangered by operation with four MOX fuel lead assemblies at Catawba.
- 17. (Nesbit) The Advisory Committee on Reactor Safeguards (ACRS) has also considered the LAR. The ACRS Subcommittee on Reactor Fuels reviewed the matter during a public meeting (in which BREDL's consultant participated) on April 21, 2004. The full ACRS reviewed the MOX fuel lead assembly LAR in a public meeting (in which BREDL's consultant again participated) on May 6, 2004. In a letter dated May 7, 2004, the ACRS concluded that "under the restricted circumstances considered in both the Duke Power application and the NRC Staff's safety evaluation, the four mixed oxide lead test assemblies in non-limiting core locations that do not contain control rods can be irradiated in either of the Catawba reactor cores with no undue risk to the public health and safety."
B. MOX Ftel Experience
- 18. (Nesbit) Standard light water reactor LEU fuel consists of sintered uranium oxide pellets enclosed in zirconium alloy cladding to form fuel rods. The fuel rods are bundled into a fuel assembly, and the fuel assemblies reside in the reactor core where they produce power through nuclear fissions in the fuel pellets.
- 19. (Nesbit) The uranium in LEU fuel is slightly enriched in 235 U (typically in the range of 4-5% 235 U). Initially, energy comes from fissions in the uranium, mainly from thermal fissions in 235 U. As LEU fuel is used, it also produces plutonium from neutron absorption in 238 U. 239Pu and 241 Pu, like 235U, are fissionable at thermal neutron energies. Later in the life of the LEU fuel, the initial 235U has been depleted and appreciable quantities of plutonium have built up, so LEU fuel energy comes from fissions in both uranium and plutonium.
7
2a1l1
- 20. (Nesbit) The plutonium that is produced in LEU fuel can be recovered by reprocessing the spent fuel. This involves discharging the fuel from the reactor, dissolving it in acid, and chemically separating the uranium, plutonium, and fission products from the resulting solution.
- 21. (Nesbit) MOX fuel is another type of nuclear fuel. MOX fuel is essentially identical to LEU fuel except that the MOX fuel pellets are comprised of a small amount of plutonium oxide mixed with the remainder uranium oxide (typically, depleted uranium in the range of 0.2-0.3% 235 U). MOX fuel is used on a large scale in some European reactors, utilizing plutonium that has been recovered from reprocessing uranium fuel. This type of plutonium is classified as reactor grade (RG) because it contains at least 20% 240 Pu. Catawba will utilize weapons grade (WG) plutonium. WG MOX fuel contains plutonium with less than 7% 240Pu.
- 22. (Ncsbit) In the United States, MOX fuel development programs were conducted in the 1960s and 1970s in the anticipation that wide-scale reprocessing of spent nuclear fuel in the United States would result in substantial quantities of separated plutonium. MOX fuel lead test assembly programs were conducted at four United States commercial nuclear power plants:
Big Rock Point, Dresden, San Onofre, and Quad Cities. In the late 1970s, in an attempt to discourage the commercial separation of plutonium, the United States government instituted a policy that the United States would not reprocess commercial reactor fuel. As a result, United States plans for reprocessing and MOX fuel use were discontinued.
- 23. (Nesbit) In 1980, the NRC approved the loading and use of four MOX fuel assemblies at Ginna Unit I (out of a total of 121 assemblies in a Ginna core). The Ginna MOX fuel assemblies were the last MOX fuel assemblies loaded into a domestic commercial nuclear power reactor in the United States.
8
- 24. (Nesbit) Despite the policy change in the United States, in the last three decades there have been substantial advances in MOX fuel technology worldwide. Other countries reprocess spent nuclear fuel on a large scale, producing plutonium than can be fabricated into MOX fuel. Nuclear power plant MOX fuel use began in German reactors in the 1970s.
Currently, more than thirty reactors in France, Germnany, Switzerland, and Belgium are using MOX fuel. European reactors operate with cores comprised of a mixture of LEU fuel assemblies and MOX fuel assemblies, with MOX fuel core fractions as high as 36%. MOX fuel fabrication and use has reached a state of industrial maturity in Europe.
- 25. (Ncsbit) European experience indicates that MOX fuel performs well in nuclear power reactors. Fuel failure rates are commensurate with those of LEU fuel. Significantly, there has never been a MOX fuel failure attributable to the fuel pellet material.3
- 26. (Nesbit) MOX fuel is fundamentally similar to LEU fuel. MOX fuel is predominantly uranium oxide, and physical characteristics such as thermal conductivity and heat capacity are very close between the fuel types. Differences such as fission gas release are accommodated by fuel design and operation. Plutonium is the primary fissionable element in MOX fuel, unlike LEU fuel which has predominantly uranium fissions at beginning of life. This difference affects nuclear characteristics such as thermal flux level and effective delayed neutron fraction. Again, these differences are well understood and accommodated by fuel assembly and core design.
- 27. (Nesbit) Irradiation of plutonium in MOX fuel is an effective means of rendering WG plutonium unattractive for theft or diversion into nuclear weapons. Irradiation of WG 3 BAW-10238P, MOX Fuel Design Report (2002). BAW-10238P, Revision 1, was issued in May 2003, as referenced below. There is also a non-proprietary version of this report that includes the pertinent informnation.
9
plutonium in MOX fuel destroys much of the original plutonium and degrades the isotopics of the remainder.
C. LAR Ovenriew'
- 28. (Nesbit) The LAR consists of a cover letter and Attachments 1-6. Attachments I and 2 are marked-up copies of the McGuire and Catawba Technical Specifications, respectively.
The Technical Specification changes are to the following areas: spent fuel storage, reactor core design features, criticality, and Core Operating Limits Report approved methodologies. LAR provides the safety justification for the proposed changes. Attachment 4 is Duke's No Significant Hazards Consideration analysis. Attachment 5 is Duke's assessment of environmental impacts associated with the MOX fuel use. Attachment 6 contains a request for exemptions from selected NRC regulations. These exemption requests clarify the applicability of current NRC regulations to MOX fuel, and provide for the use of M5Tcladding.
- 29. (Nesbit) Attachment 3 to the LAR, the safety analysis, covers four major technical areas: description of the MOX fuel lead assemblies, effects of four MOX fuel lead assemblies on plant operation, safety analyses of operation with four MOX fuel lead assemblies, and the impact of operation with four MOX fuel assemblies on risk. In addition to the information in , the LAR and associated responses to NRC RAls include by reference a number of topical reports that were developed, wholly or in part, to support the MOX fuel program. Those topical reports, reviewed separately by the NRC, are listed below:
- BAW-10238P, Revision 1,MOX Fuel Design Report (May 2003).
Background information on MOX fuel and evaluations of the impact of MOX fuel pellets on the fuel assembly and fuel rod design.
- BAW-10239P, Advanced Mk-BWV Fuel Assembly Mechanical Design (March 2002). Description of the fuel assembly design.
10
2t1l
- BAW-10231P, COPERNIC Fuel Rod Design Computer Code (September 1999). Computer code used for mechanical analyses of MOX and LEU fuel.
- DPC-NE-1005P, Nuclear Design Methodology Using CASMO-4/SIMULATE-3 MOX (August 2001). "Description of nuclear analysis methods for application to LEU and MOX fuel at McGuire and Catawba.
- DPC-NE-2005P-A, Revision 3 (September 2002), Thermal-Hydraulic Statistical Core Design Methodology. Appendix E.
Thermal-hydraulic design methodology for the MOX fuel assembly design.
The LAR also references the following topical report that is independent of fuel type:
- BAW- 10227P-A, Evaluation of.Advanced Claddingand Structural Material (M5) in PIWR Reactor Fuel (February 2000). Evaluation of M5Tm as an advanced cladding and structural material.
- 30. (Nesbit) Contention I raises issues that have been addressed in the LAR and associated correspondence. Specifically, the impact of four MOX fuel lead assemblies on a LOCA analysis is addressed in LAR Attachment 3, Section 3.7.1 and in a Duke RAI response submitted November 3, 2003. The relevant portions of these documents are included as Exhibits I and 2, respectively, to this testimony.
III. MOX FUEL APPENDIX K LOCA ANALYSIS A. Summary ofRequirements
- 31. (Han'er Nesbit) Contention I refers to the impact of MOX fuel on "LOCAs and on the DBA accident analysis" for Catawba. We read this to refer to the impact on a design basis LOCA scenario, and more specifically on the design basis LOCA analysis. As acknowledged by BREDL in response to Duke Interrogatory 4, "BREDL makes no assertions [in Contention I]
regarding design basis accident (DBA) scenarios or accident analyses other than for the design 11
basis LOCA." 4 The relevant LOCA analyses (Section 3.7.1 of the LAR and the RAI responses) that Duke has provided in connection with its application address the requirements of 10 C.F.R.
§ 50.46 and 10 C.F.R. Part 50, Appendix K.
- 32. (Hanrey) 10 C.F.R. § 50.46 states that each light-water reactor must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated design basis LOCAs conforms to specific acceptance criteria. ECCS cooling performance must be calculated in accordance with an acceptable Evaluation Model (EM) and must be calculated for a number of postulated LOCAs of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated LOCAs are evaluated. The EM must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during the postulated LOCA.
- 33. (Harvey) The 10 C.F.R. § 50.46 acceptance criteria are:
- The calculated maximum fuel element cladding temperature
("peak cladding temperature" or "PCT") shall not exceed 2200'F.
- The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation. If cladding rupture is calculated to occur, the inside surfaces of the cladding shall be included in the oxidation, beginning at the calculated time of rupture.
- The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
4 "Blue Ridge Environmental Defense League's Response to Duke Energy Corporation's First Set of Interrogatories and Requests for Production of Documents," April 14, 2004, at3.
12
i2ACl
- Calculated changes in core geometry shall be such that the core remains amenable to cooling.
- After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
- 34. Ilarvey) Appendix K defines the required and acceptable features of the EM to be utilized in a LOCA analysis. These include:
- Sources of heat during the LOCA
- Swelling and rupture of the cladding and fuel rod thermal parameters
- Blowdown phenomena
- Post-blowdown phenomena; heat removal by the ECCS
- 35. (hanvey) The results of the LOCA analyses define the "LOCA limits" (allowable core power peaking limits) which will be employed to satisfy the criteria specified in 10 C.F.R.
§ 50.46.
B. Overview of a Pressurizedmater Reactor (PITR) Design Basis LOCA
- 36. (Dunn) A large break LOCA evolves through three phases. A realistic (best estimate) analysis of the event would consider these phases as recognizable, but overlapping or mixed. However, for conservative deterministic approaches, such as that used by AREVA in its LAR Appendix K LOCA analyses for MOX fuel lead assemblies, the phases are treated as distinct. Figure 1 shows the cladding temperature response of the core hot pin during a design basis LOCA as calculated using the AREVA Appendix K methodology. This particular example is the MOX fuel case with the highest PCT (see Table Q14-1 of Exhibit 2).
13
2-(20 Figure 1 Catawba 1IOX LOCA Temperatures 220 ------- -- - - - -
21000 t X P eakaaddrgtffmh cladTenperature at I: 1/=120Ruptwe =____ Location______ ___
8000 0 200 300 400 500 600 Time. s 14
2(iz2
- 37. (Dunn) The first phase - blowdown - initiates with the opening of a break in the reactor coolant system (RCS) and lasts until the RCS has depressurized to an approximate equilibrium in pressure with the reactor building. For the Catawba LOCA shown on Figure 1, the blowdown extends from accident initiation (0 seconds) to 25 seconds. During this phase, the core is cooled primarily by the flow of water through the reactor coolant system and out of the break. With the loss of the pressure boundary, the reactor coolant flashes or boils to steam within the RCS. These areas of steam (referred to as voids) reduce the moderating ability of the coolant, effectively shutting down the nuclear reaction and reducing heat production (power) to decay heat levels. Because the power within the nuclear material has dropped significantly, the fuel pellet temperature decreases. However, the increasing voids lead to departure from nucleate boiling on the exterior of the fuel pins, dramatically reducing heat transfer from the pins. As a result, the cladding heats up quickly. As shown in Figure 1, the initial cladding temperature peak occurs as the temperatures of the cladding and the pellet approach each other. At this point the capability to transfer heat from the pellet through the cladding to the coolant is approximately equal to the decay heat rate. This typically occurs at 4 to 10 seconds after accident initiation with temperatures in the 1400 to 1800'F range. Following this initial peak in cladding temperature, the continued reduction in power allows the cladding and fuel temperatures to decrease until just before the end of the blowdown phase.
- 38. (Dunn) During blowdown, the injection of emergency coolant has not provided substantial benefit because the RCS continues to expel coolant through the break. By the end of blowdown, the RCS, as modeled in deterministic Appendix K evaluations, is nearly empty, heat transfer has further degraded due to the lack of liquid, and the cladding and fuel begin to increase in temperature. The next phase - the refill phase - is defined as the time required to inject 15
242.2w sufficient emergency coolant to fill the lower head and lower plenum of the reactor vessel and re-initiate flow of coolant into the reactor core. In Figure 1, the refill phase lasts from 25 seconds to 38 seconds. During refill the coolant is pooling below the core in the lower head, and there is no mechanism for flow up through the core. Fuel pellet and cladding heatup is nearly adiabatic during refill, with the temperatures rising perhaps 200-300'F before the end of the phase.
- 39. (Dunn) When the coolant in the reactor vessel rises to the bottom of the core, the third phase initiates -reflood. In Figure 1, the reflood phase initiates at 38 seconds and extends for the remainder of the accident simulation. During the reflood phase water in the bottom of the core is heated to boiling, creating an upward two-phase flow that restores cooling to the upper portion of the core. Early in reflood, the coolant level in the core is low with a low boiling rate, so there is insufficient two-phase flow in the upper regions of the core to remove decay heat. As a result, the upper regions of the cladding continue to increase in temperature during the initial stage of reflood.
- 40. (Dunn) It is during this early reflood phase that cladding temperature and stress develop to the point wvlhere cladding swelling and rupture are possible. The RCS is at a very low pressure, maximizing the stress pushing the cladding outward. As the temperature increases, the conditions for cladding ballooning and rupture could be met. At the rupture, the cladding would pull away from the fuel pellet to a diameter greater than the original cladding diameter in that location. Figure 2 shows two views of a sample of cladding from a LOCA test at Argonne National Laboratory. At locations A and D, which are removed from the ballooning position, the cladding has not experienced significant strain. Closer to the rupture, location B, the cladding experiences higher strain. At the rupture, location C, the cladding undergoes the largest strain and is expanded outward asymmetrically around the rupture.
16
Figure 2 A BC D A B C
- 41. (Dunn) Several important heat transfer processes ensue at the ruptured location on the pin. The pellet-to-cladding gap is increased significantly, reducing the pellet-to-cladding heat conductance and lowering, for the moment, the energy transferred from the fuel to the cladding. The cladding surface area increases, substantially increasing the rate of energy transport from the cladding to the coolant. Furthermore, the intrusion of the cladding into the coolant flow channel creates flow turbulence and mechanical interactions between entrained water droplets and the cladding, both of which act to improve the rate of heat transfer from the cladding to the coolant. The net effect, shown in Figure 1 in the plot of clad temperature at the ruptured location (or "node"), is a short-term cooling of the cladding at the rupture location.
Eventually, the cladding at the ruptured location resumes heatup at an equivalent rate to the nearby unswelled portions of the hot pin.
- 42. (Dunn) As the water level rises in the core, boiling increases, upper region two-phase flow increases, and the heat transfer approaches the ability to remove decay heat. At this 17
point, the last cladding temperature peak occurs. For deterministic Appendix K evaluations, this reflood cladding temperature peak is generally the highest of the entire event. Thereafter, cooling in the upper core continues to improve, and the cladding and pellet temperatures decrease gradually until the core is quenched (covered by a two-phase mixture) and long term cooling thereby established.
C. Overn'iew ofMOX FtelLOCA Analysis
- 43. (Dunn) Except for the MOX fuel-related changes discussed below, the LOCA analysis cases discussed in Duke's MOX fuel lead assembly LAR and in the associated responses to NRC RAls were performed using the NRC-approved AREVA deterministic LOCA EM. The EM was approved by the NRC for application to Westinghouse-designed four loop PWRs (such as Catawba) with LEU fuel. To apply the EM to MOX fuel, a review of potential differences between LEU and MOX fuel that could affect the calculation of LOCA results was conducted and, where necessary, changes were made to the evaluation techniques. These changes and the justification for the changes are also described in Section 3.7 of the LAR.
- 44. (Dunn) The LEU-based EM is easily adapted to the prediction of LOCA behavior for MOX fuel. This is because, with the exception of decay heat and initial fuel enthalpy (heat content of the fuel), the LOCA response is primarily controlled by system phenomena.
NUREG/CR-5249' captures the importance of the various phenomena impacting the results of a LOCA. The identified controlling phenomena are listed in 13 categories. Twelve of the phenomena categories are related to the reactor system and are independent of the fuel type.
Only one category includes phenomena that relate to the nuclear material (e.g., MOX vs. LEU).
5 NUREG/CR-5249, EGG-2552, "Quantifying Reactor Safety Margins - Application of Code Scaling Applicability, and Uncertainty Evaluation Methodology to a Large-Break Loss of Coolant Accident" (December 1989).
18
Within this one category, fuel pellet enthalpy at operating conditions, fuel decay heat, gap conductance, and cladding oxidation arc listed as significant contributors. Only the first three of these relate to the nuclear material. Thus, most of the approved EM is clearly already appropriate to the modeling of MOX fuel, with no adjustments.
- 45. (Dunn) That being said, some characteristics of the nuclear material that relate to the generation and conduction of energy from the pellet through the cladding to the coolant have the potential to affect the results of the Appendix K LOCA calculation. To account for these characteristics, AREVA reviewed MOX fuel characteristics for potentially LOCA-significant phenomena. Areas of concentration during the review were reactor kinetics, decay heat, and thermal and mechanical properties, including MOX pellet enthalpy. The LAR, Section 3.7.1.1, documents: (i) the phenomena that were given specific consideration for the MOX fuel lead assembly LOCA evaluation; and (ii) the disposition that was given to each phenomenon.
- 46. (Dunn) For the MOX lead assembly core at Catawba, the reactor kinetics will be dominated by the LEU fuel because the MOX lead assemblies comprise only 2 percent of the core. However, those differences that do exist, lower beta effective and a more negative moderator coefficient in MOX fuel, will act to reduce the neutron power generation in the MOX fuel relative to the surrounding LEU fuel. As is discussed in Section III.D of this testimony, a MOX/LEU decay heat comparison indicates a beneficial result for MOX fuel. For the first several thousand seconds, well beyond the time of PCT following a LOCA, the decay heat for a MOX fuel assembly, if operated at the same power as an LEU assembly, would be lower than the corresponding LEU fuel assembly. For both of these phenomena, however, the MOX assemblies were conservatively evaluated using neutron power-related characteristics and decay heat characteristics appropriate for LEU assemblies.
19
zi(k
- 47. (Dunn) Important properties associated with the conduction of energy out of the pellet are the pellet thermal properties and the fuel to cladding gap coefficient (gap conductance).
Within the pellet, only the thermal conductivity differs in any substantive way between MOX and LEU pellets of the lead assembly design. The difference, though slight, acts to increase the initial enthalpy of the MOX pellet. Therefore, the MOX thermal conductivity from COPERNIC, approved for MOX applications, was incorporated into the LOCA evaluations. There exists no significant difference in heat capacity between MOX and LEU. Composition of the gas in the pellet-cladding gap can affect the gap coefficient (thermal conductance between the pellet and the cladding). As the fuel is used, higher fission gas release rates for MOX fuel can impact the gap coefficient. To provide for this, the AREVA LOCA analysis was based on MOX-specific gap gas compositions obtained from COPERNIC.
- 48. (Dunn) As discussed above, the MOX fuel lead assemblies wvill use M5 TNI cladding. For many years Zircaloy-4 was the standard cladding material used worldwide in PWRs. Over the past decade the nuclear industry has been moving toward the use of advanced cladding materials instead of Zircaloy-4. M5TM is an advanced cladding material that provides superior corrosion resistance, lower irradiation growth, and better ductility retention than Zircaloy-4. M5TTM has been reviewed by the NRC and approved for use as a cladding material.
It is currently being used in fuel operated in nine United States reactors.. M5T` 1 is also being used in foreign nuclear power reactors, including MOX fuel applications.
- 49. (Dunn, Harvey) ZIRLOTNI is another recently-developed advanced alloy cladding material. Westinghouse developed ZIRLOTm in the same time frame that AREVA developed M5T". ZIRLOTh! is used in the predominant co-resident fuel at Catawba, the Westinghouse RFA 20
?2A Z design. Zircaloy-4, MSTNI, and ZIRLOTh are all predominantly zirconium, and the unirradiated properties of the materials in cladding applications are very similar.
- 50. (Dunn, Nesbit) The use of M5T`I for MOX fuel assemblies is not a difference "in MOX and LEU fuel behavior" as was specifically called out in Contention I. LEU fuel assemblies commonly use advanced cladding materials such as M5Tnl and ZIRLOThI. However, BREDL has alleged that the characteristics of M5T` may exacerbate MOX/LEU differences, so M5Tn cladding performance is also addressed in this testimony.
- 51. (Dunn) The modeling of M5TM cladding within the LOCA EM is specific to that material. Many characteristics (e.g., thermal conductivity, heat capacity) are very similar to Zircaloy-4. The most significant difference (relevant to a LOCA event) lies within the cladding swelling (ballooning) and rupture model.
- 52. (Dunn) Cladding ballooning occurs when the cladding is heated to elevated temperatures with an internal pressure that exceeds the external pressure (coolant pressure).
Depending on the extent of the ballooning, the cladding may rupture, relieving the driving pressure difference and terminating the ballooning effect for that rod. The cladding ballooning and rupture effects are modeled in the EM.
- 53. (Dunn) The modeling of rupture involves the prediction of the rupture temperature of the cladding as a function of cladding stress and the prediction of cladding strain as a function of the rupture temperature. Although a specific M5Th¶ rupture temperature versus stress correlation has been developed, the predictions do not differ significantly from the Zircaloy-4 correlations. The cladding strain, however, differs in magnitude, because M5T`rI is slightly less ductile prior to irradiation than Zircaloy-4. In other words, unirradiated MSTNI will 21
ZfS strain (deform) less before rupture than Zircaloy-4, for the same applied stress. The temperature distribution of strain also differs.
- 54. (Dunn) The established rule for deterministic LOCA calculational approaches, Appendix K to 10 C.F.R. § 50.46, requires that the degree of cladding swelling and incidence of rupture not be underestimated. Because all claddings tend to (i) embrittle with irradiation, and (ii) potentially accrue added strength due to pellet-cladding bonding, deterministic LOCA evaluation models use unirradiatedcladding properties to maximize the predicted strain. This same approach was incorporated within the LOCA evaluations for MOX fuel presented in the LAR.
- 55. (Dunn, Hanrey) LOCA calculations are performed at a variety of plant conditions to establish LOCA limits (allowable peaking) which ensure compliance with the criteria of 10 C.F.R. § 50.46. For the MOX lead assembly calculations, the most severe results occur for axial power peaks at the 10.3 ft elevation (i.e., for power distributions that are peaked close to the top of the 12 foot active core). The most limiting results and the corresponding acceptance criteria are shown below, with the acceptance criteria in parentheses:
Peak Cladding Temperature (PCT) 2019.5 0 F (22000 F)
PCT at Ruptured Location 1750 0 F (22000 F)
Local Oxidation 5.2% (17%)
Total Core Oxidation 0.4% (1%)
This most limiting case was evaluated at a burnup of 30 GWD/t (see Table Q14-1 of Exhibit 2).
The fuel and cladding temperatures from this case are shown in Figure 1.
- 56. (Dunn, Harvey) The maximum calculated cladding strain for this case is 51 percent and the flow blockage due to this ballooning is 52 percent of the coolant channel 22
surrounding the hot pin. This amount is well within the coolable geometry limit (specified by the AREVA LOCA evaluation model) of 90 percent.
- 57. (Dunn, Hanrey) Long term cooling is not directly assessed by the LOCA calculation. Long term cooling is a system determination requiring (i) that a cooling path is available from pumped injection and (ii) that the flow pattern within the reactor vessel is such as to prevent the precipitation of boric acid. Both of these factors are dependent on core wide phenomena and neither factor changes with the inclusion of four MOX fuel lead assemblies within the core. The conclusion that the Catawba operating procedures meet these conditions is established in the Catawba Final Safety Analysis Report (FSAR) and is not altered with four MOX fuel lead assemblies.
- 58. (Dunn, Harvey) The analyses in LAR Section 3.7.1 (Exhibit 1) show a difference of less than 40'F between the MOX fuel and LEU fuel comparison cases, with MOX fuel being the higher value. This is not a significant difference in a LOCA PCT response. The MOX and LEU results are thus essentially the same. As discussed further in Section III.D of this testimony, the MOX results are actually conservative because AREVA adopted the simplified approach of using LEU characteristics when it was obviously conservative to do so for MOX fuel. In other words, credit was not taken for beneficial aspects of MOX fuel.
- 59. (Nesbit) The NRC Staff reviewed the LOCA analyses of the MOX fuel lead assemblies. Section 2.4.1 of the April 5, 2004 NRC Safety Evaluation Report 6 states: ". . . the NRC staff concludes that the effect of four MOX LTAs has been conservatively evaluated and has been demonstrated to be in compliance with the requirements of 10 C.F.R. § 50.46."
6 R.E. Martin (NRC) to H.B. Barron (Duke Energy), "Safety Evaluation for Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies" (April 5, 2004) ("SER").
23
D. Conservatismsin the LOCA Analysis
- 60. (Dunn) The conservatisms incorporated in the MOX LOCA evaluation performed for the LAR include conservatisms (or "margin") inherent in the deterministic Appendix K approach and specific conservatisms that we employed in the MOX fuel LOCA analysis.
- 61. (Dunn) The major conservatisms included by regulation within the approved Appendix K method are:
- A decay heat approximately 10 percent higher than the 95 percent confidence level and 30 to 35 percent higher than the nominal expectation.
- The use of the Baker-Just oxidation (metal-water reaction) correlation, which incorporates a reaction rate 30 to 50 percent above the available data.
- The use of the highest allowable local power level. Plants actually operate with peaking significantly lower than the limiting values defined to be acceptable by the LOCA analyses. This conservatism is quantified for representative Catawba operation cycles in Section IV of this testimony.
- The retained use of full double-ended break areas. The NRC is currently reviewing the use of substantially reduced break areas based on the risk of the large break occurring. Leak before break analyses indicated that the probability of an instantaneous double-ended break is very low.
- Limiting ECCS bypass assumptions. Nearly all of the ECCS water delivered to the reactor coolant system prior to the end of blowdown is assumed to be bypassed, and therefore unavailable for core cooling.
- 62. (Dunn) The accrued margin provided by all of the conservatisms identified above, in terms of a change in the PCT, can be estimated from applications of "best estimate" models that include a realistic treatment for most of the conservatisms. In a recent analysis, AREVA performed both Appendix K calculations and best estimate calculations on a three-loop NVestinghouse-designed plant. The Appendix K PCT obtained was between 2100 and 2200'F, 24
2,i t.
while the nominal or best estimate PCT lay below 1500'F. Because even the best estimate approach retains some conservatisms or biases (e.g., use of full double-ended break areas), a judgment can be made that the conservatisms in the Appendix K methods provide more than 600IF margin to best estimate LOCA results.
- 63. (Dunn, Nesbit) The best example of additional conservatisms incorporated in the specific MOX fuel LOCA calculations for the LAR is the use of the LEU decay heat model, which was referred to above. A comparison of LEU and MOX decay heat on a one-to-one basis (Figure 3 below) would indicate that MOX falls 3 to 5 percent below the LEU curve during the time period of importance. Because the PCT is strongly influenced by the decay heat, this effect alone is estimated to be a conservatism of up to 750 F on PCT.
Figure 3 MOXILEU Decay Heat Ratio 1.000 r 1iI I I I 0.980 0.960
.4 0 0.940 0.920 I ===
I I W 0GW2 0.900 0.880 0 5 10 15 Time (minutes) 20 25 a30 25
- 64. (Dunn) There are also conservatisms in the MOX fuel LOCA evaluation in the use of LEU fuel neutronics coefficients. For example, the void reactivity coefficient for the MOX assemblies will be more negative than that for the LEU assemblies. Therefore, a small additional depression of the neutron power in the MOX assembly will occur as voids shut down the reactor during early blowdown. Because no multidimensional analysis of the core during LOCA has been conducted, the magnitude of the neutronic effect (beneficial for MOX fuel) has not been determined.
- 65. (Dunn) As is discussed further below, BREDL points to one factor not included in the LOCA model and cites that as a non-conservatism: fuel relocation. Fuel relocation is not modeled for either LEU fuel or MOX fuel. However, the possible impact of fuel relocation on compliance with the acceptance criteria must be considered in the context of all of the conservatisms already built into the models and the criteria. Fuel relocation is discussed in further detail in Section V of this testimony.
IV. DIFFERENCES IN LEU AND MOX FUEL BEHAVIORS
- 66. (Nesbit) Contention I focuses on "differences in MOX and LEU fuel behavior" and "the impact of such differences on LOCAs." In an attempt to obtain some understanding of what those asserted differences and impacts might be, beyond those already addressed in the LAR, Duke posed interrogatories to BREDL. In BREDL's response 7 to Duke Interrogatory 4, BREDL listed 19 MOX fuel behaviors that "... will affect a LOCA scenario or analysis in a manner different than LEU fuel." In response to Duke Interrogatory 5, BREDL listed 8 MOX fuel cladding (M5T1I) behaviors that ". . . will affect a LOCA scenario or analysis in a manner different than LEU fuel cladding behavior." BREDL added the statement that "BREDL does not 7 "Blue Ridge Environmental Defense League's Response to Duke Energy Corporation's First Set of Interrogatories and Requests for Production of Documents," April 14, 2004.
26
assert that all of these behaviors necessarily will have a significant impact." In fact, BREDL did not highlight any behavior as having a significant impact. In this section of the testimony, Duke examines each behavior listed by BREDL in some detail. Duke has determined that each behavior is either inapplicable to the design basis LOCA calculation, already addressed in the LOCA evaluation, clearly insignificant in terms of impact on the LOCA calculation, and/or specifically addressed later in the testimony.
A. Fuel-Related "Diferences" Item 1: Rod centerline temperature as afifnction of rodpower-
- 67. (1Iarvey, Dunn) Fuel rod centerline temperature is not a direct input in the LOCA analysis. The related input parameter is the fuel rod stored energy (enthalpy) or fuel pellet volumetric average temperature. This parameter is obtained using an NRC-approved fuel performance computer code/methodology. The LOCA analysis calculations model the fuel pellet properties (fuel dimensions, conductivity, and heat capacity); model inputs are adjusted to match the fuel rod stored energy determined by the approved fuel performance methodology.
For the MOX fuel LOCA analysis, the COPERNIC fuel performance code is used. COPERNIC is also used in the LEU fuel LOCA comparison analysis presented in the LAR. The COPERNIC computer code considers the difference in fuel properties between the LEU pellets and MOX pellets, and therefore adequately represents the differences in fuel centerline temperature between the MOX and LEU fuel rods.
- 68. (Harvey, Dunn) Figure 4 is taken from Figure Q21-2 in Duke's November 3, 2003 RAI response (Exhibit 2). Figure 4 shows that the MOX and LEU fuel pellet temperature profiles at the beginning of life are very close at initiation of a LOCA. This is because of the fundamental similarity between MOX and LEU fuel in the areas of thermal conductivity and pellet radial power, as shown by Figures 5 and 6, which also are reproduced from Figures Q21-1 27
and Q7-1 in the November 3, 2003 RAI response (Exhibit 2). There is no significant difference between fuel types due to the initial temperature profile.
Figu re 4 MIOX and LEU Fuel Pin Tcmperaturc Profile Comparison at Loss of Coolant Accident Initiation (Figure Q21-2) 3500 ..... I . .._
3000 - - - - -I - - - -
2500 - - - *'. *I
. - I I 2000 - - -- -I 1.000 - E I1-
- - 0 m661I 500_ _7 0 1 2 3 4 5 6 7 F 9 10 Radial MeshPoint (#)
28
Figure 5 Thernmal Conductivity Comparison for MIOX and LEU Fuel (Fuel porosity of 0.0479)
(Figure Q21-1) 9.OE-04 I I I I I I I I I I I I I I I I
, , I O n ,LEU 4 O'GWd'mtU I I I 8.OE04 - J - - - - - - - -,
LEU @ 40 GWd'mtU C - e- 4wwOXt @ OGWdmbm t I A 7.1-044 _ r l _ T_ __ _ _I_ _ _I_ _ _ I
\l " l - K - 4 w-.%MoOX@ 40
' I lI O GW'dmthm I
- e.6.0E44 I I -I - - - - - - - - - - -- _ _
U I. .
g 5.E0E44 S 4.0E-44 _ __ _ _ __ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ ___ I __________I__ S I 3.OE-04
- I I I - - - - - - - I I - - - - --lI - - --
2.0E-04 0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000 5500 Fuel Temperarure (F) 29
U2-CA Figure 6 Pellet Radial Power Profiles 0.5 Go 'D/t Item 2: Magnitude, timing, radionuclide composition and power history dependence offission gas release (duringnormal operation)-
- 69. (Harvey, Dunn) The fission gas release from the fuel pellet has two effects. First, the fission gas that is released to the pellet-cladding gap degrades the gas conductivity, which can result in a higher temperature drop across the gap. Second, the released fission gas also increases the fuel rod internal pressure. The internal pressure has a mild effect on gap conductance, and also impacts the fuel rod rupture, swelling, and flow blockage during the LOCA event. Fission gas release is modeled in the approved fuel performance code specifically 30
aI%2 for both MOX and LEU fuel. For the MOX fuel LOCA analyses, the fission gas release results from COPERNIC are included as input into the LOCA calculations. COPERNIC explicitly models MOX fuel effects on fission gas release. As mentioned above, COPERNIC fission gas release results for LEU are used as input to the LEU comparison LOCA analysis presented in the LAR. Therefore, the fission gas released during normal operation is explicitly modeled on a fuel type-specific basis (MOX or LEU) in the LAR LOCA analysis. In BREDL's response to Duke Interrogatory 32.f,8 BREDL acknowledged that this behavior is not relevant to Contention I.
Item 3: Fuel-cladinteraction-
- 70. (Harvey) This phenomenon is addressed in detail in Section V.E of this testimony.
Item 4: Peakcladding temperature(PCT)-
- 71. (Harvey) The PCT is the principal result or output of the LOCA analysis. The PCT acceptance criterion (<2200'F) is based on the performance of the cladding material, not the fuel pellet material. The MOX fuel-specific analyses summarized in Exhibit 2, Table Q14-1, demonstrate that MOX fuel meets the PCT acceptance criterion.
Item 5: Oxidation potential-
- 72. (Harney) The oxidation of the fuel pellet is not modeled in the LOCA analysis method. Therefore, this effect does not impact the LOCA analysis for either MOX or LEU fuel.
Item 6: Linearheat generationrate-
- 73. (Harvey) The linear heat generation rate (LHGR) is an input to the LOCA analysis which directly affects the sources of heat in the LOCA analysis and the resulting PCT.
The fuel initial stored energy and the decay heat in the limiting fuel rod (hot rod) are both influenced by the assumed LHGR (or peaking) value. LHGR is explicitly modeled as an input to 8 "Blue Ridge Environmental Defense League's Response to Duke Energy Corporation's Second Set of Interrogatories and Requests for Production of Documents," June 8, 2004.
31
the LOCA analysis, irrespective of fuel rod type (MOX or LEU). Thus, there is no difference in the representation of the LHGR value for a MOX fuel rod compared to a LEU fuel rod. As described in Section III of this testimony, for a given LHGR, the fuel initial stored energy was explicitly modeled for MOX fuel, and decay heat was conservatively assumed to be the same as LEU fuel. Thus, the impact of LHGR on fuel initial stored energy and decay heat are addressed appropriately. Furthermore, MOX fuel lead assembly core designs will ensure that the MOX assemblies are not the peak power assemblies (highest LHGR) in the core during the nominal depletion. The key point is that the LOCA limits (i.e., the LHGR inputs to the LOCA analyses that produce results within the PCT acceptance criterion) are established as operating limits to ensure that the Catawba unit will operate at acceptable LHGRs for both fuel types (MOX and LEU).
- 74. (Harvey, Nesbit) There is substantial margin between the actual plant operating conditions and the assumed LOCA analysis peaking (FQ) values that are used to establish operating limits. FQ is monitored monthly through core flux maps (three-dimensional measurements of the core power distribution) and compared to the LOCA analysis FQ values that define acceptable operation. Figure 7 presents the Catawba Unit I Cycle 14 measured FQ values and the FQ LOCA analysis values for the resident LEU fuel. (Note that these measured FQ values do not include an allowance for measurement uncertainty.) It is apparent that actual core operating conditions are far removed from the limiting LHGR values assumed in the LOCA analyses. While Figure 7 is for an all-LEU core, similar margin is expected to be present for MOX fuel.
32
Figure 7 2.60 2.40 4 4 U.
2.20 I 4 4 4 0
1U 2.00 I- 4 4 -t Ui.
0)
Cr 1.80 0 , ,
a.
0 g~~. .
- j. , , #*i , , . l _ a 1.60 .
0 a.
C
- C1Cl4 Core Max Fq a.
1.40
- LOCA Fq Limit 1.20 1.00 0 100 200 300 400 500 Cycle Exposure, EFPD 33
Item 7: Magnitude offission product release duringgap releasephase; Item 8: Magnitude of volatile fission product release during early in-vessel core degradation; Item 9: Rate of volatile fission product release during early in-vessel core degradation; Item 10: Magnitude ofseni-volatilefission product release during early in-vessel core degradation; Iten 11: Rate of semi-volatile fission product release during early in-vessel core degradation; Item 12: Magnitude of low-volatile fission product release during early in-vessel core degradation; Item 13: Rate of low-volatile fission product release during early in-vessel core degradation; Item 14: Radionuclide inventory offiuel-
- 75. (Harvey) The purpose of the design basis LOCA ECCS analysis is to provide reasonable assurance that the plant ECCS will prevent significant core damage in the event that a LOCA occurs. Significantly, items 7-13 are phenomena associated with hypothetical severe accidents and are not relevant to the design basis LOCA ECCS analysis, because the design basis LOCA by definition does not involve the type of core damage associated with severe accidents.
Item 14, the radionuclide inventory of the fuel rod, is not directly modeled in the design basis LOCA analysis other than as it relates to the fission gas released from the fuel pellet discussed earlier. As noted above, dose analyses (either LOCA doses or theoretical severe accident doses) are not included within the scope of Contention I.
Item 15: Radialpoterdistribution; Item 16: Axial power distribution-
- 76. (Harvey) 10 C.F.R. Part 50, Appendix K, Section I.A, requires that the maximum peaking factors allowed by the Technical Specifications (at Catawba, these factors are currently 34
contained in the Core Operating Limits Report) be bounded, and that a range of power distribution shapes and peaking factors representing power distributions that may occur over the core lifetime be studied. The selected combination of power distribution shape and peaking factor should be the one that results in the most severe calculated consequences for the spectrum of postulated breaks and single failures that are analyzed. This requirement is met for the LOCA analysis for both the LEU and the MOX fuel as the results of the LOCA analysis define the allowable peaking factors ("LOCA limits") that satisfy the acceptance criteria specified in 10 C.F.R. § 50.46.
- 77. (Nesbit) Catawba continually monitors the axial flux difference (AFD), defined as the power in the top half of the core minus the power in the bottom half of the core, divided by total core power.
AFD = Power,5,, - PowverbO..O.
Polver, 0 1 ta The core design process determines the AFD that would be required for a given power level to produce the peaking factor FQ that was assumed in the LOCA analysis. Figure 8 illustrates the margin between (i) the actual measured AFD for Catawba, Unit 1, Cycles 14 and Cycle 15 and (ii) the minimum AFD that would be required to produce core peaking at the peaking limits established by the LOCA analysis. It is evident that actual core operating conditions are far removed from the limiting conditions assumed in the LOCA analysis. Note that actual plant operating limits are more restrictive than the zero LOCA margin AFD values shown on the figure. This is due to (i) incorporation of allowances for uncertainty in the plant operating limits and (ii) the fact that other parameters (e.g., steady-state departure from nucleate boiling ratio) are more restrictive than LOCA in some cases, and therefore set some portions of the plant operating 35
limits. While the data shown below reflect cores comprised of all LEU fuel, similar margin is anticipated for cores containing MOX fuel lead assemblies.
Figure 8 110 2 10 _ _ 0__
60 50 ~* ClCl5MeasuredAFD e 40 40 - ClCl4MeasuredAFD .
IL 30 - -- AFD at Zero LOCA Margin
- 20 - AFD at Zero LOCA Margin
-50 -40 .30 -20 -10 0 10 20 30 40 50 Percent Axial Flux Difference 36
Item 17: Potentialforfiel crumbling and relocationfollowing clad ballooning; Item 18: Particlesize distributionoffiuel pellet fragmnents as afumnction of burnup; Item 19: Characteristicsoffuel relocation (filling ratio, increase in local linear power density) -
- 78. (Hanrey) BREDL has presented no evidence or basis for a postulated difference in the characteristics of fuel relocation between MOX and LEU fuel. This issue is discussed further in Section V of this testimony.
B. Cladding-Related"Differences"
- 79. (Nesbit, Dunn) It should be noted that the cladding material to be used for the MOX fuel, M5W T , is being used extensively with LEU fuel worldwide. M5TM is also being used with MOX fuel in Europe. Zircaloy-4, the traditional light water reactor (LWR) cladding material, has been used with both MOX fuel and LEU fuel worldwide.
Item 1: Extent of clad ballooning and impact onfirel relocation-
- 80. (Harvey) Cladding ballooning occurs when the cladding is heated to elevated temperatures with an internal pressure that exceeds the external pressure (coolant pressure). The deformation affects the cladding temperature, coolant flow (blockage), rod internal pressure, cladding surface area, and the pellet-to-cladding gap. As discussed above, all of these effects are modeled in the evaluation models used for both Zircaloy-4 cladding and M5TM cladding. The cladding ballooning models used for each cladding type are developed using cladding-specific experimental data. Both MOX and LEU LOCA analyses use unirradiated cladding properties (be it M5TM or Zircaloy-4) that maximize the extent of the cladding ballooning. Cladding ballooning effects are discussed further in Section V.C of this testimony.
37
Item 2: Fuel-cladinteraction-
- 81. (Harvey) This effect is not modeled in Duke's LOCA analysis. It is discussed further in Section V.E of this testimony.
Item 3: Peak clad oxidation (outer surface);
Item 4: Peak clad oxidation (inner surface) -
- 82. (Harvey) An exothermic reaction occurs when zirconium and water come into contact. This process occurs at normal operating temperatures, but the reaction rate is very small. The reaction rate increases as the temperature of the cladding material increases. 10 C.F.R. Part 50, Appendix K requires that this reaction rate be calculated during a design basis LOCA using the "Baker-Just" rate equation. The Baker-Just equation was chosen since it resulted in a conservative calculation, over-predicting the amount of oxidation over the temperature range of interest in LWR LOCA analyses. The data used to develop the Baker-Just rate equation are based on Zircaloy-4 cladding. Oxidation experiments have been performed for M5TM cladding, and those data show that the Baker-Just model also bounds the M5Tm data. The MOX LOCA analyses have been performed using the required Baker-Just model for oxidation of both the outside and inside cladding surface. These analyses demonstrate that the MOX fuel with M5TM cladding meet the 10 C.F.R. § 50.46 criteria with respect to peak and total core cladding oxidation.
Item 5: Hydrogen uptake-
- 83. (Harvey, Dunn) Hydrogen uptake into cladding occurs during normal plant operation. The uptake of hydrogen will embrittle the fuel rod cladding - decreasing its ductility. Ductility is important in a LOCA event to ensure that the cladding remains intact after core quench. Hydrogen uptake affects both LEU fuel and MOX fuel cladding, and is observable in both M5T and Zircaloy 4. Data for M5TM cladding have shown that the uptake of hydrogen is 38
more than a factor of five lower than that observed in Zircaloy-4. Given the lower hydrogen uptake, the LOCA performance of M5TM should be better than Zircaloy-4.
Itein 6: Loss of ductility (as measured by ring compression tests) as a futnction of clad oxidation and surface condition,for all burnups -
- 84. (Hartey, Dunn) Ductility is important in the LOCA event to ensure the cladding remains intact following core quench. M5TM retains its ductility better with irradiation than Zircaloy-4. With respect to cladding integrity, the important 10 C.F.R. § 50.46 acceptance criteria are the PCT limit of 2200°F and the 17% limit on local oxidation. To ensure that this limit remains applicable to the M5T' cladding, a series of high temperature, highly oxidized cladding tests were performed. The tests showed that the 2200'F PCT and 17% local oxidation limits specified as the acceptance criteria apply to M5TM cladding as well as to Zircaloy-4.
Iteiz 7: Reaction wit/l fission product releases (especially tellurium) -
- 85. (Harney, Nesbit) One difference between M5TM cladding material and Zircaloy-4 is the tin content (M5TM has none). The lower tin content in M5SI cladding has been postulated to contribute to higher tellurium releases during hypothetical severe accidents. In BREDL's response to Duke Interrogatory 33.f,9 BREDL acknowledged that this question is relevant only to source term issues that were to have been considered in Contention II, which has been dropped.
This phenomenon is not within the scope of Contention I.
Item 8: AMaximumnflow blockage consistent with core coolability-
- 86. (Hanrey, Dunn) Assembly flow blockage models are developed from single-rod burst (rupture) and pre-rupture strain tests. Bundle effects are included in the flow blockage determination through the use of multi-rod tests or through computer code simulation of the 9 "Blue Ridge Environmental Defense League's Response to Duke Energy Corporation's Second Set of Interrogatories and Requests for Production of Documents," June 8, 2004.
39
bundle. For Zircaloy-4 the single rod strain data is obtained from NUREG-0630 data.10 M5TM cladding has different high temperature creep characteristics and different a? 13 transformation temperatures than Zircaloy-4 cladding. These differences affect the cladding burst strain and resulting flow blockage, as discussed further in Section V.C of this testimony. For this reason, the NUREG-0630 data is not used to develop M5T M flow blockage models.
- 87. (Harvey, Dunnl) Tests were conducted at the CEA EDGAR facility in Saclay, France to determine the M5TM characteristics relating to maximum assembly flow blockage.
Using these data, AREVA developed new ballooning and flow blockage models for M5TM cladding similar to the methodology developed in NUREG-0630 for Zircaloy4 cladding. This is documented in an AREVA topical report." The NRC Staff has reviewed the AREVA approach and, in its safety evaluation for M5TM cladding, found that approach to be either as conservative or more conservative than the flow blockage model in NUREG-0630.' 2 The LOCA analysis for the MOX fuel uses the AREVA MSTM flow blockage model, which demonstrates that core geometry remains amenable to cooling following a design basis LOCA.
V. FUEL RELOCATION AND RELATED ISSUES
- 88. (Nesbit) As reframed by the Licensing Board, Contention I encompasses original BREDL Contention 10, which cited a purported failure by Duke to ". . . account for uncertainties in MOX fuel assembly behavior during Loss of Coolant Accidents." The basis for BREDL 10 D.A. Powers and R.O. Meyer, "Cladding Swelling and Rupture Models for LOCA Analysis," NUREG-0630 (April 1980).
Framatome ANP Topical Report, BAW-10227-A, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel" (February 2000).
12 S.A. Richards (NRC) to T.A. Coleman (Framatome Cogema Fuels), "Revised Safety Evaluation for Topical Report BAW-10227P: Evaluation of Advanced Cladding and Structural Material (MS) in PWR Reactor Fuel" (February 4, 2000) (see Section 5.3.4).
40
Contention 10 rested largely on a presentation by the Institute for Radiological Protection and Nuclear Safety (IRSN) to the NRC Office of Research on October 23, 2003. IRSN is an agency in the French government that conducts research and analyses related to radioactivity. It performs functions that are somewhat similar to those of the NRC Office of Research or to those of the national laboratories in this country. It is not the agency with actual safety oversight authority for French nuclear plants.
- 89. (Nesbit) This section of testimony addresses issues cited by BREDL and derived from the IRSN presentation and other sources. It should be noted, however, that IRSN was not addressing the United States MOX fuel project or the Duke LAR, and that IRSN has made no representations related to the safety of the MOX fuel lead assembly program at Catawba.
Moreover, the BREDL issues are derived primarily from BREDL interpretations of statements made by IRSN concerning possible research needs related to LOCAs and severe accidents.
A. Fuel Relocation - Descriptionand Regulatory History
- 90. (Haney) Fuel relocation during a LOCA is a phenomenon in which fuel pellets lose their integrity and fall (relocate) to a lower portion of the fuel rod. In order to have fuel relocation, two factors much be present. First, the pellet must be irradiated to significant burnup, such that irradiation-induced cracks have developed and the pellet can lose its integrity (break into small fragments) when the constraining influence of the cladding is removed. Second, the cladding must swell (balloon) sufficiently prior to rupture to provide room for the pellet fragments to crumble and fall.
- 91. (Hartey, Dunn) The potential for fuel relocation during LOCA occurs early in reflood as cladding temperatures rise to around 1500TF with fuel temperatures on the order of 16001F. At this time the cladding will expand and rupture providing gravity and vibration the 41
opportunity to pull pellet fragments down into the swelled region. Fuel relocation has been experimentally demonstrated in the PBF-LOC tests, FR2 tests, and the FLASH5 test. Most recently relocation was observed in the swelling and rupture tests at Argonne National Laboratory. Figure 9 shows a photograph of a cross-section of PBF LOC-6 rod 12 at the burst elevation which illustrates the fuel fragmentation in the presence of cladding rupture.
Figure 9 Cladding Epoxy Fuel-270' t .9o PA Burst tips Previously irradiated Rod 12, 74% total circumferential elongation, 1066 K estimated burst I.,:. he . _. -
1D-9A. ,9-w temperature 1.67 mm 180 42
- 92. (Harvey, Dunn) The concern over relocation stems from the possibility that the relocated fuel may generate too much power in a localized area, and thereby overheat the cladding. It should be noted, however, that this potential for increased heat generation is also mitigated by heat transfer enhancements that are associated with cladding ballooning. The swelling of the cladding does two things in this regard: first, it alters pellet-to-cladding gap in a manner that provides less efficient energy transport from the fuel to the cladding; and second, it provides for greater cladding surface and therefore enhanced heat transfer from the cladding. In addition, narrowing the flow channel outside the ballooned cladding increases turbulence, which further enhances heat transfer from the cladding. The FLECHT-SEASET flow blockage tests demonstrate this effect.13
- 93. (Harvey) Cladding ballooning is necessary for fuel relocation to occur. Cladding ballooning can occur only when the cladding is heated to an elevated temperature in the presence of a pressure difference across the cladding (internal rod pressure exceeding the external pressure, namely the coolant pressure). Given the LOCA response of the Catawba plant, cladding ballooning is only possible for a large break LOCA. The conservative licensing calculations for the small break LOCA events do not achieve cladding temperature in the range where cladding ballooning is possible. The maximum PCT for the small break LOCA event is 12250 F which is significantly below the temperature where cladding ballooning is calculated to occur. The importance of this result is that fuel relocation requires a very unlikely large break LOCA event.
- 94. (Harvey) The concern over fuel material relocation following a LOCA with fuel cladding ballooning was identified in the mid-1980's. It was recognized that the phenomenon 3~ NUREG/CR-3314, "PWR FLECHT-SEASET 163-Rod Bundle Flow Blockage Task Data Report" (October 1, 1983).
43
was not included in the LOCA evaluation models and, under certain assumptions, could result in a non-conservative assumption. This was a limitation in the ECCS performance analysis methods (LOCA analysis computer codes). However, at least one study by an NRC contractor organization concluded that known conservative features in the Appendix K models more than offset the effect. The NRC classified this item as Generic Issue (GI) 92 ("A Prioritization of Generic Safety Issues", NUREG-0933) and initially assigned this item a low priority based on the value/impact assessment. However, subsequent prioritization placed this issue in the "drop" category.
- 95. (Harvey) A combined meeting of ACRS subcommittees discussed the postulated relocation effect in a meeting on November 15, 2001. The purpose of this meeting was to discuss relaxing the Appendix K decay heat assumption. In the meeting the NRC identified the fuel relocation issue as a non-conservative assumption in the approved Appendix K methods that may need to be addressed if the Appendix K decay heat assumption is relaxed. There was no indication that this issue was something that needed to be addressed for currently approved Appendix K methods. In the meeting it was recognized that European research/analysis is being conducted in this area.
- 96. (Harey) The NRC Staff is evaluating at least the possibility of changes to 10 C.F.R. § 50.46 and Appendix K based on research that has demonstrated the existence of the large safety margin (described in Section Ill.D of this testimony) between the regulatory acceptance limits and the expected plant behavior during a LOCA.14 Staff members have noted that as known conservatism is removed from Appendix K, sources of non-conservatisms also need to be considered. One of a number of potential non-conservatisms identified is the fuel 4 Thadani, A.C. (NRC) to Collins, S. (NRC), "Research Information Letter 0202, Revision of 10 C.F.R. § 50.46 and Appendix K" (June 20, 2002) ("Thadani memorandum").
44
72<15 relocation phenomenon. However, our MOX evaluations continue to use the current deterministic acceptance criteria, with the built-in margins discussed above, so the Thadani memorandum is not directly relevant.
- 97. (Harvey, Dunn) ECCS performance analyses that do not model fuel relocation show a reduction in the rate of cladding heat up at the ruptured location and possibly a reduction in the cladding temperature following fuel swelling (ballooning) and rupture. The response was also seen in FR2 test B3.115 and is illustrated in Figure 10. The cladding temperature at the location of the rupture shows a significant drop following cladding rupture. This test was performed using an unirradiated fuel rod and therefore no pellet relocation occurred.
15 LWR Fuel Rod Behavior in the FR2 In-pile Tests Simulating the Heat-up Phase of a LOCA, KfK 3346 (March 1983).
45
2XIV52*
Figure 10 FR2 Test B3.1 for Low Exposure Fuel (No Relocation) 2000.00 -
1800.00 -
1600.00 -
1400.00 _
U. 1200.00-a, 0f 1000.00 CL E
1! 800.00_
600.00 -
400.00 -
200.00 -
0.00 _
0.00 50.00 100.00 150.00 200.00 250.00 Time, s
- 98. (Harvey, Dunn) Fuel pellet relocation can increase the heat source in the location of the ballooned cladding. However, this effect is mitigated by the fact that the fuel cannot crumble, fall, and pack itself into a mass with anything close to the original pellet density (typically around 95% theoretical density). An assumption of a very high packing factor (or filling ratio) within the ballooned section is required to develop any concern about the amount of heat generation following relocation. FR2 test E-4 was conducted on fuel of sufficient burnup that relocation occurred at the time of cladding rupture. As shown in Figure l I the ruptured 46
node temperature decreases slightly as swelling and rupture develop and then follows the excursion of the non-ballooned areas of the fuel pin.
Figure 11 FR2 Test E4 for High Exposure Fuel (Relocation)
IL aR 0
E i!
0.00 iI I4 0.00 50.00 100.00 150.00 200.00 250.00 Time, s B. VERCORS Tests
- 99. (Nesbit) BREDL used an October 23, 2003 presentation by IRSN as the basis for its concerns related to MOX fuel relocation during a LOCA.16 In particular, BREDL referred to data from the VERCORS tests as supporting the BREDL contention that fuel relocation would occur at a lower temperature in MOX fuel than in LEU fuel under LOCA conditions. This 16 A. Mailliat and J.C. Mclis (IRSN), "IRSN Source Term LOCA Program in the PHEBUS Facility," Presentation to the NRC (October 23, 2003).
47
section of the testimony summarizes the VERCORS tests and explains why the data from those tests are not applicable to this contention.
100. (Nesbit) VERCORS is a series of tests conducted by IRSN and its predecessor organization on fuel undergoing thermodynamic conditions that are consistent with severe accidents. The tests involve a single fuel rod section (three irradiated pellets) in a test rig. The test apparatus controls the temperature of the fuel and provides for variable steam and hydrogen flow past the fuel rod. The tests were carried out at Cadarache in France, beginning in 1989 and concluding recently.
101. (Nesbit) The tests investigated a number of severe accident phenomena, including fuel relocation and radionuclide release. In the context of severe accidents, fuel relocation refers to the extremely high temperature phenomenon of fuel pellets losing their integrity and collapsing downward. Severe accident fuel relocation is thermally-induced; it involves liquefaction and slumping of the fuel material following an extended loss of core cooling. In the case of the VERCORS tests with MOX fuel, the relocation temperatures were in the neighborhood of 40007F or more.
102. (Nesbit) The VERCORS RT2 and RT7 tests involved irradiated MOX fuel. The thermal conditions of the VERCORS RT2 and RT7 tests were much higher than fuel temperatures experienced during a design basis LOCA. A design basis LOCA would include restoration of core cooling within a few minutes of accident initiation, based on activation of the Emergency Core Cooling System (ECCS). In a design basis LOCA, PCT is maintained below 2200°F in accordance with 10 C.F.R.§ 50.46(b). Figure 12 is a stylized representation of LOCA temperatures along with VERCORS relocation temperatures.
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Figure 12 MOX Fuel Fuel and Cladding Temperatures During LOCA Compared to Severe Accident Fuel Melt Temperatures 0 too 200 300 400 S00 600 700 S00 900 1000 Time (seconds) 103. (Nesbit) Fuel relocation during a design basis LOCA is a distinct and different phenomenon from fuel relocation during a severe accident. Fuel relocation during LOCA is a mechanically-induced phenomenon which follows cladding ballooning and fragmentation of irradiated fuel pellets. Fuel relocation during LOCA occurs at much lower temperatures, in the neighborhood of 1500-2200'F. Temperatures are far removed from liquefaction conditions experienced in the VERCORS tests.
104. (Nesbit) To our knowledge, and based upon the publicly available materials, IRSN has made no observations about the implications of the VERCORS tests on LOCA analyses or the ability to cool MOX fuel during a LOCA.
49
105. (Nesbit) To our knowledge, and based upon the publicly available materials, IRSN has made no comments on the safety of the proposed United States MOX fuel lead assembly program or the associated LAR.
106. (Nesbit) To our knowledge, and based upon the publicly available materials, IRSN has made no effort to restrict the use of MOX fuel in French reactors due to VERCORS test results. Twenty French reactors are currently using substantial quantities of MOX fuel (generally about 30% MOX fuel in a 157 fuel assembly core).
107. (Nesbit) Based on the information from the VERCORS tests of which we are aware, the results of those tests have no implications with respect to core cooling of MOX fuel during a design basis LOCA at Catawba. LOCA performance is driven primarily by the thermal-hydraulic response of the reactor coolant system, not by the fuel pellet material.
108. (Nesbit) On April 26, 2004, Duke posed Interrogatory 45 to BREDL. Duke asked the following questions.
Does BREDL agree that the VERCORS tests and results are irrelevant to the issue of MOX fuel compliance with 10 C.F.R. § 50.46 emergency core cooling system requirements? Does BREDL agree that to the extent the VERCORS tests are relevant, it is in relation to dose analyses only? If not, describe the basis for BREDL's position.
BREDL responded "BREDL agrees that the VERCORS tests are relevant to dose analyses only when design-basis LOCAs are considered."' 7 BREDL therefore does not dispute Duke's conclusion that the VERCORS tests are not relevant to the demonstration of compliance with 10 C.F.R. § 50.46 emergency core cooling system requirements, which is the issue of Contention I.
7 Blue Ridge Environmental Defense League's Response to Duke Energy Corporation's Second Set of Interrogatories and Requests for Production of Documents, June 8, 2004.
50
2at S-(
C. M5 T,"Ballooning 109. (Dunn) The impact of fuel relocation could depend in part on the amount of strain experienced at the location of cladding swell and rupture. As explained earlier, cladding swelling just prior to rupture forms a local cavity within the cladding that can fill with fuel fragments from cracked pellets. The amount of the swelling (cladding strain) is directly proportional to the consequence of the relocation because more swelling provides more volume to be filled by fuel pellet fragments.
110. (Dunn) The amount of swelling experienced is dependent on the phase of the cladding crystalline structure at the time of rupture. As the cladding heats the structure changes from a close packed hexagonal form, termed the alpha phase, to a body centered cubic form, called the beta phase. When the cladding is in either form, cladding strain at rupture is maximized. However, during the transition from alpha phase to beta phase, the differing crystalline structures tend to retard the ability of the metal to deform, so strain at rupture is minimized. LOCA evaluations treat this phenomena with direct correlations of experimental results. Figures 13 and 14 illustrate the required information. Figure 13 correlates the rupture temperature as a function of cladding stress. A determination of rupture is made by computing the cladding stress and comparing the current cladding temperature to the temperature of rupture given by Figure 13. If the cladding temperature equals the Figure 13 temperature the cladding will rupture. The strain is then obtained from the correlation with cladding rupture temperature (Figure 14).
51
Figure 13 Figure M5 Rupture Temperature Correlation E ----------- - - - - - - - - - - -
E a'- - - - - -- - - -- - -- - - - - - - -
C~ , =
I-.
Engineering Hoop Stress, Kps]
52
Figure 14 Figure M5 Rupture Strain versus Rupture Temperature Slow Ramp Rate c
I.
A/-k i:
- 4- + f e Rupture Temperature, C 111. (Dunn) As can be observed in Figure 14, the amount of strain varies substantially with the rupture temperature. Typical rupture temperatures for the Catawba plant lie between 1500-1600'F, placing the rupture during the transition of the cladding from the alpha phase to the beta phase. Typical strains lie between 50-75%.
112. (Dunn) The strains used in the LOCA calculations are based on tests on unirradiated cladding, because this maximizes the strain and therefore maximizes the flow blockage. The maximized strain would also maximize the consequences of fuel relocation.
113. (Dunn) The impact of irradiation on cladding ductility at high temperatures is primarily due to the pickup of hydrogen during normal operation. Hydrogen tends to embrittle the cladding, thereby reducing its ability to strain prior to rupture. Because M5TM picks up 53
auto relatively little hydrogen during operation, a M5TM strain versus rupture temperature curve would show little change with irradiation. In contrast, Zircaloy4 cladding picks up substantial hydrogen during operation. Accordingly, a Zircaloy-4 cladding strain versus rupture temperature curve would change more with operation (burnup) than would a M5TM curve. In this comparison, M5TM would experience less strain at rupture than Zircaloy-4 in the unirradiated state, but the two materials have approximately equal strain potential near the end of irradiation.
114. (Dunn) Overall, therefore, there is little expected difference in the consequences of fuel relocation due to cladding differences.
D. FillingRatio of Relocated Fuel 115. (McCoy) Another factor in assessing the impact of postulated fuel relocation is the appropriate density of the material that might fall (i.e., "relocate") into the ballooned region of the cladding. This material would be made up of the fragments of pellets from the immediate vicinity of the ballooning. During LOCA, when the cladding balloons away from the fuel pellets, the pellets could break into fragments along the existing crack lines and fall into the balloon. The relevant question is whether this would be significantly different for MOX fuel relative to LEU fuel.
116. (McCoy) The potential for fragmentation can best be determined by reviewing available micrographs of irradiated LEU and MOX pellets. Figure 15 shows a LEU fuel rod cross-section following irradiation and examination. The pattern of cracks in the pellet is typical of irradiated fuel.
54
Figure 15 Micrograph of Irradiated LEU Pellet with Typical Cracking 117. (McCoy, Dunn) Fuel pellets crack during the first ascent to power. The central portion of the pellet is hotter than the periphery and therefore expands more as the pellet comes
.up to its operational temperature distribution. The result is a set of roughly radial cracks in the outer portions of the pellet. The orientation of the cracks is more variable near the center of the pellet and central "islands" may form, completely surrounded by cracks. The potential for fuel relocation during LOCA develops with burnup as gaseous fission products accumulate within the pellet on the grain boundary creating stress in the pellet leading to fragmentation along crack lines when external constraints are removed (i.e., when there is cladding ballooning). Figure 16 provides another pellet cross-section illustrating a differing crack pattern. In this case a crack has developed radially that passes through the central region of the pellet. Beyond the general pattern, the positioning and path of the cracks is stochastic.
55
Figure 16 Micrograph of Irradiated LEU Fuel Pellet with Transverse Crack 118. (McCoy, Dunn) The processes that cause cracking and the cracking pattern -
thermal stress being resolved along grain boundaries - are the same for a MOX pellet as for a LEU pellet. The crack locations are not influenced by the plutonium rich agglomerates unique to the MOX fuel. Thus, within the general cracking pattern of LEU pellets described above, no differentiation can be made between LEU and MOX pellets, and fragment formation during LOCA would be expected to be the same for both fuel types. Figure 17 is a typical micrograph of a MOX pellet. The dark black areas or spots are plutonium rich agglomerates. There is no evidence in the micrograph that the crack locations are influenced by the agglomerates. The general cracking pattern is the same as for LEU pellets as shown in Figures 15 and 16. Thus, no differentiation in pellet fragmentation should be made between LEU and MOX fuel.
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Figure 17 Micrograph of Irradiated MIOX Pellet (RG) 119. (Nesbit) In this regard, this issue was raised at an ACRS meeting on the MOX fuel LAR. Dr. Dana Powers of the ACRS reacted to the "radial cut of a high burnup MOX pellet" shown by Patrick Blanpain of Framatome ANP, as follows:
CHAIRMAN POWERS: This hits to a point that Dr. Lyman raised, a question that he raised in his presentation. I think it is fair to characterize it as a question. Was there any inherent difference in the fragmentation of this MOX fuel relative to what we have experienced in uranium dioxide 57
?A&L fuel a(s) we go up to these burnups? I'll have to admit, had you not told me this was MOX fuel, I probably would not have known othervise.i8 120. (McCoy, Dunn) The density of the fragmented fuel material after falling into the cladding balloon will depend on the shapes of the fragments, the size of the fragments, and the efficiency with which the fragments settle into the ballooned region of the cladding. IRSN presented a distribution of densities based on the limited experimental evidence available at a May 25-26, 2004 meeting' 9 in Argonne, Illinois. The sources were the PBF and FR2 experiments, which used LEU fuel. The results are reported as filling ratio, which is the mix density of relocated material divided by the original pellet density. For the cases reported the ratio falls between 0.55 and 0.8. However, the upper portion of this range may be discounted.
The most reliable relocation data is from the FR-2 tests, and those values lie in the 0.55 to 0.65 range.
121. (McCoy) In the end, in my opinion, the breakup of either LEU or MOX fuel during LOCA would occur by the same process and should produce nearly identical fragmentation. The available measures of relative fuel material density within the relocation area indicate that a filling ratio less than 0.7 would be expected for LOCA conditions. There would be, in this aspect, no significant difference between MOX fuel and LEU fuel.
Is Advisory Committee on Reactor Safeguards Subcommittee on Reactor Fuels, April 21, 2004, Tr. at 69.
19 Grandjean, Claude and Hache, Georges (IRSN), "LOCA Issues Related to Ballooning, Fuel Relocation, Flow Blockage and Coolability," Presentation to the Meeting of the Special Experts Group on Fuel Safety Margins, Organization for Economic Co-operation and Development's Committee on the Safety of Nuclear Installations, Argonne, Illinois, May 24-26, 2004.
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E. Futel-CladdingInteraction 122. (AcCoy) Bonding of the fuel to the cladding is a complex process that occurs during operation of the fuel. The first requirement for bonding is intimate contact between the fuel pellets and the cladding. Fresh fuel at room temperature has a substantial diametral gap between the pellets and cladding, nominally about 150 pim. The gap size changes during irradiation as a result of irradiation-induced and pressure-induced creep of the cladding, and fracturing, densification, and swelling of the pellets. The general trend is for the gap to shrink until the pellets and cladding are in contact. The chemistry inside the fuel rod also changes during irradiation, with fission products being produced and the oxygen potential increasing.
When the gap is sufficiently small, the interior surface of the cladding can oxidize by athermal diffusion. In this process, oxygen is dislodged from fission spikes near the surface of the pellet and implanted in the cladding. Finally, cesium, which is produced as a fission product, can react with the fuel pellet and the zirconium dioxide layer of the cladding to form a variety of cesium uranates, zirconates, and plutonates. It is these phases that bond the pellet to the cladding.
123. (AcCoy) Despite the complexity of bond formation, the processes that lead to bonding are the same for LEU and MOX fuels, and the bond strengths are expected to be similar because of the similarities in fuel chemistry and operating conditions. Both fuels are predominantly uranium dioxide, and they have similar yields of cesium as a fission product.
Therefore, it is entirely reasonable to expect that pellet-cladding bonding will have similar effects on LOCA performance in MOX and LEU fuels (if indeed there is any pellet-cladding bonding effect on LOCA at all).
124. (McCoy, Dunn) BREDL has speculated that the extent of fuel-cladding bonding may be important to the fuel relocation phenomenon. It has also been speculated, however, that bonding is beneficial in controlling relocation. Although there is insufficient information to 59
quantify the benefits of bonding, we list potential benefits below. Nonetheless, AREVA has taken the conservative approach by taking no credit for the possible benefits in the MOX fuel lead assembly LOCA analysis.
125. (McCoy, Dunn) One possible benefit of bonding is that it may provide some additional strength to the cladding and therefore may reduce the extent of ballooning. Reducing the extent of ballooning might not prevent relocation, but it would reduce the volume into which fuel fragments could relocate and thus lower the local decay heat at that location. At present, there is insufficient data to confirm this effect.
126. (McCoy, Dunn) A second possible benefit is that, under some circumstances, the bond and the pellet structure near the pellet surface may be strong enough to allow a pellet fragment to adhere to the cladding interior surface during swelling and rupture. If this occurs, the bonded fragment may obstruct the motion of the other pellet fragments, slowing the filling of the balloon during fuel relocation, and possibly even reducing the final amount of relocated fuel.
This effect also has not been observed in experiments to date.
127. (McCoy, Dunn) Fuel bonding might also provide a coating that would remain with the cladding after thermal expansion causes the cladding to separate from the pellet during the first seconds of a LOCA. This is not specifically a fuel relocation effect, but if present any coating effect would be seen following cladding rupture. Such a coating effect might partially protect the cladding interior from high temperature oxidation. While there is no direct experimental confirmation of the phenomena, if present it would provide additional mitigation against local oxidation approaching the 17 percent acceptance criterion.
128. (McCoy, Dunn) At bottom, there is no evidence that pellet-cladding bonding differs significantly between MOX and LEU fuel. Even if one were to assume that it did, the 60
discussion above makes it clear that pellet-cladding bonding might act to improve LOCA performance by reducing the extent of cladding ballooning. Furthermore, the most recent studies of fuel relocation made no provisions to include postulated effects of pellet-cladding bonding.
These reports are included with this testimony as Exhibits 3, 4, and 5.20 Therefore, hypothetical MOX/LEU differences in this area do not affect the current regulatory position that relocation effects, if present at all, are adequately bounded by the inherently conservative nature of Appendix K LOCA analyses.
F. AMOXFuel Relative Power at High B urn up 129. (Nesbit) BREDL hypothesizes that relocation in the MOX fuel lead assemblies may be a more significant impact than relocation in LEU fuel because the power in MOX fuel is higher at end-of-life than the power in LEU fuel. This theory appears to be based on no more than a statement by IRSN that "... this question is particularly important for end-of-life MOX fuel where power generation is not reduced, unlike for U02 fuel."2 '
130. (Nesbit) Higher power at end-of-life is adverse for fuel relocation because fuel that is operating at higher power prior to the postulated accident gives off more power (decay heat) during the accident. The relocated fuel fragments would have a higher power, contributing to higher PCTs.
20 M. Lambert, et al. "Synthesis of an EDF and Framatome ANP Analysis on Fuel Relocation Impact in Large Break LOCA", proceedings of the Topical Meeting on LOCA Fuel Safety Criteria, Aix-en-Provence (March 2001) (Exhibit 3); C. Grandjean, et al. "High Burnup UO 2 Fuel LOCA Calculations to Evaluate the Possible Impact of Fuel Relocation after Burst", Proceedings of the Topical Meeting on LOCA Fuel Safety Criteria, Aix-en-Provence March (2001) (Exhibit 4); and V. Guillard, et al., "Use of CATHARE2 Reactor Calculations to Anticipate Research Needs", presentation at the SEGFSM Topical Meeting on LOCA Issues, Argonne National Laboratory (May 2004)
(Exhibit 5).
21 "IRSN Source Term LOCA PROGRAM IN THE PHEBUS FACILITY," IRSN presentation to the NRC (October 23, 2003), Slide 21.
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131. (Nesbit) However, the IRSN statement about higher power at end-of-life is apparently based on European MOX fuel experience. For the reasons discussed below, the Catawba MOX fuel lead assemblies will operate at lower power at end-of-life than LEU fuel.
This will provide a MOX fuel benefit (not a penalty) with respect to fuel relocation (relative to the LEU fuel at Catawba).
132. (Nesbit) First, we consider the type of plutonium in the fuel. European reactors have been operating for decades with "reactor grade" (RG) MOX fuel. The fuel is referred to as "reactor grade" because it is initially comprised of RG plutonium blended with depleted (or natural) uranium. As noted in Section I1.B of this testimony, RG MOX fuel contains plutonium with more than 20% 240Pu. In contrast, the Catawba MOX fuel lead assemblies will initially contain "weapons grade" (WG) plutonium, which is less than 7% 240pu. The initial isotopic content of the plutonium (RG vs. WG) affects the nuclear characteristics of the MOX fuel, including the profile of power vs. burnup, as discussed in the paper "Basis for the Design of Reactor Cores Containing Weapons Grade MOX Fuel."22 133. (Nesbit) WG MOX fuel is more similar neutronically to LEU fuel than is RG MOX fuel, as illustrated below. Figure 18 shows k-infinity (k4)for three mechanically identical fuel assemblies as a function of burnup, as calculated by the CASMO-4 computer code for an infinite lattice configuration. K, correlates to the ability of the fuel assembly to generate power.
As the fissile material depletes with burnup, k. decreases.
22 Steve Nesbit and Jim Eller, "Basis for the Design of Reactor Cores Containing Weapons Grade MOX Fuel," Advances in Nuclear Fuel Management III, American Nuclear Society, October 5-8, 2003.
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Figure 18 1.30 1.25 =
<~~R Ns>D~IlOX 1.15
- 1. 10-1.05 1.00 0.95 0.90 0.80 0 10 20 30 40 s0 60 Fuel Exposure, GNlD/t 134. (Nesbit) At about 22 GWD/t burnup, the curves intersect; at that point the three fuel assembly types would have about the same relative power if loaded in the same core location.
135. (Nesbit) The slope of the RG MOX kt vs. bumup curve is flatter than the curve for LEU fuel. This means that, all other things being equal, at burnups higher than 22 GWDAt, the RG MOX fuel would have a higher relative power than LEU fuel.
136. (Nesbit) The WG MOX fuel k. vs. burnup curve is much closer to the LEU fuel k1. vs. burnup curve. Therefore, at higher burnups (e.g., 40-50 GWD/t), the WG MOX fuel relative power would be much closer to LEU fuel. In other words, the effect noted by IRSN (a 63
tendency for higher end-of-life power in MOX fuel) is much less important for Catawba, which will use WG MOX fuel lead assemblies, than it is for French reactors that use RG MOX fuel.
137. (N'esbit) Second, we consider the nature of the lead assembly program itself. The purpose of lead test assemblies is to gather additional data on new fuel assembly designs that may not have extensive prototypical operating experience. Accordingly, lead test assemblies are placed in "nonlimiting core regions."2 3 The Catawba cores containing MOX fuel lead assemblies will be designed to ensure that the MOX fuel assembly power during the nominal depletion is not the highest assembly power in the core.
138. (Nesbit) Duke has developed a fuel cycle design for ClC16 containing four fresh MOX fuel assemblies and 72 fresh LEU assemblies. Those MOX and LEU assemblies have been carried forward into preliminary scoping analyses for CIC17 and CIC18. The results from those analyses were used to calculate anticipated fuel assembly peaking for each of the 76 assemblies loaded into CIC16, as shown in Figure 19.
23 Catawba Nuclear Station Technical Specifications, Section 4.2.1.
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Figure 19 1.6 1A L% AAA A A A AIAAt~
0I g ° o a; ' > cA ]AAA - A -A -A- E
- 0. 0 0a a a0 E~l U"Q 1
8 0 0 3M 1.0 c
. -1MOX LTA
> 0.8 A Maximum LEU E 0.6 E OAverage LEU "A A A A A A A A A AA 4) 40 OA (Oo 0 0 0 0 0 s0 0 0.2 0.0 0 200 400 600 800 1000 1200 1400 1600 Accumulated Cycle Irradiation In Effective Full Power Days 139. (Vesbit) As shown in Figure 19, the MOX fuel lead assembly relative power (squares) is lower than the corresponding average LEU assembly relative power (circles) at all times except at the very beginning of each of the first two cycles.
140. (Nesbit) The MOX fuel lead assembly relative power (squares) is lower than the maximum LEU assembly relative power (triangles) at all times.
141. (Nesbit) Third, we consider the fact that the decay heat from a MOX fuel assembly will be lower than the decay heat from a LEU fuel assembly with the same bumup and same operating power level. Figure 3 appearing earlier in this testimony shows that the MOX fuel decay heat should be 3-5% lower than the corresponding LEU fuel decay heat during the time period of interest for a LOCA. It appears that the IRSN did not acknowledge the decay heat effect, which is a benefit for MOX fuel.
142. (Nesbit) In summary, the IRSN statement about higher power in MOX fuel at end-of-life is not applicable to the Catawba MOX fuel lead assemblies. LEU fuel assembly operating power will be higher than MOX fuel lead assembly operating power at end-of-life at 65
Catawba, due both to the nature of WG MOX fuel and to core design. Power in relocated fuel fragments is also affected by decay heat. LEU fuel decay heat will be higher than MOX fuel lead assemblies decay heat at the same power level and burnup. To the extent that assembly relative power and decay heat might affect LOCA performance through fuel relocation, that effect will be beneficial for MOX fuel relative to LEU fuel. In other words, these MOX/LEU differences would act to mitigate, not exacerbate, the impact of MOX fuel relocation during a postulated LOCA.
G. Conclusions on Futel Relocation 143. (Dunn) As discussed above, the issue of fuel relocation during LOCA was identified and evaluated by the NRC in the mid-1980s. The potential for relocation affects all nuclear plants using fuel within a sealed cylindrical cladding - that is, the majority of nuclear reactors world-wide. The issue was reviewed by the NRC, assigned a low priority, and subsequently given a "drop" designation. The NRC's review cited the conservatisms inherent in a 10 C.F.R. § 50.46 evaluation as more than bounding the potential impact of relocation.
144. (Dunn) As also discussed above, in the Thadani memorandum the NRC Staff has recognized that any removal of the substantial existing embedded conservatisms (or margin) in the regulations that might be associated with risk-informing the regulations should also be accompanied by either (i) a quantitative assessment that the retained conservatisms remain sufficient to bound fuel relocation or (ii) an explicit relocation model included in the modeling.
In this respect fuel relocation is just one of several potential conservatisms and non-conservatisms. Aside from the concern over removal of current conservatisms, the NRC's position on fuel relocation remains that the issue is well-bounded and does not require explicit modeling under the current, deterministic regulatory approach (such as was utilized for the MOX fuel lead assemblies).
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145. (Dunn, McCoy) The industry has generally held that the impact of fuel relocation on calculated cladding temperatures or cladding oxidation predictions would be small or inconsequential if incorporated within comprehensive LOCA modeling. We do not expect fuel relocation effects to be significantly different for MOX versus LEU fuel.
146. (Dunn) At the meeting of the Organization for Economic Co-operation and Development's Committee on the Safety of Nuclear Installations in Aix, France in 2001, two evaluations were presented. One, by Electricite de France (EDF) and Framatome ANP, produced only a limited increase in cladding temperature (301C or 54 0 F), resulting in the presenter's conclusion that relocation should not be taken into account in safety regulation 24 .
The other, from IRSN and using more conservative assumptions, indicated a 130'C (234 0 F) potential impact on cladding temperatures.
147. (Dunn) Neither calculation took credit for increased heat transfer due to the rupture-induced geometric perturbation of the flow field. Even a partial accounting for these effects indicates a potential reduction of cladding temperatures of 100-150 0 C (180-270'F).
Thus, the presentations at Aix indicate that a comprehensive evaluation of the effect of fuel relocation would either reduce the cladding temperature beyond that calculated by current models or make essentially no difference in the predicted temperature.
148. (Dunn) Experimentally, the most comprehensive treatment of the phenomena are the KfK tests performed at the FR2 reactor in Germany. Here testing was done for fuel that 24 M. Lambert, ct al., "Synthesis of an EDF and Framatome ANP Analysis of Fuel Relocation Impact in Large Break LOCA," Proceedings of the Topical Meeting on LOCA Fuel Safety Criteria, Special Experts Group on Fuel Safety Margins, Organization for Economic Co-operation and Development's Committee on the Safety of Nuclear Installations, Aix, France, March 22-23, 2001. As noted above, this paper is Exhibit 3 to this testimony.
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cracked and relocated and for fuel that did not relocate, providing a unique opportunity for comparison with limited cross test complications. Figures 10 and 11 appearing earlier in this testimony provide the results for two such tests.
149. (Dunn) When the fuel does not relocate, Figure 10, the cladding temperature near the ballooned and ruptured location cools substantially followed by a delayed heatup and a peak approximately 120'C (2167F) lower than that of cladding above and below the rupture location.
This cooling is caused by the combined effects of the increase in cladding surface area, the enhancement of heat transfer due to geometry effects, and the decrease in cladding-to-pellet gap conductance.
150. (Dunn) For the test with fuel relocation, Figure I 1, there is only a slight reduction in temperature at the time of rupture followed by a heat up that is very similar to the other locations on the fuel pin. This indicates that the benefits of increased surface area and increased heat transfer from geometric effects are a near match for the detrimental effect of the increased decay heat and lack of expanded gap associated with the fuel repositioning.
151. (Dunn) Thus, the most comprehensive experimental evidence available for the evaluation of fuel relocation indicates that the phenomenon is not of substantial consequence with respect to PCT.
152. (Dunn) Finally, the most pessimistic of the relocation impacts calculated by IRSN can be added to the results of the AREVA LOCA analyses that were used to establish LOCA limits for the MOX fuel lead assemblies (Exhibit 2, Table Q14-1), and the results still do not exceed the LOCA acceptance criteria of 10 C.F.R. § 50.46.
153. (Dunn) At the conference in Aix in 2001, the IRSN presented a calculation of fuel relocation showing an increase in cladding temperature at the swelling location (fuel relocation 68
site) of slightly less than 320°F. The increase in local oxidation was 7% at the ruptured location.
The IRSN calculation was for a fillin fraction of 0.7.
1t0 care~s flc 4 ktilrijh le- ue, e1 Iiw"% cy ie 1 e, 154. (Dunn) the highest PCT at the ruptured location ird tezLOA cIlultin fui Catawba dzesribed in the MOX fuel lead assembly license amendment request was approximately 1750TF and the local oxidation on that fuel pin is 3%. Adding the IRSN predictions to the Catawvba MOX fuel results gives an estimated PCT of 2070'F and a local oxidation of 10%. Thus, even if the pessimistic IRSN predictions are simply added to the current Catawba MOX fuel LOCA evaluations, the results remain well below the acceptance criteria of 10 C.F.R. § 50.46.
155. (Dunn) At a conference in Argonne in May 2004, IRSN presented calculations of a fuel relocation PCT impact as high as 1501C (270TF) for LEU fuel, and a further fuel relocation PCT impact of I00 C (18TF) for MOX fuel.25 A 18'F difference between MOX and LEU fuel is negligible with respect to calculated LOCA PCTs, and supports the position that fuel relocation is not primarily a MOX fuel issue.
156. (Dunn) The IRSN presentation at Argonne notes a total PCT increase of 160'C (288TF) for MOX fuel during relocation, and traces the 100 C MOX fuel increment to higher initial stored energy in MOX fuel. As shown earlier, the MOX fuel lead assemblies at Catawba will have lower initial stored energy than LEU fuel due to operation at lower peaking and to lower decay heat. Accordingly, the MOX fuel lead assemblies at Catawba would actually see a benefit, not a penalty, relative to LEU fuel and potential fuel relocation impacts.
25 Guillard, et al., "Use of CATHARE2 Reactor Calculations to Anticipate Research Needs," Presentation to the Meeting of the Special Experts Group on Fuel Safety Margins, Organization for Economic Co-operation and Development's Committee on the Safety of Nuclear Installations, Argonne, Illinois, May 24-26, 2004. As noted above, this paper is included with this testimony as Exhibit 5.
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157. (Dunn) Summing up, the impact of fuel relocation is generally held to be inconsequential. This view is supported by the most comprehensive experimental results available, the KfK tests in the FR2 reactor. Moreover, MOX/LEU differences at end-of-life would act to mitigate any adverse fuel relocation impact in the MOX fuel lead assemblies. Even so, superimposing IRSN's conservative evaluation of the impact of fuel relocation on top of the conservative Catawba MOX fuel LOCA predictions would result in consequence predictions within those allowable under 10 C.F.R. § 50.46. Any uncertainty associated with fuel relocation in MOX fuel is sufficiently bounded by the existing analyses.
VI. UNCERTAINTIES 158. (ANesbit) Contention I reduces to a question of what constitutes acceptable "uncertainty." Uncertainty is inherent in LOCA analyses. The computer calculations used to simulate a LOCA involve rapidly changing thermal-hydraulic conditions coupled with neutronic and mechanical phenomena. Recognizing this inherent uncertainty, the NRC established the conservative LOCA acceptance criteria in 10 C.F.R. § 50.46 and prescribed conservative LOCA modeling approaches in Appendix K to 10 C.F.R. Part 50.
159. (Nesbit) Duke used these conservative modeling approaches to demonstrate that the conservative acceptance criteria are met for the four MOX fuel lead assemblies following a postulated large break LOCA at Catawba.
160. (Nesbit) 10 C.F.R. § 50.46 and Appendix K were promulgated in 1974. In the ensuing years, NRC and others have developed more and more sophisticated analytical tools for analyzing design basis LOCAs. In addition, NRC and others have invested substantial resources in performing separate effects tests and integral tests for the purpose of improving and validating LOCA computer models. As discussed earlier in this testimony, one of the results of this work has been to confirm the fundamental overall conservatism inherent in both the 10 C.F.R. § 50.46 70
acceptance criteria and the Appendix K models such as those used to analyze the four MOX fuel lead assemblies.
161. (Nesbit) The October 23, 2003 IRSN presentation provided no evidence that the MOX fuel lead assembly LOCA analyses do not provide reasonable assurance that the LOCA acceptance criteria are met. In fact, the IRSN presentation addressed neither the Catawba MOX fuel lead assembly LAR nor the broader program to use MOX fuel to dispose of weapons grade plutonium.
162. (Nesbit) Based on remarks by IRSN, BREDL is speculating that fuel relocation may be different in MOX fuel than LEU fuel, and that such a difference, if present, may cause fuel relocation in MOX fuel to have an adverse impact on a design basis LOCA. Again based on remarks by IRSN, BREDL claims the M5 cladding used on the MOX fuel lead assemblies is likely to have higher blockage ratios during a LOCA, resulting in coolant blockage that could lead to an unacceptable loss of core geometry and an uncontrolled core melt. BREDL states that these effects cannot be fully assessed in the absence of integral LOCA MOX fuel-bundle tests, such as those proposed by IRSN. BREDL claims that this alleged uncertainty is sufficient cause for denying the Duke application to use four MOX fuel lead assemblies until the testing proposed by IRSN is carried out.
163. (Nesbit) As discussed in Sections IV and V of this testimony, BREDL's concern that fuel relocation and M5T 1' cladding ballooning during design basis LOCA are especially significant uncertainties for MOX fuel is fundamentally wrong. However, even if one were to assume for the sake of argument that the concern has merit, BREDL's position presents a classical "chicken vs. egg" paradox. In order to conduct the tests suggested by IRSN, irradiated MOX fuel rods would be required. In order to have irradiated fuel rods, a reactor would have to 71
operate with MOX fuel. However, BREDL would not allow a reactor to operate with MOX fuel until the tests are performed.
164. (Nesbit) In response to Duke Interrogatory 15, BREDL reaffirmed its position that the use of MOX lead assemblies at Catawba requires prior execution of a LOCA test program such as the PHEBUS tests proposed by IRSN.26 BREDL addressed the "chicken vs. egg" paradox by allowing for the possibility that PHEBUS tests using M5-clad reactor grade MOX fuel (presumably obtained from European reactors) may be adequate for understanding the relevant phenomena, if accompanied by additional analyses or separate effects tests to understand the additional impact of plutonium isotopic composition on design basis LOCAs.
(Note: BREDL has postulated no mechanism for the isotopic composition (WG vs. RG) of the MOX fuel to adversely impact the response to a LOCA.) PHEBUS tests are complex and require significant preparation, and the reduction and analysis of the data would require additional and substantial time. Even if the necessary arrangements could be made, the BREDL prescription would obviously delay the MOX fuel lead assembly program (and therefore the large-scale disposition of excess weapons plutonium) for a number of years in the name of "reducing uncertainty." Finally, the BREDL response to Duke Interrogatory 15 ignores the question - what should be done to resolve the "chicken vs. egg" paradox if European RG MOX fuel turns out to be unavailable, or if the isotopic issues cannot be resolved to BREDL's satisfaction?
165. (Nesbit) In making provisions for lead test assembly programs at commercial reactors, NRC envisioned the need to conduct limited irradiations of fuel assemblies such as are proposed for the MOX fuel lead assemblies. NRC implicitly recognizes that a small number of 26 Blue Ridge Environmental Defense League's Response to Duke Energy Corporation's First Set of Interrogatories and Requests for Production of Documents, April 14, 2004.
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lead test assemblies, located in non-limiting core locations, can be used without posing an undue risk to the health and safety of the public. Over the past decades, the nuclear industry has made substantial improvements in fuel design that have contributed to improved fuel performance and utilization. Lead test assembly programs have contributed to improvements in cladding material, development of debris-resistant features, and implementation of integral neutron absorbers in fuel pellets, among other advancements. These types of improvements are beneficial from the standpoint of public health and safety, and they provide for more economical utilization of the uranium in reactor fuel.
166. (Nesbit) The logical extension of the BREDL position, if adopted wholesale by the NRC, would be that lead test assembly programs cannot be carried out unless there is sufficient data such that the lead test assembly program is not needed in the first place. This is obviously not a desirable result because it would inhibit future fuel design improvements that might provide additional safety and economic benefits.
167. (Nesbit) In alleging that differences between LEU fuel and MOX fuel could potentially exacerbate MOX fuel relocation during a LOCA relative to LEU fuel, and that this potential impact introduces an unacceptable amount of uncertainty, BREDL does not acknowledge that there are actually already differences among LEU fuel pellet designs (approved and operating in reactors) that equally could be postulated to have similar effects on fuel relocation. This inherent "uncertainty" remains well within the acceptable margins. The MOX fuel lead assemblies fall within a similar range and are well-bounded by conservatisms in the analysis and acceptance criteria.
168. (Nesbit) For example, some LEU fuel pellets contain integral burnable absorbers (e.g., gadolinium, erbium) within the fuel matrix. Other LEU fuel 'pellets are coated with 73
-2jiY absorber material (e.g., zinc diboride). These materials can impact pellet properties such as fuel conductivity and fuel pellet internal gas generation.
169. (Nesbit) Even without different materials, LEU fuel pellets are designed with different sizes and shapes (e.g., length, diameter, dish, chamfer), and such differences could be postulated to affect the characteristics of fuel fragments resulting from a design basis LOCA with fuel relocation.
170. (Nesbit) LEU fuel pellets are also manufactured in different plants using different processes. LEU pellets have different physical characteristics (e.g., theoretical density). These physical differences too could be postulated to affect fuel relocation as well.
171. (Nesbit) At bottom, it is not necessary in a LOCA model to address and to quantify the impact of every difference in order to reduce uncertainty to an acceptable value.
The fundamentally conservative nature of Appendix K LOCA analyses provides reasonable assurance that postulated LOCA relocation effects for either LEU or MOX fuel do not threaten the health and safety of the public. BREDL's desire for perfect certainty with respect to relocation effects of MOX fuel should be considered in the context of the inherent differences among the various LEU fuel designs operating in the United States and abroad.
VIL. CONCLUSIONS 172. (All) As described above, the LAR and the supplements in the RAT responses demonstrate that Duke has satisfied 10 C.F.R. § 50.46 by performing LOCA analyses in accordance with 10 C.F.R. Part 50, Appendix K. The AREVA LOCA analysis uses an approved methodology that has been appropriately modified to address differences between MOX and LEU fuel behaviors. The results demonstrate conservative compliance with conservative acceptance criteria.
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173. (All) At the April 21, 2004 meeting of the Advisory Committee on Reactor Safeguards Subcommittee on Reactor Fuels (Tr. 246), Dr. Ralph Meyer of the NRC Staff stated:
The loss of coolant accident, the effect of MOX - Well, first off, let me say that there cleatly are neutron physics effects of MOX, and these can be and are being handled. But when you talk about the fuel part of that, for the loss of coolant accident any connection, any difference between MOX and LEU at this time is purely speculative, and I don't think there is any evidence that there is a difference, although we are, of course, interested in looking.
174. (All) We agree with Dr. Meyer's statement. Contention 1, as explained by BREDL, is pure speculation based on statements made by research organizations addressing issues outside the context of the Catawba MOX fuel lead assembly LAR. There is no evidence for an assertion that MOX/LEU differences might significantly and adversely impact design basis LOCA analyses or LOCA fuel performance.
175. (All) Indeed, the BREDL contention was initially, based on certain VERCORS tests at severe accident conditions that clearly do not establish a relocation issue at design basis LOCA conditions. BREDL offers no other data or testing and its arguments now pertain primarily to "uncertainty" related to the possibility of fuel relocation at design basis LOCA temperatures.
176. (All) In our testimony we have described the analyses performed to demonstrate that the MOX fuel lead assemblies will meet the relevant regulatory acceptance criteria for LOCA. We have shown that these analyses address the pertinent MOX/LEU differences through explicit modeling or conservative assumptions.
177. (All) In our testimony we have explained the substantial conservatisms (and associated margin) embedded in the NRC's criteria that more than compensate for any uncertainty related to fuel relocation. In addition, we have explained the conservatisms (and associated margin) in the MOX fuel LOCA analyses. We have also explained why, in our 75
opinion, MOX fuel relocation effects should not be significantly different from postulated effects for LEU fuel. In fact, we have shown that the MOX/LEU difference in end-of-life power would be beneficial to MOX fuel with respect to any fuel relocation effect. Therefore, taken in context, any "uncertainty" related to fuel relocation does not pose a significant issue.
178. (Al) We have also shown that even superimposing IRSN's most pessimistic estimates of the impact of fuel relocation on top of the conservative Catawba MOX fuel LOCA predictions, the results would be within the acceptance criteria of 10 C.F.R. § 50.46.
179. (AII) Finally, the inherent assertion in Contention I that all uncertainty must be eliminated prior to approving a lead assembly program is entirely inconsistent with the fundamental purpose of lead assembly programs and would have the undesirable effect of stifling future fuel advancements and delaying an important nuclear non-proliferation initiative.
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Attachment 1 STEVEN P. NESBIT Duke Power 526 South Church Street Charlotte, NC 28202 QUALIFICATIONS:
Air. Nesbit has 24 jears nuclear engineering and management experience in the commintercialsector and on DepartmentofEEnergj (DOE)projects. ie is the Alred Oxide ("M\IOX") Fuel Project Alanager for Duke Power, which is playing a k-ey role in the DOEprogram to dispose of surplus weapons plutonitu,. le has 22 years experience with Duke Power. In addition, AMr. Nesbit has managed actiritiesforthe Mlanagingand OperatingContractorto DOE's Office of Civilian Radioactive JY'aste AManagemzent. He also has expertise in nuclear safet, analysis technology. Ar. Nesbit has extensive experience interacting with the Nuclear Regulatory Commission and he has authored numerous topical reports and technicalpapers.
EDUCATION/TRAINING:
ME, Nuclear Engineering, University of Virginia, 1982 BS, Nuclear Engineering, University of Virginia, 1980 Graduate course work, Environmental Science Supervisory Development Program, Duke Power PROFESSIONAL AFFILIATIONS/CERTIFICATIONS:
Registered Professional Engineer, North Carolina Registered Professional Engineer, South Carolina American Nuclear Society EXPERIENCE:
3/99-Present Engineering Supervisor 11- Duke Power Manages Duke Power's activities as part of the project to dispose of surplus United States weapons plutonium using mixed oxide (MOX) fuel. Directs technical, licensing, and business activities. Serves as a public spokesperson on the MOX fuel project.
09196-3199 Consulting Engineer - Duke Power 1
Steven P. Nesbit Led Duke Power's feasibility investigations regarding using MOX fuel at the company's three nuclear plants to support DOE's surplus weapons plutonium disposition program. Served as a representative on the Nuclear Energy Institute's Working Group on Surplus Weapons Plutonium Disposition. Interacted with external groups (Congress, DOE, and the public) in support of the MOX fuel project.
11/95-09/96 Engineering Supervisor II - Duke Engineering & Services (DE&S)
Supervised the Design Basis and Project Integration Section of the DOE Office of Civilian Radioactive Waste Management (OCRWVM) Management and Operating Contractor. Developed environmental design criteria and performed design basis accident evaluations for an interim storage facility for spent nuclear fuel.
05/94-11/95 Manager, Regulatory Interactions Section - DE&S Manager of the Las Vegas Regulatory Interactions Section of the Regulatory and Licensing Department of the Management and Operating contractor for the DOE OCRWM. Responsibilities of the seven-person section included interactions with the Nuclear Regulatory Commission (NRC) staff and on-site representatives, the Advisory Committee on Nuclear Waste, and the Nuclear Waste Technical Review Board; development of regulatory positions; regulatory reviews; Site Characterization Analysis comment responses; regulatory commitments; and NRC issue resolution activities.
12/92-04/94 Engineering Consultant - DE&S Licensing Engineer in the Las Vegas Regulatory and Licensing Department of the Management and Operating contractor for the DOE OCRWM.
Provided nuclear power plant licensing experience and general support to the DOE Yucca Mountain Site Characterization Office. Assisted with interactions between the DOE, the National Academy of Sciences, the Environmental Protection Agency, and the NRC, related to the development of an environmental standard for the potential repository at Yucca Mountain.
1991-1992 Utility Engineering Group (UEG) Site Engineer - DE&S Site Engineer in Washington, D.C., for the DE&S Utility Engineering Group. Provided utility perspective and experience to the DOE for the New Production Reactor Project. Served on the staff of the Chief Engineer of the project. Provided day-to-day liaison with the various project areas. Served as Project Engineer for the UEG. Managed the DE&S Washington, D.C.,
office.
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Steven P. Nesbit 1990-1991 Senior Engineer- Duke Engineering & Services Worked in the safety review area of the UEG. Provided utility perspective and experience to the New Product Reactor Project in the area of nuclear reactor safety.
1988-1990 Design Engineer - Duke Power Lead engineer in the area of nuclear safety analysis technology, a work group comprised of five engineers. Worked on developing mass and energy release analysis capability for high energy line breaks at Oconee, McGuire, and Catawba Nuclear Stations. Used the RELAP5/MOD002 transient analysis computer code and wrote in-house analytical codes. Worked to develop reactor building analysis capability for large dry and ice condenser containments, including applications of the FATHOMS (COBRA-NC) and CONTEMPT computer codes. Tested the upgraded Oconee training simulator and evaluated vendor performance. Represented the Babcock and Wilcox Owners Group (B&WOG) on the Project Management Group of the Multi-Loop Integral System Test Facility, a thermal-hydraulic research project sponsored by the B&WOG, the Electric Power Research Institute (EPRI) and the NRC. Served on the Duke Power Crisis Management Team.
1982-1988 Design Engineer/Assistant Engineer/Junior Engineer - Duke Power Lead safety analysis engineer for the Oconee Nuclear Station, a work group of up to five engineers. Served as Duke Power representative on the B&WOG Analysis Committee. Participated in the Technical Advisory Group, a committee comprised of B&WOG, EPRI and NRC representatives, which evaluated the need for thermal hydraulic testing related to once-through steam generators. Helped develop symptom-oriented emergency procedures for Oconee. Performed extensive RETRAN benchmarks of plant transients and helped prepare a safety analysis methods topical report for submission to the NRC. Served as one of 12 auditors for the inaugural Duke Power Self-initiated Technical Audit, patterned after the NRC Safety System Functional Inspections. Participated in fuel loading and start-up physics testing at McGuire Nuclear Station. Participated in zero power physics testing at Oconee. Performed system and containment analyses of the Oconee plant. Prepared technical justifications for emergency Technical Specification changes which prevented -unnecessary unit shutdowns.
3
Steven P. Nesbit 1979-1982 Rcactor Operator/Reactor Operator Trainee - University of Virginia Reactor Facility Reactor Operator Trainee and licensed Reactor Operator for the 2-MW research reactor in Charlottesville, Va. Duties included shift operation work, training and fuel handling.
AWVARDS/HONORS:
"Doer of Deeds," Yucca Mountain Site Characterization Office, February 2, 1994.
NewcombMiomton Fellowship, University of Virginia, 1980-1981.
Bachelor of Science with Highest Distinction, University of Virginia, 1980.
PUBLICATIONS:
Nesbit, S. P., Scott, M. W., Eller, J. L., Verbos, F. J., and Costello, M. V., "Non-LOCA Safety Analysis for Operation with Weapons Grade MOX Fuel Lead Assemblies," American Nuclear Society Winter Meeting 2003, New Orleans, LA, November 2003.
Nesbit, S. P. and Eller, J. L., "Basis for the Design of Reactor Cores Containing Weapons Grade MOX Fuel," Advances in Nuclear Fuel Management 111, Hilton Head, SC, October 2003.
Anderson, S. L., Gilreath, J. D., Nesbit, S. P., and Laubam, T. J, "Mixed Oxide Fuel Effects on the Integrity of the McGuire and Catawba Reactor Vessels," Fifth Topical Meeting on Spent Nuclear Fuel and Fissile Materials Management, Charleston, SC, September 18, 2002.
Buckner, M. R., Bengelsdorf, H. D., and Nesbit, S. P., "American Nuclear Society Nonproliferation Position Statement," Fiflh Topical Meeting on Spent Nuclear Fuel and Fissile Materials Management, Charleston, SC, September 18, 2002.
Clark, R. H., Dziadosz, D., and Nesbit, S. P., "MOX Fuel Irradiation Program for Disposition of Surplus United States Plutonium," Fourth Topical Meeting on Department of Energy Spent Nuclear Fuel and Fissile Materials Management, San Diego, SC, June 7,2000.
Nesbit, S. P. and Bengelsdorf, H. D., "A Comparison of Surplus Weapons Plutonium Disposition Technologies," Third Topical Meeting on Department of Energy Spent Nuclear Fuel and Fissile Materials Management, Charleston, SC, September 9, 1998.
S. P. Nesbit, "A Utility Perspective on Surplus Weapons Plutonium Disposition in Existing United States Light Water Reactors," Advances in Nuclear Fuel Management II, Myrtle Beach, S.C.,
March 1997.
S. P. Nesbit, S. J. Brocoum, M. A. Lugo, J. A. Duguid, P. M. Krishna, "Regulatory Perspective on NAS Recommendations for Yucca Mountain Standards," 7th Annual International High-Level Radioactive Waste Management Conference, Las Vegas, NV, May 1, 1996.
4
Steven P. Nesbit J. Carl Stepp, Silvio Pezzopane, Quazi Hossain, Michael Hardy, Steven P. Nesbit, "Criteria for Design of the Yucca Mountain Structures, Systems, and Components for Fault Displacement,"
FOCUS '95 - Methods of Seismic Hazards Evaluation, Las Vegas, NV, September 20, 1995.
J. Carl Stepp, Michael P. Hardy, Quazi A. Hossain, Steven P. Nesbit, J. Timothy Sullivan, "Seismic Design Methodology for a Geologic Repository at Yucca Mountain," 6th Annual International High-Level Radioactive Waste Management Conference, Las Vegas, NV, May 4, 1995.
D. Stahl, S. P. Nesbit, L. Berkowitz, "Approach to Compliance with the NRC Substantially Complete Containment Requirement at the Potential Repository at Yucca Mountain," 6th Annual International High-Level Radioactive Waste Management Conference, Las Vegas, NV, May 3, 1995.
S. P. Nesbit, S. J. Brocoum, "New Public Health and Safety Standards for Yucca Mountain and Their Impact on the Carbon-14 Issue," Waste Management '95 Conference, Tucson, AZ, February 26, 1995.
S. P. Nesbit, R. J. Gerling, and G. B. Swindlehurst, "Qualification of the Oconee RETRAN Model by Comparison with Plant Transient Data," Nuclear Technology, Volume 83, December 1988.
TOPICAL REPORTS:
DPC-NE-1005P, "Duke Power Nuclear Design Methodology Using CASMO4/SIMULATE-3 MOX," Duke Energy, August 2001.
YMPITR-003-NP, "Seismic Design Methodology for a Geologic Repository at Yucca Mountain,"
U. S. Department of Energy, October 1995.
DPC-NE-3003-P, "Mass and Energy Release and Containment Response Methodology," Duke Power Company, August 1993.
BAW-2079, "Technical Advisory Group Investigation of Once-Through Steam Generator Therrnal-Hydraulic Data Requirements," Babcock and Wilcox, March 1989.
DPC-NE-3000, 'Thermal-Hydraulic Transient Analysis Methodology," Duke Power Company, July 1987.
SECURITY CLEARANCE:
DOE "L" Clearance (active)
REFERENCES:
DOE and commercial references available upon request 5
Attachment 2 ROBERT C. IIARVEY Duke Power 526 Soutfi Church St Charlotte, NC 28202 QUALIFICATIONS:
25 years of Thermal Hydraulic and Safety Analysis experience supporting the reload and licensing of pressurized water reactors. Mr. Harvey has performed numerous safety analysis calculations using the RELAP4, RELAP5, RETRAN, TOODEE-2, CONTEMPT-LT, and MAAP computer codes.
EDUCATION:
Nuclear Engineering Graduate Studies, University of Lowell (1980-1982)
BS, Nuclear Engineering, University of Lowell, 1979 Supervisory Training, Yankee Atomic Electric Company, 1994 MAAP Code Utilization and Phenomena seminar, Fauske & Associates Two-Phase Gas-Liquid Flow Seminar, University of Houston Nuclear Power Reactor Safety Seminar, Massachusetts Institute of Technology (MIT)
Simulator Training, Combustion Engineering (CE)
Two-Phase Flow and Heat Transfer, Rensselaer Polytechnic Institute EXPERIENCE:
Senior Engineer - Duke Power Company 2/99 - present Lead Engineer responsible for the LOCA analysis supporting the Oconee, McGuire, and Catawba nuclear plants. Responsibilities include providing interface and oversight of the vendor analyses. In addition, performs LOCA mass and energy release calculation used as input to the containment analysis and performs UFSAR Chapter 15 non-LOCA safety analysis. Specific accomplishments include supporting the Oconee reanalysis to support steam generator replacement and the transition to best-estimate LOCA analysis methods for the McGuire and Catawba units. Serves as a member of the Emergency Operations Facility (EOF) in the position of Accident Assessment Manager.
Provided an independent assessment of the Texas Utilities LOCA analysis supporting the transition to Westinghouse fuel.
I
Robert C. Harvey Engineer - Duke Engineering & Services - (12/97 - 1/99); Senior Nuclear Engineer -
Yankee Atomic Electric Company (5/91 - 11/97)
Lead Engineer for pressurized water reactor (PWR) LOCA analyses supporting licensing for the Yankee Rowe, Maine Yankee and Seabrook Nuclear Power Stations. Areas of involvement included LOCA and containment analyses and severe accident analyses related to Individual Plant Evaluations (IPEs). Specific accomplishments included supporting the Maine Yankee small break loss of coolant accident (SBLOCA) analysis to justify a return to 2440 MWth operation, and providing oversight of Siemens Power Corporation SBLOCA re-analysis. Served as a response team member to Maine Yankee 1996 Independent Safety Assessment.
In addition, supported General Electric (GE) in severe accident analysis for simplified boiling water reactor (SBWR) certification and provided consulting to the Siemens fuel user group in the area of LOCA analysis.
Provided support to Northeast Utilities on severe accident management guidelines (SAMGs) for Millstone Units 2 & 3 and the Seabrook Nuclear Power Station and performed a technical review of ABB/CE reload analysis of St. Lucie Unit 2 for FP&L.
Senior E-ngineer - Yankee Atomic Electric Company 5/88 - 5/91 Lead Engineer for Yankee Rowe and Seabrook LOCA analysis related activities and for all severe accident analysis activities. Duties involved reload licensing analysis for the Yankee Rowe plant and vendor oversight of Seabrook LOCA analysis activities.
Supported the plant specific model development and certification of the Yankee Rowe plant simulator. Participated in the Yankee Rowe plant life extension (PLEX) effort providing support in severe accident evaluations and pressurized thermal shock (PTS) analysis. Also, provided training to Texas Utilities personnel in LOCA analysis method applications.
Ntclear Engineer - Yankee Atomic Electric Company 5/85 - 5/88 Lead Engineer for Yankee Rowe LOCA analysis related activities. Activities included large break loss of coolant accident (LBLOCA) model development and applications related to reload licensing, steam generator tube rupture (SGTR) analysis and plant request responses. Participated in the development of Yankee Rowe, plant specific emergency operating procedures (EOPs) based on the generic Westinghouse Owners Group (WOG) emergency response guidelines (ERGs). Performed a plant specific analysis to support deviations from the generic WOG guidelines. Also, provided training to Korean Power (KEPCO) engineers in LOCA analysis methods.
2
Robert C. Harvey Engineer - Yankee Atomic Electric Company 6/79 - 5/85 Performed LBLOCA analyses in support of reloads for the Yankee Rowe and Maine Yankee plants. Contributed to model enhancements of the LBLOCA methods used for Yankee Rowe. Participated in developing and assessing the RELAP5YA computer code used for PWR SBLOCA analysis.
PROFESSIONAL AFFILIATIONS/CERTIFICATIONS:
American Nuclear Society (ANS), Member The Research Society of Sigma Xi, Associate Member Registered Professional Engineer North Carolina (Registration # 027387)
South Carolina (Registration # 22237)
SELECTED PUBLICATIONS:
- 1. Maine Yankee Steam Generator Tube Sleeving Thermal-Hydraulics and Safety Analysis Impacts, co-authors K. R. Rousseau, S. Palmer, P. A. Bergeron, presented at the American Power Conference, Chicago, Ill., 1995.
- 2. Maine Yankee Cycle 15 Core Performance Analysis, YAEC- 1907, co-authors, January 1995.
- 3. Yankee Rowe Pressurized Thermal Shock, Thermnal-Hydraulic Analysis, International Heat Transfer Conference, co-authors P. A. Bergeron, N. Fujita, August 1993.
- 4. Maine Yankee Level II PRA Results, ASME/JSME International Conference on Nuclear Engineering, co-author K. E. St. John, March 1993.
- 5. Thermal Hydraulics Analysis of the Yankee Plant Due to a Stuck Open PORV Using RELAP5/MOD3 Computer Code, RELAP5/TRAC-B International Users Seminar, co-authors W. S. Yeung, R. K. Sundaram, November 1991.
- 6. Yankee Plant Small Break LOCA Analysis, YAEC-1732, co-authors S. Mihaiu-Westerlind, R. K. Sundaram, July 1990.
- 7. Yankee Nuclear Power Station Core 21 Performance Analysis, YAEC-1731, co-authors, July 1990.
- 8. Yankee Nuclear Power Station Severe Accident Closure Submittal, YAEC-171 1, co-authors, December 1989.
3
qjql Robert C. Harvey
- 9. Plant-Specific Analysis to Support the Yankee Emergency Operating Procedures, YAEC-1663, co-authors, April 1989.
- 10. Seabrook Station Risk and Plant Response for Low Power Operating Conditions, YAEC-1623, co-authors, March 1988.
- 11. RELAP5YA Simulation of LOFT Small Break Experiments L3-6 and L5-1, Transactions American Nuclear Society, Volume 55, co-authors, L. Schor, November 1987.
- 12. Estimate of Peak Clad Temperature and Its Uncertainty in a Large Break LOCA at Yankee Nuclear Power Station, YAEC-1431P, co-authors, R. K. Sundaram, K. E. St.
John, May 1984.
- 13. RELAP5YA - A Computer Program for Light Water Reactor System Thermal-Hydraulic Analysis, YAEC-1300P, co-authors, R. T. Fernandez, R. K. Sundaram, J.
Ghaus, A. Husain, J. N. Loomis, L. Schor, R. Habert, October 1982.
- 14. RELAP4 and RELAP5 Calculation of LOFT L3-5 and L3-6 Experiments: Comparison to Data, ANS Specialists Meeting on Small Break Loss-of-Coolant Accident Analyses in LWRs, co-authors, L. Schor, J. N. Loomis, A. Husain, August 1981.
- 15. RELAP4 Analysis of CREARE Flashing Transients with Reverse Core Steam Flow, Transactions American Nuclear Society, Volume 38, co-authors G. J. Brown, A.
Husain, August 1981.
- 16. Applications of a Lower Plenum Phase Separation Model to Yankee Rowe Large Break LOCA Analysis, YAEC-1231, Revision 1, co-authors, March 1981.
- 17. Maine Yankee Cycle 5 Core Performance Analysis, YAEC-1202, co-authors, December 1979.
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Attachment 3 BERT M. DUNN AREVA 3315 Old Forest Road Lynchburg,'VA 24501 EDUCATION: BS in Physics, Washington State University, 1968 MS in Physics, Lynchburg College, 1973 EXPERIENCE: AREVA Framatome ANP Inc. (formerly Framatome Nuclear Technologies Inc., Nuclear Power Division of the Babcock & Wilcox Company), Lynchburg, Virginia 3/84-Present Advisory Engineer AREVA representative to high burnup industry forums responsible for providing recommendations on licensing criteria to NRC. Lead on AREVA consultation with NRC on irradiated fuel LOCA testing at Argonne National Laboratory.
Member of U.S. NRC High Bumup Fuel Design Basis Accident Technical Evaluation PIRT (Phenomena Identification and Ranking Table) Panel.
Technical lead for the development of LOCA and Safety Analysis techniques for the licensing of new fuel cladding materials.
Technical lead for the development of evaluation techniques to determine the outcome of inherent boron dilution events.
Technical co-lead for the evaluation of best estimate LOCA licensing techniques.
Project lead for development of Loss-of-Coolant-Accident analysis capability for Westinghouse and Combustion Engineering plants.
2/83-3/84 Senior Product Manager Project Manager for Pressurized Thermal Shock (PTS) B&W Owners Group Task Force. Lead development of a probabilistic risk assessment for PTS in B&W plants.
I
vq13 Bert M. Dunn 1975-1980 Unit Manager. Emergency Core Cooling System Analysis Responsible for all ECCS evaluations of the performance of B&W-designed nuclear power plants.
1970-1975 Engineer/Supervisorv Engineer Licensing of B&W's ECCS evaluation models and development of techniques for evaluating reactor building subcompartment pressure forces.
1968-1970 Engineer - Douwlas United Nuclear Corporation, Richland. Washington Reactor physics, fuel engineering, operations.
PUBLICATIONS: Major contributor or the principal author of:
- a. BAW-1 0034, "Multinode Analysis of B&W's 2568 MWt Nuclear Plants During a Loss-of-Coolant Accident," October 1971.
- b. BAW-10045, "Multinode Analysis of B&W's 205-Fuel Assembly Nuclear Plants During Loss-of-Coolant Accident,"
May 1972.
- c. BAW-10052, "Multinode Analysis of Small Breaks for B&W's 2568 MWt Nuclear Plants," September 1972.
- d. BAW-10064, "Multinode Analysis of Core Flooding Line Break for B&W's 2568 MWt Internals Vent Valve Plants," April 1973.
- e. BAW-10091, "B&W's ECCS Evaluation Model Report With Specific Application to 177-FA Class Plants With Lowered-Loop Arrangement," August 1974.
- f. BAW-10091, Supplement 1, "Supplementary and Supporting Documentation for B&W's ECCS Evaluation Model Report With Specific Application to 177-FA Class Plants With Lowered-Loop Arrangement," December 1974.
2
AqN Bert M. Dunn
- i. BAW-10106, "- QUENCH - Digital Program for Analysis of
- Core Thermal Transients During Loss-of-Coolant Accident,"
May 1975.
3
(iqS Attachment 4 Di: J. Kevin McCoy Advisory Engineer - Materials FrainatoineANP, Inc.
General Backeround Dr. McCoy, a veteran scientist/engineer, has more than 20 years' experience, largely in the nuclear industry. He has proposed, planned, and executed research projects for the Nuclear Regulatory Commission (NRC), the Air Force Office of Scientific Research (AFOSR), the National Science Foundation (NSF), and the Electric Power Research Institute (EPRI). 'Dr.
McCoy's research included developing models for waste glass degradation, identifying a new mechanism for densification of ceramics, vectorizing computer codes to run 500 times faster, and studying boundaries between quasi-periodicity and chaos as it applies to hydrogen embrittlement. In his research, Dr. McCoy developed methods for calculating the positions of atoms at a crack tip. He even found a method for calculating the energy barrier that must be surmounted to break one atomic bond.
Dr. McCoy worked on the Yucca Mountain Project (YMP) for nearly eight years, providing leadership and technical expertise on materials selection and nuclear waste behavior. He coordinated the efforts of Framatome ANP, Inc. and Lawrence Livermore National Laboratory to identify material properties and make material selection for the waste packages slated for use in the proposed geologic repository. Also, Dr. McCoy led the revision of a comprehensive report on nuclear waste behavior. This report summarizes $25 million worth of research and is one of nine major documents used to support the site recommendation for the nation's first high-level radioactive waste repository.
His recent work has focused on the behavior of commercial nuclear fuel. In the past three years he has written or revised topical reports on dry fuel storage and the performance of a mixed oxide fuel assembly. He has also written several reports on poolside postirradiation examinations of irradiated fuel.
Education
- Ph.D., Materials Engineering, Purdue University
- M.S., Metallurgical Engineering, Purdue University
- B.S. (with Highest Distinction), Metallurgical Engineering, Purdue University Oualifications/Certifications
- Good conversational and written French
- Listed in Who's Who in Science and Engineering
- Member, American Nuclear Society Publications
- Civilian Radioactive Waste Management System Management & Operating Contractor, W1aste Form Degradation Process AModel Report, TDR-WIS-MD-000001 REV 00 ICN 01, July 2000.
1
,iAq (a Dr. J. Kevin McCoy Representative of more than 30 peer-reviewed journal articles authored or co-authored are:
- D. Stahl, J. K. McCoy, and R. D. McCright, "Impact of Thermal Loading on Waste Package Material Performance", in Scientific Basis for Nuclear Waste Management XVIII, 671-678, ed. T. Murakami and R. C. Ewing, Materials Research Society, Pittsburgh (1995).
- J. K. McCoy, D. Stahl, and T. A. Buscheck, "A Corrosion Model for Waste Package Corrosion-Allowance Materials", in Proceedings of the Sixth Annual International Conference on High Level Radioactive Waste Management, 565-567, American Nuclear Society, La Grange Park, Illinois, and American Society of Civil Engineers, New York (1995).
- J. K. McCoy, "Fuel and Cladding Oxidation Under Expected Repository Conditions", in Proceedings of the Seventh Annual International Conference on High Level Radioactive Waste Management, 396-397, American Nuclear Society, La Grange Park, Illinois, and American Society of Civil Engineers, New York (1996).
Publications - continued
- J. K. McCoy, "Mechanical Failure of Commercial Spent Nuclear Fuel Cladding",
in Proceedings of the Sixth International Conference on Nuclear Engineering, 632-633, American Society of Mechanical Engineers, New York (1998).
- J. A. Blink, T. W. Doering, J. K. McCoy, R. W. Andrews, J. H. Lee, D.
Sevougian, V. Vallikat, D. G. McKenzie, and J. N. Bailey, "Factors Affecting Performance of Engineered Barriers", in Proceedings of the Eighth Annual International Conference on High-Level Radioactive WVaste Management, 290-292, American Nuclear Society, La Grange Park, Illinois (1998).
Applicable Work Experience Advisory Engineer II, Framatome ANP, 1993 - Present
- Revised topical report on performance of a mixed oxide fuel assembly. Report compares materials and performance of mixed oxide and low-enriched uranium fuels
- Wrote topical report on behavior of spent nuclear fuel in dry storage. Provided justification for increasing the allowable burnup limit
- Prepared six reports on poolside postirradiation examinations of nuclear fuel assemblies
- Predicted behavior of high-level radioactive wastes and support selection of waste container material 2
SlI 41-1 Dr. J. Kevin McCoy
- Led revision of a comprehensive report on nuclear waste behavior. Report summarizes $25 million worth of research and is one of nine major documents that supported site recommendation for the nation's first high-level nuclear waste repository
- Critically reviewed models for creep rupture of spent fuel cladding. Determined that previous models did not adequately account for cladding texture. Showed that inclusion of the effects of texture increases predicted creep life by a factor of six
- Developed novel mathematical approach for describing degradation of spent nuclear fuel. The new approach is based on a deep insight into the similarities in the degradation behavior of fuel rods that start to degrade at different times.
Lengthy performance calculations are now performed more than ten times faster
- Developed computer model to simulate bending and breaking of fuel rods during earthquakes. Proved that earthquakes strong enough to break fuel rods occur only once in a million years and would have very little effect on repository performance.
Principal Research Scientist, Battelle, Metals and Ceramics Department, Columbus, Ohio (1981-1992)
- Proposed, planned, and executed research projects. Analyzed physical processes in materials, mostly with C code. Provided own research support. Served as system administrator for departmental UNIX computer system with twenty users
- Developed methods for calculating the positions of atoms at a crack tip. Found method to calculate the energy barrier that must be surmounted to break one atomic bond. Knowing the height of the barrier is critical for calculating crack growth rates. This breakthrough came after several years of unsuccessful efforts by other scientists
- Developed technique for instrumenting hot isostatic pressing. Instrumentation provides a continuous record of material behavior and increases the amount of data obtained from costly experiments by three to five times
- Managed personal computer resources for department with fifty users. Planned hardware acquisitions and allocated resources.
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July 8, 2004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of: )
)
DUKE ENERGY CORPORATION )
) Docket Nos. 50-41 3-OLA (Catawba Nuclear Station, ) 50-414-OLA Units I and 2) )
)
)
REBUTTAL TESTIMONY OF STEVEN P. NESBIT, ROBERT C. HARVEY, BERT M. DUNN, AND J. KEVIN McCOY ON BEHALF OF DUKE ENERGY CORPORATION ON CONTENTION I
TABLE OF CONTENTS
- 1. INTRODUCTION .I II. GENERAL OBSERVATIONS (A.4 - A.6)........................................................................2 III. LOCAL LINEAR HEAT GENERATION RATE (A. 8, A.1O) .7 IV. PELLET FRAGMENT SIZE (A.l 1) .8 V. FUEL-CLADDING INTERACTION (A.12). . 13 VI. CLADDING BALLOON SIZE (A.12 - A.14) .15 VII. COOLABLE CORE GEOMETRY (A.15) .22 VIII. SAFETY MARGINS (A.16) .23 IX. "GAPS IN THE EXPERIMENTAL DATABASE" (A. 17) .26 X. CONCLUSION .27 i
July 8, 2004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of: )
)
DUKE ENERGY CORPORATION )
) Docket Nos. 50413-OLA (Catawba Nuclear Station, ) 50-414-OLA Units l and 2) )
)
REBUTTAL TESTIMONY OF STEVEN P. NESBIT, ROBERT C. HARVEY, BERT AM. DUNN, AND J. KEVIN McCOY ON BEHALF OF DUKE ENERGY CORPORATION ON CONTENTION I I. INTRODUCTION
- 1. (Nesbit) 1, Steven P. Nesbit, am an Engineering Supervisor II employed by Duke Energy Corporation (Duke). I currently serve as the Duke Mixed Oxide (MOX) Fuel Project Manager. A full statement of my Professional Qualifications was included with Duke's initial written testimony in this proceeding.
- 2. (Harvey) I, Robert C. Harvey, am a Senior Engineer employed by Duke, responsible for the Loss of Coolant Accident (LOCA) analyses supporting the Oconee Nuclear Station (Oconee), McGuire Nuclear Station (McGuire), and Catawvba Nuclear Station (Catawba).
A full statement of my Professional Qualifications was included with Duke's initial written testimony in this proceeding.
- 3. (Dunn) 1, Bert Dunn, am an Advisory Engineer employed by AREVA Framatome ANP, Inc. A full statement of my Professional Qualifications was included with Duke's initial written testimony in this proceeding.
I
pap t
- 4. (McCoy) I, J. Kevin McCoy, am an Advisory Engineer in the fields of metallurgy and materials engineering, employed by AREVA Framatome ANP, Inc. A full statement of my Professional Qualifications was provided with Duke's initial written testimony in this proceeding.
- 5. (All) Based on our specific job responsibilities, we are all very familiar with Duke's License Amendment Request (LAR), dated February 27, 2003. The LAR seeks NRC approval for Duke's proposal to use four MOX fuel lead assemblies at Catawba. The MOX fuel lead assemblies will be included in cores that will be predominantly comprised of Low Enriched Uranium (LEU) fuel assemblies.
- 6. (Alli) The purpose of this rebuttal testimony is to specifically address the written testimony of Dr. Edwin S. Lyman regarding Contention I in this proceeding. Dr. Lyman's testimony was submitted on behalf of the Blue Ridge Environmental Defense League (BREDL) on July 1, 2004. We show how the issues raised by Dr. Lyman were previously addressed in our initial direct testimony in this proceeding, also filed on July 1, 2004. We also provide some additional responses to those issues. In general, as we concluded previously, the issues (or "uncertainties") raised by Dr. Lyman are well-bounded by the conservatisms in the NRC's requirements and in Duke's analyses of a postulated Loss of Coolant Accident (LOCA).
II. GENERAL OBSERVATIONS (A.4 - A.6)
- 7. (Nesbit) Dr. Lyman summarizes the BREDL case by stating in his Answer to Question 4 (A.4)1 that: "In my professional judgment, Duke's design-basis loss of coolant ('DB-LOCA') analysis is inadequate because it does not address the uncertainties associated with relocation effects that M5-clad MOX fuel may experience under LOCA conditions." He goes on References to specific Answers in the BREDL or NRC Staff testimony are referenced herein as "A.x," where x is the Question/Answer number.
2
11,02' in A.4 to state that: "Duke has failed to address these uncertainties in MOX fuel behavior, and therefore its [lead test assembly] application is unacceptable to satisfy the requirements of 10 C.F.R. § 50.46 with respect to [peak cladding temperature (PCT)], maximum cladding oxidation, and coolable geometry of fuel. In addition, by failing to address the uncertainties in MOX fuel behavior, Duke has not demonstrated compliance with the general reasonable assurance standard in 10 C.F.R. § 50.40(a)."
- 8. (Nesbit) First, Dr. Lyman provides no expert opinions, data, or analyses to support his professional judgment. Dr. Lyman claims to be a "qualified expert on nuclear safety and safeguards issues." However, he has no experience performing LOCA analyses or conducting experiments related to LOCA phenomena. He has written and spoken extensively on nuclear issues, but his work is predominantly in the policy arena. Dr. Lyman's judgments are based entirely on statements made by other people. BREDL offers none of those people as expert witnesses so they can be questioned by Duke. None of the sources referenced by Dr. Lyman specifically address the safety of the Duke MOX fuel lead assembly license amendment request.
The sources of most of the BREDL "uncertainties" are proposals for experiments by a French research agency, the Institute for Radiological Protection and Nuclear Safety (IRSN). IRSN's desire to conduct experiments related to MOX fuel must be viewed in context - French nuclear safety regulators have taken no action to curtail MOX fuel use in the twenty French reactors that use substantial quantities of MOX fuel.
- 9. (Nesbit) Dr. Lyman has made no definitive statement that the MOX fuel lead assemblies would be unsafe - he claims in A.4 only that ". . . the impact of fuel relocation effects during a LOCA may be more severe for MOX fuel rods than for LEU fuel rods" (emphasis added). However, the presence of "uncertainty" is not surprising and certainly does 3
not alone preclude approval for a lead assembly program. There is no such thing as perfect certainty, particularly for a lead assembly program that has, as its fundamental purpose, the goal of acquiring more data.
- 10. (Nesbit) Moreover, a proposal for research by a foreign organization cannot be equated with the existence of Unacceptable uncertainty. Would BREDL apply this same requirement across the board - must every nuance of every possible difference associated with every change to every nuclear plant structure, system, or component be exhaustively tested at accident conditions prior to implementing that change? WAe addressed the uncertainty issue in detail in Section VI (paragraphs 158-171) of our initial testimony. We continue to take issue with Dr. Lyman's judgment in this regard.
- 11. (Nesbit) Most significantly, uncertainty is routinely addressed by conservatism (or "margin") in regulatory requirements and licensing analyses. Dr. Lyman does not acknowledge the inherent margin in Appendix K LOCA calculations. In fact, there is substantial conservatism in Appendix K design basis LOCA calculations, as we discussed in detail in Section III.D (paragraphs 60-65) of our initial testimony. These conservalisnmsadd up to approximately 600 "F PCTmarginrelative to best estimate calculations.
- 12. (Nesbit) Dr. Lyman also does not acknowledge beneficial characteristics of the MOX fuel lead assemblies with respect to Appendix K LOCA calculations. Section III.C (paragraphs 43-55) of our initial testimony summarized the conservative modeling techniques employed by AREVA in the MOX fuel Appendix K evaluation model. These include using LEU models for decay heat and neutron power, rather than taking credit for beneficial MOX fuel characteristics for those parameters (see paragraphs 63-65 of our initial testimony). Dr. Lyman claims in A.4 that "calculations in Duke's LAR indicate that MOX fuel is generally more 4
limiting than LEU fuel with respect to DB-LOCAs." In fact, however, as discussed in paragraph 58 of our previous testimony, we showed that an "apples to apples" comparison of MOX and LEU fuel shows essentially no difference (less than 40'F in PCT) between the fuel types, without taking any credit for the beneficial aspects of MOX fuel.
- 13. (Harvey) In his response to Question 6 (A.6) Dr. Lyman alleges the "non-conservatism" in Appendix K based on the "omission of fuel relocation effects" as supposedly reflected in various regulatory documents and NRC Generic Issue (GI) 92. We addressed this in our initial testimony at paragraphs 94-96 and 144. Dr. Lyman goes on to state". . . the NRC has acknowledged that omission of fuel relocation effects is a non-conservatism in Appendix K with a very large potential impact on PCT. . . ." To our knowledge, however, the NRC has never made such a statement - it is certainly not present in BREDL Exhibits B and C referenced by Dr. Lyman in A.4. Those NRC documents only state that fuel relocation remains an issue of interest in Europe, and that fuel relocation during a LOCA is one potential non-conservatism in present Appendix K methods (evaluation models). BREDL Exhibit C, Attachment 4 at 5 (the "Thadani memorandum"), suggests an impact on PCT of at least +46 0 F but does not endorse the higher PCT values calculated by Grandjean, et al. (BREDL Exhibit E; Duke Exhibit 4).
- 14. (Harvey) As we stated in Section III.D of our initial testimony (paragraphs 60-62), in addition to any non-conservatism related to fuel relocation, there are many clearly identified conservatisms in the Appendix K methods. In fact, the Thadani memorandum (BREDL Exhibit C), identifies one proposed regulatory option which would allow licensees to continue to use the presently-accepted Appendix K methods as grandfathered and therefore no reanalysis would be required. In the other proposed option, a licensee could choose to use a revised Appendix K rule that would relax some of the known conservative requirements. The 5
LV?~5 licensee would then need to account for the potential non-conservatisms in the approved evaluation model. Clearly, the current Appendix K approach relies on the known conservatisms to more than offset any of the potential non-conservatisms. If a licensee would choose to apply the revised Appendix K requirements, better understanding of the non-conservatisms (including any fuel relocation effect) would be required in order to accurately represent those effects. This is not the case in the LAR, which is based on presently-accepted Appendix K methods and includes all of the inherent Appendix K conservatisms.
- 15. (Nesbit) In the second paragraph of A.6, Dr. Lyman makes other conclusory comments on the relocation issue. He suggests that "certain characteristics of MOX fuel appear to exacerbate the effects of fuel relocation, thus leading to higher PCTs and greater maximum cladding oxidation." However, he offers no quantitative assessment to dispute our quantitative analysis or to show how a MOX relocation effect could challenge the margin relative to a best estimate calculation. Furthermore, we have addressed in Sections IV and V of our initial testimony each and every one of the suggested, qualitative "differences" between MOX and LEU fuel and concluded that they do not involve significant negative impacts. We address these again in this rebuttal testimony, to the extent warranted.
- 16. (Nesbit) Finally, BREDL Contention I is also fundamentally a challenge to the application of Appendix K LOCA models to all nuclear fuel, including LEU fuel. Relocation can occur in LEU fuel during design basis LOCA conditions, and no Appendix K LOCA models address fuel relocation. Even if BREDL's supposition that relocation may be worse in MOX fuel were true, it is our professional judgment that the difference would only be a matter of a relatively small degree. At bottom, we do not believe that fuel relocation needs to be specifically modeled in LOCA analyses for either MOX or LEU fuels.
6
III. LOCAL LINEAR HEAT GENERATION RATE (A. 8, A.10)
- 17. (Dunn) Dr. Lyman starts from the premise, in his A.8, that postulated fuel relocation increases the local linear heat generation rate within the ballooned area and therefore could increase the severity of a LOCA by increasing PCT and maximum cladding oxidation relative to an analysis that does not model relocation. He specifically states that: ". . . the greater local linear heat generation rate requires a greater coolant flow around the ballooned area to ensure long-term core coolability."
- 18. (Dunn) With respect to the first part of this testimony in A.8, in Section V.G of our initial testimony (paragraphs 145-157) we have already shown that the impact of fuel relocation on calculated PCT and local cladding oxidation is not significant, even with very conservative assumptions.
- 19. (Dunn) With respect to the second part of the testimony in A.8, the fact is that there is no significant impact on "long-term coolability" around a ballooned area. Long-term cooling is established following initiation of core quench through the use of extended service pumped injection. For the most severe LOCA scenarios (breaks in the Reactor Coolant System cold legs), the core transitions to a boiling pot mode following reflood and core quench. The core inlet flow does not change due to fuel relocation - core flow is based on total core power (decay heat), not on the distribution of that power. Within the core region cooling is by pool boiling, a heat transfer mechanism that is dependent on coolant availability, but not on flow rate. The pool boiling mechanism is very efficient and is stable for heat fluxes more than an order of magnitude higher than those required by the long term core power. Coolant availability can be supplied from below, laterally across assemblies, or from down flow without difficulty. Cooling is ample at greatly reduced flow areas, even assuming a larger local heat source. Thus, there will be no 7
degradation in the long term cooling process for either LEU or MOX fuel, even if the local power generation were to increase due to fuel relocation.
- 20. (Dunn) Accordingly, both aspects implied by Dr. Lyman in A.8 are incorrect: the increase in local linear heat generation rate within the ballooned area will not lead to significant increases in PCT in that area and long-term core coolability will be maintained.
- 21. (Nesbil) In A.10 Dr. Lyman states: "MOX fuel may experience more severe relocation effects than [LEU] fuel at the same burnup because several characteristics that are important for relocation may be less favorable for MOX fuel. These include pellet fragment size and fuel-clad interaction." Again, Dr. Lyman makes no definitive statement that MOX fuel would be worse in this regard. As we have shown in our testimony and will show further in this rebuttal, there is no data indicating that MOX fuel would perform worse. However, Dr. Lyman has ignored the clear evidence that the Catawvba MOX fuel lead assemblies would be better than their co-resident LEU fuel assemblies in one key aspect - peak linear heat generation rate (see Section V.4, paragraphs 129-142 of our initial testimony).
IV. PELLET FRAGMENT SIZE (A.1 1)
- 22. (McCoy:) In paragraph 2 of A. 11, Dr. Lyman states that: "The fuel relocation phenomenon has been observed in LEU fuel for rod bumups exceeding around 48 GWD/t. See Grandjean, Hache and Rondier [sic] at 2 (2001) [BREDL Exhibit E; Duke Exhibit 4]. This suggests that vulnerability to fuel relocation is associated with the development of the high-burnup 'rim' region known to emerge in LEU fuel for burnups exceeding about 40-45 GWD/t."
- 23. (AMcCoy) Dr. Lyman's theory is that the MOX fuel pellets may experience greater fragmentation than LEU fuel pellets, and thus MOX fuel relocation will involve greater "fill fractions" with commensurately greater effects. We disagree with this theory.
8
- 24. (McCoy) Relocation during LOCA occurs when fuel pellet fragments are dislodged. Processes that might influence relocation include cracking of the pellets, accumulation of released fission gas, and bonding of the fuel to the cladding. In Section V.D of our initial testimony (paragraphs 117-121), we explained why no differentiation in pellet cracking and fragmentation should be made between MOX and LEU fuel. This was supported by the micrographs of irradiated fuel.
- 25. (McCoy) Dr. Lyman's argument appears to hinge on speculation that the susceptibility of fuel to relocation during LOCA is somehow related to the formation of a "rim" microstructure in the fuel. The "rim" microstructure develops in fuel that is irradiated at relatively low temperatures until a high local bumup is reached. It is characterized by a marked refinement of the grains and increased porosity. The "rim" microstructure is so named because in LEU fuels it appears first at the extreme periphery of the pellets. This is the region of an LEU fuel pellet where the local burnup is highest and the temperature is lowest.
- 26. (AMcCoy) Dr. Lyman suggests a relationship between relocation and microstructure of the fuel that is based only on the fact that susceptibility to fuel relocation and the formation of a "rim" structure occur at roughly comparable burnups. No mechanistic argument is given to relate relocation with the development of a "rim" structure.
- 27. (McCoy) Relocation requires that the fuel pellets lose their mechanical integrity.
Therefore, if the "rim" microstructure is to affect relocation, it must do so by affecting the mechanical properties of the fuel. The mechanical properties of the "rim" regions are in fact different from those of the remainder of the fuel. Specifically, the "rim" regions are tougher, that is, more resistant to cracking. Therefore, the appearance of "rim" regions will not contribute to a loss of mechanical integrity or increase the susceptibility to relocation.
9
- 28. (McCoy) In paragraph 3 of Dr. Lyman's A. 11, he states that: "... it is reasonable to expect that the onset of fuel relocation in MOX fuel may occur at lower rod-average burnups than in LEU fuel. This would imply that MOX fuel will be vulnerable earlier in its irradiation history (and consequently for a longer time) than LEU fuel." We do not agree with the logic of this testimony.
- 29. (McCoy) Dr. Lyman's discussion follows the same logic as his previous testimony: the "rim" microstructure appears earlier in MOX fuel than in LEU fuel, so if the "rim" microstructure embrittles the fuel, MOX fuel may be susceptible to relocation earlier than LEU fuel. But in fact the "rim" regions are tougher than the balance of the fuel, so the reasoning fails.
- 30. (McCoy) In paragraph 3 of A.1 1, Dr. Lyman next states that: "...the particle size distribution in MOX fuel vill be smaller than in LEU fuel at the same rod-average burnup, to the extent that fine fragments are generated in the ultra-high burnup plutonium agglomerate regions." The apparent meaning of this statement is as follows: for fuel that forms large fragments of a certain size, plus fine particles of a smaller size, the mean particle size will decrease if the fraction of small particles increases.
- 31. (McCoy) We agree with Dr. Lyman's implied statement that the large fragments formed in MOX and LEU fuels with a given rod-average burnup will be of similar size.
However, Dr. Lyman has not produced any evidence for his implied assertion that the plutonium-rich agglomerates in MOX fuel will produce a larger fraction of fine particles than will LEU fuel. The extra toughness imparted by the formation of the "rim" structure actually suggests that the agglomerates will not yield fine particles.
10
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- 32. (McCoy, Nesbit) The issue of whether or not MOX fuel has a greater propensity to produce smaller fragments was discussed at the Advisory Committee on Reactor Safeguards (ACRS) meeting on MOX fuel (April 24, 2004). At that meeting, Dr. Lyman admitted that he has no evidence that MOX fuel has a smaller fragment size: "I would guess, to the extent that MOX starts fragmenting at lower bum-ups than LEU and a greater part of the fuel pellet is affected and fragmented, that may mean the mean particle size - or fragment size is lower for MOX. But I don't have any - I haven't seen anything." (Tr. at 27.) Dr. Dana Powers, the ACRS Subcommittee Chairman, responded: "What I am not aware of, and maybe you can help me there, is a tendency for MOX to fragment more extensively than LEU. In fact, one would think that MOX would have inherently a low fragmentation tendency, because crack tips get blunted." (Tr. at 17.) Later in the discussion, Dr. Lyman stated: "I don't know what it [the MOX/LEU difference in particle size distribution following fuel relocation] is, but it is certainly a difference that should be considered. It's possible that it is favorable [to MOX fuel]. (Tr. at 28.)
- 33. (McCoy) In paragraph 4 of Dr. Lyman's A. I I, he next states that: "Because MOX fuel has a lower thermal conductivity and a higher radial temperature gradient than LEU fuel, it could experience greater fuel fragmentation during the blowdown and more severe relocation effects as a result." The point of the argument appears to be that MOX fuel has a lower thermal conductivity than LEU fuel, and therefore it is more susceptible to fragmentation by thermal shock.
- 34. (McCoy) Paragraph 68 of our initial direct testimony shows that the MOX/LEU difference cited by Dr. Lyman is not a significant effect. Figure 5 in that testimony shows that the difference between MOX and LEU thermal conductivity is small and that this difference does 11
not change significantly with burnup. The MOX pellet temperature profile should be very close to the LEU pellet temperature profile; accordingly, there will be no significant differences in thermal gradients or thermal stresses during blowdown. Therefore, the effect of these stresses on pellet fragmentation will be similar to that in LEU fuel.
- 35. (McCoy, Dunn) Finally, even if it could be shown that MOX fuel produces a larger fraction of fine particles than does LEU fuel, we expect that a substantial portion of such fine particles would not be retained in the cladding balloon and therefore would not be relevant to LOCA performance. In this regard, the testimony of Dr. Ralph Meyer on behalf of the NRC Staff (in his response to Question 40) indicates his view, based on recent high burnup integral tests, that the fine particles would not be retained: "It ... appears that the small particles or fines are blown out of the burst opening when the rod depressurizes. Thus, there would be few or no small particles in the ballooned region, and it is these small particles that have been postulated to make a difference between the mass of fuel in the balloon in MOX fuel and LEU fuel."
- 36. (Harve;) In the third paragraph on page 6 in Dr. Lyman's A. 11, he next refers to differences of opinion among "NRC experts" who participated in the 2001 PIRT panel on LOCAs and high burnup fuel. The PIRT panel actually looked at a number of aspects of high burnup fuel with respect to a LOCA event. The testimony of Dr. Lyman fails to point out that, under the category of "plant transient analysis," none of the PIRT panel rated fuel relocation as important; two rated it as "low medium" importance; and four rated it as low importance, stating that it has a small effect on the system analysis and it could make the burst (ruptured) node limiting. The last statement clearly reflects our point that the ruptured node is not necessarily the location of PCT. This latter point is demonstrated in the FR-2 tests without relocation illustrated by Figure 10 in our initial direct testimony. For the Catawba MOX fuel lead assemblies, the 12
q/1o V PCT is not at the ruptured location, as shown by the Catawba MOX calculation (LAR Table 3-5)(Duke Exhibit 1).
- 37. (Harvey) Under the category of "transient fuel rod analysis," one expert rated fuel relocation as high importance, but qualified that by stating it was plant-dependent (that is, if the ruptured node was limiting it could make the event worse). Again, however, the ruptured node is not the highest PCT node for the Catawba MOX fuel LOCA analysis (as also described in paragraphs 36-40 and shown on Figure 1 of our initial testimony). The five other experts rated fuel relocation as having medium importance, stating only that it has a modest impact on the local heat rate.
- 38. (Harvey) In total context, there is little support from the PIRT expert panel that fuel relocation is an important phenomenon with respect to LOCA analyses, or more particularly with respect to the Catawba LOCA analyses. The fact that the ruptured node at Catawba is not the limiting PCT node renders speculation from various members of the PIRT panel largely academic.
V. FUEL-CLADDING INTERACTION (A.12)
- 39. (McCoy, Dunn) In paragraph I of A.12, Dr. Lyman, states that: "According to IPSN (now IRSN), tight fuel-clad bonding may delay the onset of fuel relocation. Mailliat and Schwarz at 433 [BREDL Exhibit G]. Tight bonding has also been observed at the Halden reactor in Nonvay to retard the rate of balloon formation."
- 40. (McCoy, Dunn) The question of fuel-cladding bonding is discussed at length in Section V.E of our initial direct testimony (paragraphs 122 through 128). Our conclusion is that "hypothetical MOX/LEU differences in this area do not affect the current regulatory position that relocation effects, if present at all, are adequately bounded by the inherently conservative nature 13
1,I.A%
of Appendix K LOCA analyses" (paragraph 128). This conclusion remains valid. Nonetheless, we will make a few additional points.
- 41. (McCoy, Dunn) In paragraph I of Dr. Lyman's A.12, he states: "It has been confirmed that MOX fuel is more resistant to clad failures due to pellet-clad mechanical interaction (PCMI) than LEU fuel, even at high burnups. Nuclear Energy Agency, NEAINSC/DOC(2004)8, InternationalSeminar on Pellet-CladInteractionswith Water Reactor Fuels, at 20 (May 6, 2004) [BREDL Exhibit J] ..... This phenomenon is not wvell-understood but may imply that the pellet-clad bond is weaker for MOX fuel, in which case MOX fuel may have a greater propensity to earlier and more extensive fuel relocation than LEU."
- 42. (AMcCoy Dunn) With regard to this testimony, it is worthwhile to refer directly to the document cited (BREDL Exhibit J). The relevant text reads as follows: "The reasons why MOX fuel and [chromium (Cr)-doped] fuel appear to behave better with respect to conventional
[LEU fuel] under PCI conditions must be tackled further: is it fuel cracking propensity by itself, and/or is it enhanced viscosity reducing the hour-glass effect by dish filling and perhaps favouring peripheral cracking?" The original report makes no suggestion that pellet-cladding bonding is a possible explanation. Dr. Lyman's connection betveen PCMI performance and pellet-cladding bonding is speculation.
- 43. (McCoy, Dunn) Even if it were supposed that PCMI performance is influenced by pellet-cladding bonding, the strength of the bond is expected to be similar for MOX and LEU fuels. This was discussed in paragraph 123 of our direct testimony.
- 44. (Nesbit) In paragraph 2 of A.12, Dr. Lyman challenges Duke for failing to quantify the amount of conservatism in the MOX fuel LOCA analysis. He states that "... there is no way of knowing the degree to which this assumption (no fuel-cladding bonding) is 14
conservative for MOX fuel" and that this ". . contributes another uncertainty to the safety margin associated with Duke's design basis LOCA calculation."
- 45. (Nesbit) In this specific instance, Duke's analysis takes no credit for fuel-cladding bonding to mitigate cladding swelling. This is a standard conservative assumption in Appendix K LOCA analyses of LEU fuel. There is no requirement that the impact of the "no fuel-cladding bonding" assumption be quantified for LEU fuel, and to our knowledge, no one has ever done so.
- 46. (Nesbit ) Dr. Lyman specifically argues that Duke has failed to quantify .. . the degree to which this assumption is conservative for MOX fuel." This argument reflects a lack of understanding of the governing requirements for LOCA emergency core cooling system analyses, 10 C.F.R. § 50.46 and 10 C.F.R. Part 50, Appendix K. Appendix K LOCA analyses provide a conservative estimation of PCT and cladding oxidation, as discussed in Section IHI.D (paragraphs 60-65) of our initial testimony. There is no requirement that licensees quantify the margin present in every individual aspect of an Appendix K LOCA analysis. In the overall context of Appendix K conservatisms, we do not see the need to quantify the value of one specific conservatism.
VI. CLADDING BALLOON SIZE (A. 12 - A. 14)
- 47. (Harvest Dunn) In the last paragraph of his response to Question 12, and in A. 13, Dr. Lyman raises the issue of the ballooning of the fuel cladding, specifically considering (i) the assumed ductility of the cladding in the MOX fuel LOCA analysis and (ii) the relative ballooning of M5TNI cladding. These issues are addressed in Section V.C of our initial testimony (paragraphs 109-114).
- 48. (Harvey) Dr. Lyman's A.12 states: "According to IPSN (now IRSN), results from the PBF-LOC experiments found that irradiated rods experienced greater clad deformation than unirradiated rods during design-basis LOCA conditions. . . . There is simply no way to 15
determine whether Duke's design-basis LOCA analysis underestimates or overestimates the degree of clad swelling (and hence the degree of fuel relocation) for MOX LTAs without additional experimental data from integral LOCA tests of high-burnup MOX fuel rods." I disagree.
- 49. (Harvey) The PBF test was conducted with irradiated fuel rods that had relatively little burnup. The burnup of the fuel rods ranged from 10.8 to 17.7 GWD/t. In addition, there
,were a limited number of PBF test rods and only one set of rods (unirradiated v. irradiated) for which the test conditions were similar. It is not appropriate to draw a conclusion about irradiation effects based on this limited set of data. In any case, the PBF test results have been discussed in multiple reports, one of which is NUREG-1230.2 These reports (including NUREG-1230) attribute the increased cladding strain observed in the PBF test to more uniform temperature distribution around the circumference for the irradiated rods, and not to irradiation effects. It should be pointed out that the Catawba MOX LOCA analysis assumes a uniform temperature distribution around cladding circumference. This assumption conservatively maximizes the calculated cladding strain (ballooning) for all fuel conditions (irradiated and unirradiated).
- 50. (Harvey) Other in-reactor fuel deformation tests were conducted in the FR-2 reactor. The FR-2 tests were conducted for irradiated rods with fuel rod bumups ranging from 2.5 - 35 GWD/t. These tests showed no influence of irradiation damage (i.e., there was no shift in cladding strain with fuel rod exposure). In addition, it has been suggested that the effect of irradiation damage on the cladding are annealed out when the temperatures exceed 7001C (about 1300'F) during the accident.
16
- 51. (Harvey) In conclusion, the differences between the irradiated and unirradiated tests results at PBF have been explained. Irradiation tests conducted at FR-2 have concluded that there is no significant effect of irradiation damage on cladding strain (ballooning).
- 52. (Dunn) Next, in A.13, Dr. Lyman raises the issue of M5TM cladding that is addressed in Section V.C of our initial testimony. Dr. Lyman testifies that: "According to IRSN, M5 will form larger balloons than Zircaloy-4 in a design-basis LOCA because it remains more ductile during irradiation. October 2003 IRSN presentation to NRC at 24. The greater retained ductility of M5 as a function of burnup compared to Zircaloy-4 can result in a greater M5 balloon size during a design-basis LOCA for fuel rods of the same burnup. Larger balloons increase the space available for fuel fragments to fall and hence result in a greater propensity for fuel relocation during a LOCA, with an associated increase in PCT and local clad oxidation."
We have already testified, however, that "there is little expected difference in the consequences of fuel relocation due to cladding differences" (paragraph 114 of our initial direct testimony).
- 53. (Dunn) The ballooning characteristics of the cladding need to be determined from LOCA-relevant ramp tests with increasing cladding temperatures. The IRSN statements quoted by Dr. Lyman are based on constant temperature creep tests. The strain results between these two test types or test conditions vary substantially for either Zircaloy-4 or MSTM. This point was the thesis of a paper delivered by Nicolas Waeckel at the meeting at Argonne in May 2004.
(NRC Staff Exhibit 4).
- 54. (Dunn) In addition to demonstrating that creep tests cannot be, employed to simulate a LOCA, the WVaeckel paper compared the available ramp test data from the EDGAR 2 NUREG-1230, "Compendium of ECCS Research for Realistic LOCA Analysis" (December 1988).
17
1/I'lI test program involving simulated irradiated Zircaloy-4 and M5Th¶ cladding. Irradiated cladding can be simulated by doping fresh cladding with hydrogen. Zircaloy-4 is more ductile for ramp testing than M5TM in its fresh or unirradiated state. However, as shown by Waeckel, when simulated end-of-life (EOL) cladding is tested, the strain data shows only a small reduction in strain for M5 TM cladding relative to virgin cladding. There is a larger reduction in strain for Zircaloy-4 cladding in the alpha phase when irradiation effects are simulated. As a result, at EOL conditions the strain developed by the two claddings at rupture is very similar, without significant difference except for phase transition temperatures.
- 55. (Dunn) The NRC Staff addresses this issue in its answer to Question 35 of the Staff direct testimony, and also cites the Waeckel paper (NRC Staff Exhibit 4). The Staff conclusion and associated rationale are consistent with our own.
- 56. (Dunn) Thus, Zircaloy-4 cladding will produce higher strains than M5TN for low irradiation levels but will approach the M5TN1 strain as the claddings approach EOL conditions.
As such, Zircaloy-4 produces a larger ballooned volume into which fuel fragments can relocate for low irradiation cladding and essentially the same relocation volume for highly irradiated cladding. Thus, at the burnup most commonly associated with a real potential for fuel relocation, there is essentially no difference in the ballooned volumes between the two cladding types.
- 57. (Nesbil) In Dr. Lyman's answer to Question 14, he tries to rebut the NVaeckel presentation (NRC Staff Exhibit 4). Dr. Lyman acknowledges that a ramp test is more relevant than a creep test, but nonetheless questions the validity of using pre-hydrided but unirradiated cladding material to simulate the performance of irradiated cladding. He notes that M5Th1 cladding has less oxidation (corrosion) than Zircaloy4 during normal operation, and he posits that the potential for spalling of Zircaloy-4 at high burnups ". . . will cause spatial 18
inhomogeneities in the clad temperature that negatively affect ductility, leading to earlier cladding ruptures during a LOCA and hence smaller balloon sizes." He argues, therefore, that:
"I don't believe that the EDF presentation fully addresses the differences that would be observed in actual irradiated fuel with regard to the ductility and the balloon size of M5 compared to that of zircaloy-4."
- 58. (Nesbit)Section III.C (paragraphs 48-54) of our testimony points out that AREVA used M5TM-specific properties in its evaluation of MOX fuel cladding ballooning and rupture during a design basis LOCA. As we note in Section V.C (paragraph 113) of our testimony, MS"' ductility does not change much with irradiation, relative to Zircaloy-4.
Nevertheless, AREVA used unirradiated (most ductile) M5TNM properties in order to maximize the extent of ballooning considered in the analysis. Therefore, the ARE VA analysis of the MOX fuel lead assemblies maximizes the calculated ballooning and associated flow blockage, consistent with Appendix K. The LOCA analysis showed that the regulatory acceptance criteria are met. Duke fundamentally disagrees with the implication that M5TNM cladding introduces unacceptable additional uncertainty with respect to fuel relocation during LOCA.
- 59. (Nesbit) Dr. Lyman in A.14 is clearly - but unnecessarily - concerned about Duke's proposed use of M5TM with MOX fuel. The characteristics that make MTS an attractive cladding material - low corrosion, retention of ductility with irradiation - seem to be undesirable to BREDL. Dr. Lyman notes that spalling in Zircaloy-4 might lower cladding ductility and lead to an earlier cladding rupture during a postulated design basis LOCA. Such a rupture would limit balloon size and thereby provide unquantified margin with respect to fuel relocation impacts. However, it is not preferable to use cladding that might corrode and spall simply because such spalling could be hypothesized to be beneficial during a postulated design 19
basis LOCA for which regulatory limits are already met. In total context, we do not believe that such an approach would be very prudent.
- 60. (Dunn) Citing BREDL Exhibit K, Dr. Lyman next claims in A.14 that: "The Electric Power Research Institute (EPRI) and Areva (parent company of Framatome ANP) apparently continue to deny NRC access to samples of irradiated high-burnup M5-clad LEU fuel for testing at Argonne National Laboratory.... This lack of cooperation can only cause further delays in the ability of NRC to obtain the experimental data it needs to confirm the safety of high-burnup M5-clad fuel (whether LEU or MOX)." While this characterization is largely immaterial to the technical issues raised in Co'ntention I, some response seems warranted.
- 61. (Dunn) First, I find mystifying the assertion concerning a lack of cooperation by AREVA that is developed by Dr. Lyman from the NRC letter to EPRI that is BREDL Exhibit K.
The text of the letter alludes to no such "lack of cooperation." AREVA was in fact the first vendor to agree to supply non-irradiated advanced alloy cladding to NRC for LOCA testing.
Further, AREVA is currently working with the NRC to develop a mutually acceptable memorandum of understanding (MOU) prescribing the management and conduct of LOCA testing on irradiated M5TM fuel rods. AREVA expects to complete the MOU such that irradiated M5nTh fuel rods will be available to the NRC for LOCA research.
- 62. (Dunn) In A. 14 Dr. Lyman next claims that: "For some reason, France is reluctant to use M5-clad MOX fuel domestically and is primarily producing it for export to Germany (and now to the United States). However, even in Germany the use of M5-clad MOX has been extremely limited." While again largely immaterial to Contention 1, this testimony warrants a response to clarify the record.
20
I'V/0
- 63. (Dunn, Nesbit) Contrary to the implication in Dr. Lyman's testimony, there is no nefarious reason that French reactors do not extensively use M5TN¶ cladding. The French nuclear industry ("France") is a not a single entity. Framatome ANP is a worldwide fuel vendor, while EDF is the French utility (and a Framatome customer) that operates the French nuclear power plants. The two corporations are separate and have distinct goals. Framatome ANP developed the M5TM advanced cladding alloy to provide for enhanced performance of pressurized water reactor fuel, with an emphasis on high burnups. EDF's current fuel management scheme does not incorporate high burnup to the extent of other Framatome customers. Accordingly, EDF has not yet chosen to deploy advanced cladding materials, including M5TM, on a large scale in its nuclear fuel (either LEU or MOX).
- 64. (Dunn, Nesbit) Furthermore, since 1998, 193 MOX fuel assemblies with M5ST M cladding have been delivered to four reactors in Germany. This is comparable to the number of M5TMCjad LEU fuel assemblies (192) that have been supplied to German reactors in the same time period. While this may seem "extremely limited" to BREDL, it is far more than the number of M5TM-clad MOX fuel assemblies that Duke proposes to use at Catawba (four). The total experience base for MSm-clad fuel assemblies is substantial and growing, with more than 3070 such assemblies supplied worldwide, to 41 reactors, through mid-2004.
- 65. (Dunn, Nesbit) To our knowledge, as Dr. Lyman states at the end of A.14, there have been no integral LOCA tests performed on irradiated M5TM-clad MOX fuel. For that matter, to our knowledge, there have been no integral LOCA tests performed on irradiated M5TM_
clad LEU fuel, irradiated Zirlo-clad LEU fuel, irradiated Zircaloy-4-clad LEU fuel with integral absorbers, etc. The performance of integral LOCA tests with irradiated fuel has never been considered a prerequisite for the deployment of fuel designs. If there were such a requirement, 21
the irradiated fuel would need to come from somewhere (like a lead test assembly program), as we pointed out in Section VI of our initial testimony.
VII. COOLABLE CORE GEOMETRY (A. 15)
- 66. (Harvey, Dunn) In our initial testimony, in Section III.C (paragraphs 55-56) we presented the results of relevant MOX fuel LOCA analyses. We specifically testified that: "The maximum calculated cladding strain for the most limiting case is 51 percent and the flow blockage due to this ballooning is 52 percent of the coolant channel surrounding the hot pin.
This amount is well within the coolable geometry limit (specified by the AREVA LOCA evaluation model) of 90 percent."
- 67. (Harves) Dr. Lyman in his A.15 states: "The maximum flow blockage that will preserve the coolable geometry depends on the assumed heat source and the heat transfer properties of the fuel bundle. As IRSN points out, acceptable bundle blockage ratios were derived based upon arrays of unirradiated fuel rods, and did not take into account fuel relocation and its associated impacts on the redistribution of the decay heat source within the fuel rods.
IRSN presentation to NRC at 29 (October 23, 2003).... Thus, any analysis that does not take this [the impact of relocation in the fuel rod balloon in the flow blockage area] into account is incomplete and is likely to be non-conservative. Lack of consideration of this phenomenon will be of greater concern for the MOX [lead assembly(LTA)] core to the extent that the MOX LTAs have a smaller margin to regulatory limits than LEU fuel." I disagree with this assessment.
- 68. (Harvey) The question about coolability of the blocked region given fuel relocation is not unique to MOX fuel. The only specific tie being made by Dr. Lyman to MOX fuel assemblies is based on the inaccurate statement that the MOX lead assemblies have less margin to the LOCA acceptance criteria than LEU fuel assemblies. In fact, the calculated results for the MOX lead assemblies have more margin to the 10 C.F.R. § 50.46 acceptance criteria than 22
Ubr/q LEU fuel. The limiting PCT for the MOX fuel LOCA limits calculations is 2019.50 F and the maximum local oxidation value is 5.2%. These values are below the limiting cases for the co-resident LEU fuel in the Catawba core as analyzed by Westinghouse. Thus, this issue raised by Dr. Lyman is not a MOX fuel issue.
- 69. (Harvey) Furthermore, the maximum calculated blockage value for the MOX analysis -52 percent (see paragraph 56 of our initial testimony) is well below the values where core cooling has been demonstrated (90%). Dr. Lyman offers no evidence that fuel relocation will result in a significant change in the ability to cool the fuel assembly following a LOCA. We have addressed the matter of long-term coolability in paragraph 19 of this rebuttal testimony.
VIII. SAFETY MARGINS (A.16)
- 70. (Nesbit) In A. 16, Dr. Lyman alleges that safety margins for MOX fuel are smaller than for LEU fuel with respect to PCT following a LOCA. He reiterates many of the arguments made before; as before, however, the discussion is fundamentally flawed in several key respects.
- 71. (Nesbit) First, Dr. Lyman states that "As Duke's calculations have demonstrated, the PCT in a design-basis LOCA is higher for a MOX rod than for an LEU rod in the same position in the core.... The margin to the 10 C.F.R. § 50.46 PCT limit of 22000 F is therefore smaller for a MOX rod than for an LEU rod in the same position." As we point out in Section III.C (paragraph 58) of our initial testimony, the difference in the "apples to apples" PCT (less than 401F) is insignificant in the context of Appendix K design basis LOCA calculations.
Furthermore, as we have noted several times, we chose not to take credit for some MOX/LEU differences that would, in all likelihood, reduce the MOX fuel PCT below that of LEU fuel.
Fundamentally, Duke's analyses demonstrate that MOX fuel and LEU fuel PCTs following LOCA are essentially the same.
23
- 72. (Aresbit) Second, Dr. Lyman alleges that: "At high bumups, the linear heat generation rate for MOX fuel is generally higher than that for LEU fuel. This, in turn, results in increased centerline temperature and stored energy, therefore reducing the margin to design-basis LOCA regulatory limits." This "fact" is apparently based on statements by IRSN researchers; again, however, BREDL has not offered the originator of this statement as a witness, so we have no opportunity to critically examine the basis for the statement. The statement appears to be based on an aspect of European MOX fuel that is different from the weapons grade MOX fuel that is proposed for use at Catawba. As we make clear in Section V.X (paragraphs 129-142) of our initial testimony, the MOX fuel lead assemblies will operate at a linear heat generation rate that is lower than the peak (and at almost all bumups, the average) for co-resident LEU fuel assemblies in the same cycle of operation. The MOX/LEU difference in linear heat generation rate cited by BREDL is actually a benefit, not a penalty, for the Catawvba MOX fuel lead assemblies.
- 73. (Nesbit) Third, Dr. Lyman's approach of simplistically superimposing a PCT increase of 313'F on a reported MOX fuel PCT of 2018'F is incorrect. The PCT increase of 313'F was obtained from an IRSN calculation reported at Aix-en-Provence in 2001, and it was based on a high filling fraction (70%). The presenter of the information admitted that the calculation took no credit for the cooling benefits of ballooning and rupture 3 , which, as noted in Section V.A (paragraphs 90-98) of our previous testimony, are substantial. Furthermore, any impact of relocation on cladding temperature would necessarily take place at the location 3 Grandjean, C., et al., "High Burnup U02 Fuel LOCA Calculations to Evaluate the Possible Impact of Fuel Relocation After Burst," NEA/CSNIIR(2001)18, Aix-en-Provence, March 2001 (Duke Exhibit 4; see "Discussion" following paper and presentation material).
24
(elevation) of the ballooning and rupture. Modeling relocation effects would not adversely affect cladding temperatures at other (non-ruptured) elevations on the fuel rod. It is inappropriate to add a APCT estimated for relocation to a calculated PCT at a non-ruptured location, as Dr.
Lyman did in paragraph 4 of A.16. For the Appendix K LOCA analyses that actually formed the basis for the MOX fuel lead assembly LOCA limits 4 , the highest rupturednode temperature was 17501F, as stated in Section III.C (paragraph 55) of our testimony. Therefore, an appropriate bound of the potential relocation effect would add the very conservative IRSN APCT to the highest rupturednode temperature. As noted in Section V.G (paragraph 154) of our testimony, this produces a value of 2070°F, well below the 2200WF PCT regulatory limit in 10 C.F.R.
§ 50.46. Again, BREDL has misconstrued the available information on fuel relocation and PCT, and provided an erroneous assessment of the potential impact on MOX fuel lead assemblies.
The statement in A. 16 that ".... the MOX LTAs could well be limiting with respect to LOCA compliance if relocation is fully accounted for" has no basis.
- 74. (Nesbit) Finally, contrary to Dr. Lyman's point in the last paragraph of A.16, there are no "significant uncertainties" that warrant NRC disapproval of the MOX fuel lead assembly program. Our testimony makes it clear that the uncertainty in calculated, post-LOCA PCT and cladding oxidation for the MOX fuel lead assemblies is essentially the same as the corresponding uncertainty in LEU fuel calculations. Furthermore, our testimony points out (i) the inherent conservatism in Appendix K LOCA analyses, and (ii) the substantial margin between actual plant operation and the LOCA limits.
4 Duke Energy response to NRC Requests for Additional Information, November 3, 2003, (Duke Exhibit 2; see Table Q14-1).
25
IX. "GAPS IN THE EXPERIMENTAL DATABASE" (A. 17)
- 75. (Nesbit) Dr. Lyman states in A.17 that: "The only way to fully address the uncertainties associated with the behavior of high-burnup, M5-clad MOX fuel during LOCAs is to conduct integral LOCA tests of such fuel, fabricated with the same specifications as the lead test assemblies that are under consideration here, and irradiated to a range of burnups, including the maximum of 60 GWD/t that Duke has requested in its LAR. The proposed Phdbus test series would likely make a substantial contribution to reducing the level of uncertainty associated with MOX fuel behavior during LOCAs." Dr. Lyman goes on to suggest that these integral tests could be supplemented by separate effects tests such as are currently planned at Halden for LEU fuel, but he believes that ". . . similar (separate effects) tests on mixed oxide fuel will also be needed." Dr.
Lyman has also previously stated5 that the proposed MOX fuel tests could be carried out with European reactor grade MOX fuel, provided that some combination of tests and analyses were to address unspecified effects associated with the isotopic difference between European reactor grade MOX fuel and the weapons grade MOX fuel that will be used at Catawba. It is evident that, in Dr.
Lyman's view, a major experimental and analytical program must be completed prior to using MOX fuel, even in a limited demonstration exercise (e.g., four lead assemblies at Catawba).
- 76. (Nesbit) The arguments against the use of MOX fuel lead assemblies at Catawba are based solely on purported uncertainties associated with fuel relocation and advanced cladding materials, which we have already discussed. Despite BREDL attempts to tie the fuel relocation issue to alleged differences between MOX and LEU fuel, the relocation issue is just as applicable to the range of currently-deployed LEU fuel designs as it is to the MOX fuel lead assemblies.
Dr. Lyman's cladding based concerns are clearly applicable to the numerous United States 5 BREDL Response to Duke Energy Corporations First Set of Interrogatories and Requests for Production of Documents, April 14, 2004 (see response to Interrogatory 15).
26
nuclear power plants that use M5TM and ZirloP cladding with LEU fuel. The NRC has quite obviously concluded that these purported uncertainties are not too great with respect to the ongoing use of LEU fuel in millions of fuel rods at more than one hundred United States nuclear power reactors.
- 77. (N'esbit) BREDL would establish an impossibly high and completely unnecessary standard for conducting lead test assembly programs. We address this point in some detail in Section VI (paragraphs 158-171) of our prior testimony. The purpose of lead test assemblies is to gather information prior to large-scale deployment of a new or revised fuel design in nuclear power reactors. This process has worked well over the past decades, and nuclear fuel designs have evolved greatly, with substantial performance and environmental benefits. MOX fuel lead assemblies are a key element in the overall program to dispose of surplus weapons grade plutonium in the United States and in Russia. Thus, this lead assembly program will provide benefits associated with achieving nonproliferation policy objectives of the United States government and the international community.
X. CONCLUSION
- 78. (All) We have shown that the MOX fuel lead assemblies can be used in conformance with regulatory requirements and without posing an undue threat to the health and safety of the public.
27 DC:365699.1
2227 1 CHAIRPERSON YOUNG: Do you have a copy for 2 the court reporter? Does she have a copy already or 3 do you?
4 MR. REPKA: I don't believe so. Of all of 5 the testimony?
6 CHAIRPERSON YOUNG:- Of your direct and 7 rebuttal prefiled testimony.
8 MR. REPKA: We will send that up shortly.
9 CHAIRPERSON YOUNG: All right. Then let's 10 continue. After the place in the transcript where the 11 direct and prefiled will be bound into the transcript 12 -- and whoever is transcribing should understand from 13 what I have just said how to do that. Are we waiting 14 for that or shall we go ahead with these?
15 MR. REPKA: No. We'll hand it to you 16 right now.
17 CHAIRPERSON YOUNG: Okay.
18 JUDGE BARATTA: And these do, of course, 19 have those corrections and additions?
20 MR. REPKA: Those have been marked.
21 CHAIRPERSON YOUNG: And this is just the 22 testimony, not the exhibits, correct? Let's hold off.
23 We don't want the exhibits at this point. Ms.
24 Cottingham, we don't want the exhibits at this point.
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2228 1 separately. Only the testimony gets bound into the 2 transcript.
3 So there's one document entitled "July 4 1st, 2004 Testimony of Stephen P. Nesbit, Robert C.
5 Harvey, Bert M. Dunn, and J. Kevin McCoy on Behalf of 6 Duke Energy Corporation on Contention 1." And then 7 there's another document entitled "Rebuttal Testimony 8 of" the same people. Those are the two documents that 9 would be bound into the transcript at the point I 10 designated earlier.
11 Okay. Now let's go ahead and mark and 12 enter those exhibits. And you can indicate that this 13 is the place in the transcript that they are entered, 14 but they will not be bound into the transcript. They 15 will just be attached at the end.
16 We will start with 1 through 5. And then 17 the figures 1 through 19 will be official exhibit 18 numbers 6 through 24. So it will probably take you 19 some time to mark each one of those. So while the 20 court reporter and Ms. Lynn, our law clerk, are 21 marking those exhibits, Judge Baratta and I are going 22 to take a short break to confer. And we will 23 reconvene as soon as the court reporter and Ms. Lynn 24 are finished.
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2229 1 the panel would like to address Judge Baratta's 2 question about figure 14.
3 CHAIRPERSON YOUNG: Are we with you? We 4 need to make sure that we don't burden the court 5 reporter with doing two things at once.
6 JUDGE BARATTA: Yes. I think we'd better 7
8 MR. REPKA: Understood.
9 JUDGE BARATTA: That can be addressed 10 during the cross-examination. I just wanted to alert 11 you that I had a question that would require some 12 discussion and that you may want to think how you want 13 to handle it.
14 MR. REPKA: That's fine.
15 CHAIRPERSON YOUNG: Okay. So can we 16 excuse them for the moment?
17 JUDGE BARATTA: I would suggest we excuse 18 19 CHAIRPERSON YOUNG: Okay. We will excuse 20 you for the moment. And then we will come back after 21 our break to move to BREDL and do the same procedure 22 with you and then to the staff.
23 (Whereupon, the foregoing matter went off 24 the record at 1:49 p.m. and went back on 25 the record at 2:10 p.m.)
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2230 1 (Whereupon, the aforementioned 2 documents were marked for 3 identification as Exhibits 4 Number 1 through 24, 5 respectively.)
6 CHAIRPERSON YOUNG: Based on the parties' 7 earlier comments, none of the parties had nay 8 objections to any of the other parties exhibits.
9 Official exhibits 1 through 24 are admitted into the 10 record.
11 (Whereupon, the aforementioned 12 documents, having previously 13 been marked for identification 14 as Exhibits Number 1 through 15 24, respectively, were received 16 in evidence.)
17 CHAIRPERSON YOUNG: That's all from Duke 18 at this point, correct?
19 MR. REPKA: At this point, yes.
20 (Whereupon, the witness was excused.)
21 CHAIRPERSON YOUNG: Okay. In a moment, we 22 are going to move on to BREDL. While we are going 23 through the process of binding the prefiled testimony 24 into the record and admitting the exhibits, I want to 25 share with you our thoughts on taking testimony today.
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2231 1 Based on our reading of the prefiled 2 testimony, it appears that section 5 of Duke's -- we 3 would be starting with the cross-examination of Duke's 4 panel. It would appear that section 5 of Duke's 5 prefiled testimony is the most critical part of it.
6 We feel hesitant to go forward with 7 cross-examination on that part of it without Judge 8 Elleman being here. We would like to have the 9 parties' thoughts on whether we could go forward with 10 parts 1 through 4, whether that could be separated out 11 to take testimony on that today and whether there are 12 similar issue that may not involve as significant 13 areas of dispute but which any of the parties may wish 14 to conduct cross-examination on. So be thinking about 15 that.
16 And if you want to ask for any 17 clarification of what I said of either of us, go ahead 18 at this point. And then we'll move on with the 19 transcript and exhibits.
20 Does the way I described this potential 21 way of proceeding make sense to all counsel? And do 22 you have any thoughts on it you would like to -- we 23 would like your feedback on how reasonable a process 24 this would be to follow to use today's time as 25 efficiently and productively and meaningfully as NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2232 1 possible.
2 MS. CURRAN: Do you want our reactions 3 now, Judge Young?
4 CHAIRPERSON YOUNG: Yes. Well, you can 5 give your reactions now. You can think about it. But 6 we wanted to let you know what we were thinking so 7 that we could try to approach this in a meaningful way 8 that will not deprive Judge Elleman of the ability to 9 follow the testimony on the most critical, what appear 10 to us at this point to be the most critical, issues.
11 Go ahead.
12 MS. CURRAN: Well, I would like to 13 respond. We look at the testimony as an integrated 14 whole. While there may be some parts that are more 15 important than others, we want to be able to question 16 the witnesses on the whole thing with Judge Elleman 17 present.
18 Once again -- and maybe I misinterpreted 19 your e-mail, and I didn't bring a copy with me. But 20 when I read it, it was my understanding that the Board 21 had made a decision --
22 CHAIRPERSON YOUNG: Let me just interrupt 23 you. We had made a decision. And we are taking Ms.
24 Uttal's request as a request to reconsider our earlier 25 decision. So we want to hear what all of you had to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE.. N.W.
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2233 1 say on this.
2 We want to proceed in a manner that will 3 be most meaningful for all concerned and that will 4 make the best use of the time that we have.
5 And I guess what I am asking you is, are 6 there areas -- and you may want to confer with Dr9.
)7 Lyman before you give your complete response, but are 8 there areas that we could go ahead with 9 cross-examination on, some of the background areas 10 that the Duke panel provides in its testimony that 11- could be separated out? And I'm asking all of you to 12 take testimony on those today.
13 We will have one day turn around for us on 14 the transcripts. Assuming we get enough, we will be 15 glad to let the parties use copies as necessary 16 tomorrow. We will have a copy for Judge Elleman, 17 although they may not come in until around 10:00 or 18 10:30 in the morning. We're asking that they be 19 provided to us as soon as possible.
20 So that's what we want you to think about.
21 You can respond to the degree you want to now. You 22 can think about it, and we can come back to it after 23 we have taken care of some of these more housekeeping 24 types of issues.
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2234 1 issue of the sort of res judicata issue, I left my 2 office at about 11:30 to come to the NRC, meet my 3 clients, and we had a sandwich. And I didn't get Ms.
4 Uttal's letter.
5 I didn't bring the clean copies of Duke's 6 testimony that I could hand Dr. Lyman on the stand 7 because I didn't think we were going to do that today.
8 I didn't come prepared -- well, and that would have 9 been for Dr. Lyman.
10 I didn't make the choice of, instead of 11 eating a sandwich, to continue preparing my 12 cross-examination. And this case has been that 13 expedited that sometimes that is the choice.
14 And I chose to have my sandwich today.
15 And I would really appreciate being able to start in 16 the morning.
17 CHAIRPERSON YOUNG: Let me just raise 18 another issue. There is a provision in the NRC rules 19 that allows for some cross-examination of experts by 20 other experts. And at times in the oral argument, 21 various oral arguments, we have heard in this case 22 there have been instances where having the parties 23 experts speak directly to us, rather than through 24 counsel, has proven more efficient and meaningful.
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2235 1 you can discuss whether there are some things in the 2 nature of background that don't go so directly to the 3 critical areas of dispute that might be able to be 4 addressed today so that we can move tomorrow morning 5 if this is the way we decide to proceed more directly 6 and focus on the critical areas of dispute. I won't 7 describe them at this point, but I have indicated to 8 you that we think that section 5 appears to be the 9 most critical part of the testimony.
10 MS. CURRAN: I would just want to add one 11 more thing, which is that my cross-examination is 12 going to be fairly limited. And it's going to be 13 limited to what we think are some of the crucial 14 points.
15 JUDGE BARATTA: So, in other words, you 16 think that the questions that you would be asking are 17 the ones that we would really want Dr. Elleman to hear 18 the responses on?
19 MS. CURRAN: Yes.
20 JUDGE BARATTA: Okay. Thank you.
21 CHAIRPERSON YOUNG: Another thing that you 22 can be thinking about -- and you may want to hear from 23 the other parties as well and from BREDL on whether 24 there any other parts of any witnesses' 25 cross-examination that would be of the nature that are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2236 1 more perhaps background or preliminary information 2 that we could do in such a way that it doesn't make it 3 difficult for Judge Elleman to sort of step into the 4 middle of critical testimony on highly disputed areas.
5 And then, secondly, if your 6 cross-examination is going to be very limited, it may 7 be that we don't have a problem finishing if we do 8 start tomorrow morning with the actual 9 cross-examination.
10 So if you could mull those things over in 11 your mind while we proceed with these more 12 housekeeping issues? And then we'll come back to it 13 after we have done that.
14 So next would be BREDL's prefiled direct 15 and rebuttal testimony and exhibits. You can stay 16 there if it's easier, but let me ask you to raise your 17 right hand, Dr. Lyman.
18 Whereupon, 19 EDWIN S. LYMAN 20 was called as a witness by counsel for the petitioner 21 and, having been first duly sworn, was examined and 22 testified as follows:
23 CHAIRPERSON YOUNG: All right. Go ahead, 24 Ms. Curran.
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2237 1 MS. CURRAN: Dr. Lyman, do you have before 2 you a copy of the document entitled "Prefiled Written 3 Testimony of Dr. Edwin S. Lyman Regarding Contention 4 1 "?
5 DR. LYMAN: I do.
6 MS. CURRAN: The document dated July 1st, 7 2004?
8 DR. LYMAN: That's correct.
9 MS. CURRAN: Is this a copy of the 10 testimony that you submitted in this proceeding 11 regarding contention 1?
12 DR. LYMAN: Yes, it is.
13 MS. CURRAN: And it was prepared by you or 14 prepared under your supervision?
15 DR. LYMAN: Yes, it was.
16 MS. CURRAN: Do you have any corrections 17 you would like to make or clarifications to this 18 testimony?
19 DR. LYMAN: Yes. I have two 20 clarifications. The first is in answer 5 on page 3.
21 The third sentence under answer 5, I would like to 22 make this change. After the word "forth" as in "sets 23 forth," I would like add "required and acceptable 24 features of." Then the sentence reads, "ECCS 25 evaluation models." I'd like to delete the comma.
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2238 1 CHAIRPERSON YOUNG: Slow down just a 2 second.
3 DR. LYMAN: Sorry.
4 CHAIRPERSON YOUNG: In the sentence 5 starting with "Appendix K," after the word "forth,"
6 "required and"?
7 DR. LYMAN: "And acceptable features of."
8 CHAIRPERSON YOUNG: Okay. Go ahead.
9 DR. LYMAN: Then after "models," as in 10 "evaluation models' in parentheses, I would delete the 11 comma. And then I would delete the phrase "i.e.,
12 assumptions about the behavior of reactor fuel."
13 Then I would delete after "determining" 14 the phrase "whether it." And I would change the word 15 "complies" to "compliance." And then I will read the 16 whole sentence back to you.
17 CHAIRPERSON YOUNG: Okay.
18 DR. LYMAN: The sentence should read, 19 "Appendix K to part 50, whose requirements are 20 referenced in 10 CFR 50.46(a) (1), sets forth required 21 and acceptable features of ECCS evaluation models that 22 are to be used in determining compliance with the 23 criteria in 10 CFR 50.46."
24 CHAIRPERSON YOUNG: Have you made a copy 25 with those changes written in?
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2239 1 MS. CURRAN: Yes.
2 DR. LYMAN: Yes, we have.
3 CHAIRPERSON YOUNG: Okay. Great.
4 DR. LYMAN: The second change I would like 5 to make is the answer to question 8 on page 4. The 6 sentence which is the fourth sentence, begins with 7 "finally," I would like to delete the words 8 "long-term" before the "core coolability" at the end 9 of the sentence.
10 So the sentence should read, "Finally, the 11 greater local linear heat generation rate requires a 12 greater coolant flow around the ballooned area to 13 ensure core coolability."
14 And those are the only corrections I have 15 on the prefiled testimony.
16 MS. CURRAN: Should I move on to --
17 CHAIRPERSON YOUNG: Let's go ahead and go 18 do the rebuttal.
19 MS. CURRAN: Okay.
20 CHAIRPERSON YOUNG: And then we'll bind 21 them in together.
22 MS. CURRAN: All right. Dr. Lyman, do you 23 have before you a copy of the document entitled 24 "Rebuttal Testimony of Dr. Edwin S. Lyman Regarding 25 BREDL Contention 1"?
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2240 1 DR. LYMAN: Yes, I do.
2 MS. CURRAN: Now, I believe this document 3 is undated. Am I correct?
4 DR. LYMAN: It's undated, yes.
5 MS. CURRAN: Do you recall the date that 6 you submitted this testimony?
7 DR. LYMAN: I believe it was July 8th.
8 MS. CURRAN: Yes.
9 DR. LYMAN: July 8th.
10 MS. CURRAN: Is this the testimony that 11 you prepared and submitted regarding contention 1?
12 DR. LYMAN: Yes, it is.
13 MS. CURRAN: And this testimony is your 14 work or was prepared under your supervision?
15 DR. LYMAN: Yes, it is.
16 MS. CURRAN: Do you have any corrections 17 that you would like to make to your rebuttal 18 testimony?
19 DR. LYMAN: Yes, I have two corrections in 20 the transcription that are typographical errors. The 21 first is the answer to rebuttal question 3 on page 2.
22 The third sentence that reads, "This can be seen from 23 figures 10 and 11 in Duke's testimony in response to 24 questions." It reads, "46 to 47." I would like to 25 replace "46 to 47" with "97 to 98," so "in response to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2241 1 questions 97 to 98."
2 The second change I would like to make is 3 the last paragraph on page 2, the very last sentence, 4 which reads, "The oxidation rate for M-5 is 5 substantially greater at 2,174 degrees Fahrenheit than 6 at 1,841 degrees Fahrenheit." The first number should 7 actually read "2,154 degrees Fahrenheit." So I would 8 like to strike "2,174" and replace it with "2,154."
9 And those are all the corrections I have.
10 MS. CURRAN: I would like to request this 11 piece of the testimony be bound into the record.
12 CHAIRPERSON YOUNG: All right. Then I 13 would ask the court reporter and the transcriber to 14 take a break in the transcript as soon as I finish my 15 sentence and bind BREDL's prefiled direct and rebuttal 16 testimony into the transcript. So we'll stop here and 17 then start a new page after that.
18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON. D.C. 20005-3701 www.nealrgross.com
July 1, 2004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of: )
DUKE ENERGY CORPORATION } 50-4N4-OLA (Catawba Nuclear Station, )
Units 1 and 2) )
PREFILED WRITTEN TESTIMONY OF DR. EDWIN S. LYMAN REGARDING CONTENTION I On behalf of Blue Ridge Environmental Defense League ("BREDL"), Dr. Edwin S.
Lyman hereby submits the following testimony regarding BREDL's Contention I.
Q.1. Please state your name and describe your professional qualifications to give this testimony.
A.1. My name is Dr. Edwin S. Lyman. I am a Senior Scientist with the Global Security Program at the Union of Concerned Scientists, 1707 H Street, NW, Suite 600, Washington, D.C.
20006. My education and experience are described in my curriculum vita, which is attached to my testimony as Exhibit A.
I am a qualified expert on nuclear safety and safeguards issues. I hold a Ph.D., a master of science degree, and a bachelor's degree in physics. For over eleven years, I have conducted research on security and environmental issues associated with the management of nuclear materials and the operation of nuclear power plants. My research has included the safety and security implications of using mixed oxide fuel as a substitute for uranium fuel in nuclear power plants. I have also published articles on this topic. A list of my publications is included in my C.V.
Q.2. What is the purpose of your testimony?
A.2. The purpose of my testimony is to discuss my views on BREDL Contention I, which was admitted for litigation by the Atomic Safety and Licensing Board ("ASLB") in LBP-04-04, Memorandum and Order (Ruling on Standing and Contentions) (March 5, 2004). BREDL Contention I asserts that Duke Energy Corporation's ("Duke's") license amendment request
("LAR") to test plutonium mixed oxide ("MOX") fuel at the Catawba nuclear power plant is inadequate because Duke has failed to account for the differences between MOX and low enriched uranium ("LEU") fuel behavior; nor has Duke accounted for the impact of these differences on Duke's analysis of loss of coolant accidents ("LOCAs").
Q.3. What materials lave you revieved in preparation for your testimony?
A.3. I have reviewed Duke's LAR and related correspondence, including Duke's responses to Requests for Additional Information ("RAls") by the NRC Staff. I have also reviewed the body of literature which has been developed regarding the behavior of MOX and other types of reactor fuel under LOCA conditions. I am also familiar with relevant NRC documents, including correspondence regarding this license amendment application, reports and correspondence concerning characteristics and behavior of MOX fuel, and correspondence and reports concerning the behavior of LEU fuel under LOCA conditions. In addition, I am familiar with regulations and guidance of the U.S. Nuclear Regulatory Commission ("NRC") and the U.S.
Department of Energy ("DOE") governing plutonium processing facilities. Finally, I am familiar with U.S. and foreign government reports regarding testing of LEU fuel under accident conditions.
Q.4. Please summarize the conclusions you have reached regarding the adequacy of Duke's LAR application to account for the differences between MOX and LEU fuel.
A.4. In my professional judgment, Duke's design-basis loss of coolant ("DB-LOCA") analysis is inadequate because it does not address the uncertainties associated with relocation effects that M5-clad MOX fuel may experience under LOCA conditions. These uncertainties relate to Duke's assertion that the action proposed in the MOX LTA LAR will not result in a violation of the emergency core cooling system (ECCS) acceptance criteria in 10 C.F.R. § 50.46: peak cladding temperature ("PCT"), maximum cladding oxidation, and the preservation of a coolable core geometry.
The phenomenon of fuel relocation has been observed in experiments with irradiated LEU fuel underLOCA conditions. While to my knowledge no similar experiments have been done on MOX fuel, there are technical reasons to believe that the impact of fuel relocation effects during a LOCA may be more severe for MOX fuel rods than for LEU fuel rods of the same bumup, due to differences in characteristics such as fuel fragment sizes and fuel-clad interactions. Moreover, calculations in Duke's LAR indicate that MOX fuel is generally more limiting than LEU fuel with respect to DB-LOCAs. Therefore, the consequences of fuel relocation, and the non-conservatism associated with neglecting them, may be of greater concern for MOX fuel rods than for LEU fuel rods with respect to compliance with LOCA regulatory criteria.
Duke has failed to address these uncertainties in MOX fuel behavior, and therefore its LTA application is unacceptable to satisfy the requirements of 10 C.F.R. § 50.46 with respect to PCT, maximum cladding oxidation, and coolable geometry of fuel. In addition, by failing to address the uncertainties in MOX fuel behavior, Duke has not demonstrated compliance with the general reasonable assurance standard in 10 C.F.R. § 50.40(a).
2
I do not believe, however, that these uncertainties can be addressed with mere calculations or analyses based on LEU performance. In my professional opinion, the only satisfactory way to address these uncertainties would be to conduct integral tests of MOX fuel assemblies under LOCA conditions in such a manner that the impacts of the phenomena I have previously described can be measured with reasonable accuracy and precision.
Q 5: Please explain how the regulations in 10 C.F.R § 50.46 apply to Contention I.
A.5: NRC regulations at 10 C.F.R. § 50.46 establish acceptance criteria for emergency core cooling systems for light-water nuclear reactors. Essentially, the regulation sets design limits for behavior of the reactor fuel under LOCA conditions. Appendix K to Part 50, whose requirements are referenced in 10 C.F.R. § 50.46(a)(1), sets fo ECCS "evaluation models," i.e. assumi about the behavior of reactor fuel that are to be use in determining the d1 criteria in 10 C.F.R. § 50.46. J s 10 C.F.R. § 50.46 and Appendix K apply only to uranium-based fuel, but Duke has requested an exemption from this limitation so that these requirements will apply to MOX fuel. I believe that it is generally appropriate to apply the requirements of 10 C.F.R. § 50.46 to MOX fuel, as long as Appendix K is not strictly applied to exclude consideration of relocation of the fuel during LOCAs.
The regulations in 10 C.F.R. § 50.46 sets forth fuel performance limits in three categories that have importance with respect to performance of MOX fuel: peak cladding temperature ("PCT'),
maximum cladding oxidation, and coolable geometry. Section 50.46(b)(1) requires that that PCT "shall not exceed 22000 F." Section 50.46(b)(2) provides that the "calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation." Section 50.46(b)(4) also requires that "[c]alculated changes in core geometry shall be such that the core remains amenable to cooling."
Q.6: Please explain why you think the Appendix K evaluation models for the MOX LTA core should include consideration of fuel relocation during LOCAs.
A.6: Appendix K does not include consideration of fuel relocation. The NRC did contemplate including fuel relocation as a criterion in Appendix K, but claimed to have resolved the question in Generic Issue 92. Memorandum from Ralph Meyer, NRC Office of Nuclear Regulatory Research, to John Flack, NRC Regulatory Effectiveness and Human Factors Branch, re: Update on Generic Issue 92, Fuel Crumbling During LOCA (February 8, 2001) (NRC ACN #
ML010390163) (hereinafter "Meyer Memorandum'). A copy is attached to my testimony as Exhibit B. See also Memorandum from Ashok C. Thadani, Office of Nuclear Regulatory Research, to Samuel J. Collins, Office of Nuclear Reactor Regulation, re: Information Letter 0202, Revision of 10 CFR 50.46 and Appendix K (June 20,2002) (hereinafter "Thadani Memorandum") (NRC ACN # ML021720690). A copy is attached to my testimony as Exhibit C. More recently, the NRC has acknowledged that omission of fuel relocation effects is a non-conservatism in Appendix K with a very large potential impact on PCT, and that an early "resolution" of this issue (i.e., Generic Issue 92) may have been in error or is no longer 3
applicable because of new information. See; Meyer Memorandum and Thadani Memorandum, Attachment 4 at 4-5.
As I wvill discuss in more detail later in my testimony, certain characteristics of MOX fuel appear to exacerbate the effects of fuel relocation, thus leading to higher PCTs and greater maximum cladding oxidation. While there are several other known non-conservatisms in Appendix k, this one in particular appears to be relevant to the MOX LTA LAR because of its disproportionately large impact on the MOX LTAs compared to the LEU assemblies that comprise the remainder of the core. Given the potential impact on PCT of relocation effects, it is not appropriate to omit consideration of this phenomenon from the Appendix K models that Duke uses to establish that loading of the MOX LTAs into Catawba will result in compliance with 10 CFR § 50.46 criteria.
Q.7: Please explain what you mean by fuel relocation during a LOCA.
A.7: According to NRC, "fuel relocation refers to the movement of fuel pellet fragments into regions of the fuel rod where the cladding has ballooned during a LOCA transient." Thadani Memorandum, Attachment 4 at 4-5.
Q.8: Please explain why fuel relocation could increase the severity of a LOCA.
A.8: Fuel relocation increases the local linear heat generation rate within the ballooned area. Thus it could increase the severity of a LOCA by resulting in a greater fuel rod peak cladding temperature (PCT) than in a situation in which fuel relocation did not occur. Because transient oxidation during a LOCA increases with an increase in PCT, fuel relocation could also result in a greater maximum cladding oxidation. Finally, the greater local linear heat generation rate requires a greater coolant flow around the ballooned area to ensure long-effricore coolability. See slides presented by A.
Mailliat and J.C. M6lis, IRSN, at "PHEBUS STLOC Meeting" with NRC Staff (October 23, 2003) (NRC ACN # ML032970624) (hereinafter October 2003 IRSN Presentation"). A copy is attached to my testimony as Exhibit D.
Q.9: Plcase discuss the potential magnitude of the impact of fuel relocation on PCT and maximum cladding oxidation for uranium oxide (U0 2 ) fuel.
A.9: The most recent calculations of the impact of fuel relocation on PCT of which I am aware were conducted by the Institut de Protection et de Sfzret6 Nucleaire (IPSN, now IRSN) and published in 2001. In that study, the authors used the CATHARE2 computer code to calculate the impact of fuel relocation on the large-break LOCA PCT for a high-bumup U0 2 fuel rod as a function of the "filling ratio," or the ratio of the volume of the relocated fuel material to the volume of the ballooned region. For the scenario evaluated, the authors found that the PCT in the absence of relocation effects was 970TC. For a filling ratio of 70%, the maximum considered, the PCT was 1144 0 C. For a filling ratio of 40%, the PCT was about 200C greater than for the no-relocation case. Thus the maximum impact on PCT of relocation in this study was a APCT of +
1740C (313'F) for high-burnup U02 fuel. It is not clear from the study whether higher filling ratios, and hence larger impacts on PCT, are possible. C. Grandjean, G. Hache and C. Rongier, "High Bumup U0 2 Fuel LOCA Calculations to Evaluate the Possible Impact of Fuel Relocation After Burst," OECDINEA Proceedings of the Topical Meeting on LOCA Fuel Safety Criteria, Aix-en-4
Provence (March 22-23, 2001) (hereinafter "Grandjean, Hache, and Rongier' T). A copy of this paper is attached to my testimony as Exhibit E. The NRC staff appears to be familiar with this result. See Thadani Memorandum, Attachment 5 at 4.
The study also evaluated the impact on the maximum cladding oxidation for the ruptured region (two-sided oxidation). The equivalent cladding reacted (ECR) calculated by the Cathcart-Pawel rate law (a surrogate for "maximum cladding oxidation')¶was 12.6% for the no-relocation case, and 19.7% for the 70% filling ratio case. Thus the maximum impact on ECR resulting from relocation was calculated as AECR = 7.1%.
Q.10: Please explain why the impact of fuel relocation on the severity of a LOCA could be greater for MNOX fuel than for U0 2 fuel at the same burnup.
A.10: Experts have concluded that MOX fuel may experience more severe relocation effects than U0 2 fuel at the same bumup. The IPSN study above did not explicitly consider MOX fuel, but stated that "it must be pointed out that that results of corresponding calculations with ... high bumup MOX fuels would have been more severe with regard to acceptance limits." Grandjean, Hache and Rongier at 7.
IRSN, the successor to IPSN, has reiterated these concerns, stating in a recent presentation that for MOX fuel, a "higher initial energy" and an "enhance [sic] of fuel relocation impact" results in greater increases in PCT and ECR associated with relocation. V Guillard, C. Grandjean, S.
Bourdon and P. Chatelard, "Use of CATHARE2 Reactor Calculations to Anticipate Research Needs," SEGFSM Topical Meeting on LOCA Issues, Argonne National Laboratory, slides at 8-9 (May 25-26, 2004) (NRC ACN # ML041600261). A copy of this paper is attached to my testimony as Exhibit F. In the abstract for this presentation, the authors state that "a lack of knowledge on theses [sic] parameters [important for relocation] for irradiated U02 and particularlyMOXfuel [emphasis added] may lead to reduce [sic] safety margins."
MOX fuel may experience more severe relocation effects than U0 2 fuel at the same burnup because several characteristics that are important for relocation may be less favorable for MOX fuel. These include pellet fragment size and fuel-clad interaction.
Q.11: Please explain the basis for your concern regarding the pellet fragnient size of MOX fuel and its impact on fuel relocation in a LOCA.
A.11: The IPSN calculations cited above demonstrate the high sensitivity of fuel relocation-induced increases in PCT and ECR to the filling ratio. The filling ratio, in turn, is a function of the average particle size of the relocated fuel fragments, in that smaller particles will in general result in greater packing of the relocated area and hence higher filling ratios.
The fuel relocation phenomenon has been observed in LEU fuel for rod bumups exceeding around 48 GWD/t. See Grandjean, Hache and Rondier at 2 (2001). This suggests that vulnerability to fuel relocation is associated with the development of the high-burnup "rim" region known to emerge in LEU fuel for bumups exceeding about 40-45 GWD/t. IPSN states 5
/
that "fuel fragmentation is clearly associated to [sic] burnup, with finer fragments at higherBU."
See Grandjean, Hache and Rondier at 2 (2001).
For During manufacture of MOX fuel using the MIMAS process (which will be used for the Duke LTAs), plutonium agglomerates --- macroscopic clumps of plutonium-rich particles ---
occur in the fuel. Because the fissile material is concentrated in these clumps, very high local bumups result, due to the fact that the fission is occurring in a heterogeneous fashion. The ratio of local burnup within the agglomerates is on the order of 4-6 times the rod-average bumup, depending on the irradiation time. For instance, the agglomerate bum-up reaches about 60 GWD/t when the rod average is only around 18 GWD/t, and reaches 100 GWD/t when the rod average is only 28.4 GWD/t As a result, high-bumup rim-like regions emerge in the outer layers of the plutonium agglomerates for much lower rod-average bumups than 40-45 GWD/t, because the local bumups within the plutonium agglomerates increase much more rapidly than the rod-average burnups. Thus it is reasonable to expect that the onset of fuel relocation in MOX fuel may occur at lower rod-average bumups than in LEU fuel. This would imply that MOX fuel will be vulnerable earlier in its irradiation history (and consequently for a longer time) than LEU fuel.
Also, the particle size distribution in MOX fuel will be smaller than in LEU fuel at the same rod-average bumup, to the extent that fine fragments are generated in the ultra-high burnup plutonium agglomerate regions.
Fuel fragmentation can also be caused by the stress induced by the stored-energy redistribution during the blowdown phase of a LOCA. A. Mailliat and M. Schwarz, "Need for Experimental Programmes on LOCA Issues Using High Bum-Up and MOX Fuels," NUREG/CP-0 176, Proceedings of the Nuclear Safety Research Conference at 436 (May 2002) (NRC ACN #
ML021710793) (hereinafter "Mailliat and Schwarz"). A copy of this paper is attached to my testimony as Exhibit G. Because MOX fuel has a lower thermal conductivity and a higher radial temperature gradient than LEU fuel, it could experience greater fuel fragmentation during the blowdown and more severe relocation effects as a result.
According to two out of four NRC experts who participated in the 2001 PIRT panel on LOCAs and high-bumup fuel, the composition of fuel (i.e. a specified MOX composition) is of "high importance" for consideration of fuel relocation effects because it "may affect the amount of fine grain material after relocation. Fuel structure and mechanical properties are influenced by fuel type." See NUREG/CR-6744, "Phenomenon Identification and Ranking Tables for Loss-of-Coolant Accidents in Pressurized and Boiling Water Reactors Containing High-Bumup Fuel,"
Appendix D, Table D-1 at D-67 (December 2001) (NRC ACN # 013540623) (hereinafter "NUREG/CR-6477"). Relevent portions of this report are attached to my testimony as Exhibit II. One expert concluded that fuel composition was of moderate importance to relocation, stating that "the consequence of fuel fragments relocation (higher local decay heat and higher cladding temperature) could be more effective with MOX fuel than with U02 fuel" but that "the viscoelastic properties of the MOX should impair the fuel fragments relocation at high bumup."
Id. at D-67. A fourth expert concluded that fuel composition would be of only low importance to relocation. Id. at D-67. This difference of expert opinion highlights the inadequacies of the experimental database with regard to integral tests of MOX fuel under design-basis LOCA conditions, and underscores the significant uncertainties in Duke's design-basis LOCA analysis.
6
Q.12: Please explain the basis for your concern regarding the effects of fuel-clad interaction within the MOX LTAs and their impact on fuel relocation in a LOCA.
A.12: I am concerned about differences betveen MOX and LEU fuel with respect to fuel-clad bonding and the impact of such differences on fuel relocation behavior during a design-basis LOCA. According to IPSN (now IRSN), tight fuel-clad bonding may delay the onset of fuel relocation. Mailliat and Schwarz at 433. Tight bonding has also been observed at the Halden reactor in Norway to retard the rate of balloon formation. Nuclear Energy Agency, NEA/CSNIIR(2003)9, Ongoing and PlannedFuel Safety Research in NEA Member States at 79 (March 5, 2003). Relevant excerpts of this report are attached to my testimony as Exhibit 1. During NRC's recent expert elicitation (PIRT) process on LOCA issues for high-burnup fuel, all four participating experts agreed that "chemical and mechanical bonding between the fuel pellet and the cladding ... " was of high importance to the fuel relocation phenomenon, because "bonding could significantly affect the relocation characteristics by impeding pellet fragment movement."
NUREG/CR-6744, Table D-l at D-69. It has been confinned that MOX fuel is more resistant to clad failures due to pellet-clad mechanical interaction (PCMI) than LEU fuel, even at high burrups. Nuclear Energy Agency, NEA/NSC/DOC(2004)8, International Seminar on Pellet-CladInteractionswith Water Reactor Fuels, at 20 (May 6, 2004).
Relevant excerpts of this report are attached to my testimony as Exhibit J. This phenomenon is not well-understood but may imply that the pellet-clad bond is weaker for MOX fuel, in which case MOX fuel may have a greater propensity to earlier and more extensive fuel relocation than LEU.
In Duke's April 14, 2004, Response to BREDL's first set of discovery requests, Duke stated that the Framatome design-basis LOCA analysis for the MOX LTAs did not assume any fuel-clad bonding and was therefore "conservative" with respect to the requirement that the degree of cladding swelling not be underestimated. Id. at 14. However, in the absence of an assessment of whether and to what extent the pellet-clad interaction is weaker in MOX fuel than in LEU fuel, there is no way of knowing the degree to which this assumption is conservative for MOX fuel. Therefore, Duke's failure to properly account for this phenomenon contributes another uncertainty to the safety margin associated with Duke's design basis LOCA calculation.
Moreover, there is evidence to contradict Duke's assertion that "deterministic LOCA evaluations typically based on data taken from unirradiated cladding" are conservative with respect to clad swelling. According to IPSN (now IRSN), results from the PBF-LOC experiments found that irradiated rods experienced greater clad deformation than unirradiated rods during design-basis LOCA conditions. See Mailliat and Schwartz at 432. There is simply no way to determine whether Duke's design-basis LOCA analysis underestimates or overestimates the degree of clad swelling (and hence the degree of fuel relocation) for MOX LTAs without additional experimental data from integral LOCA tests of high-burnup MOX fuel rods. Given the lack of data, BREDL finds unpersuasive the NRC's 1999 speculation, quoted by Duke in its April 14, 2004 set of responses to BREDL's discovery requests, that "a major effect is not expected" with regard to differences in pellet-clad bonding between MOX and LEU. Id. at 15.
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Q.13: Please explain the basis for your concern regarding clad balloon size and its impact on the severity of fuel relocation affecting the MIOX LTAs in a LOCA.
A.13: The MOX LTAs avill use M5 cladding, as compared to the Zircaloy4 orZIRLO cladding that is extensively used in US PWRs. According to IRSN, M5 will form larger balloons than Zircaloy-4 in a design-basis LOCA because it remains more ductile during irradiation. October 2003 IRSN presentation to NRC at 24. The greater retained ductility of M5 as a function of burnup compared to Zircaloy-4 can result in a greater M5 balloon size during a design-basis LOCA for fuel rods of the same burnup. Larger balloons increase the space available for fuel fragments to fall and hence result in a greater propensity for fuel relocation during a LOCA, with an associated increase in PCT and local clad oxidation.
Q.14: A group of experts from Electricit6 de France (EDF), Framatome ANP and the French CEA recently challenged IRSN's assertion that M5 cladding would form bigger balloons during a LOCA than zircaloy4 in a presentation at Argonne National Laboratory. Please explain your view of this position.
A. 14: I do not believe the EDF presentation responds adequately to the issue that IRSN has raised. Their claim is that the Edgar creep tests -- which indicated a greater ductility and a larger balloon size for M5 than for zircaloy are not the appropriate tests to actually evaluate balloon size during LOCAs. Ramp tests utilizing pre-hydrided cladding samples, which EDF asserts are more representative ofLOCA conditions, indicate that the balloon size for M5 is not actually greater than for zircaloy4.
Obviously, a ramp test would be more similar to the conditions experienced during a LOCA than a steady-state creep test. However, neither creep tests nor ramp tests utilizing pre-hydrided but unirradiated cladding materials adequately simulate all the relevant phenomena that could affect balloon formation during a LOCA involving high-bumup fuel. For example, a well-known property of M5 cladding is that it generates a thinner oxidation layer during normal irradiation as a function of burnup than zircaloy-4. Zircaloy-4 at high burnups tends to generate a thick oxidation layer that's prone to spalling. Spalling will cause spatial inhomogeneities in the clad temperature that negatively affect ductility, leading to earlier cladding ruptures during a LOCA and hence smaller balloon sizes. I don't think that the ramp tests described by EDF take that effect into account. Therefore, I don't believe that the EDF presentation fully addresses the differences that would be observed in actual irradiated fuel with regard to the ductility and the balloon size of M5 compared to that of zircaloy-4.
8
7Z-Sb This question remains unresolved because there is an absence of experimental data on the performance of high-burnup, M5-clad fuel, under design-basis LOCA conditions. The Electric Power Research Institute (EPRI) and Areva (parent company ofFramatome ANP) apparently continue to deny NRC access to samples of irradiated high-burnup MS-clad LEU fuel for testing at Argonne National Laboratory. Letter from Ashok C. Thadani, NRC, to David Modeen, EPRI (April 21, 2004) (ADAMS ACN # ML04 1l30490). A copy of this letter is attached to my testimony as Exhibit K. This lack of cooperation can only cause further delays in the ability of NRC to obtain the experimental data it needs to confirm the safety of high-burnup M5-clad fuel (whether LEU or MOX).
I would underscore the admission of M. Blanpain of AREVA during the ACRS Reactor Fuels Subcommittee Meeting on April 21, 2004 that MOX fuel irradiated in France is predominantly clad in Zircaloy-4, and only "some MS fuel rods with MOX for experimental purposes" have been used in France. See Transcript at 61-62. For some reason, France is reluctant to use MS-clad MOX fuel domestically and is primarily producing it for export to Germany (and now to the United States). However, even in Germany the use of M5-clad MOX has been extremely limited. And I am unaware of any integral LOCA tests performed with irradiated M5-clad MOX fuel.
Q.15: Please explain the basis for your concern regarding the impact of fuel relocation on the ability of the MOX LTA core to satisfy the regulatory requirement for coolable core geometry.
A.l5: As stated above, fuel relocation increases the local linear heat generation rate The maximum flow blockage that wvill preserve a coolable geometry depends on the assumed heat source and the heat transfer properties of the fuel bundle. As IRSN points out, acceptable bundle blockage ratios were derived based upon arrays of unirradiated fuel rods, and did not take into account fuel relocation and its associated impacts on the redistribution of the decay heat source within the fuel rods. IRSN presentation to NRC at 29 (October 23, 2003). IRSN restated its concern in a recent presentation:
"The impact of fuel relocation in fuel rod balloons, as was observed in all in-reactor tests with irradiated fuel, leading to an increase in local power (lineic and surfacic) ..., on the coolability of the blocked region, is still fully questionable and should be addressed by specific analytical tests with a simulation of fuel relocation."
C. Grandjean and G. Hache, "LOCA Issues Related to Ballooning, Fuel Relocation, Flow Blockage and Coolability," SEGFSM Topical Meeting on LOCA Issues, Argonne National Laboratory at 23 (May 25-27, 2004) (emphasis in original). A copy of this paper is attached to my testimony as Exhibit L.
Thus, any analysis that does not take this into account is incomplete and is likely to be non-conservative. Lack of consideration of this phenomenon will be of greater concern for the MOX LTA core to the extent that the MOX LTAs have a smaller margin to regulatory limits than LEU fuel.
9
Q. 16: Please explain the basis for your concern regarding the smaller safety margins for MOX fuel with respect to peak clad temperature in a LOCA.
A.16: As Duke's calculations have demonstrated, the PCT in a design-basis LOCA is higher for a MOX rod than for an LEU rod in the same position in the core. Duke MOX LTA LAR at 3-43 (February 27, 2003). The margin to the 10 CFR §50.46 PCT limit of 2200'F is therefore smaller for a MOX rod than for an LEU rod in the same position.
At high burnups, the linear heat generation rate for MOX fuel is generally higher than that for LEU fuel. This, in turn, results in increased centerline temperature and stored energy, therefore reducing the margin to design-basis LOCA regulatory limits. BREDL maintains that every reduction in margin associated with MOX fuel use, coupled with the non-conservatism of ignoring fuel relocation effects, reduces confidence in Duke's design-basis LOCA analysis of the MOX LTA core.
Because there is little or no experimental data to conclusively validate the impact of relocation on either LEU or MOX fuel, a design-basis MOX LTA LOCA analysis that takes relocation into account would be highly uncertain --- with a resulting large uncertainty in the calculation of the relocation-associated increase in PCT of a MOX LTA fuel rod compared to the relocation-associated increase in PCT of an LEU fuel rod. For instance, if the MOX filling ratio is 70% and the LEU filling ratio is only 40%, because of a greater quantity of fine fragments in the MOX fuel, the increase in PCT could be nearly three hundred degrees Fahrenheit greater for MOX than for LEU (assuming that no other MOX-related effect, such as a greater initial linear heat generation rate, results in an even more severe increase in PCT associated with relocation).
The PCT calculated by Duke for the MOX LTA is 201 80 F. Obviously, a relocation-associated increase in PCT of, say 313'F (associated with a 70% filling ratio for LEU fuel), would result in an exceedance of the 2200'F limit by 1310 F. On the other hand, if the LEU filling fraction is closer to 40%, the increase in PCT would only be about 40'F, and the LEU fuel would still be in compliance with the regulatory limit. Thus the MOX LTAs could well be limiting with respect to LOCA compliance if relocation is fully accounted for.
These significant uncertainties should be reflected in Duke's analysis, and NRC approval should be contingent upon a demonstration that uncertainties of this magnitude do not undermine reasonable assurance of adequate protection of the public health and safety. I do not believe that such a finding can be made, given the potential severity of the relocation phenomenon and its associated uncertainties.
Q.17: Please discuss how, in your opinion, the gaps in the experimental database for the behavior of highi-burnup, M5-clad MOX fuel during LOCAs can be reduced.
A.17: The only way to fully address the uncertainties associated with the behavior of high-burnup, M5-clad MOX fuel during LOCAs is to conduct integral LOCA tests of such fuel, fabricated with the same specifications as the lead test assemblies that are under consideration here, and irradiated to a range of bumups, including the maximum of 60 GWD/t that Duke has requested 10
2-7SZ.
in its LAR. The proposed Phebus test series would likely make a substantial contribution to reducing the level of uncertainty associated with MOX fuel behavior during LOCAs.
These integral tests could be supplemented with separate-effects tests specifically designed to look at fuel relocation as a function of bumup for both MOX and LEU fuel, and to measure the relative susceptibility to relocation of the two types of fuels. The Halden IFA-650 test, which I understand is being designed to examine fuel relocation effects in LEU fuel, could help to resolve some of these questions. But similar tests on mixed oxide fuel will also be needed. And separate effects tests cannot reproduce the complex, interrelated set of thermal-hydraulic and mechanical phenomena that would occur during a LOCA and would affect fuel relocation.
Q.18: Does the Staffs Safety Evaluation Report (SER) provides any insight into the issues raised by Contention I?
A.18: The SER doesn't address the issues that we've raised concerning the impact of relocation.
So to that extent, it doesn't affect my conclusions at all. Members of the Staff admitted during the ACRS subcommittee meeting on the LTA LAR application that they have not done their own independent calculations to confirm Duke's LOCA analyses. The Staff has only checked Duke's results for internal consistency, rather than doing any of its own simulations. Therefore, to the extent that the Staff claims to have independently verified the adequacy of Duke's LOCA analysis, I do not believe that claim is correct.
Q.19: Does this conclude your testimony?
A.19. Yes.
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CERTIFICATE OF SERVICE I hereby certify that on July 1, 2004, copies of the Written Prefiled Testimony of Dr. Edwin S.
Lyman Regarding Contention I were served on the following by e-mail, as indicated below. The
- exhibits to Dr. Lyman's testimony, plus a hard copy of his testimony, were served by hand or by Federal Express on the following day, as indicated below:
Ann Marshall Young, Chair Susan L. Uttal, Esq.
Administrative Judge Antonio Femandez, Esq.
Atomic Safety and Licensing Board Margaret J. Bupp, Esq.
U.S. Nuclear Regulatory Commission Office of the General Counsel 1 1545 Rockville Pike 11555 Rockville Pike Rockville, MD 20852 Rockville, MD 20852 E-mail: AMY@nrc. gov 11545 Rockville Pike EXHIBITS/HARD COPY BY HAND 7/2/04 Rockville, MD 20852 E-mail: slu6nrc.gov axf2@nrc.gov, Anthony J. Baratta mjb5@nrc.gov Administrative Judge EXHIBITS/HARD COPY BY HAND 7/2/04 Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Mary Olson 11545 Rockville Pike Southeast Office, Nuclear Information and Rockville, MD 20852 Resource Service E-mail: AJB5@nrc.gov P.O Box 7586 EXHIBITS/HARD COPY BY HAND 7/2/04 Asheville, NC 28802 E-mail: nirs. sermindspring. com ALSO BY FIRST-CLASS MAIL Office of Commission Appellate Adjudication U.S. Nuclear Regulatory Commission Lisa F. Vaughn, Esq.
Mail Stop: 0-16C1 Legal Dept. (PBO5E)
Washington, D.C. 20555 Duke Energy Corporation ALSO BY FIRST-CLASS MAIL ON 7/2/04 526 South Church Street (ECI IX)
Charlotte, NC 28201-1006 Thomas S. Elleman 704/382-8134 Administrative Judge E-mail: 1fVaughn~duke-energy.com Atomic Safety and Licensing Board EXHIBITS AND HARD COPY SERVED BY 11545 Rockville Pike FEDERAL EXPRESS ON JULY 2, 2004 Rockville, MD 20852 E-mail: ellemaneeos.ncsu.edu Janet Marsh Zeller, Executive Director EXHIBITS/HARD COPY BY HAND 7/2/04 Blue Ridge Environmental Defense League P.O. Box 88 Glendale Springs, NC 28629 E-mail: BREDL@skybest . corn ALSO BY FIRST-CLASS MAIL
2 David A. Repka, Esq._
Anne W. Cottingham, Esq.
Winston & Strawn, LLP 1400 L Street, N.W.
Washington, D.C. 20005-3502 11545 Rockville Pike Rockville, MD 20852 E-mail: drepka6wins ton.comr acotting~winston.com EXHIBITS/HARD COPY BY HAND 7/2/04 Office of the Secretary (original and two copies)
ATTN: Docketing and Service U.S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852 301/415-1966 11545 Rockville Pike Rockville, MD 20852 E-mail: HEARINGDOCKET@nrc. gov EXHIBITS/HARD COPY BY HAND 7/2/04 (Diane Curran
July 1, 2004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of: )
Docket Nos. 50-413-OLA DUKE ENERGY CORPORATION 50-414-OLA (Catawba Nuclear Station, )
Units I and 2)
PREFILED WRITTEN TESTIMONY OF DR. EDWIN S. LYMAN REGARDING CONTENTION I On behalf of Blue Ridge Environmental Defense League ("BREDL"), Dr. Edwin S.
Lyman hereby submits the following testimony regarding BREDL's Contention I.
Q.1. Please state your name and describe your professional qualifications to give this testimony.
A.1. My name is Dr. Edwin S. Lyman. I am a Senior Scientist with the Global Security Program at the Union of Concerned Scientists, 1707 H Street, NW, Suite 600, Washington, D.C.
20006. My education and experience are described in my curriculum vita, which is attached to my testimony as Exhibit A.
I am a qualified expert on nuclear safety and safeguards issues. I hold a Ph.D., a master of science degree, and a bachelor's degree in physics. For over eleven years, I have conducted research on security and environmental issues associated with the management of nuclear materials and the operation of nuclear power plants. My research has included the safety and security implications of using mixed oxide fuel as a substitute for uranium fuel in nuclear power plants. I have also published articles on this topic. A list of my publications is included in my C.V.
Q.2. What is the purpose of your testimony?
A.2. The purpose of my testimony is to discuss my views on BREDL Contention I, which was admitted for litigation by the Atomic Safety and Licensing Board ("ASLB") in LBP-04-04, Memorandum and Order (Ruling on Standing and Contentions) (March 5, 2004). BREDL Contention I asserts that Duke Energy Corporation's ("Duke's") license amendment request
("LAR") to test plutonium mixed oxide ("MOX") fuel at the Catawba nuclear power plant is inadequate because Duke has failed to account for the differences between MOX and low enriched uranium ("LEU") fuel behavior; nor has Duke accounted for the impact of these differences on Duke's analysis of loss of coolant accidents ("LOCAs").
Q.3. Wlhat materials have you reviewed in preparation for your testimony?
A.3. I have reviewed Duke's LAR and related correspondence, including Duke's responses to Requests for Additional Information ("RAIs") by the NRC Staff. I have also reviewed the body of literature which has been developed regarding the behavior of MOX and other types of reactor fuel under LOCA conditions. I am also familiar with relevant NRC documents, including correspondence regarding this license amendment application, reports and correspondence concerning characteristics and behavior of MOX fuel, and correspondence and reports concerning the behavior of LEU fuel under LOCA conditions. In addition, I am familiar with regulations and guidance of the U.S. Nuclear Regulatory Commission ('NRC") and the U.S.
Department of Energy ("DOE") governing plutonium processing facilities. Finally, I am familiar with U.S. and foreign government reports regarding testing of LEU fuel under accident conditions.
Q.4. Please summarize the conclusions you have reached regarding the adequacy of Duke's LAR application to account for the differences between MOX and LEU fuel.
A.4. In my professional judgment, Duke's design-basis loss of coolant ("DB-LOCA') analysis is inadequate because it does not address the uncertainties associated with relocation effects that M5-clad MOX fuel may experience under LOCA conditions. These uncertainties relate to Duke's assertion that the action proposed in the MOX LTA LAR will not result in a violation of the emergency core cooling system (ECCS) acceptance criteria in 10 C.F.R. § 50.46: peak cladding temperature ("PCT"), maximum cladding oxidation, and the preservation of a coolable core geometry.
The phenomenon of fuel relocation has been observed in experiments with irradiated LEU fuel under LOCA conditions. While to my knowledge no similar experiments have been done on MOX fuel, there are technical reasons to believe that the impact of fuel relocation effects during a LOCA may be more severe for MOX fuel rods than for LEU fuel rods of the same burnup, due to differences in characteristics such as fuel fragment sizes and fuel-clad interactions. Moreover, calculations in Duke's LAR indicate that MOX fuel is generally more limiting than LEU fuel with respect to DB-LOCAs. Therefore, the consequences of fuel relocation, and the non-conservatism associated with neglecting them, may be of greater concern for MOX fuel rods than for LEU fuel rods with respect to compliance with LOCA regulatory criteria.
Duke has failed to address these uncertainties in MOX fuel behavior, and therefore its LTA application is unacceptable to satisfy the requirements of 10 C.F.R. § 50.46 with respect to PCT, maximum cladding oxidation, and coolable geometry of fuel. In addition, by failing to address the uncertainties in MOX fuel behavior, Duke has not demonstrated compliance with the general reasonable assurance standard in 10 C.F.R. § 50.40(a).
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7-7<S7 I do not believe, however, that these uncertainties can be addressed with mere calculations or analyses based on LEU performance. In my professional opinion, the only satisfactory way to address these uncertainties would be to conduct integral tests of MOX fuel assemblies under LOCA conditions in such a manner that the impacts of the phenomena I have previously described can be measured with reasonable accuracy and precision.
Q 5: Please explain how the regulations in 10 C.F.R. § 50.46 apply to Contention 1.
A.5: NRC regulations at 10 C.F.R. § 50.46 establish acceptance criteria for emergency core cooling systems for light-water nuclear reactors. Essentially, the regulation sets design limits for behavior of the reactor fuel under LOCA conditions. Appendix K to Part 50, whose requirements are referenced in 10 C.F.R. § 50.46(a)(1), sets forth ECCS "evaluation models)' Iaemptiows abatthba ear that are to be use in determining whethri s with the criteria in 10 C.F.R. § 50.46. IC rifLo 1/kct ncckcp&AL(o.
10 C.F.R. § 50.46 and Appendix K apply only to uranium-based fuel, but Duke has requested an exemption from this limitation so that these requirements will apply to MOX fuel. I believe that it is generally appropriate to apply the requirements of 10 C.F.R. § 50.46 to MOX fuel, as long as Appendix K is not strictly applied to exclude consideration of relocation of the fuel during LOCAs.
The regulations in 10 C.F.R. § 50.46 sets forth fuel performance limits in three categories that have importance with respect to performance of MOX fuel: peak cladding temperature ("PCT"),
maximum cladding oxidation, and coolable geometry. Section 50.46(b)(1) requires that that PCT "shall not exceed 22000 F." Section 50.46(b)(2) provides that the "calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation." Section 50.46(b)(4) also requires that "[c]alculated changes in core geometry shall be such that the core remains amenable to cooling."
Q.6: Please explain why yhou think the Appendix K evaluation models for the MOX LTA core should include consideration of fuel relocation during LOCAs.
A.6: Appendix K does not include consideration of fuel relocation. The NRC did contemplate including fuel relocation as a criterion in Appendix K, but claimed to have resolved the question in Generic Issue 92. Memorandum from Ralph Meyer, NRC Office of Nuclear Regulatory Research, to John Flack, NRC Regulatory Effectiveness and Human Factors Branch, re: Update on Generic Issue 92, Fuel Crumbling During LOCA (February 8, 2001) (NRC ACN #
ML010390163) (hereinafter "Meyer Memorandum"). A copy is attached to my testimony as Exhibit 13. See also Memorandum from Ashok C. Thadani, Office of Nuclear Regulatory Research, to Samuel J. Collins, Office of Nuclear Reactor Regulation, re: Information Letter 0202, Revision of 10 CFR 50.46 and Appendix K (June 20,2002) (hereinafter "Thadani Memorandum") (NRC ACN # ML021720690). A copy is attached to my testimony as Exhibit C. More recently, the NRC has acknowledged that omission of fuel relocation effects is a non-conservatism in Appendix K with a very large potential impact on PCT, and that an early "resolution" of this issue (i.e., Generic Issue 92) may have been in error or is no longer 3
-Zzf.6 applicable because of new infoirmation. See; Meyer Memorandum and Thadani Memorandum, Attachment 4 at 4-5.
As I will discuss in more detail later in my testimony, certain characteristics of MOX fuel appear to exacerbate the effects of fuel relocation, thus leading to higher PCTs and greater maximum cladding oxidation. While there are several other known non-conservatisms in Appendix K, this one in particular appears to be relevant to the MOX LTA LAR because of its disproportionately large impact on the MOX LTAs compared to the LEU assemblies that comprise the remainder of the core. Given the potential impact on PCT of relocation effects, it is not appropriate to omit consideration of this phenomenon from the Appendix K models that Duke uses to establish that loading of the MOX LTAs into Catawba will result in compliance with 10 CFR § 50.46 criteria.
Q.7: Please explain what you mean by fuel relocation during a LOCA.
A.7: According to NRC, "fuel relocation refers to the movement of fuel pellet fragments into regions of the fuel rod where the cladding has ballooned during a LOCA transient." Thadani Memorandum, Attachment 4 at 4-5.
Q.8: Please explain wvliy fuel relocation could increase the severity of a LOCA.
A.8: Fuel relocation increases the local linear heat generation rate within the ballooned area. Thus it could increase the severity of a LOCA by resulting in a greater fuel rod peak cladding temperature (PCT) than in a situation in which fuel relocation did not occur. Because transient oxidation during a LOCA increases with an increase in PCT, fuel relocation could also result in a greater maximum cladding oxidation. Finally, the greater local linear heat generation rate requires a greater coolant flow around the ballooned area to ensure long-term core coolability. See slides presented by A.
Mailliat and J.C. M6lis, IRSN, at "PHEBUS STLOC Meeting" with NRC Staff (October 23, 2003) (NRC ACN # ML032970624) (hereinafter October 2003 IRSN Presentation"). A copy is attached to my testimony as Exhibit D.
Q.9: Please discuss the potential magnitude of the impact of fuel relocation on PCT and maximum cladding oxidation for uranium oxide (U0 2 ) fuel.
A.9: The most recent calculations of the impact of fuel relocation on PCT of which I am aware were conducted by the Institut de Protection et de Suret6 Nucl6aire (IPSN, now IRSN) and published in 2001. In that study, the authors used the CATHARE2 computer code to calculate the impact of fuel relocation on the large-break LOCA PCT for a high-burnup U0 2 fuiel rod as a function of the "filling ratio," or the ratio of the volume of the relocated fuel material to the volume of the ballooned region. For the scenario evaluated, the authors found that the PCT in the absence of relocation effects was 970'C. For a filling ratio of 70%, the maximum considered, the PCT was 11440C. For a filling ratio of 40%, the PCT was about 20'C greater than for the no-relocation case. Thus the maximum impact on PCT of relocation in this study was a APCT of +
1741C (313'F) for high-burnup U02 fuel. It is not clear from the study whether higher filling ratios, and hence larger impacts on PCT, are possible. C. Grandjean, G.Hache and C. Rongier, "High Burmup U0 2 Fuel LOCA Calculations to Evaluate the Possible Impact of Fuel Relocation After Burst," OECD/NEA Proceedings of the Topical Meeting on LOCA Fuel Safety Criteria, Aix-en-4
T 2 2fC Provence (March 22-23, 2001) (hereinafter "Grandjean, Hache, and Rongier"). A copy of this paper is attached to my testimony as Exhibit E. The NRC staff appears to be familiar with this result. See Thadani Memorandum, Attachment 5 at 4.
The study also evaluated the impact on the maximum cladding oxidation for the ruptured region (two-sided oxidation). The equivalent cladding reacted (ECR) calculated by the Cathcart-Pawel rate law (a surrogate for "maximum cladding oxidation') was 12.6% for the no-relocation case, and 19.7% for the 70% filling ratio case. Thus the maximum impact on ECR resulting from relocation
,vas calculated as AECR = 7.1%.
Q.10: Please explain why the impact of fuel relocation on the severity of a LOCA could be greater for MOX fuel than for U0 2 fuel at the same burnup.
A. 10: Experts have concluded that MOX fuel may experience more severe relocation effects than U0 2 fuel at the same bumup. The IPSN study above did not explicitly consider MOX fuel, but stated that "it must be pointed out that that results of corresponding calculations with ... high burnup MOX fuels would have been more severe with regard to acceptance limits." Grandjean, Hache and Rongier at 7.
IRSN, the successor to IPSN, has reiterated these concerns, stating in a recent presentation that for MOX fuel, a "higher initial energy" and an "enhance [sic] of fiel relocation impact" results in greater increases in PCT and ECR associated with relocation. V Guillard, C. Grandjean, S.
Bourdon and P. Chatelard, "Use of CATHARE2 Reactor Calculations to Anticipate Research Needs," SEGFSM Topical Meeting on LOCA Issues, Argonne National Laboratory, slides at 8-9 (May25-26, 2004) (NRC ACN# ML041600261). A copy of this paper is attached to my testimony as Exhibit F. In the abstract for this presentation, the authors state that "a lack of knowledge on theses [sic] parameters [important for relocation] for irradiated U02 and particularlyMOXfitel [emphasis added] may lead to reduce [sic] safety margins."
MOX fuel may experience more severe relocation effects than U0 2 fuel at the same burnup because several characteristics that are important for relocation may be less favorable for MOX fuel. These include pellet fragment size and fuel-clad interaction.
Q.11: Please explain the basis for your concern regarding the pellet fragment size of MOX fuel and its impact on fuel relocation in a LOCA.
A.11: The IPSN calculations cited above demonstrate the high sensitivity of fuel relocation-induced increases in PCT and ECR to the filling ratio. The filling ratio, in turn, is a function of the average particle size of the relocated fuel fragments, in that smaller particles will in general result in greater packing of the relocated area and hence higher filling ratios.
The fiel relocation phenomenon has been observed in LEU fuel for rod bumups exceeding around 48 GWD/t. See Grandjean, Hache and Rondier at 2 (2001). This suggests that vulnerability to fuel relocation is associated with the development of the high-bumup "rim" region known to emerge in LEU fuel for bumups exceeding about 40-45 GWD/t. IPSN states 5
that "fuel fragmentation is clearly associated to [sic] burnup, with finer fragments at higher BU."
See Grandjean, Hache and Rondier at 2 (2001).
For During manufacture of MOX fuel using the MIMAS process (which will be used for the Duke LTAs), plutonium agglomerates --- macroscopic clumps of plutonium-rich particles ---
occur in the fuel. Because the fissile material is concentrated in these clumps, very high local bumups result, due to the fact that the fission is occurring in a heterogeneous fashion. The ratio of local bumup within the agglomerates is on the order of 4-6 times the rod-average bumup, depending on the irradiation time. For instance, the agglomerate bum-up reaches about 60 GWD/t when the rod average is only around 18 GWD/t, and reaches 100 GWD/t when the rod average is only 28.4 GWD/t As a result, high-bumup rim-like regions emerge in the outer layers of the plutonium agglomerates for much lower rod-average burnups than 40-45 GWD/t, because the local bumups within the plutonium agglomerates increase much more rapidly than the rod-average bumups. Thus it is reasonable to expect that the onset of fuel relocation in MOX fuel may occur at lower rod-average burnups than in LEU fuel. This would imply that MOX fuel will be vulnerable earlier in its irradiation history (and consequently for a longer time) than LEU fuel.
Also, the particle size distribution in MOX fuel will be smaller than in LEU fuel at the same rod-average bumup, to the extent that fine fragments are generated in the ultra-high bumup plutonium agglomerate regions.
Fuel fragmentation can also be caused by the stress induced by the stored-energy redistribution during the blowdown phase of a LOCA. A. Mailliat and M. Schwarz, "Need for Experimental Programmes on LOCA Issues Using High Bum-Up and MOX Fuels," NUREG/CP-0 176, Proceedings of the Nuclear Safety Research Conference at 436 (May 2002) (NRC ACN #
ML021710793) (hereinafter "Mailliat and Schwarz"). A copy of this paper is attached to my testimony as Exlhibit G. Because MOX fuel has a lower thermal conductivity and a higher radial temperature gradient than LEU fuel, it could experience greater fuel fragmentation during the blowdown and more severe relocation effects as a result.
According to two out of four NRC experts who participated in the 2001 PIRT panel on LOCAs and high-bumup fuel, the composition of fuel (i.e. a specified MOX composition) is of "high importance" for consideration of fuel relocation effects because it "may affect the amount of fine grain material after relocation. Fuel structure and mechanical properties are influenced by fuel type." See NUREG/CR-6744, "Phenomenon Identification and Ranking Tables for Loss-of-Coolant Accidents in Pressurized and Boiling Water Reactors Containing High-Bumup Fuel,"
Appendix D, Table D-1 at D-67 (December 2001) (NRC ACN # 013540623) (hereinafter "NUREG/CR-6477"). Relevent portions of this report are attached to my testimony as Exhibit II. One expert concluded that fuel composition was of moderate importance to relocation, stating that "the consequence of fuel fragments relocation (higher local decay heat and higher cladding temperature) could be more effective with MOX fuel than with U02 fuel" but that "the viscoelastic properties of the MOX should impair the fuel fragments relocation at high bumup."
Id. at D-67. A fourth expert concluded that fuel composition would be of only low importance to relocation. Id. at D-67. This difference of expert opinion highlights the inadequacies of the experimental database with regard to integral tests of MOX fuel under design-basis LOCA conditions, and underscores the significant uncertainties in Duke's design-basis LOCA analysis.
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U& I Q.12: Please explain the basis for your concern regarding the effects of fuel-clad interaction within the MOX LTAs and their impact on fuel relocation in a LOCA.
A.12: I am concerned about differences between MOX and LEU fuel with respect to fuel-clad bonding and the impact of such differences on fuel relocation behavior during a design-basis LOCA. According to IPSN (now IRSN), tight fuel-clad bonding may delay the onset of fuel relocation. Mailliat and Schwarz at 433. Tight bonding has also been observed at the Halden reactor in Nonvay to retard the rate of balloon formation. Nuclear Energy Agency, NEA/CSNI/R(2003)9, Ongoing andPlannedFuel Safety Research in NEA Member States at 79 (March 5, 2003). Relevant excerpts of this report are attached to my testimony as Exhibit I. During NRC's recent expert elicitation (PIRT) process on LOCA issues for high-burnup fuel, all four participating experts agreed that "chemical and mechanical bonding between the fuel pellet and the cladding ... " was of high importance to the fuel relocation phenomenon, because "bonding could significantly affect the relocation characteristics by impeding pellet fragment movement."
NUREG/CR-6744, Table D-1 at D-69. It has been confirmed that MOX fuel is more resistant to clad failures due to pellet-clad mechanical interaction (PCMI) than LEU fuel, even at high burnups. Nuclear Energy Agency, NEAINSC/DOC(2004)8, International Seminar on Pellet-CladInteractionswvith Mater ReactorFufels, at 20 (May 6, 2004).
Relevant excerpts of this report are attached to my testimony as Exhibit J. This phenomenon is not well-understood but may imply that the pellet-clad bond is weaker for MOX fuel, in which case MOX fuel may have a greater propensity to earlier and more extensive fuel relocation than LEU.
In Duke's April 14, 2004, Response to BREDL's first set of discovery requests, Duke stated that the Framatome design-basis LOCA analysis for the MOX LTAs did not assume any fuel-clad bonding and was therefore "conservative" with respect to the requirement that the degree of cladding swelling not be underestimated. Id. at 14. However, in the absence of an assessment of whether and to what extent the pellet-clad interaction is weaker in MOX fuel than in LEU fuel, there is no way of knowing the degree to which this assumption is conservative for MOX fuel. Therefore, Duke's failure to properly account for this phenomenon contributes another uncertainty to the safety margin associated with Duke's design basis LOCA calculation.
Moreover, there is evidence to contradict Duke's assertion that "deterministic LOCA evaluations typically based on data taken from unirradiated cladding" are conservative with respect to clad swelling. According to IPSN (now IRSN), results from the PBF-LOC experiments found that irradiated rods experienced greater clad deformation than unirradiated rods during design-basis LOCA conditions. See Mailliat and Schwartz at 432. There is simply no way to determine whether Duke's design-basis LOCA analysis underestimates or overestimates the degree of clad swelling (and hence the degree of fuel relocation) for MOX LTAs without additional experimental data from integral LOCA tests of high-burnup MOX fuel rods. Given the lack of data, BREDL finds unpersuasive the NRC's 1999 speculation, quoted by Duke in its April 14, 2004 set of responses to BREDL's discovery requests, that "a major effect is not expected" with regard to differences in pellet-clad bonding between MOX and LEU. Id. at 15.
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Q.13: Please explain the basis for your concern regarding clad balloon size and its impact on the severity of fuel relocation affecting the MOX LTAs in a LOCA.
A.13: The MOX LTAs will use M5 cladding, as compared to the Zircaloy-4 or ZIRLO cladding that is extensively used in US PWRs. According to IRSN, M5 will form larger balloons than Zircaloy-4 in a design-basis LOCA because it remains more ductile during irradiation. October 2003 IRSN presentation to NRC at 24. The greater retained ductility of M5 as a function of burnup compared to Zircaloy-4 can result in a greater M5 balloon size during a design-basis LOCA for fuel rods of the same burnup. Larger balloons increase the space available for fuel fragments to fall and hence result in a greater propensity for fuel relocation during a LOCA, with an associated increase in PCT and local clad oxidation.
Q.14: A group of experts from Electricit6 de France (EDF), Framatome ANP and the French CEA recently challenged IRSN's assertion that M5 cladding would form bigger balloons during a LOCA than zircaloy4 in a presentation at Argonne National Laboratory. Please explain your view of this position.
A. 14: I do not believe the EDF presentation responds adequately to the issue that IRSN has raised. Their claim is that the Edgar creep tests --- which indicated a greater ductility and a larger balloon size for MS than for zircaloy-4 -- are not the appropriate tests to actually evaluate balloon size during LOCAs. Ramp tests utilizing pre-hydrided cladding samples, which EDF asserts are more representative of LOCA conditions, indicate that the balloon size for M5 is not actually greater than for zircaloy-4.
Obviously, a ramp test would be more similar to the conditions experienced during a LOCA than a steady-state creep test. However, neither creep tests nor ramp tests utilizing pre-hydrided but unirradiated cladding materials adequately simulate all the relevant phenomena that could affect balloon formation during a LOCA involving high-bumup fuel. For example, a well-known property of M5 cladding is that it generates a thinner oxidation layer during normal irradiation as a function of bumup than zircaloy-4. Zircaloy-4 at high burnups tends to generate a thick oxidation layer that's prone to spalling. Spalling will cause spatial inhomogeneities in the clad temperature that negatively affect ductility, leading to earlier cladding ruptures during a LOCA and hence smaller balloon sizes. I don't think that the ramp tests described by EDF take that effect into account. Therefore, I don't believe that the EDF presentation fully addresses the differences that would be observed in actual irradiated fuel with regard to the ductility and the balloon size of M5 compared to that of zircaloy-4.
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%n&3 This question remains unresolved because there is an absence of experimental data on the performance of high-burnup, M5-clad fuel, under design-basis LOCA conditions. The Electric Power Research Institute (EPRI) and Areva (parent company of Framnatome ANP) apparently continue to deny NRC access to samples of irradiated high-bumup M5-clad LEU fuel for testing at Argonne National Laboratory. Letter from Ashok C. Thadani, NRC, to David Modeen, EPRI (April 21, 2004) (ADAMS ACN # ML041130490). A copy of this letter is attached to my testimony as Exhibit K. This lack of cooperation can only cause further delays in the ability of NRC to obtain the experimental data it needs to confirm the safety of high-burnup M5-clad fuel (whether LEU or MOX).
I would underscore the admission of M. Blanpain of AREVA during the ACRS Reactor Fuels Subcommittee Meeting on April 21, 2004 that MOX fuel irradiated in France is predominantly clad in Zircaloy-4, and only "some M5 fuel rods with MOX for experimental purposes" have been used in France. See Transcript at 61-62. For some reason, France is reluctant to use M5-clad MOX fuel domestically and is primarily producing it for export to Germany (and now to the United States). However, even in Germany the use of M5-clad MOX has been extremely limited. And I am unaware of any integral LOCA tests performed with irradiated M5-clad MOX fuel.
Q.15: Please explain the basis for your concern regarding the impact of fuel relocation on the ability of the MIOX LTA core to satisfy the regulatory requirement for coolable core geometry.
A.15: As stated above, fuel relocation increases the local linear heat generation rate The maximum flow blockage that will preserve a coolable geometry depends on the assumed heat source and the heat transfer properties of the fuel bundle. As IRSN points out, acceptable bundle blockage ratios were derived based upon arrays of unirradiated fuel rods, and did not take into account fuel relocation and its associated impacts on the redistribution of the decay heat source within the fuel rods. IRSN presentation to NRC at 29 (October 23, 2003). IRSN restated its concern in a recent presentation:
"The impact of fuel relocation in fuel rod balloons, as was observed in all in-reactor tests with irradiated fuel, leading to an increase in local power (lineic and surfacic) ..., on the coolability of the blocked region, is still fully questionable and should be addressed by specific analytical tests with a simulation of fuel relocation."
C. Grandjean and G. Hache, "LOCA Issues Related to Ballooning, Fuel Relocation, Flow Blockage and Coolability," SEGFSM Topical Meeting on LOCA Issues, Argonne National Laboratory at 23 (May 25-27, 2004) (emphasis in original). A copy of this paper is attached to my testimony as Exhibit L.
Thus, any analysis that does not take this into account is incomplete and is likely to be non-conservative. Lack of consideration of this phenomenon will be of greater concern for the MOX LTA core to the extent that the MOX LTAs have a smaller margin to regulatory limits than LEU fuel.
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Q. 16: Please explain the basis for your concern regarding the smaller safety margins for MOX fuel with respect to peak clad temperature in a LOCA.
A. 16: As Duke's calculations have demonstrated, the PCT in a design-basis LOCA is higher for a MOX rod than for an LEU rod in the same position in the core. Duke MOX LTA LAR at 3-43 (February 27, 2003). The margin to the 10 CFR §50.46 PCT limit of 22000 F is therefore smaller for a MOX rod than for an LEU rod in the same position.
At high burnups, the linear heat generation rate for MOX fuel is generally higher than that for LEU fuel. This, in turn, results in increased centerline temperature and stored energy, therefore reducing the margin to design-basis LOCA regulatory limits. BREDL maintains that every reduction in margin associated with MOX fuel use, coupled with the non-conservatism of ignoring fuel relocation effects, reduces confidence in Duke's design-basis LOCA analysis of the MOX LTA core.
Because there is little or no experimental data to conclusively validate the impact of relocation on either LEU or MOX fuel, a design-basis MOX LTA LOCA analysis that takes relocation into account would be highly uncertain --- with a resulting large uncertainty in the calculation of the relocation-associated increase in PCT of a MOX LTA fuel rod compared to the relocation-associated increase in PCT of an LEU fuel rod. For instance, if the MOX filling ratio is 70% and the LEU filling ratio is only 40%, because of a greater quantity of fine fragments in the MOX fuel, the increase in PCT could be nearly three hundred degrees Fahrenheit greater for MOX than for LEU (assuming that no other MOX-related effect, such as a greater initial linear heat generation rate, results in an even more severe increase in PCT associated with relocation).
The PCT calculated by Duke for the MOX LTA is 2018'F. Obviously, a relocation-associated increase in PCT of, say 31 3'F (associated with a 70% filling ratio for LEU fuel), would result in an exceedance of the 2200'F limit by 1310 F. On the other hand, if the LEU filling fraction is closer to 40%, the increase in PCT would only be about 40'F, and the LEU fuel would still be in compliance wvith the regulatory limit. Thus the MOX LTAs could well be limiting with respect to LOCA compliance if relocation is fully accounted for.
These significant uncertainties should be reflected in Duke's analysis, and NRC approval should be contingent upon a demonstration that uncertainties of this magnitude do not undermine reasonable assurance of adequate protection of the public health and safety. I do not believe that such a finding can be made, given the potential severity of the relocation phenomenon and its associated uncertainties.
Q.17: Please discuss how, in your opinion, the gaps in the experimental database for the behavior of high-burnup, M5-clad MIOX fuel during LOCAs can be reduced.
A. 17: The only way to fully address the uncertainties associated with the behavior of high-burnup, MS-clad MOX fuel during LOCAs is to conduct integral LOCA tests of such fuel, fabricated with the same specifications as the lead test assemblies that are under consideration here, and irradiated to a range of bumups, including the maximum of 60 GXD/t that Duke has requested 10
in its LAR. The proposed Phebus test series would likely make a substantial contribution to reducing the level of uncertainty associated with MOX fuel behavior during LOCAs.
These integral tests could be supplemented with separate-effects tests specifically designed to look at fuel relocation as a function of burnup for both MOX and LEU fuel, and to measure the relative susceptibility to relocation of the two types of fuels. The Halden IFA-650 test, which I understand is being designed to examine fuel relocation effects in LEU fuel, could help to resolve some of these questions. But similar tests on mixed oxide fuel will also be needed. And separate effects tests cannot reproduce the complex, interrelated set of thermal-hydraulic and mechanical phenomena that would occur during a LOCA and would affect fuel relocation.
Q.18: Does the Staff's Safety Evaluation Report (SER) provides any insight into the issues raised by Contention I?
A. 18: The SER doesn't address the issues that we've raised concerning the impact of relocation.
So to that extent, it doesn't affect my conclusions at all. Members of the Staff admitted during the ACRS subcommittee meeting on the LTA LAR application that they have not done their own independent calculations to confirm Duke's LOCA analyses. The Staff has only checked Duke's results for internal consistency, rather than doing any of its own simulations. Therefore, to the extent that the Staff claims to have independently verified the adequacy of Duke's LOCA analysis, I do not believe that claim is correct.
Q.19: Does this conclude your testimony?
A. 19. Yes.
11
I CERTIFICATE OF SERVICE r hereby certify that on July 1, 2004, copies of the Written Prefiled Testimony of Dr. Edwin S.
Lyman Regarding Contention I were served on the following by e-mail, as indicated below. The exhibits to Dr. Lyman's testimony, plus a hard copy of his testimony, were served by hand or by Federal Express on the following day, as indicated below:
l Ann Marshall Young, Chair Susan L. Uttal, Esq.
Administrative Judge Antonio Femandez, Esq.
Atomic Safety and Licensing Board Margaret J. Bupp, Esq.
U.S. Nuclear Regulatory Commission Office of the General Counsel 11545 Rockville Pike 11555 Rockville Pike Rockville, MD 20852 Rockville, MD 20852 E-mail: AMYenrc.gov 11545 Rockville Pike EXHIBITS/HARD COPY BY HAND 7/2/04 Rockville, MD 20852 E-mail: slusnrc.gov axf2@nrc.gov, Anthony J. Baratta mjb5@nrc.gov Administrative Judge EXHIBITS/HARD COPY BY HAND 7/2/04 Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Mary Olson 11545 Rockville Pike Southeast Office, Nuclear Information and Rockville, MD 20852 Resource Service E-mail: AJB5@nrc .gov P.O Box 7586 EXHIBITS/HARD COPY BY HAND 7/2/04 Asheville, NC 28802 E-mail: nirs. seemindspring.com ALSO BY FIRST-CLASS MAIL Office of Commission Appellate Adjudication U.S. Nuclear Regulatory Commission Lisa F. Vaughn, Esq.
Mail Stop: 0-16C1 Legal Dept. (PBO5E)
Washington, D.C. 20555 Duke Energy Corporation ALSO BY FIRST-CLASS MAIL ON 7/2/04 526 South Church Street (ECI lX)
Charlotte, NC 28201-1006 Thomas S. Elleman 704/382-8134 Administrative Judge E-mail: 1 fVaughneduke-energy. com Atomic Safety and Licensing Board EXHIBITS AND HARD COPY SERVED BY 11545 Rockville Pike FEDERAL EXPRESS ON JULY 2,2004 I Rockville, MD 20852 E-mail: ellemaneeos .ncsu.edu Janet Marsh Zeller, Executive Director EXHIBITS/HARD COPY BY HAND 7/2/04} Blue Ridge Environmental Defense League P.O. Box 88 Glendale Springs, NC 28629 E-mail: BREDL@skybest . corn ALSO BY FIRST-CLASS MAIL
2 I
David A. Repka, Esq.
Anne W. Cottingham, Esq.
Winston & Strawvn, LLP 1400 L Street, N.W.
Washington, D.C. 20005-3502 11545 Rockville Pike Rockville, MD 20852 E-mail: drepka~wins ton. com acottingqwinston.com EXHIBITS/HARD COPY BY HAND 7/2/04 Office of the Secretary (original and two copies)
ATTN: Docketing and Service U.S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852 301/415-1966 11545 Rockville Pike Rockville, MD 20852 E-mail: HEARINGDOCKETonrc. gov EXHIBITS/HARD COPY BY HAND 7/2/04 (Diane Curran
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of Docket No's. 50-413-OLA, DUKE ENERGY CORPORATION 50-414-OLA (Catawba Nuclear Station, Units I and 2)
REBUTTAL TESTIMONY OF DR. EDWIN S. LYMAN REGARDING BREDL CONTENTION I Blue Ridge Environmental Defense League ("BREDL") hereby presents the following rebuttal testimony by Dr. Edwin S. Lyman regarding BREDL's Contention I.
Q. R-1: Have you read the testimony submitted by Duke Energy Corporation
("Duke") and by the U.S. Nuclear Regulatory Commission ("NRC") Staff on BREDL's Contention 1?
A.R-1: Yes, I have.
Q.R-2: Has the testimony of either Duke or of the NRC Staff caused you to reconsider your opinion regarding the validity of BREDL's safety contention I?
A.R-2: No, it hasn't. Contention I is based on the fact that the experimental database on the behavior of M5-clad MOX fuel under design-basis LOCA conditions is inadequate to resolve issues such as the impact of fuel relocation on MOX fuel. Neither the testimony of Duke nor the testimony of the NRC staff offers any experimental evidence to support their claims regarding the ability of the MOX LTAs to comply with 10 CFR § 50.46 criteria in the presence of relocation effects. Instead, Duke offers a single micrograph of an irradiated, reactor-grade MOX fuel pellet of unspecified burnup. Thus the testimony of Duke and the NRC Staff tends to support, rather than refute, BREDL's contention that the LOCA analysis in Duke's LTA LAR lacks adequate experimental support.
In fact, the NRC Staff testimony confirms several of the assertions that BREDL has made in the course of this proceeding. In response to Question 47, the Staff confirms that fuel relocation could cause the cladding temperature in the balloon to increase by several hundreds of degrees F. In response to Questions 38-39, the Staff also confirms the possibility that the amount of fuel that is relocated in MOX rods may be greater than the amount in LEU rods.
Q.R-3: In response to Question 42, NRC Staff witness Meyer states that "if fuel relocation has any effect, it would increase the temperature only in the ballooned
2zt4 .
region of the fuel rod ... the ballooned region is seldom the location of the calculated peak cladding temperature when relocation is ignored." Do you agree with this statement? Ca \ \I A.R-3: No. According to the results of the FR2 test seri s, fuel relocation has a significant impact on the clad temperature botlf at the rup ured, ballooned location on the I
fuel pin and at non-ballooned locations on the fuel pin. Tjlis can be seen from Figures 10 and 11 in Duke's testimony in response to Questions 47 For low-exposure fuel where relocation does not occur (Figure 10), the maximum temperature at the ballooned location is indeed about 2007F lower than the peak clad temperature, which occurs at a non-ballooned location. However, for high-exposure fuel where relocation does occur, the maximum clad temperature at the ruptured, ballooned location is about 300'F higher than the ruptured, ballooned location for the no-relocation case, and about 200 F lower than the peak clad temperature at the non-ballooned location. Therefore, I would expect from this data that fuel relocation in high-burmup fuel would have a substantial impact on the clad temperature not only at the ruptured location but also at the location where the peak clad temperature occurs.
This conclusion is also consistent with Exhibit C attached to my prefiled testimony, in which NRC states that "the fuel relocation effect on PCT may be significantly larger than that assumed in GS-92 [+46 0F]" (Exhibit C, Attachment 4 at 5) and states that fuel relocation may have a +313 0 F impact on PCT. Exhibit C, Attachment 5 at 4.
While the phenomena involved in these processes are too complicated for simple back-of-the-envelope assessments, some observations can be made. According to the MOX LTA LAR at 3-43, the peak temperature at the hot pin rupture location is 1841 F. (Note that the Duke testimony in response to Question 154 is erroneous in stating that "the highest PCT at the rupture location in the LOCA calculations for Catawba described in the MOX fuel lead assembly license amendment request was approximately 1750'F ... ").
If the 313TF increase in clad temperature associated with fuel relocation with a filling ratio of 0.7 is added to this value, the resulting clad temperature at the rupture location is 21541F. From Figure 11 in Duke's testimony, the PCT in a rod where relocation occurs appears to be about 200 F greater than the maximum temperature at the rupture location.
Therefore, the peak clad temperature associated with an LEU rod with 0.7 filling ratio due to relocation could be as high as 2174 0 F --- a value with substantially less margin to the 10 CFR § 50.46 limit. Consideration of additional MOX effects, such as a greater filling ratio, could shrink this margin even further.
In addition to limits on PCT, the maximum clad oxidation must also be limited to less than 17%. Since the maximum clad oxidation typically occurs at the ruptured location, as a result of double-sided oxidation, a significant increase in clad temperature at the ruptured location due to relocation would result in a significant increase in maximum clad oxidation, with the potential to exceed the 10 CFR § 50.46 limit. The oxidation rate for M5 is substantially greater at 21-74 0F than at 1841 'F.
2
Q.R-4: In response to Question 98, Duke's witnesses Harvey and Dunn state that the impact of fuel relocation on the increase of the heat source in the ballooned region is mitigated by the lower density of the relocated fuel compared to the original pellet density, so that only very high filling ratios are a cause for concern.
Do you agree with this statement?
A. R-4: No. As the Staff testifies in response to Question 32, the diameter increase in the balloon can be as great as 100%, so the cross-sectional area can increase by a factor of 4. Therefore, if the entire area fills with fuel, the linear heat source would double for a filling ratio of only 50%. For the FR2 observed filling ratio of 0.615, the linear heat source would increase by a factor of 2.5. The significant impact of relocation on PCT and maximum clad oxidation for these filling ratios is demonstrated in Exhibit E attached to my pre-filed testimony.
Q.R-5: In response to Question 40, NRC Staff witness Meyer claims that, according to the results of recent lhigh-burnup integral tests at Argonne National Laboratory ("ANL") "it appears that the small particles or fines are blown out of the burst opening when the rod depressurizes," implying that "there would be few or no small particles in the ballooned region" of the type that could make a difference in the relocated fuel mass in M1OX fuel and LEU fuel. Does this test resolve your concern regarding the potentially more severe impact of fuel relocation with the MOX fuel LTAs?
A.R-5: No. First, the tests in question were performed on BWR fuel rods and not PWR fuel rods. The two types of fuel are sufficiently different that it is difficult to come to any conclusions about the behavior of PWVR fuel rods during a LOCA from experiments on BWR rods. This is why ANL plans to repeat these tests using rods from the H.B.
Robinson PWR. See NRC Staff Proposed Exhibit 3 at 32.
Second, BREDL's claim, confirmed by the Staff in response to Question 39, is that the excess fine particles generated in MOX fuel originates in the rim regions surrounding the plutonium agglomerates, which are distributed throughout the fuel pellet. The Staff provides no argument as to how'fine particles generated near plutonium agglomerates but away from the periphery of the fuel rod would be selectively blown out of the burst opening, leaving only large particles. On the other hand, the rim material in LEU fuel is generated only around the circumference of the fuel pellets (Staff response to Question 39), and thus is adjacent to the clad and the burst opening. The apparent absence of fine particles in the vicinity of the burst opening does not provide evidence that fine particles escape from the fuel rod at circumferential locations away from the burst opening (in the case of LEU fuel) or from locations throughout the entire fuel rod cross-section (in the case of MOX fuel).
In fact, to the extent that the phenomenon observed in this test indicates that fine-grain rim material near the burst opening is blown out of the rod during depressurization of LEU fuel rods, the result only strengthens BREDL's contention that MOX fuel rods 3
Sizs' I contain greater quantities of fine-grain material than LEU fuel and hence may be subject to more severe relocation effects.
Q.R-6: In response to Question 63, Duke's vitnesses assert that the effect on Duke's LOCA analysis of using the LEU decay heat curve instead of the MOX decay heat curve is a conservatism of "tip to 751F on ICT." What is your opinion of this statement?
AR-6: While I agree that this is a conservatism, I would point out that the effect of using the LEU decay heat curve on PCT is considerably smaller than the effect of relocation on PCT, which could be on the order of several hundred Fahrenheit degrees. Both effects should be considered in any LOCA analysis involving MOX fuel, to ensure that any interactions between the two effects are properly accounted for.
Q.R-7: In response to Question 163, Duke's witnesses claim that BREDL is presenting a "chicken and egg" paradox by arguing that LOCA testing of irradiated MOX fuel rods is necessary to support the safety case for the MOX LTA program at Catawba. Is this a fair characterization of BREDL's position?
A.R-7: No, it is not. BREDL has pointed out that the necessary testing can be conducted with MOX fuel irradiated in Europe at European test facilities, of which there are several.
The MOX LTAs that Duke proposes to load at Catawba are not merely incremental modifications of the LEU fuel that is in use at US reactors, but are radically different fuels with completely different microstructures. BREDL believes it would be irresponsible to use the Catawba station as a test reactor for these novel fuel assemblies.
BREDL also notes that a loss-of-coolant accident at Catawba when MOX LTAs are present in the core would likely put an end to the MOX program in the United States.
Thus it would be prudent to ensure that the likelihood of such an occurrence is low.
Q.R-8: Is this your entire rebuttal to the testimony of Duke and the NRC Staff?
A.R-8: No, these are just some initial observations, which I have been able to make in the relatively brief period of time available to prepare a written rebuttal. I am continuing to review their testimony.
4
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of Docket No's. 50-413-OLA, DUKE ENERGY CORPORATION 50-414-OLA (Catawba Nuclear Station, Units 1 and 2)
REBUTTAL TESTIMONY OF DR EDWIN S. LYMAN REGARDING BREDL CONTENTION I Blue Ridge Environmental Defense League ("BREDL") hereby presents the following rebuttal testimony by Dr. Edwin S. Lyman regarding BREDL's Contention I.
Q. R-1: hlave 3'Ou read the testimony submitted by Duke Energy Corporation
("Duke") and by the U.S. Nuclear Regulatory Commission ("NRC") Staff on BREDL's Contention 1?
A.R- 1: Yes, I have.
Q.R-2: Has the testimony of either Duke or of the NRC Staff caused you to reconsider your opinion regarding the validity of BREDL's safety contention I?
A.R-2: No, it hasn't. Contention I is based on the fact that the experimental database on the behavior of M5-clad MOX fuel under design-basis LOCA conditions is inadequate to resolve issues such as the impact of fuel relocation on MOX fuel. Neither the testimony of Duke nor the testimony of the NRC staff offers any experimental evidence to support their claims regarding the ability of the MOX LTAs to comply with 10 CFR § 50.46 criteria in the presence of relocation effects. Instead, Duke offers a single micrograph of an irradiated, reactor-grade MOX fuel pellet of unspecified burnup. Thus the testimony of Duke and the NRC Staff tends to support, rather than refute, BREDL's contention that the LOCA analysis in Duke's LTA LAR lacks adequate experimental support.
In fact, the NRC Staff testimony confirms several of the assertions that BREDL has made in the course of this proceeding. In response to Question 47, the Staff confirms that fuel relocation could cause the cladding temperature in the balloon to increase by several hundreds of degrees F. In response to Questions 38-39, the Staff also confirms the possibility that the amount of fuel that is relocated in MOX rods may be greater than the amount in LEU rods.
Q.R-3: In response to Question 42, NRC Staff witness Meyer states that "if fuel relocation has any effect, it would increase the temperature only in the ballooned
region of the fuel rod ... the ballooned region is seldom the location of the calculated peak cladding temperature when relocation is ignored." Do you agree with this statement? c(j -cig A.R-3: No. According to the results of the FR2 test seri s, fuel relocation has a significant impact on the clad temperature both at the rup ured, ballooned location on the fuel pin and at non-ballooned locations on the fuel pin. his can be seen from Figures 10 and 11 in Duke's testimony in response to Questions 46-47. For low-exposure fuel where relocation does not occur (Figure 10), the maximum temperature at the ballooned location is indeed about 200'F lower than the peak clad temperature, which occurs at a non-ballooned location. However, for high-exposure fuel where relocation does occur, the maximum clad temperature at the ruptured, ballooned location is about 300'F higher than the ruptured, ballooned location for the no-relocation case, and about 20'F lower than the peak clad temperature at the non-ballooned location. Therefore, I would expect from this data that fuel relocation in high-burnup fuel would have a substantial impact on the clad temperature not only at the ruptured location but also at the location where the peak clad temperature occurs.
This conclusion is also consistent with Exhibit C attached to my prefiled testimony, in which NRC states that "the fuel relocation effect on PCT may be significantly larger than that assumed in GS-92 [+46WF]" (Exhibit C, Attachment 4 at 5) and states that fuel relocation may have a +313'F impact on PCT. Exhibit C, Attachment 5 at 4.
While the phenomena involved in these processes are too complicated for simple back-of-the-envelope assessments, some observations can be made. According to the MOX LTA LAR at 3-43, the peak temperature at the hot pin rupture location is 1841 0F. (Note that the Duke testimony in response to Question 154 is erroneous in stating that "the highest PCT at the rupture location in the LOCA calculations for Catawba described in the MOX fuel lead assembly license amendment request was approximately 1750'F ... ").
If the 313°1 increase in clad temperature associated with fuel relocation with a filling ratio of 0.7 is added to this value, the resulting clad temperature at the rupture location is 21540 F. From Figure 11 in Duke's testimony, the PCT in a rod where relocation occurs appears to be about 20'F greater than the maximum temperature at the rupture location.
Therefore, the peak clad temperature associated with an LEU rod with 0.7 filling ratio due to relocation could be as high as 2174WF --- a value with substantially less margin to the 10 CFR § 50.46 limit. Consideration of additional MOX effects, such as a greater filling ratio, could shrink this margin even further.
In addition to limits on PCT, the maximum clad oxidation must also be limited to less than 17%. Since the maximum clad oxidation typically occurs at the ruptured location, as a result of double-sided oxidation, a significant increase in clad temperature at the ruptured location due to relocation would result in a significant increase in maximum clad oxidation, with the potential to exceed the 10 CFR § 50.46 limit. The oxidation rate for M5 is substantially greater at 2+4WF than at 1841'F.
2
Q.R-4: In response to Question 98, Duke's witnesses Harvey and Dunn state that the impact of fuel relocation on the increase of the heat source in the ballooned region is mitigated by the lower density of the relocated fuel compared to the original pellet density, so that only very high filling ratios are a cause for concern.
Do you agree with this statement?
A. R4: No. As the Staff testifies in response to Question 32, the diameter increase in the balloon can be as great as 100%, so the cross-sectional area can increase by a factor of 4. Therefore, if the entire area fills with fuel, the linear heat source would double for a filling ratio of only 50%. For the FR2 observed filling ratio of 0.615, the linear heat source would increase by a factor of 2.5. The significant impact of relocation on PCT and maximum clad oxidation for these filling ratios is demonstrated in Exhibit E attached to my pre-filed testimony.
Q.R-5: In response to Question 40, NRC Staff witness Meyer claims that, according to the results of recent lhighi-burnup integral tests at Argonne National Laboratory ("ANL") "it appears that the small particles or fines are blown out of the burst opening when the rod depressurizes," implying that "there would be few or no small particles in the ballooned region" of the type that could make a difference in the relocated fuel mass in MOX fuel and LEU fuel. Does this test resolve your concern regarding the potentially more severe impact of fuel relocation with the MOX fuel LTAs?
A.R-5: No. First, the tests in question were performed on BWR fuel rods and not PWR fuel rods. The two types of fuel arc sufficiently different that it is difficult to come to any conclusions about the behavior of PWR fuel rods during a LOCA from experiments on BWR rods. This is why ANL plans to repeat these tests using rods from the H.B.
Robinson PWR. See NRC Staff Proposed Exhibit 3 at 32.
Second, BREDL's claim, confirmed by the Staff in response to Question 39, is that the excess fine particles generated in MOX fuel originates in the rim regions surrounding the plutonium agglomerates, which are distributed throughout the fuel pellet. The Staff provides no argument as to how fine particles generated near plutonium agglomerates but away from the periphery of the fuel rod would be selectively blown out of the burst opening, leaving only large particles. On the other hand, the rim material in LEU fuel is generated only around the circumference of the fuel pellets (Staff response to Question 39), and thus is adjacent to the clad and the burst opening. The apparent absence of fine particles in the vicinity of the burst opening does not provide evidence that fine particles escape from the fuel rod at circumferential locations away from the burst opening (in the case of LEU fuel) or from locations throughout the entire fuel rod cross-section (in the case of MOX fuel).
In fact, to the extent that the phenomenon observed in this test indicates that fine-grain rim material near the burst opening is blown out of the rod during depressurization of LEU fuel rods, the result only strengthens BREDL's contention that MOX fuel rods 3
contain greater quantities of fine-grain material than LEU fuel and hence may be subject to more severe relocation effects.
Q.R-6: In response to Question 63, Duke's witnesses assert that the effect on Duke's LOCA analysis of using the LEU decay heat curve instead of the AIOX decay heat curve is a conservatism of "up to 751F on PCT." What is your opinion of this statement?
AR-6: While I agree that this is a conservatism, I would point out that the effect of using the LEU decay heat curve on PCT is considerably smaller than the effect of relocation on PCT, which could be on the order of several hundred Fahrenheit degrees. Both effects should be considered in any LOCA analysis involving MOX fuel, to ensure that any interactions between the two effects are properly accounted for.
Q.R-7: In response to Question 163, Duke's witnesses claim that BREDL is presenting a "chicken and egg" paradox by arguing that LOCA testing of irradiated MOX fuel rods is necessary to support the safety case for the MOX LTA program at Catawba. Is this a fair characterization of BREDL's position?
A.R-7: No, it is not. BREDL has pointed out that the necessary testing can be conducted with MOX fuel irradiated in Europe at European test facilities, of which there are several.
The MOX LTAs that Duke proposes to load at Catawba are not merely incremental modifications of the LEU fuel that is in use at US reactors, but are radically different fuels with completely different microstructures. BREDL believes it would be irresponsible to use the Catawba station as a test reactor for these novel fuel assemblies.
BREDL also notes that a loss-of-coolant accident at Catawba when MOX LTAs are present in the core would likely put an end to the MOX program in the United States.
Thus it would be prudent to ensure that the likelihood of such an occurrence is low.
Q.R-8: Is this your entire rebuttal to the testimony of Duke and the NRC Staff?
A.R-8: No, these are just some initial observations, which I have been able to make in the relatively brief period of time available to prepare a written rebuttal. I am continuing to review their testimony.
4
2276 1 CHAIRPERSON YOUNG: Okay. So continuing 2 on on the new page, do you wish to present your 3 exhibits for admission or offer those?
4 MS. CURRAN: Okay. Dr. Lyman, do you also 5 have before you a list of exhibits to your prefiled 6 written testimony on contention 1?
7 DR. LYMAN: Yes, I do.
8 MS. CURRAN: Are these exhibits A through 9 L?
10 DR. LYMAN: Yes, they are.
11 MS. CURRAN: Are these the exhibits that 12 you attached to your testimony?
13 DR. LYMAN: Yes, they are.
14 MS. CURRAN: I would like to go through 15 and identify each one of these exhibits, if I may.
16 Exhibit A is the curriculum vitae for Dr. Edwin S.
17 Lyman.
18 CHAIRPERSON YOUNG: Just to save a little 19 time, as we go through to help everyone in keeping 20 track, the other ones we know go through 24. So as we 21 go through, I am going to indicate what the official 22 exhibit number will be when it gets marked up here.
23 Exhibit A will then be official exhibit 24 25. Go ahead. And I'll just interrupt in between 25 each one.
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2277 1 MS. CURRAN: Okay. Or I could do that 2 myself if you'd like.
3 CHAIRPERSON YOUNG: Okay. Go ahead. Yes.
4 That would be easier. Yes.
5 MS. CURRAN: Exhibit B, which would be 6 global exhibit 26, is a memorandum from Ralph Meyer, 7 NRC Office of Nuclear Regulatory Research, to John 8 Flack, NRC Regulatory Effectiveness and Human Factors 9 Branch, "Re: Update on Generic Issue 92, Fuel 10 Crumbling During LOCA." It's dated February 8th, 11 2001.
12 Exhibit C, which is global exhibit 27, is 13 a memorandum from Ashook C. Thadani, T-h-a-d-a-n-i, 14 NRC Office of Nuclear Regulatory Research, to Samuel 15 J. Collins, NRC Office of Nuclear Reactor Regulation, 16 regarding "Research Information Letter 0202, Revision 17 of 10 CFR 50.46 and Appendix K." It is dated June 18 20th, 2002.
19 Exhibit D, which is global exhibit 28, 20 consists of slides presented by A. Mailliat and J. C.
21 Melis from IRSN at PHEBUS, S-T-L-O-C meeting -- PHEBUS 22 is spelled P-H-E-B-U-S -- dated October 23rd, 2003.
23 Exhibit E, which is global exhibit 29, is 24 a paper by C. Grandjean, C. Hache, and C. Rongier 25 entitled "High Burnup U02 Fuel LOCA Calculations to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2278 1 Evaluate the Possible Impact of Fuel Relocation After 2 Burst." That paper is dated March 22nd and 23rd, 3 2001.
4 CHAIRPERSON YOUNG: Does that have a cover 5 sheet as well indicating proceedings of the topical 6 meeting?
7 MS. CURRAN: Yes. Would you like me to 8 read all of that?
9 CHAIRPERSON YOUNG: No, no. I just wanted 10 to make sure I had that right.
11 MS. CURRAN: Yes, it does.
12 CHAIRPERSON YOUNG: Okay.
13 MS. CURRAN: Exhibit F, which is global 14 exhibit 30, is a paper by V. Guillard, C. Grandjean, 15 S. Bourdon, and P. Chetelard entitled "Use of 16 C-A-T-H-A-R-E 2," all in caps, "Reactor Calculations 17 to Anticipate Research Needs" from SEGSFM topical 18 meeting on LOCA issues, Argonne National Laboratory, 19 May 25th to 26th, 2004.
20 Exhibit G, which is global exhibit 31, is 21 a paper by A. Mailliat and M. Schwarz entitled "Need 22 for Experimental Programmes on LOCA Issues Using High 23 burnup and MOX Fuels." The word "Programmes" is 24 spelled P-r-o-g-r-a-m-m-e-s. Its number is 25 NUREG/CP-0176, proceedings of the nuclear safety NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE.. N.W.
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2279 1 research conference. It's dated May 2002.
2 Exhibit H, or global exhibit 32, is 3 NUREG/CR-6744 entitled "Phenomenon Identification and 4 Ranking Tables for Loss of Coolant Accidents in 5 Pressurized and Boiling Water Reactors Containing High 6 Burnup Fuel." We are attaching excerpts of this 7 document. The excerpts consist of appendix D, table 8 D-1 at D-67. The document is dated December 2001.
9 Exhibit I, which is global exhibit 33, 10 consists of excerpts from NEA/CSN/R(2003)9, "Ongoing 11 and Planned Fuel Safety Research in NEA Member States, 12 Compiled From SEGFS Members' Contributions" dated 13 October 2002.
14 Exhibit J, which is global exhibit 34, 15 consist of excerpts from NEA/NSC/DOC(2004)8, 16 "International Seminar on Pellet-Clad Interaction in 17 Water Reactor Fuels, Organized by CEA 18 Cadarache/DEN/DEC, in Cooperation with OECD/NEA, IAEA, 19 EDF, Framatome, ANP, COGEMA" dated March 9 to 11, 20 2004.
21 Exhibit K is a letter from Ashook C.
22 Thadani, NRC, to David Modeen, M-o-d-e-e-n, EPRI, 23 E-P-R-I, dated April 21st, 2004.
24 CHAIRPERSON YOUNG: And that will be 25 exhibit 35?
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2280 1 MS. CURRAN: That is exhibit 35.
2 Exhibit L, which is global exhibit 36, is 3 a paper by C. Grandjean, Georges Hache, H-a-c-h-e, 4 entitled "LOCA Issues Relating to Ballooning, Fuel 5 Relocation, Blockage and Coolability: Main Findings 6 from a Review of Past Experimental Programs, Presented 7 at SEGSFM Topic Meeting on Local Issues" -- I think 8 that is probably a typo; it should be "Topical" --
9 dated May 25th through 27th, 2004.
10 CHAIRPERSON YOUNG: All right. Those will 11 be admitted as official exhibits 25 through 36 in a 12 moment, we will have them marked.
13 Do you have any other materials?
14 MS. CURRAN: Well, we may in connection 15 with our surrebuttal but not at the moment.
16 (Whereupon, the witness was excused.)
17 CHAIRPERSON YOUNG: All right. If you 18 could provide your originals to the court reporter?
19 And while those are being marked, we would like to ask 20 you all to discuss the approach that we talked about 21 earlier, the possibility of taking testimony or having 22 cross-examination of actually not just Duke's but any 23 of the witnesses on issues that may not be quite as 24 critical or quite as focused on the main areas of 25 dispute and whether we could use our time this NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2281 1 afternoon to good use or to good effect in that 2 manner. And then when we come back, we will hear from 3 you on that.
4 All right. Then let's break and mark the 5 exhibits.
6 (Whereupon, the foregoing matter went off 7 the record at 2:35 p.m. and went back on 8 the record at 2:45 p.m.)
9 (Whereupon, the aforementioned 10 documents were marked for 11 identification as Exhibits 12 Number 25 through 36, 13 respectively, and were received 14 in evidence.)
15 CHAIRPERSON YOUNG: Let's go to the 16 staff's, and then we'll hear back from all of you on 17 the question that we asked you to think about and make 18 a decision on that.
19 Would you raise your right hands, please?
20 Whereupon, 21 RALPH MEYER, UNDINE SHOOP, and RALPH LANDRY 22 were collectively called as witnesses by counsel for 23 the staff and, having been first duly sworn, were 24 examined and testified as follows:
25 CHAIRPERSON YOUNG: Thank you. Go ahead.
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2282 1 DIRECT EXAMINATION 2 MS. UTTAL: Good afternoon. Could you 3 please each identify yourselves and your jobs in the 4 NRC? Start with Dr. Meyer.
5 DR. MEYER: Okay. I'm Ralph Meyer from 6 NRC Office of Research.
7 MS. UTTAL: Ms. Shoop?
8 MS. SHOOP: I am Undine Shoop from the 9 Office of Nuclear Reactor Regulation. I specialize in 10 fuels.
11 DR. LANDRY: And I'm Ralph Landry from the 12 Office of Nuclear Reactor Regulation.
13 MS. UTTAL: Do you have before you a copy 14 of the document entitled "NRC Staff Testimony of 15 Undine Shoop, Dr. Ralph Landry, and Dr. Ralph Meyer 16 Concerning BREDL Contention 1" dated July 1st, 2004?
17 Do you have that document?
18 MS. SHOOP: Yes.
19 DR. MEYER: Yes.
20 DR. LANDRY: Yes.
21 MS. UTTAL: Do you recognize that 22 document?
23 MS. SHOOP: Yes.
24 DR. LANDRY: Yes.
25 DR. MEYER: Yes.
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2283 1 MS. UTTAL: All right. Is that the 2 testimony that you prepared for this proceeding?
3 MS. SHOOP: Yes.
4 DR. LANDRY: Yes, it is.
5 DR. MEYER: Yes, it is.
6 MS. UTTAL: Was it prepared by you o: r 7 under your direction?
8 MS. SHOOP: Yes.
9 DR. LANDRY: Yes.
10 DR. MEYER: Yes.
11 MS. UTTAL: Have you also prepared 12 statements of your professional qualifications?
13 MS. SHOOP: Yes.
14 DR. LANDRY: Yes.
15 DR. MEYER: Yes.
16 MS. UTTAL: And are they attached to the 17 testimony?
18 MS. SHOOP: Yes.
19 DR. LANDRY: Yes, they are.
20 DR. MEYER: Yes, they are.
21 MS. UTTAL: Dr. Meyer, do you have any 22 corrections, revisions, additions, or deletions you 23 wish to make to your testimony?
24 DR. MEYER: Yes, I have one correction.
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2284 1 says that the balloon was located almost a foot below 2 the location of the peak cladding temperature. That's 3 a mistake. It's a foot above the location of the peak 4 cladding temperature.
5 MS. UTTAL: Any other corrections to the 6 testimony?
7 DR. MEYER: No.
8 MS. UTTAL: Ms. Shoop, do you have any 9 corrections to the testimony?
10 MS. SHOOP: No.
11 MS. UTTAL: Dr. Landry, do you have any 12 corrections to the testimony?
13 DR. LANDRY: Yes, I do. To answer 17 on 14 page 7 of the prepared testimony, approximately midway 15 through the answer, the sentence begins "Duke stated,"
16 "Duke stated in their February 27, 2003 submittal, 17 section 3.7.1.1.2, that Framatome ANP utilized a decay 18 heat curve." That is, strike the words "based on."
19 Insert the words "compared with."
20 MS. UTTAL: Anything else?
21 DR. LANDRY: Then go down to the sentence 22 that reads, "Framatome ANP added conservatism by 23 increasing." Strike that --
24 CHAIRPERSON YOUNG: Page 8.
25 DR. LANDRY: I'm sorry?
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2285 1 CHAIRPERSON YOUNG: Page 8?
2 DR. LANDRY: I'm sorry. Yes.
3 CHAIRPERSON YOUNG: Thanks. Go ahead.
4 Excuse me for interrupting.
5 DR. LANDRY: The sentence that reads, 6 "Framatome ANP added conservatism by increasing the 7 curve by a factor of 1.2." Scratch that entire 8 sentence.
9 MS. UTTAL: Anything else?
10 DR. LANDRY: And in the following 11 sentence, "The resulting decay heat curve bounds the 12 1994 standard," scratch "by a factor of 1.2."
13 And that is all the corrections that I 14 have.
15 MS. UTTAL: Judge, what the staff has done 16 in this case is prepared a new document making these 17 changes. We dated it July 14th. We have copies for 18 everyone.
19 Dr. Meyer, do you have any corrections, 20 revisions, additions, or deletions you wish to make to 21 your statement of professional qualifications?
22 DR. MEYER: Yes. I made some additions to 23 that and one correction. There was a typographical 24 error in the date of my Ph.D. And that was corrected.
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2286 1 remarks about what I had done in the previous and 2 current employment positions and also went ahead and 3 put in my complete publication list.
4 CHAIRPERSON YOUNG: So you have also added 5 your achievements?
6 DR. MEYER: Say it again.
7 CHAIRPERSON YOUNG: Your achievements?
8 DR. MEYER: I can't hear you.
9 CHAIRPERSON YOUNG: I'm sorry.
10 MS. UTTAL: Achievements, I think she 11 said.
12 CHAIRPERSON YOUNG: Achievements. You 13 added a list of achievements.
14 DR. MEYER: Achievements. Yes, I did. I 15 added a section on achievements. Sorry.
16 CHAIRPERSON YOUNG: Thank you.
17 MS. UTTAL: All right. Ms. Shoop, do you 18 have any changes to your professional qualifications?
19 MS. .SHOOP: Yes, I do. Because of changeovers by the year, in the professional affiliation section, I am no longer the past president 22 of North American Young Generation of Nuclear. I am 23 also no longer the chairman of the Nuclear 24 Installation Safety Division Program Committee.
25 MS. UTTAL: Thank you.
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2287 1 CHAIRPERSON YOUNG: Do you have new 2 versions of this one?
3 MS. UTTAL: I have new versions of Dr.
4 Meyer's professional qualifications. I do not have a 5 new version of Ms. Shoop's. It's the last paragraph 6 on her professional qualifications. Apparently there 7 were new elections or a changeover and --
8 CHAIRPERSON YOUNG: I wasn't on that page.
9 Could you tell me again which parts to mark out?
10 MS. SHOOP: Certainly. In the second 11 sentence, where it says, "Currently serving as past 12 president," that is to be deleted.
13 CHAIRPERSON YOUNG: That sentence, okay.
14 MS. SHOOP: Yes because with the 15 elections, I am no longer the past president.
16 CHAIRPERSON YOUNG: Okay.
17 MS. SHOOP: And the last sentence, where 18 it says, "Chairman of the Nuclear Installation Safety 19 Division Program Committee," that is also to be 20 deleted.
21 CHAIRPERSON YOUNG: Thank you.
22 MS. UTTAL: Dr. Landry, do you have any 23 changes to your professional qualifications?
24 DR. LANDRY: No, I do not.
25 MS. UTTAL: Now, with the corrections and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2288 1 revisions, are your written testimony and your 2 attached statements of professional qualifications 3 true and correct, to the best of your information and 4 belief?
5 MS. SHOOP: Yes.
6 DR. LANDRY: Yes, they are.
7 DR. MEYER: Yes, they are.
8 MS. UTTAL: Do you adopt the written 9 testimony as now revised as your sworn testimony in 10 this proceeding?
11 MS. SHOOP: Yes.
12 DR. LANDRY: Yes.
13 DR. MEYER: Yes.
14 CHAIRPERSON YOUNG: Why don't you go ahead 15 to your rebuttal testimony? Actually, when we start 16 with the exhibits, I think we need to see these the 17 CVs off the testimony and make them exhibits, rather 18 than part of the testimony.
19 MS. UTTAL: Okay. Do you also have before 20 you a document entitled "NRC Staff Rebuttal Testimony 21 of Dr. Ralph Landry and Dr. Ralph Meyer Concerning 22 BREDL Contention 1"?
23 MS. SHOOP: Yes.
24 DR. LANDRY: Yes.
25 DR. MEYER: Yes.
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2289 1 MS. UTTAL: These questions are to Dr.
2 Meyer and Dr. Landry. Did you prepare that document 3 presentation in this proceeding?
4 DR. LANDRY: Yes.
5 DR. MEYER: Yes.
6 MS. UTTAL: Was it prepared by you or 7 under your direction?
8 DR. LANDRY: Yes.
9 DR. MEYER: Yes.
10 MS. UTTAL: Do you have any corrections or 11 changes?
12 DR. LANDRY: No.
13 DR. MEYER: No.
14 MS. UTTAL: Is your written testimony, 15 your written rebuttal testimony, true and correct, to 16 the best of your information and belief?
17 DR. LANDRY: Yes.
18 DR. MEYER: Yes.
19 MS. UTTAL: And do you adopt this written 20 rebuttal testimony as your sworn testimony in this 21 proceeding?
22 DR. LANDRY: Yes.
23 DR. MEYER: Yes.
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2290 1 as evidence and bound into the transcript.
2 CHAIRPERSON YOUNG: All right. Then the 3 corrected version of the prefiled direct testimony 4 followed by the prefiled rebuttal testimony of the NRC 5 staff will be inserted and bound into the transcript 6 at this point. And then we'll start a new page with 7 the exhibits.
8 9
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2Ace July 14, 2004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
DUKE ENERGY CORPORATION ) Docket Nos. 50-413-OLA
) 50-414-OLA (Catawba Nuclear Station )
Units 1 and 2) )
NRC STAFF TESTIMONY OF UNDINE SHOOP, DR. RALPH LANDRY AND DR. RALPH 0. MEYER CONCERNING BREDL CONTENTION I Ql. Please state your name, occupation, and employer.
Ala. (US) My name is Undine Shoop. I am employed as a Reactor Systems Engineer in the office of Nuclear Reactor Regulation at the U.S. Nuclear Regulatory Commission (NRC).
A statement of my professional qualifications is attached hereto.
Al b. (RL) My name is Ralph Landry, I am a Senior Reactor Engineer employed by the NRC in the Office of Nuclear Reactor Regulation. A statement of my professional qualifications is attached hereto.
Alc. (ROM) My name is Ralph 0. Meyer. I am employed as a Senior Technical Advisor for Core Performance and Fuel Behavior in the Office of Nuclear Regulatory Research at the NRC. A statement of my professional qualifications is attached.
Q2. Please describe your current responsibilities.
A2a. (US) In my position as a Reactor Systems Engineer at the NRC, I currently serve as the lead fuels reviewer for several projects involving technical evaluation of fuel designs, in-reactor fuel use, and core components. This work includes reviewing new fuel designs, fuel transition methodologies, core component changes (such as control elements), fuel pellet modifications, fuel assembly component changes, and cladding material.
22R#2-A2b. (RL) I am currently assigned responsibility for leadership in the reviews of the thermal hydraulic analysis computer codes. This includes review of the advanced computing methodologies, Appendix K methodologies, advanced nuclear reactorsystem design analyses, and specific Loss-of-Coolant Accident (LOCA) application analyses.
A2c. (ROM) I am responsible for the technical content of all of NRC's research on fuel behavior under conditions of design-basis accidents. This work is currently being performed at three national laboratories and six cooperative international programs.
Q3. Please explain whatyourduties have been in connectionwiththe NRCstaff's (Staff) review of the license amendment request (LAR) filed by Duke Energy Corp. (Duke) for a license amendment to insert fourmixed oxide (MOX) fuel lead test assemblies (LTAs) into the reactorcore at Duke's Catawba Nuclear Station, Units 1 or 2.
A3a. (US) My duties in connection with the Staff's review of the LAR filed by Duke relative to the insertion of MOX fuel LTAs into the core at Catawba have been focused on the fuel rod design features, the fuel assembly design, and the exemptions for using MOX fuel and the M5 cladding material.
A3b. (RL) My duties in connection with the Staff's review of the LAR filed by Duke relative to the insertion of MOX LTAs into the core at Catawba have been focused on the LOCA analysis performed pertaining to the MOX LTAs.
A3c. (ROM) I was not involved in the Staff's review, but I am familiar with the technical issues.
Q4; Are you familiar with Contention I?
A4. Yes. As admitted by the Licensing Board, Contention I reads as follows:
The LAR is inadequate because Duke has failed to account for differences in MOX and LEU fuel behavior (both known differences and recent information on possible differences) and for the impact of such differences on LOCAs and on the DBA analysis forCatawba.
32-
- 05. Do you agree with the assertion in the contention that the Licensing Amendment Request (LAR) is inadequate because of failure to account for differences in MOX and LEU behavior and for the impact that such differences might have on design-basis LOCAs?
A5. No.
- 06. Do you agree that the LAR is inadequate because of failure to account for uncertainties in MOX fuel assembly behavior during LOCAs?
A6. No.
Q7. BREDL has stated that the experimental database for MOX fuel performance during LOCAs is woefully inadequate. Do you agree?
A7. (ROM) No.
- 08. What is the purpose of your testimony?
A8a. (US) The purpose of this testimony is to provide the NRC Staff's views concerning the acceptability of Duke's LAR, which is the subject of Contention I.
A8b. (RL) The purpose of this testimony is to provide the NRC Staff's views concerning the adequacy of Duke's LOCA analysis and the acceptability of the LAR, which is the subject of Contention I.
A8c. (ROM) The purpose of this testimonyis to provide the bases formyanswers to Q5, 6 and 7.
A9. (RL, US) Four.
010. Inaddition to the four LTAs, how many other assemblies will be in the core?
Al 0. (RL, US) There will be 189 other assemblies in the core.
Q11. The contention raises concerns about aspects of fuel behavior during a LOCA.
Could you give us a brief description of fuel behavior during a LOCA?
2-u@LF Al1. (RL, ROM) Pressurized water reactors like Catawba use circulating water to take heat from the fuel, and they generate steam with the hot water. This removal of heat from the fuel keeps the fuel relatively cool in relation to temperatures that would cause fuel damage. For a licensing analysis, it is assumed that a large pipe breaks and the water (i.e., the coolant) starts escaping when the reactor is at full power. This loss of coolant automatically shuts down the reactorbecause the nuclearchain reaction cannot be sustained without thewater, and powerbeing produced by the fuel decays rapidly toverylow levels. Aftersufficient coolant has boiled away from the core region, the fuel cladding begins to heat up because heat is no longer being adequately removed from the cladding surface. During the heatup, the cladding will soften, balloon, and burst because the internal pressure is high. As the cladding continues to heat up beyond the temperature for bursting, the cladding begins to oxidize rapidly. Eventually, cold water is injected into the core by an emergency core cooling system (ECCS) and the fuel cladding is cooled back down. Heat removal systems keep the reactor cool from that time on.
Q1 2. Did the Staff conduct an evaluation of Duke's LOCA analysis?
A12. (RL) Yes. The details of the evaluation are found in the Staff's Safety Evaluation for Proposed Amendments to the Facility Operating License and Technical Specifications to Allow Insertion of Mixed Oxide Fuel Lead Assemblies (SE), sections 2.1.2 and 2.4.1, issued April 5,2004.
(NRC Staff's Proposed Exhibit 1,"Safety Evaluation by the Office of Nuclear Reactor Regulation Renewed Facility Operating License NPF-35 and NPF-52," April 5, 2004).
013. Please summarize the Staff's evaluation of the LOCA analysis.
Al 3. (RL) The Staff's evaluation iscontained in section2.4.1 of theSE. Briefly, the ECCS performance of the Catawba nuclear plant Is contained in the analysis done for the current operating core. That analysis was performed by Westinghouse using theirNRC approved realistic large-break loss-of-coolant accident analysis (LBLOCA) program, WCOBRANTRAC. The analysis of record demonstrates that the Catawba nuclear plant complies with the acceptance criteria
?
delineated in 10 CFR 50.46(b). When Westinghouse performed the analysis-of-record LBLOCA analysis, a series of sensitivity studies were performed. One study concerned the effect of co-resident fuel from another vendor. At the time of the Westinghouse study, Catawba was in transition from Framatome Mark-BW fuel to Westinghouse Robust Fuel Assemblies (RFA) and the core contained fuel from both vendors.
The Staff's review included a review of the analysis-of-record (NRC Staff's Proposed Exhibit 2, Letter from M.S. Tuckman (Duke) to the NRC, "License Amendment to Request, Implementation of Best-Estimate Large Break Loss of Coolant Analysis Methodology," August 10, 2000),,the sensitivity studies, and the MOX LTA LOCA analysis, as discussed in the answers to questions 15-20, below.
Q14. Against what regulatory requirements was the LAR LOCA analysis evaluated?
A14. (RL) The regulatory requirements are provided in 10 CFR § 50.46. That is, an evaluation model must be used which either realistically describes the behavior of the reactor system during a LOCA such that the uncertainty in the calculated results can be estimated, or conforms with the required and acceptable features of 10 CFR 50, Appendix K. Whichever approach to the evaluation model is followed, the results must meet the acceptance criteria stated in 10 CFR 50.46(b). Specifically, the peak cladding temperature must not exceed 2200F, the maximum local oxidation must not exceed 17%, the hydrogen generated must not exceed that which could be produced by oxidation of 1% of the total cladding, the core must remain in a coolable geometry, and the core temperature must remain at an acceptable level foran extended period of time.
The following questions and answers will explain infurtherdetail how the LAR has shown compliance with the requirements of 10 CFR 50.46.
015. What is the Peak Cladding Temperature (PCT)?
Al 5. (RL) The PCT is the highest temperature calculated to occur in the reactors core and is specified by 10 C.F.R. § 50.46 to not exceed 22000 F. Compliance with this criterion, along with the oxidation limit, assures that the cladding will not become embrittled and lose its rod-like geometry during and after a LOCA. The PCT predicted by Westinghouse, for Catawba, is below the acceptance criterion of 22000 F specified in 10 CFR 50.46(b)(1).
Q16. Please describe the LOCA analysis that was performed for the MOX LTAs.
A16. (RL) The LOCA behavior of the proposed Framatome ANP MOX LTAs was evaluated In two ways. First, the analysis of record was shown to still be valid for the Catawba nuclear plant with the MOX LTAs in core. This was done by comparison of the hydraulic behavior of the MOX assembly, noted as Mark-BW/MOX1, with the Mark-BW assembly used in the analysis of record study. Comparison of the hydraulic behavior of the Mark-BW/MOX1 fuel assembly as referenced in the Duke February.27, 2003 submittal with the analysis of record performed by Westinghouse shows that the Mark-BW/MOX1 fuel assemblyis muchcloserin hydraulic behavior, such as pressure drop, to the Westinghouse RFA fuel than is the Mark-BW fuel design. Thus, the effect of the Mark-BW/MOX1 fuel on the performance of the RFA fuel under LBLOCA conditions would be less than the effect of the Mark-BW fuel that was resident at the time of the transition to the RFA fuel. (NRC Staff's Proposed Exhibit 2). Further discussion of the comparisons between the AOR and the MOX LTAs is found in the LAR, section 3.7.1.7., submitted by Duke February27, 2003, and in Duke's November3,2003 response to Staff RAI 14. The November3,2003 response to RAI 14 states that "...the MOX fuel lead assemblies are more similar hydraulically to the RFA fuel than the Mark-BW design fuel...' In addition, the mixed core sensitivities performed for the Westinghouse RFA fuel showed that '...the presence of the Mark-BW fuel assemblies had an insignificant impact on the calculated results."
The second evaluation performed was a LBLOCA analysis of the Framatome ANP Mark-BW/MOX1 fuel itself. Framatome ANP performed that analysis using their NRC approved
10 C.F.R. Part 50, Appendix K-computer code, RELAP5/MOD2-B&W. As a sensitivity study, Framatome ANP also analyzed the Mark-BW/MOX1 assembly loaded with low enriched Uranium fuel pellets rather than the MOX pellets, thus obtaining a comparison for MOX versus LEU when installed in the same non-limiting core location. The results of those studies are that the Appendix K code analyzed MOX peak cladding temperature is 2018 eF, while that of the same fuel assembly design containing LEU fuel is 1981 IF. LAR, Table 3-5.
Q17. How did the Appendix K analysis account for MOX fuel?
Al 7. (RL) The requirements specified in 10 CFR Part 50, Appendix K for an acceptable evaluation model are not dependent upon the content of the fuel pellet except in limited areas:
initial stored energy, fission heat, and decay heat. First, the fuel stored energy, which is a measure of the initial temperature of the fuel, was calculated by Framatome ANP using their NRC approved COPERNIC fuel code which has been modified and approved to include MOX properties. Stored energy will be discussed by Ms Shoop. Second, the rate at which the fuel continues to produce heat after the nuclear reaction has been shut down is determined by the decay heat model. The applicability of the Framatome ANP decay heat model, which provides the amount of heat generated in the fuel after the nuclear chain reaction has been stopped, to the MOX fuel was reviewed by the staff. The requirement of Appendix K, 10 CFR 50, Appendix K 1.A.4, is that the ANSIIANS-5.1-1971 decay heat curve multiplied by 1.2 be used to predict the heat generation by uranium dioxide fuel following cessation of the nuclear chain reaction. Duke stated in their February 27, 2003 submittal, Section 3.7.1.1.2, that Framatome ANP utilized a decay heat curve that is compared with the ANSI/ANS-5.1-1994 standard. (American National Standard for Decay Heat Powerin Light Water Reactors, ANSI/ANS-5.1-1994, American Nuclear Society, 1994.) That standard accounts for the fact that for long exposure times, LEU fuel produces the majority of its energy from the fission of plutonium. Thus, the 1994 standard would be expected to be the more appropriate model to use for the decay heat production of MOX fuel than the 1971 model since the
R MOX fuel is producing heat from the fission of plutonium throughout its lifetime. The resulting decay heat curve bounds the 1994 standard and bounds the 1971 standard that has been multiplied by a factor of 1.2. The staff concluded that the decay heat model used is conservative with respect to the decay heat generated by the MOX fuel as well as being more conservative than the regulation specifies in bounding the Appendix Kspecified 1971 curve multiplied by a factor of 1.2.
018. Does the LOCA calculation include the effect of fission heat?
A18. (RL) Yes, as specified by Appendix Kfission heat is calculated based on the known reactivity (how readily fission takes place, that is,if the rate of fission increases or decreases) of the fuel which is well understood for both LEU and MOX, and the effects on reactivity that occur during a LOCA. Such phenomena as voiding of the reactor cooling water and fuel temperature increase cause the fission reaction to slow down and stop. Each of these factors is assumed to be at its worst value, that is, the fission heat is required to be maximized.
Q19. Did the staff perform independent LOCA calculations for the MOX LTAs?
A19. (RL) The staff did not perform independent LOCA calculations for the MOX LTAs.
The staff normally performs those analyses as part of the review of the vendors' generic evaluation models and methodologies. Our practice is to review the documentation supporting the evaluation models, sample calculations, and perform calculations using the vendors' evaluation models as well as the staff's independent computercodes. Once that has been done, we review the plant-specific analyses that have been performed and submitted by the licensee. Inthe case of the MOX LTAs, we reviewed the evaluation model input description, the assumptions made by the vendor, Framatome ANP, and the results. The input and assumptions were found to be consistent with the staff's review of the generic evaluation model in the NRC staff "Safety Evaluation of Framatome
.9 vvq Technologies Topical Report BAW-10164P, Revision 4, 'RELAP5/MOD2-B&W, An Advanced Computer Transient Analysis,"' April 9, 2002.
Q20. What is your conclusion regarding the LOCA analysis performed in support of the Duke LAR?
A20. (RL) The staff concludes on the basis of its review of the LOCA analyses which form the analysis-of-record, and its review of the analyses in support of the use of MOX LTAs at Catawba in the LAR and supplements, that the analyses have been performed in an acceptable manner, are conservative, and demonstrate compliance with the stated acceptance criteria contained in 10 CFR 50.46.
021. Do you have an opinion regarding the safety of permitting the use of fourMOX LTAs at Catawba Nuclear Station?
A21. (RL) Yes.
022. What is that opinion?
A22. (RL) The information submitted by Duke, including their responses to requests for additional information (RAls) asked by the Staff, demonstrate that there is reasonable assurance that the four MOX LTAs to be installed at the Catawba NuclearStation would not adversely affect the health and safety of the public. That opinion is based on the LOCA analyses that demonstrate that there is reasonable assurancethatthe MOX LTAswould behave in compliance with the NRC's regulations.
Q23. Earlier you stated that you do not agree with the assertion in Contention I that the LAR is inadequate. Please provide the basis for that conclusion.
A23. (RL) Duke has accounted for the effect of MOX on the LOCA through use of an approved method of determining the fuel stored energy at the initiation of the LOCA and also use of a conservative MOX-applicable decay heat model to determine the heat production of the fuel
during the LOCA analysis. Use of these sources of heat ensure a conservative prediction of the behavior of the MOX LTAs in the unlikely event of a LOCA. In addition, Duke has performed specific analyses which provide a direct comparison of the response of MOX versus LEU fueled assemblies at the same core location. This was done by taking the physical design of the MOX-specific fuel assembly and performing the LOCA analysis with the assembly fueled with MOX pellets, and then repeating the analysis with the assembly fueled with LEU pellets. Those analyses were performed using the appropriate properties of each fuel pellet material and demonstrate that there is a difference in predicted PCT of only 370 F for the MOX versus LEU fueled assemblies.
Q24. Could you please describe your review of initial stored energy to ensure that MOX fuel was treated properly?
A24. (US) During the review of the COPERNIC code for MOX fuel, the staff used prediction to measurement comparisons at Linear Heat Generation Rate (LHGR) levels for LOCA stored energycalculations to estimate uncertainty including standard deviation infuel performance codes. The uncertainty is applied to code predictions to obtain a conservative stored energy prediction at a 95/95 tolerance level for LOCA analysis. The staff used the FRAPCON-3.2 code to compare the results from the COPERNIC code for stored energy calculations.
COPERNIC supplies the stored energy and thermal conductivityvalues to the LOCA code; therefore, the MOX specific parameters are used in the analysis and the results account for the differences between uranium and MOX fuel in the stored energy and thermal conductivity calculations.
Q25. Did Framatome consider the clad ballooning properties of M5?
A25. (US) Yes. Framatome developed a model to specifically describe the clad ballooning properties of M5 and submitted it with topical report BAW-1 0227. The topical report was approved in December of 1999 and revised in February 2000. "Revised Safety Evaluation by The Office of
l30(
Nuclear Reactor Regulation, Topical Report BAW-1 0227P, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel, Framatome Cogema Fuels, Inc," Feb. 4, 2000.
Q26. Do you have an opinion regarding the safety of permitting the use of four MOX LTAs at Catawba Nuclear Station?
A26. (US) Yes.
027. What is that opinion?
A27. (US) On the basis of the review of the fuel design, as articulated in the LAR and supplements, the staff concludes that the fuel has been analyzed in an acceptable manner with sufficient conservatism to demonstrate compliance with General Design Criteria 10 and that there is reasonable assurance that public health and safety will be protected.
Q28. Earlier you stated that you do not agree with the assertion in Contention I that the LAR is inadequate. Please provide the basis for your conclusion.
A28. (US) The parameters provided by the fuel performance code to the LOCA analysis in the LAR are MOX specific parameters, so the differences introduced from the use of MOX on the LOCA calculation are accounted for in the analysis.
029. BREDL has alleged that the LAR is based on out-of-date information relating to (a) the M5 Cladding, (b)the behavior of MOX fuel and (c)the interaction between the fuel and the cladding. Do you agree?
A29. (US, ROM) No.
030. Please provide the bases for this conclusion.
A30. (US, ROM) (a)The review of M5 cladding was done several years ago. A lot of work on niobium-bearing cladding alloys like M5 has been done since then and the NRC has been involved in that work. Although the Staff and the scientific community know more than they did then, nothing that has been learned invalidates the Staff's overall conclusion in 2000 that M5 is an
acceptable cladding material. (b)The Staff used current data on MOX fuel for the review of the LAR. (c) Dr. Meyer will discuss the interaction between the fuel and the cladding.
Q31. One of the issues raised byBREDLconcemed ballooned rods during a LOCA. Can you describe the ballooning and rupturing process?
A31. (ROM) Fuel rods are pressurized with helium during fabrication and further with fission gas during operation, and at normal full-power operation the internal rod pressure is high.
Nearthe end of the fuel lifetime, the pressure on the inside is roughly balanced by the pressure on the outside (i.e., the reactor system pressure). When the coolant is lost in a LOCA, the system pressure is also lost, so the resulting large fuel rod pressure differential tries to expand the cladding. When the cladding temperature reaches 600 to 700C (approximately 1100 to 1300F),
cladding expansion becomes rapid, and around 800C (approximately 5OOF) the fuel rod bursts just like a rubber balloon would burst. The fuel rod's internal pressure isthen lost, and the deformation or ballooning ceases.
032. How large a diameter do these balloons have?
A32. (ROM) The diameter increase can be as big as 100%, but the balloons are usually smaller. Their size will depend on the open volume within the fuel rod, the amount of gas within the fuel rod, and the temperature at which the rupture occurs.
033. Will the diameter of the balloon be different for MOX fuel and LEU fuel?
A33. (ROM) No.
034. Please provide the basis for your answer.
A34. (ROM) During the last 18 months, researchers at Argonne National Laboratory completed the first four ballooning tests ever performed with actual high-bumup fuel rods (NRC Staff's Proposed Exhibit 3, Y. Yan et al., "LOCA Test Results for High-Burnup BWR Fuel and Cladding," Organization for Economic Cooperation and Development (OECD) Topical Meeting on LOCA Issues, May 25-26, 2004, p. 17). Ineach case, the size of the balloon was the same as the
size of the balloon in the control test with an unirradiated specimen. In my opinion, the high cladding temperature during the LOCAtransientloosens the pellet-to-cladding bond that develops at high burnup just like hot water loosens the cap on a jelly jar, and there is no effect of fuel-to-cladding bonding on ballooning. Therefore, the size of the balloon is determined by gas quantity, volume and temperature as described earlier. There would be no difference in ballooning between MOX fuel and LEU fuel.
Q35. Will the diameter of the balloon be larger for M5 cladding than for Zircaloy cladding?
A35. (ROM) No, but it is important to use the appropriate data in reaching a conclusion.
Creep tests, for example, are conducted with pressurized tubes that are held at a constant temperature fora long time (10-1000 sec) until the tubes burst. During a LOCA, on the other hand, fuel rods would be experience a temperature that is rapidly increasing at the rate of about 5 C/sec at the time of bursting. These different temperature conditions can have a significant effect on the deformation process. Therefore, rod burst tests for LOCA applications are always done with increasing temperatures that are typical of a LOCA. N.Waeckel presented such test data for M5 and Zircaloy at a recent conference (NRC Staff's Proposed Exhibit 4, N.Waeckel et al., "Does M5 Balloon More that Zircaloy-4 Under LOCA Conditions?," OECD Topical Meeting on LOCA Issues, May 25-26, 2004, p. 10), and those data show that M5 does not develop larger balloons than Zircaloy under LOCA conditions.
036. After ballooning occurs during a LOCA, can fuel material move around inside the fuel rod?
A36. (ROM) Yes. Fuel pellets, which are about the size of little marshmallows, develop cracks during normal operation such that they can easily break apart. Tests have shown that broken pieces of fuel pellets can move down into the ballooned region of the cladding.
Q37. What, if any, is the significance of this fuel movement or relocation?
A37. (ROM) If extra fuel particles or fragments move into the ballooned region, they will increase the mass of fuel in that region and thereby increase the heat generated in that region.
The increased heat generation would increase the cladding temperature in the balloon and thus increase the amount of cladding oxidation, which causes embrittlement.
038. Would there be any difference in the amount of fuel relocation for MOX fuel and LEU fuel?
A38. (ROM) It is possible, but in my opinion it would not matter. If the MOX fuel fragments were smaller than LEU fuel fragments, then you might be able to get more MOX fuel into the balloon.
Q39. Why might the fuel fragments be smaller?
A39. (ROM) There is probably a little more rim material in MOX fuel than in LEU fuel with the same bumup. This rim material, which forms inhigh-bumup regions around the circumference of LEU fuel and also around the agglomerates in MOX fuel, is the result of fission gas migration within the uranium-plutonium oxide crystalline grains. Fission gas migrates, coalesces, and precipitates insmall bubbles, which attach themselves to the grain boundaries. As the number of bubbles increases with bumup, the grain boundaries subdivide to form more surface area to accommodate the bubbles, thus producing the smaller grained rim material. Because fission gas release, which is also related to the migration process, is a little higher inMOX fuel than in LEU fuel (e.g., 5%in MOX and 4% in LEU), the volume of rim material might be roughly 25% greater in MOX fuel than in LEU fuel. On the other hand, MOX fuel has a little more plasticity than LEU fuel, so I would expect fewer of the larger fragments in MOX fuel.
Q40. You said that in your opinion, the difference in the amount of relocated fuel between MOX and LEU fuel would not make any difference. Why is that?
A40. (ROM) In recent high-burnup integral tests in our program at ANL, we have observed a black deposit on the quartz tube of the apparatus just opposite the burst opening.
! O° Large fuel fragments are also visible through the burst opening, and these particles have no small particles or fines around them (NRC Staff's Proposed Exhibit 3, Y. Yan et al., OECD Topical Meeting on LOCA Issues, May 25-26, 2004, p. 17). It thus appears that the small particles orfines are blown out of the burst opening when the rod depressurizes. Thus, there would be few or no small particles in the ballooned region, and it is these small particles that have been postulated to make a difference between the mass of fuel in the balloon in MOX fuel and LEU fuel.
041. For a given amount of fuel relocation, would the heat source in the balloon be greater for MOX fuel than for LEU fuel?
A41. (ROM) No. For the Catawba plant, the peak cladding temperature occurs a couple of minutes after the loss of coolant has shut down the power. By that time, most of the stored heat in the fuel has been dissipated and the chemical heat from the metal-water reaction is small, so the heat source is dominated by decay heat. Decay heat for MOX fuel is lower than it is for LEU Fuel; therefore, the heat source in the balloon for MOX fuel would be less than it would be for LEU fuel.
Q42. In his ACRS presentation of May 6, 2004, Dr. Lyman used an average figure of 105C (190F) increase in peak cladding temperature and added that to the reported peak cladding temperature of 201 8F for the MOX fuel in Catawba. He then concluded that this would bring the MOX peak cladding temperature well over the regulatory limit of 2200F. Do you agree with that conclusion?
A42. (ROM) No. If fuel relocation has any effect, it would increase the temperature only in the ballooned region of the fuel rod. Because of the largersurface area of the ballooned region, its cooling is enhanced and the ballooned region is seldom the location of the calculated peak cladding temperature when relocation is ignored. For the MOX fuel in Catawba, Duke reported a maximum cladding temperature in the balloon of only 1841 F, and the balloon was located almost a foot above the location of the peak cladding temperature of the fuel rod. If you add 190F -- the
number used by Dr. Lyman -- to the maximum cladding temperature in the balloon, you get2031 F, which is just 13F over the peak cladding temperature of 201 8F reported by Duke. This is a small increase in peak cladding temperature, which would still be well below the allowable temperature of2200F.
Q43. Was fuel relocation taken into account in the LOCA analysis for the MOX lead test assemblies?
A43. (ROM) There was no specific accounting for fuel relocation in the LOCA analysis submitted by the applicant, since it is not required by Appendix K.
Q44. How long are the ballooned regions in relation to the fuel rod length?
A44. (ROM) Fuel rods are 12-feet long. Because of local temperature variations, a localized bulge or balloon develops in a fuel rod under LOCA conditions and then ruptures. These ruptured balloons are only a few inches long.
Q45. Are the ballooned regions of the fuel rod treated in a special way in a safety analysis?
A45. (ROM) Yes. Appendix K requires that the inside of the cladding must also be considered to oxidize over a 3-inch length on the assumption that steam will enter the ruptured opening of the balloon and react with the inside of the cladding. Of course, the outside of the cladding is considered to oxidize over its full length. In the balloon, the thickness of the cladding is also reduced for the oxidation calculation, so the maximum oxidation that must be compared with the 17% limit is almost always in the balloon, as it is for the MOX fuel in Catawba. The bottom line is that only about 3 inches of the 12-foot fuel rod are threatened by embrittlement. Protecting these most vulnerable three inches of the limiting fuel rod ensures coolable geometry following a LOCA.
046. Earlier, you said that you did not agree that the database for MOX fuel performance during LOCAs is woefully inadequate. What is the basis for that statement?
A46. (ROM) As can be seen from the above testimony, fuel performance during LOCAs is almost entirely controlled by cladding behavior, which is unaffected by fuel type. The database for cladding behavior is the same for MOX and LEU fuel, and it is substantial. The only important fuel property that is different for MOX fuel and thus affects LOCA performance is the fuel thermal conductivity. Many measurements of MOX fuel thermal conductivity have been made during the past 35 years because of extensive research forbreeder reactors as well as forMOX utilization in LWRs. However, the most critical measurements in relation to fuel temperatures and LOCA behavior are direct measurements of centerline temperature in instrumented fuel rods taken from commercial reactors. To this end, there have been 7 MOX instrumented fuel assemblies in the Halden test reactor, some of which tested fuel that operated to 65 GWd/t bumup (and some are continuing to test fuel above that burnup level). Each test assembly has produced hundreds of data points on fuel temperature and rod pressure over a long period of time. These results have provided an adequate database for validation of fuel rod codes for application to MOX fuel (NRC Staff's Proposed Exhibit 5, Memorandum from Farouk Eltawila, RES, to Suzanne C. Black, NRR, RE: Response to User Need for Development of Radiological Source Terms for Review of Mixed Oxide Fuel Lead Test Assemblies, February 23, 2004, Attachment B, Figure 1). Although additional data are being taken, especially at higher bu mups, it is incorrect to say that the database for MOX fuel performance during LOCAs is inadequate.
047. Would you summarize your testimony and give us your perspective on the possible effects of fuel relocation in MOX fuel versus LEU fuel?
A47. (ROM) NRC is treating fuel relocation during a LOCA as a significant issue and we are investigating it actively in our research programs. If relocation is found to occur during that relatively small window of time between the bursting of the balloon and the rapid cooling (quenching) of the core, and if the average densities in the balloons are found to be relatively high, then the cladding temperature in the balloon might be increased by several hundreds of degrees
2jgo2' F. However, the peak cladding temperature of a fuel rod does not usually occur in the balloon, so the actual increase in the peak cladding temperature would probably be much smaller. Further, we should not lose sight of the objective of the LOCA analysis: to preserve long-term cooling of the core following a LOCA. Our regulatory logic is that we would be able to continue cooling the core if the fuel pellets stay within the cladding, and that we could accomplish this even if the cladding is bent or has holes in it as long as the cladding does not embrittle and fragment. Even if the balloon were to exceed the embrittlement criteria, the remainderof the 12-foot fuel rod would have lots of ductility because the embrittlement in the balloon is twice that in other comparable locations due to the doubling of the oxidation (inside oxidation plus outside oxidation). Realistically, the embrittlement and fragmentation of 3 inches of cladding out of 12 feet would not result in a core melt, and without core melt there could be no large releases and no major consequences. So in my opinion, the regulatory approach is conservative. The assumption that the relocation effect would be more severe for MOX fuel than for LEU fuel is speculative. Hypotheses have been put forth by Dr. Lyman that the balloons would be bigger, that the density of relocated fuel would be higher, and that the heat source for a given amount of fuel would be greater during a LOCA for MOX fuel than it would be for LEU fuel. Ibelieve, based on the data discussed above and on my experience, training and knowledge, that those hypotheses are incorrect. I do not see any reason to believe, that potential relocation effects would be worse for MOX fuel than for LEU fuel.
Q48. What is your opinion regarding the safety of using four MOX LTAs at Catawba Nuclear Station?
A48. (ROM). Ibelieve that the use of fourMOX LTAs does not adversely affect the health and safety of the public.
049. Please provide a basis for your answer.
A49. (ROM) NRC has an aggressive research program that probes into potential
weaknesses inreactorsafety analyses and licensing procedures. We are investigating fuel-related issues associated with several accidents, including LOCA, and these issues are well known from documents that we have issued. The question is not whether we have issues, but, rather, the question is whether MOX fuel exacerbates these issues compared with LEU fuel. Based upon my experience, knowledge and training and the testimony I have given above, it is my opinion that these issues are not exacerbated by the use of MOX. Therefore, it is my opinion that the use of MOX LTAs will not have a deleterious effect on public health and safety.
Q50. What is your conclusion regarding Duke's LAR?
A50. Based upon the staff's evaluation and the testimony provided above, the Staff concludes that there is reasonable assurance that the public health and safety will be protected if the amendment is granted.
Q51. Does this conclude your testimony?
A51. Yes.
July 8, 2004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )
)
DUKE ENERGY CORPORATION ) Docket Nos. 50-413-OLA
) 50-414-OLA
)
(Catawba Nuclear Station )
Units 1 and 2) )
NRC STAFF REBUTTAL TESTIMONY OF DR. RALPH LANDRY AND DR. RALPH 0. MEYER CONCERNING BREDL CONTENTION I W1. Please state your name.
Al a. (RL) My name is Ralph Landry. I am a Senior Reactor Engineer employed by the NRC in the Office of Nuclear Reactor Regulation. A statement of my professional qualifications was attached to the NRC Staff Testimony of Undine Shoop, Dr. Ralph Landry and Dr. Ralph 0. Meyer Concerning Contention I, (Staff Prefiled Testimony),' filed July 1, 2004.
Alb. (ROM) My name is Ralph O. Meyer. I am employed as a SeniorTechnical Advisor for Core Performance and Fuel Behavior inthe Office of Nuclear Regulatory Research at the NRC. A statement of my professional qualifications was attached to the Staff Prefiled Testimony, filed July 1, 2004.
' The NRC Staff submitted several proposed exhibits with the Staff Prefiled Testimony.
The Staff intends to introduce only the following portions of the documents into evidence: Exhibit 3, "LOCA Test Results for High-Bumup BWR Fuel and Cladding," Y. Yan, et al., pages 1, 17 (unpaginated); Exhibit 4, "Does M5 Balloon More Than Zircaloy-4 Under LOCA Conditions?," N.
Waeckel, et al., pages 1-2, 10; Exhibit 5, Memorandum from F. Eltawila to S. Black, "Response to User Need for Development of Radiological Source Terms for Review of Mixed Oxide Fuel Lead Test Assemblies," February 23, 2004, pages 1-2 and Attachment B, page B-5, Figure 1.
Q2. What is the purpose of this testimony?
A2. The purpose of this rebuttal testimony is to address the Prefiled Written Testimony of Dr. Edwin Lyman Regarding Contention I,submitted on behalf of the Blue Ridge Environmental Defense League (BREDL) on July 1, 2004.
03 In the answer to Question 5 of his prefiled written testimony, Dr. Lyman states that "Appendix K to Part 50...sets forth ECCS 'evaluation models', i.e. assumptions about the behavior of reactor fuel that are to be used in determining whether it complies with ihe criteria in 10 C.F.R. § 50.46." Is this statement correct?
A3. (RL) No, this is not a correct statement. Appendix K to 10 C.F.R. Part 50 provides the descriptions of the required and acceptable features of the evaluation models as well as the required documentation. Appendix K does not provide assumptions about the behavior of reactor fuel. The topics covered include the sources of heat during the LOCA, the swelling and rupture of the cladding and fuel rod thermal parameters, the blowdown phenomena, and the post-blowdown phenomena. These are all descriptions and specifications placed on phenomena that must be modeled by an acceptable evaluation model. The majority of the specifications provided by Appendix K deal with thermal hydraulic phenomena, that is, heat transfer and fluid flow behavior.
More specifically, only the first few paragraphs of Appendix K address matters related to fuel, such as how the decay heat is to be calculated, how stored energy is to be calculated, and how the heat from the reaction of the cladding material with the cooling water, or steam, is to be calculated. The remainderof the appendix gives specific details and requirements on how the heat removal by the coolant water is to be calculated, and how the movement of the coolant water through the reactor system is to be calculated. Thus, the majority of Appendix K provides details on how cooling of the reactor fuel is to be calculated, along with details on how the movement of the cooling water around the system is to be calculated. Only the early part of the.appendix is
concerned with specifying how the amount of heat contained and produced by the fuel are to be calculated.
Q4 In his answerto Question 18 of his prefiled written testimony, Dr. Lyman states that
"...the Staff claims to have independently verified the adequacy of Duke's LOCA analysis..." Is this statement correct?
A4. (RL) No, this is not a correct statement. The staff does not claim to have performed independent analyses or calculations to verify the submittal of Duke. The staff has stated in its Safety Evaluation (SE): "Based on the NRC review of the information provided, the NRC staff concludes that the effect of four MOX LTAs has been conservatively evaluated and has been demonstrated to be in compliance with the requirements of 10 C.F.R. 50.46." That statement was based on the material submitted by Duke that was reviewed and found to be consistent with the approval the staff has granted with regard to the fuel vendor's LOCA evaluation model and effect on the LOCA analysis of record for the Catawba Nuclear Plant.
Q5. In the answer to Question 6 of his prefiled written testimony, Dr. Lyman explains why he thinks the Appendix-K evaluation models should include consideration of fuel relocation during LOCAs. He bases his explanation partly on your memorandum of February 8, 2001, and on Mr. Thadani's memorandum of June 20, 2002. Do you agree with his response?
A5. (ROM) No. Appendix K has been in effect since 1974, and over the years some extra conservatisms and some non-conservatisms have been identified. To the best of my knowledge, the NRC never contemplated including fuel relocation in Appendix K as mentioned by Dr. Lyman. However, by using Appendix Kwith its compensating extra conservatisms, Duke has adequately accounted for any non-conservatisms. Furthermore, based on my experience and knowledge, and as is demonstrated in my testimony, I do not agree that certain characteristics of MOX fuel exacerbate the effects of fuel relocation.
- 06. Inthe answer to Question 12 of his prefiled written testimony, Dr. Lyman states that "Tight bonding has also been observed at the Halden reactor in Norway to retard the rate of balloon formation." Is that statement correct?
A6. (ROM) No. Dr. Lyman accurately interprets a statement made in NEA/CSNI/R(2003)9, but the statement in that report is not correct. No ballooning tests with high bumup fuel rods have been performed at Halden as of this date, and the statement in NEA/CSNI/R(2003)9 was merely a suggestion of what might happen rather than a report of what has been experimentally observed. I have verified that this statement is in error by an e-mail exchange with Dr. Wolfgang Wiesenack, who is the general manager of the Halden Project (Exhibit A, Wolfgang Wiesenack e-mail to Miroslav Hrehor, "Re: Statement in one of ourSEGFSM Reports," 6/14/04). Miroslav Hrehor, who is also mentioned in that e-mail, isthe scientific secretary at NEA who is responsible for that report.
- 07. Inthe answer to Question 12 of Dr. Lyman's prefiled written testimony, he states that "Ithas been confirmed that MOX fuel is more resistant to clad failures due to pellet-clad mechanical interaction (PCMI) than LEU fuel, even at high burnups." Is that observation relevant to the behavior of MOX fuel under LOCA conditions?
A7. (ROM) No. First, there is no PCMI during a LOCA. PCMI occurs when the power is increased and thermal expansion of the pellet, which is greater than that of the cladding, causes the pellet to push on the cladding. During a LOCA, power is decreased and the cladding expands faster than the pellet -- actually moving away from the pellet. More fundamentally, though, the additional resistance of MOX fuel to cladding failure by PCMI is the result of the greater plasticity of the MOX pellets. They are softer than LEU pellets. The MOX fuel pellets are thus able to deform somewhat and relax the stress they apply to the cladding. This has nothing to do with bonding between the pellets and the cladding.
Q8. In the answer to Question 12 of Dr. Lyman's pref iled written testimony, he claims that LOCA test results from the Power Burst Facility (PBF) have shown that irradiated rods experience greater deformation (swelling) than unirradiated rods. He then states that there is no way to determine whether Duke's LOCA analysis underestimates oroverestimates the degree of cladding swelling. Do you agree with this conclusion?
A8. (ROM) No. As discussed by the IRSN authors in the reference cited by Dr. Lyman (Mailliat and Schwartz at 432), it would appearthat the PBF tests showed an enhancement of more than a factor of 2 in balloon size for irradiated rods compared with unirradiated rods. This was said to be the result of more uniform temperatures in the irradiated rods. This is probably not an accurate interpretation of the test results. After the PBF tests were performed, more work was done on the effects of temperature uniformity at the Karlsruhe nuclear research centerin Germany by F.J. Erbacher and coworkers. In a review article by Erbacher, the PBF results are discussed along with additional German test results with irradiated and unirradiated rods (Exhibit B, F. J. Erbacher, "Cladding Tube Deformation and Core Emergency Cooling in Loss of Coolant Accident of a Pressurized Water Reactor," NuclearEngineering and Design, 1987, pp. 55-64). In Figure 2 of this paper, the PBF data are plotted along with other data, and the temperature at the time of bursting for the PBF data is seen to vary from about 800 to 1100 C. Also shown in this figure are curves that show the strong variation of balloon size associated with the temperature at the time of bursting. This well known variation in balloon size is the result of changes in the crystal structure of the cladding, which switches from an alpha-phase to a beta-phase between 800 and 1000 C. The rod that produced the largest balloon ruptured in the alpha phase just below 800 C, right at a temperature where balloon sizes are expected to be their maximum. Thus, it was probably the differences in temperature at the time of rupture of the PBF rods that produced the different balloon sizes ratherthan the difference in bumup. In Figure 5 of Erbacher's paper, results are shown forasubstantial numberof ballooning testin the same facility. Nosystematicdifference
is seen between fresh fuel and the irradiated rods. High-bumup effects are being studied in the ongoing NRC research program to furtherclarify LOCA behavior, but itcan be noted again that the variation in balloon sizes just discussed is not related to the use of MOX fuel pellets.
Q9. Inthe answer to Question 14 of Dr. Lyman's prefiled written testimony, he concludes that the recent Electricit: de France (EDF) presentation at Argonne National Laboratory does not fully address the differences in the size of balloons between M5 and Zircaloy cladding. Do you agree with that statement?
A9. (ROM) I agree that the EDF presentation does not entirely address the differences in the size of balloons between M5 and Zircaloy, but it clearly shows that the large difference claimed by IRSN is a consequence of using inappropriate data. Further, Dr. Lyman's comment about spalling, or flaking of a thick oxide coating, is not relevant. To the best of my knowledge, none of the ballooning tests utilized cladding with spalled oxide, and certainly no spalling is expected in the Catawba core with its modem cladding materials (ZIRLO and M5). With regard to the size of the balloons, it should not be forgotten that ballooning is an M5 cladding issue; it is not a MOX issue. Based on my knowledge and experience, and the testimony I have given, there is no valid reason to expect that the size of the balloons will be affected by the type of fuel inside.
Although confirmatory research on M5 cladding under LOCA conditions is continuing, it is my opinion that the specific concerns raised by Dr. Lyman are not valid. The staff believes the ballooning size has been adequately accounted for in the analysis.
Q10. In the answer to Question 15 of Dr. Lyman's prefiled written testimony, he quotes an IRSN presentation as saying, 'The impact of fuel relocation in fuel rod balloons ... is still fully questionable and should be addressed by specific analytical tests with a simulation of fuel relocation." Do you agree?
A10. (ROM) No. As discussed in my answer to Questions 33 and 34 of the Staff's Pref iled Testimony, the diameter of the balloons will not be different for MOX fuel and LEU fuel.
Further, the diameter of the balloons will not be affected by fuel relocation because fuel relocation would occur afterthe balloons are formed. Thus neitherMOX fuel norfuel relocation will affect flow blockage, which Is calculated by the models used by Duke. Therefore, the Duke analysis is not incomplete and is not likely to be non-conservative.
Q11. Inthe answer to Question 16 of Dr. Lyman's prefiled written testimony, he points out that Duke's calculations have demonstrated a peak cladding temperature (PCT) that is higher for a MOX fuel rod than for an equivalent LEU rod in the same position in the core. He then concludes that the margin is therefore smaller for a MOX rod than for an LEU rod in the same position. Do you agree with that conclusion?
Al1. (ROM) No. Duke's analysis shows a higher PCT for MOX fuel only because they used the same decay heat curve for MOX and for LEU fuel. In fact, the decay heat is lower for MOX fuel than for LEU fuel at the time of importance for LOCA (Exhibit C, 'Decay Heat Power in Light Water Reactors,' ANS standard ANS/ANS-5.1-1994, pp.1, 2, 14, 16). Therefore, in reality, the PCT for MOX fuel should be a little lower than the PCT for LEU fuel and the margin will not be reduced.
Q12. In the answerto Question 17 of Dr. Lyman's prefiled written testimony, he states that the onlyway to fully address the uncertainties associated with the behaviorof high-burnup, M5-clad MOX fuel during LOCAs is to conduct integral LOCA tests of such fuel. Do you agree with that statement?
A12. (ROM) No, I do not agree that integral LOCA tests of high-bumup MOX fuel with M5 cladding are needed. The effect of plutonium on LOCA behavior is almost entirely the result of small changes in initial stored energy, fission heat, and decay heat as discussed in the Staff's prefiled testimony in answer to Questions 17, 18, and 24. These changes are well known and
adequately modeled. There has been speculation that MOX fuel would enhance the effect of fuel relocation into balloons during a LOCA, but it is my opinion that there will be no such enhancement (see answer to Question 40 in the Staff's prefiled testimony). It has also been claimed that cladding behavior will be altered by MOX fuel in comparison with LEU fuel, but I have offered testimony that shows there will be no effect of the type of fuel pellets inside on cladding behavior (see answer to Question 34 in the Staff's prefiled testimony). In my opinion, the uncertainties associated with the MOX LTAs for Catawba are adequately understood.
Q. Does this conclude your REBUTTAL testimony?
A. Yes.
2318 1 CHAIRPERSON YOUNG: Let me just make sure 2 you have everything you need, Ms. court reporter.
3 MS. UTTAL: Your Honor, regarding the 4 staff's professional qualification, we would ask that 5 it be bound to the testimony because it essentially is 6 part of the testimony, their qualifications.
7 CHAIRPERSON YOUNG: Well, is there any 8 problem with making it an exhibit? It's not in 9 question and answer form like a transcript ordinarily 10 is. And so our ordinary procedure is to include 11 actual testimony in the transcript. And everything 12 else would be exhibits to the transcript.
13 MS. UTTAL: Okay.
14 CHAIRPERSON YOUNG: It's still in the 15 record either way.
16 MS. UTTAL: Okay. That's acceptable, Your 17 Honor.
18 CHAIRPERSON YOUNG: That's actually why we 19 ask Duke to bring the copies of its figures because 20 those normally wouldn't be in the testimony per se.
21 But since we're there, we're not asking you to take 22 them out. We're just wanting to make sure that they 23 also get --.
24 MR. REPKA: Right. I would note that our 25 professional qualification statements are included NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com
2319 1 with the testimony as attachments.
2 CHAIRPERSON YOUNG: They were?
3 MR. REPKA: Not as exhibits.
4 CHAIRPERSON YOUNG: We probably should 5 have done those as exhibits also. We can put those in 6 as exhibits after the staff's. And they will be in 7 the record. I did consult about this prior to make 8 sure I knew that we were doing it the right way. So 9 I think --
10 MR. REPKA: And we may be making it harder 11 than we need to but whatever the Board would like.
12 CHAIRPERSON YOUNG: Yes. Well, let's make 13 them exhibits. And if they somehow get bound into the 14 transcript also, then they will be there twice. And 15 then after the staff's exhibits, we will put Duke's as 16 whatever the next numbers are.
17 So do you want to put a collective, all 18 three CVs as a collective, exhibit?
19 MS. UTTAL: That would be fine.
20 CHAIRPERSON YOUNG: Okay. So that would 21 be exhibit 37.
22 MS. UTTAL: Exhibit 37 is the professional 23 qualifications of Undine Shoop as corrected on the 24 record, professional qualifications of Dr. Ralph 25 Landry, and the updated professional qualifications of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com
2320 1 Dr. Ralph Meyer.
2 The next exhibit, which would be number --
3 CHAIRPERSON YOUNG: Thirty-eight.
4 MS. UTTAL: -- 38, is the safety 5 evaluation for the proposed amendments to the facility 6 operating license and technical specifications to 7 allow insertion of mixed oxide fuel lead assemblies, 8 tax number MB-7863, MB-7864, MC-0824, and MC-0825, 9 documents dated April 5th, 2004.
10 Next exhibit, exhibit 39, --
11 CHAIRPERSON YOUNG: Yes.
12 MS. UTTAL: -- is a letter from M. S.
13 Tuckman, Duke Power, addressed to the'NRC, "
Subject:
14 Duke Energy Corporation Catawba Nuclear Station Units 15 1 and 2, Docket Numbers 50-413 and 50-414, and 16 "McGuire Nuclear Station Units 1 and 2, Docket Numbers 17 50-369 and 50-370, Implementation of Best Estimate 18 Large Break LOCA Methodology," document dated April 19 10th, 2000.
20 The next exhibit, which would be exhibit 21 40, is two pages: the cover page and one page, the 22 document entitled "LOCA Test Results for High Burnup 23 BWR Fuel and Cladding" authored by Y. Yan and several 24 other people of the Energy Technology Division, the 25 Radiation Performance Section, Argonne National NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2321 1 Laboratory. It is undated.
2 The next exhibit, which would be exhibit 3 -
4 CHAIRPERSON YOUNG: Forty-one.
5 MS. UTTAL: -- 41, is entitled "Does M-5 6 Balloon More than Zircaloy-4 Under LOCA Conditions?"
7 authored by Nicholas Waeckel of EDF and several other 8 people. It is dated May 27, 2004 and consists of a 9 cover page and pages 2 and 10 from a slide 10 presentation.
11 CHAIRPERSON YOUNG: Two through ten you 12 mean?
13 MS. UTTAL: No. Two and ten.
14 CHAIRPERSON YOUNG: I have more pages.
15 JUDGE BARATTA: Yes. We have two through 16 ten.
17 CHAIRPERSON YOUNG: Is this one of the 18- ones you're taking the one page out of?
19 MS. UTTAL: The one immediately prior to 20 it, 40 and 41, are both ones that I have redacted 21 pursuant to my --
22 CHAIRPERSON YOUNG: So you are going to 23 provide us with substitutions for those?
24 MS. UTTAL: Yes.
25 CHAIRPERSON YOUNG: Okay.
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2322 1 MS. UTTAL: We will be providing that.
2 CHAIRPERSON YOUNG: And we will give you 3 back your originals since they are not in the record 4 -
5 MS. UTTAL: Thank you.
6 CHAIRPERSON YOUNG: -- at this point 7 unless somebody else --
8 MS. UTTAL: The next exhibit, which would 9 be exhibit 42, is also one that we have redacted. It 10 is a memorandum to Suzanne C. Black, Director of the 11 Division of System Safety and Analysis, Office of 12 Nuclear Reactor Regulation, from Farouk Eltawila, 13 Director, Division of Systems Analysis and Regulatory 14 Effectiveness, Office of Nuclear Regulatory Research, 15 "
Subject:
Response to User Need for Development of 16 Radiological Source Terms for Review of Mixed Oxide 17 Fuel Lead Test Assemblies." Document is dated 18 February 23rd, 2004. And a redacted version exists, 19 a four-page memorandum, and one attachment numbered 20 B-5.
21 CHAIRPERSON YOUNG: I just probably ought 22 to say we will not consider the unredacted portions as 23 part of the evidentiary record, but they were filed 24 with SECY as prefiled. And so they will be in the 25 overall docket as prefiled.
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2323 1 MS. UTTAL: I understand. Okay. I've 2 lost track of the numbers.
3 CHAIRPERSON YOUNG: The next one would be 4 43. And I assume you're going to now go to the 5 rebuttal exhibits.
6 MS. UTTAL: Yes.
7 CHAIRPERSON YOUNG: Okay.
8 MS. UTTAL: The rebuttal exhibits, 9 document 43 is an e-mail from Wolfgang Wiesenack to 10 Miroslav and to ROM, which is Ralph Meyer, dated 11 Monday, June 14th, 2004, "Re: Statement in One of our 12 SEGSFM Reports."
13 The next document, document number 44, is 14 an article from Nuclear Engineering and Design, volume 15 103. It's entitled "Cladding Tube Deformation and 16 Core Emergency Cooling in a Loss of Coolant Accident 17 of a Pressurized Water Reactor" by F. J. Erbacher.
18 It's dated November 4th, 1986.
19 The next document, document 45, is an 20 excerpt from American Nuclear Society standard for 21 decay heat power in lightwater reactors, 22 ANSI/ANS-5.1-1994. And the exhibit consists of two 23 cover pages and two charts.
24 We have one more to add that was not 25 pre-submitted. It's excerpts from a published work NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2324 1 entitled "KfK In-Pile Tests on LWR Fuel Rod Behavior 2 During the Heatup Phase of a LOCA." It's authored by 3 E. H. Karb, K-a-r-b, et al. It's dated October 1980.
4 And the exhibit consists of a cover sheet, a chart, 5 and a drawing.
6 CHAIRPERSON YOUNG: And that would be 46.
7 And that is all of your exhibits?
8 MS. UTTAL: That's all that we have.
9 CHAIRPERSON YOUNG: All right. Based on 10 the earlier statement that there were no objections to 11 any of the exhibits, those will be admitted. And in 12 a moment, we will take some time to mark those.
13 (Whereupon, the aforementioned 14 documents were marked for 15 identification as Exhibits 16 Number 37 through 46, 17 respectively, and were received 18 in evidence.)
19 CHAIRPERSON YOUNG: Mr. Repka, do you want 20 to make your attachments 1 through 4 exhibits 46 21 through 49?
22 MR. REPKA: Yes, that would fine.
23 CHAIRPERSON YOUNG: Forty-seven through 24 50.
25 MR. REPKA: Forty-seven through 49.
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2325 1 CHAIRPERSON YOUNG: Fifty.
2 MR. REPKA: Attachment 1 to the prefiled 3 testimony dated July 1st, 2004 was the professional 4 qualifications statement as to Stephen Nesbit. And 5 that would be exhibit --
6 CHAIRPERSON YOUNG: Forty-seven.
7 MR. REPKA: -- 47. Attachment 2 to that 8 same prefiled direct testimony was the professional 9 qualifications statement of Robert C. Harvey. And 10 that would become exhibit 48.
11 Attachment 3 to the same prefiled direct 12 testimony was the professional qualifications 13 statement of Bert M. Dunn. And that would become 14 exhibit 49.
15 And the last attachment, 4, was the 16 professional qualifications statement of Dr. J. Kevin 17 McCoy. And that would become exhibit 50.
18 (Whereupon, the aforementioned 19 documents were marked for 20 identification as Exhibits 21 Number 47 through 50, 22 respectively, and were received 23 in evidence.)
24 MS. UTTAL: Your Honor, can I let the 25 panel go now since we are done?
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2326 1 CHAIRPERSON YOUNG: Yes. Thank you.
2 (Whereupon, the witness was excused.)
3 MS. CURRAN: Just a point of 4 clarification. Do we have your exhibit 46?
5 MS. UTTAL: We will be giving it. We have 6 copies for everyone.
7 MS. CURRAN: Okay.
8 CHAIRPERSON YOUNG: Go ahead.
9 MR. REPKA: With respect to those 10 professional qualifications statements, does the court 11 reporter need additional copies of those? They were 12 provided with the prefiled testimony.
13 CHAIRPERSON YOUNG: We can pull them out.
14 And probably we can actually print out clean sheets 15 without the holes in the side.
16 MR. REPKA: That would be easier for us.
17 CHAIRPERSON YOUNG: Okay. Once everybody 18 gets all of their copies, before we take a break to 19 have all of the exhibits marked, we would like to hear 20 back from you on the question we left you with before 21 the last break.
22 MS. CURRAN: We're not prepared at this 23 point to say that we have identified issues that we 24 could concede. It may be that tomorrow morning we 25 could.
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2327 1 CHAIRPERSON YOUNG: Okay.
2 MS. CURRAN: We will think about it and 3 see if there is anything we can do.
4 CHAIRPERSON YOUNG: Okay. Or anything in 5 the nature of background information that --
6 MS. CURRAN: We don't have any 7 cross-examination of that nature.
8 CHAIRPERSON YOUNG: Okay. Do either of 9 the other two parties have anything that you think you 10 could do in terms of cross-examining any other witness 11 on material in the nature of background or less 12 critical issues?
13 MR. REPKA: What I have for Dr. Lyman that 14 would be in the nature of background and less critical 15 would be fairly brief. And I'm not sure we would gain 16 a whole lot by beginning that today. We could, but it 17 wouldn't be substantial.
18 CHAIRPERSON YOUNG: Okay.
19 MR. REPKA: With respect to our panel, I 20 know Ms. Curran is not prepared to proceed. We are 21 certainly prepared to proceed on anything the Board 22 wants to proceed with if the Board would like to ask 23 questions on background or the staff.
24 I also spoke earlier about some brief 25 surrebuttal testimony. I'm prepared to proceed with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2328 1 that. But you may view that as more substantive and 2 critical. But I would be prepared to proceed with 3 that as well.
4 CHAIRPERSON YOUNG: What about the staff?
5 MS. UTTAL: We have no questions of Duke's 6 panel that could be regarded as background. And 7 questions we have for Dr. Lyman on background again 8 will not take that long.
9 CHAIRPERSON YOUNG: Okay. All right. We 10 have discussed this and given your responses, go ahead 11 as we have discussed. We went back and we looked at 12 the e-mail messages.
13 Mine was sent at 10:24 and Ms. Uttal's was 14 sent at 12:13, almost two hours later. And in light 15 of what we have just been told and in light of the 16 fact that Ms. Uttal did not send her e-mail before 17 probably both Duke and BREDL were on the way here, we 18 think that the fairest and the most appropriate thing 19 to do is to start with the testimony tomorrow morning.
20 And so we will do that starting at 8:00 o'clock with 21 BREDL's cross-examination of Duke's panel.
22 As I said earlier, the rules do permit if 23 it would be more efficient and useful for 24 cross-examination of experts by other experts. And 25 obviously we will be monitoring how that goes, but NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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2329 1 that is an option that you may want to avail 2 yourselves of tomorrow to move things along if it 3 would achieve that.
4 Is there anything else?
5 JUDGE BARATTA: Just one other procedural 6 item. I understand that we were advised that this 7 hearing I guess, which normally would have been in 8 addition to the Federal Register and such, would have 9 appeared on the NRC Web page, most of the hearings.
10 Is that --
11 CHAIRPERSON YOUNG: Apparently, although 12 it was in the Federal Register and there was a press 13 release, it may not have gotten onto the Web site.
14 JUDGE BARATTA: Some people had expected 15 that. And we apologize for that. In fact, I had 16 thought it was there and told some people to look 17 there myself. So we'll try to make sure in the future 18 that that does get there. We normally aren't 19 responsible for that, but we'll follow up with that 20 just to make sure --
21 CHAIRPERSON YOUNG: Right.
22 JUDGE BARATTA: -- because we don't want 23 anybody to be inconvenienced by that.
24 MS. CURRAN: We appreciate that.
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2330 1 from any of the parties? After we go off the record, 2 we will mark the exhibits. And so you might want to 3 stay around to make sure that we have copies of 4 everything because we-didn't get the staff's yet.
S JUDGE BARATTA: Do we want to allow for 6 the possibility that we might have to have an 7 additional day of testimony at some point in the 8 future? Over this year, if they could look at their 9 calendars to see if we can square it up.
10' CHAIRPERSON YOUNG: Yes. Look at your 11 calendars and bring your calendars tomorrow just in 12 case we need that. But given what we have been told, 13 it sounds like there is a chance that we might be able 14 to finish in the time that we have set aside tomorrow.
15 MS. CURRAN: I think that's a reasonable 16 prospect.
17 CHAIRPERSON YOUNG: Okay. All right.
18 Then we will go off the record for now and reconvene 19 tomorrow morning at 8:00 o'clock. Off the record now.
20 (Whereupon, at 3:13 p.m., the foregoing 21 matter was recessed, to reconvene at 8:00 22 a.m. on Thursday, July 15, 2004.)
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CERTIFICATE This is to certify that the attached proceedings before the United States Nuclear Regulatory Commission in the matter of:
Name of Proceeding: Duke Energy Corporation Docket Number: 50-413-OLA; ASLBP No. 03-815-03-OLA Location: Rockville, MD were held as herein appears, and that this is the original transcript thereof for the file of the United States Nuclear Regulatory Commission taken by me and, thereafter reduced to typewriting by me or under the direction of the court reporting company, and that the transcript is a true and accurate record of the foregoing proceedings.
Rebecca ilberman Official Reporter Neal R. Gross & Co., Inc.
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