ML20138H853
| ML20138H853 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 12/10/1985 |
| From: | Atomic Safety and Licensing Board Panel |
| To: | |
| References | |
| CON-#485-534 OLA-1, NUDOCS 8512170351 | |
| Download: ML20138H853 (217) | |
Text
,
j n UN11ED STATES U
NUCLEAR REGULATORY COMMISSION IN THE MATTER OF:
DOCKET NO: 50-250 OLA-1 50-251 OLA-1 FLORIDA POWER AND LIGHT COMPANY (Turkey Point Nuclear Generating Plant, Units 3 and 4)
ORJGWAL EVIDENTIARY HEARING O
LOCATION: MIAMI, FLORIDA PAGES: 234 - 440 DATE: TUESDAY, DECEMBER 10, 1985 b,
o,'
ACE-FEDERAL REPORTERS, INC.
p OfficialReporters
\s_. l 444 North CapitolStreet Washington, D.C. 20001 ,
(202)347-3700 8512170351 PDR 851210 ADOCK 0500 0 NATIONWIDE COVERACE ,
m a
Jm Walch 234
- UNITED STATES 0F AMERICA I
/ 2 l NUCLEAR REGULATORY COMMISSION 3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 1
li 5
l ---------------------------------X 6 In the matter of: :
7 TURKEY POINT NUCLEAR GENERATING : Docket Nos. 50-250-OLA-1 PLANT, UNITS 3 and 4 : 50-251-OLA-1 8
FLORIDA POWER AND LIGHT :
9 :
COMPANY :
10 - ---------------------------------X k
- i.
11 United States District i Courthouse l 12 i 2nd Floor Courtroom !
g-) Old Building s
) 13 300 NE First Avenue ,
Miami, Florida '
14 Tuesday, December 10, 1985 15 16 l
Hearing in.the above-entitled matter was convened j 17 ,
at 9:29 a.m., ROBERT M. LAZO, presiding. I 18 l i
BEFORE: i io ROBERT M. LAZO, Chairman 20 !
Nuclear Regulatory Commission l Atomic Safety and Licensing Board I y' ,
RICHARD F. COLE, Member Nuclear Regulatory Commission A m afe y an ensing Board 23 g ,- 24 j EMMETH A. LUEBKE, Member Lce F[~ J Herorters, Inc. g y ommission 25
! Atomic Safety and Licensing Board l
~ Jco-Walch 235 1 APPEARANCES:
.f~h'
\/ 2 On behalf.of Applicant, Florida Power and Light Company:
3 NO,RMAN A. COLL, Esq.
Steel Hector & Davis 1- , 4000 Southeast Financial Center
!l Miami, Florida 33131-2398 5 .
MICHAEL A. BAUSER, Esq.
6 Newman & Holtzinger, P. C.
1615 L St., N. W., Suite 1000 7 Washington, D. C., 20036 8 On behalf of Interveners 9 MARTIN H. H0DDER, Esq.
1131 N. E. 86th Street 10 . Miami, Florida 33138 11 On behalf of NRC Staff 12 MITZI A. YOUNG, Esq.
and 1() 13 ! MARY C. WAGNER, Esq. .
Nuclear Regulatory Commission !
14 Washington, D. C.
f 15 i 16 17
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20 ; j 21 ,
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/" 24 b FE m Reporters, Inc.
25 I
- Jco Walch- 236 i CgEIENIS
(_/ 2 OPENING STATEMENTS:
3 Mr. Bauser Page 239 4 , Mr . Hodder Page 241 5 i
! VOIR 6 WITNESS DIRECT DIRE CROSS REDIRECT RECROSS BOARD 7 Gordon D. J. . Edwards 8 Examination by Mr. Hodder'251 Examination by Mr. Coll 254 9 Examination by Ms. Wagner 269 4
10 Edward A. Dzenis f, l
11 j Examination by Mr. Bauser 293 l Examination-by Mr. Hodder 295 l 12 i Examination by Mr. Edwards 303 l
,, Examination by Ms. Young 381 !
13 ' Examination by Judge Colo !
). 407 Examination.bv Judge Lazo 417 14 Examination by Judge Luebke 419 Examination by Mr. Edwards 424 15 Examination by Ms. Young 432 ,
Examination by Judge Cole 435 '
- 16. Examination by Ms. Young 437 Examination by Mr. Edwards 438 17 RECESSES:
18 10:45 to 11:02 a.m. Page 292 I t,
'O Luncheon recess - 12:34 to 2:01 p.m. Page 349 20 3:12 to 3:31 p.m. Page 395 21 !
l ******
22 { '
LIMITED APPEARANCE BY: Ms. BRENDA MEYERSON Page 397 23 24 DOCUMENTS: AFTER PAGE:
nar t. a,pon n, ane. Prefiled testimony of Mr. Dzenis 302 25 Public-appearance documents 402
237
- 1-1-SueW 1 P_ R,O C,E E_ D,I,N_ G S, i 2 (9:29 a.m.)
3 JUDGE LAZO: Good morning, ladies and gentlemen.
4 The clock on the courtroom wall is obviously stopped. It 5 will be correct twice a day. But it's just about 9:30, so 6 I Propose that.we proceed.
7 This is an administrative- proceeding before 'an 8
Atomic Safety and Licensing Board of the United States Nuclear 9 Regulatory Commission in the matter of Florida Power and 10 Light Company, Turkey Point. Plant, Unit' Numbers 3 and 4, 11 hereinafter referred to as Licensee.
12 The proceeding is identified as Nuclear Regulatory Oq) 13 Commission Docket Numbers 50-250-OLA-1 and 50-251-OLA-1.
14 The Licensee, Florida Power and Light Company, 15 is authorized to possess, use and operate Turkey Point Plant 16 Units 3 and 4, two pressurized water nuclear reactors located 17 in'Dade County, Florida, pursuant to the provisions of 18 Facility Ocerating Licenses Numbers DPR-31 and DPR-41, issued 19 July 19, 1972 and April 10, 1973, respectively.
20 The parties to the proceeding are the Licensee, l
21 Florida Power and Light Company; the technical staff of 22 the Nuclear Regulatory Commission; and, the joint intervenors, 23 the Center for Nuclear Responsibility, Incorporated and Joette 24 Lorion.
WF Reno,ters, Inc.
25 Now, the parties have all met the Licensing Board l
238
- 1-2-SueWl' before during the prehearing conferences which we held in FT
\/
2 Miami, but for the benefit of the spectators let me just. .
3 briefly introdui*e the Judges. Judge Righard F. Cole is 4 seated at my left, your right. Dr. Cole is an engineer and 5 an environmental'iscientist.- Judge Emmeth A. Leubke is seated 6 at my right, your left. He is a nuclear physicist. My name 7 is Robert M. La,.zo.
I am an attorney.
-8 I am going to aak for appearances by the parties, 9 and I will ask that if each'of them, when they entdr'an 10 appearance, would indicate t?hether they wish to make an open-11 ing statement. I think also after we've had appearances 12 entered, I will ask counsel for the License to just briefly '
n-(_) 13 summarize the previous history of this proceeding beginning 14 with the notice of opportunity for hearing' through the pro-15 cedures that took place up until today at'the openiitg of 16 the evidentiary hearing. 3 17 Now, may we have the appearances of the parties, 18 please? -
19 MR. BAUSER: For thh Licensee, my name is Michael 20 Bauser.. I am with the Washington, D. C. law' firm of Newman 21 and Holtzinger. With me at the tdble today also on behalf of 22 Florida Power and Light Compuny is Norman A. Coll with the 23 Miami law firm of Steel, Hector and. Davis.
s <
24 JUDGE LAZO: Thank you, Mr. Bauser'. And for the f
w-Fu~). Ceoorters, sne. '
25 Nuclear Regulatory Commission staff.
I ,
L 4
239 ,
'#1-3-SueWl MS. YOUNG: Good morning. My name is Mitzi Young..
O k/ 2 I am representing the NRC Staff. Seated at counsel table 3 with me is Mary Wagner, also entering an appearance in this 4 proceeding.
5 Also seated at counsel table but not entering an 6 appearance in this proceeding is lur. Daniel Mcdonald, who 7 is the NRC Plant Project Manager for Turkey Point.
8 I will not be making an opening statement.
9 JUDGE LAZO: You will not make an opening state-10 ment? Thank you, Mrs. Young.
11 And for the Intervenors.
r-12 MR. HODDER: My name is Martin H. Hodder. I am
() 13 the lawyer for the Intervenors. Seated at my table with me :
14 is my client, Joette Lorion and Dr. Gordon Edwards, who will 15 address technical aspects of the case.
16 I intend to make a brief opening statement with 17 ' the permission of the Board. I also have some cuestions --
18 a preliminary question I wanted to ask the Board.
19 JUDG" LAZO: Well, thank you, Mr. Hodder. Mr.
20 Bauser, would you please provide the Board and the parties 21 with a brief summary of the proceeding to date?
22 MR. BAUSER: Yes, Judge Lazo, I would like to do 23 that and also add a brief statement at the end of that, of 24 what Licensee intends to do at these hearings and this ko-Fg Reorwes,1N.
k 25 proceeding.
240 fl-4-SueW 1 MR. BAUSER: A notice of the proposed amendments
(-
\_/ which are the subject.of this proceeding was published in 2
3 the Federal Register on August 7th, 1983.IIn response to 4 that notice, the Center for Nuclear Responsiblity and Joette 5 Lorion filed a petition to intervene in this proceeding.
6 The NRC Staff issued the amendments in question 7 on December 23rd, 1983 stating that the amendments posed -
8 no significant hazards under 10 CFR, Section 50.91.A.4 and 9 could be issued without a prehearing on Contentions filed 10 by the Intervenors.
11 Licensee has filed a motion for summary disposition 12 in this proceeding. By Orders dated August 16 and November 8,
?
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(_) 13 1985 the Board has granted Licensee's motions for summary
-14 disposition in this proceeding with respect to one of the two e
15 Contentions admitted, that is Contention B but denied summary 16 disposition with respect to Contention D.
17 In denying summary disposition with respect to 18 Contention D, however, the Board has limited the scope of 19 litigation to three narrow issues presented in the form of 20 three specific questions. The purpose of Florida Power and 21 Light company's appearance today is to address those three 22 questions submitted by the Board.
23 We will be offering the testimony of a witness 24 this morning to address those contentions and then will be b F8 Coporters, Inc.
25 presenting that witness for examination by the Board and other
E 241-i 31-7-SueW I parties.
p)\- '2 l JUDGE LAZO: Thank you, Mr. Bauser. Let me just
, 3 say that I was remiss in introducing the Board members, not l
, 4 to introduce Ms. Ellen Ginsberg, who is the law clerk. Ms.
Z, 5 Ginsberg is a lawyer and is serving a clerkship with the 6 Atomic Safety and Licensing Board panel.
7 Mr. Hodder, you indicated -- Mr. Bauser, had you I# 8 completed? Is that the end of your opening statement?
9 MR. BAUSER: Yes, it is.
10 JUDGE LAZO: Mr. Hodder, would you procee d, 11 please?
12 ,
MR. HODDER: Yes. Before I even make an opening
() 13 statement, I would' like to ask for some guidance from the 14 Board.
15 JUDGE LAZO: Surely.
16 MR. HODDER: Today, as Mr. Bauser has indicated, 17 we are going to limit the scope of this proceeding to Con-l 18 " tention D and questions pertaining to it, identified by the 19 Board. These questions are highly technical considerations 20 that I, as an attorney, may not profess the expertise that
- 2) would fully be developed in my cross-examination.
22 Sitting with me, as I've indicated earlier, is
'd.) ,
23 Dr. Gordon Edwards who has special expertise by way of back-24 ground and training and also has special experience, wherein ham-F Reporters, Inc.
25 previously he has examined witnesses on technical points such
242
- 1-6-SueW 1 as that before the Board today.
/
I am going to ask the Board to considor allowing 2
3 Dr. Edwards to conduct cross-examination of the technical 4
witnesses,-because I feel that his special ability which 5 transcends my own would help the Board to develop a fuller ,
6 record and a more complete understanding of the problems here.
7 If the Board is unwilling to do that, of course, Dr. Edwards 8 is sitting with me and he could suggest questions which I, 9 of course, will ask.
10 I thought though it might be a smoother handling, 11 especially if I get into a lack of understanding on my part 12 of responses to questions, if the Board considered letting k_) 13 Dr. Edwards handle cross. Therefore, it's my suggestion that 14 you do this and my request.
15 There is precedent for it.
16 JUDGE LAZO: Mr. Hoddar, our rules of practice 17 do permit a technicalointerrogator to assist a party by 18 cross-examination. What we would want to avoid is having two a
19 separate persons questioning the same witness. That has an 20 element of unfairness to it.
21 MR. HODDER: I fully understand that.
22 JUDGE LAZO: So, if you were to delegate that 23 responsibility to Dr. Edwards then at least for that witness 24 or that portion of the testimony. We would not want two W-F Reporters, Inc.
25 people cross-examining.
l
243
- 1-7-SueW I And I also want to say that I think we should first 1
i' 2
hear from the other parties to see if there are any objections.
3 MR. HODDER: Well, first in response to the Board, 4 before we address the other parties, I agree with the Board 5
that there is a potential for abuse by having two people in- ,
6 volved in the conduct of the case on cross. Therefore, I 7 would assure the Board in the event that I turn cross- i 8 examination over to Dr. Edwards on a given witness, that then 9 he would be the one that would exclusively conduct that cross, 10 for better or for worse.
II Now, that doesn't mean that maybe on reexamination I2 of some other matter such as voir dire that I might not cross-() 13 examine. But, once I give it to Dr. Edwards on a technical Id question, such as those identified by the Board, the three 15 questions, I wouldn't seek to then interfere.
16 JUDGE LAZO: Licensee, do you have any problems?
17 MR. BAUSER: In connection with the provisions 18 referred to by the Board providing for special interrogators a
19 we, I think, will have a few questions that we would like to 20 raise prior to the Board's making the required findings speci-21 fled in the pertinent regulation.
ll END #1 22 Mary flws 23 (x 24 Acm-Fm. __.) Reporters, Inc.
25
244
( i JUDGE LAZO: That should present no problems.
2 Mrs. Young, do you have any comments?
3 MRS. YOUNG: The procedure Mr. Hodder describes is 4 provided for in 10 CFR 2.73B, and that provision states that 5 a party may request that the Presiding Officer permit a 6 qualified Individual who has scient111c or technical 7 training or experience to participate on behalf of a party 8 in examining and cross-examination of expert witnesses.
9 The finding required for that procedure to be to followed is, first, and I think most importantly that the 11 individual is qualified by scientific or technical training 12 or experience to contr1bute to the development of an 13 adequate decisional record in the proceeding by the conduct 14 of such examination or cross-examination.
15 So I think Mr. Hodder, while the procedure may be 16 easy to accommodate and the confusion that the witness may i 17 have by the potential for two Interrogators, may be able to 18 accommodate that, it it important that some showing be made 19 that Dr. Edwards is qualified to act as an expert 20 interrogator.
21 JUDGE LAZO: Yes. Perhaps we could resolve that 22 problem by having Mr. Hodder identify Dr. Edwards and put 23 his qualifications into the record, and then you and 24 licensee's counsel would have an opportunity to question the 25 technical interrogator.
245 Mr. Hodder, do you have something in addition?
(s/~')
v y
2 You had mentioned that you had some questions.
3 MR. HOODER: Well, Chairman Lazo, I would like to 4 point to the Board that in the outline of testimony prefiled 5
by Dr. Edwards and submitted to this Board,.there was 6
accompany.ing that testimony as Exhibit A the professional 7
qualifications and experience of Gordon D. J. Edwards.
8 On the second page of that, which is in the 9 possession of the parties, Dr. Edwards states that in 1977 10 through 1978 he was a consultant to the Ontario Royal 1; Commission on-Electric Power Planning where he 12 cross-examinec exper.ts in reactor safety from Ontario Hydro, y3 the utility, the Atomic Energy Commission of Canada,
[) ja Limited, a research and development organization and the 15 Atomic Energy Control Board, the regulatory agency, over a.
t 16 perioo of.several months.
17 He goes on to outline other experience as a 18 consultant ani txpert, and I think this testimony of Dr.
i9 Edwards tends to qualify him as a technical interrogator in 20 these proceedings here today.
21 I have a copy, if the Board requires it, of his 22 technical background.
23 JUDGE COLE: I think we have some copies of it.
24 Thank you.
25 JUDGE LAZO: Well, are there any other opening gvf})
v
F I 246 1 matters that we should discuss before we proceed?
2 (No response.)
3 Hearing no response ---
4 MR. HODOER: I did indicate that I also had a 5 brief opening statement.
6 JUDGE LAZO: Yes. Why don't you present that now, 7 Mr. Hodder.
8 MR. HODDER: Mr. Bauser certainly identified the 9 issues that are before the Board today, and very briefly 10 identified the history of these proceed 1ngs.
11 I do wish to point out and remind the Board that 12 the license amendments being sought to be vindicated here j3 today by the licensee, the utility company, are efforts on L- 14 their part to reduce radiation or flux at the Turkey Point 15 reactors.
16 And these particular license amendments wh1ch we 17 are addressing here today are one small increment of a 18 program being implemented by the utility company to achieve 19 that vessel flux reduction.
20 We nave indicated that there is leakage between 21 the e a r,11 e r amendments requesting the use of the OFA fuel 22 and of course the utility company contests this.
23 We have also indicated that we feel it is L
24 appropriate for this Board to take into account the impaired 25 condition of this reactor. as it consider the matters before
247 hl 'N_/.
i it today.
2 Unfortunately for our side, the Board has ruled 3
that it would be inappropriate for the purpose of these 4 proceedings to take testimony of to consider the present 5 condition of the reactor.
6 We'. feel that this-should be taken into account 7 and we wish to remind the Board of our view on that subject.
8 But we are certainly prepared to address this case within 9 the parameters established by the Board, and Dr. Edwards yo will address those three Board questions.
11 We think, however, that one other consideration 12 is appropriate for the Board, and we will attempt in our 13
(< amination of witnesses to determine what the consequences
[p vlm ) ja are of the utility company not getting the license 15 amendment. In other words', we feel.that it is important.for
-16 the Board to consider the consequences of a decis1on not to i7 allow issuance of this amendment, and we-may seek to inquire 18 of witnesses:what those consequences might be. A possible-39 consequence of course would be deration of the unit.
20 We would like to know more about why it is that 21 this amendment is so necessary in light' of the testimony 22 submitted by'the utility company saying.that although they v
23 have DNBR at 1.17, they have a calculated DNBR of 1.34. We 24 don't really understand this and we are going to seek to 25 grain a broader understan. ding. We really want to know the
- p r R).
248 7
_j g g I consequences of their'not getting this amendment.
2 JUDGE LAZO: Counsel, let us assure you that if 3 you'do not ask those questions, that the Licensing Board 4 intends to.
5 MR. HODDER: I am encouraged, and I thank.you.
6 JUDGE LAZO: Now in our notice scheduling the 7 evidentiary hearing here today, and that notice was entered 8 on September 18th, we did indicate that we would be prepared 9 to receive statements from members of the public who to requested the right to make a limited appearance.
11 So let us ask now before we proceed. It is 12 generally the practice of the Board to take those limited 13 appearances at a very early stage in the' proceeding. We have
[J/T)
- c- m 14 not received any written r'equests for limited appearances, 15 which can be written or oral, but- are there present in the 16 hearing room today any members of the public who desire to 17 make a limited appearance in this proceeding?
18 (No response.)
19 Well, hearing no response ---
20 MR. HODDER: Mr. Chairman, Ms. Lorion-tells me 21 that someone asked her to deliver their written statement'. I 22 know nothing about this, but I would let her explain.
'23 JUDGE LAZO: Yes. Ms. Lorion.
24 MS. LORION: Somebody gave me a written 25 statement, and I also probably misled people in thinking j ,. e xl)-
249
/^
'b} j th.at statements would be taken in the afternoon, and told 2 them that they would probably, you know, if they couldn't be
~
3 here all day, they should come in the afternoon. But-I do 4 have-one written statement, and I can just submit it for the 5 record?
6 ~ JUDGE LAIO: Ms. Lorion, we would be pleased to.
7 If you give it to our Law Clerk, please, we will see that it 8
becomes a part of the record.
9 (The Limited Appearance Statement follows:
10 11 12 13 14' 15 16 17 18 19 20 21 22 23 24 25
('N m Dj
1 6830 S. W. 127th Place Miami, Florida 33183 V Tel. (305) 385-3714 W N r 9, 1985 Statement Directed to:
The Atcnic Safety and Licensing Board of the U.S. Nuclear Regulatory Ccrrmission Re: 12/10/85 Hearing on the Safety of Fuel Core Redesign at Turkey Point Nuclear Power Plant Gentlemen:
Although the present . hearing regarding the safety of the fuel core redesign at the Turkey Point Nuclear Power Plant is to be ccrrmended, I would like to state I believe this is not enough.
The Plant's last test 10 years ago, determined that the weld metal in Reactor No. 4 was '30% more brittle than Reactor No. 3. Yet, Florida Power & Light Ccurpany has managed to receive an extension period for the re-testing of Reactor No. 4, of 14 additional years!
!)
k 24 years will have passed between the testing of Reactor No. 1, if it is not tested this year as originally planned. Not to test this particular i
i reactor as originally intended, . shows blatent disregard for public safety.
I find this appalling.
When originally . built, I understard that it was. determined that the reactors should be tested every 10 years for safety reasons. .%ny problems the Nuclear Regulatory Ccrrmission is facing today, (such as the Three Mile Island incident, ultdowns and pressure vessel embrittlement) were never considered, since these were problems - no cne could foresee.
I believe what you are acccuplishing with this hearing today is only the
- l. very beginning of an extremly needed full hearing en the safety and operations of the individual nuclear reactors at Turkey Point.
l I w>uld demand that the testing of the weld metal in Reactor No. 4 be done within the next 6 months and that a citizen's right to participate in discussions, information, and regulation of a nuclear utility that serves the public and can threaten life itself, within the large l surreurding area of nuclear energy plants, is an undeniable democratic right of every Amrican citizen.
Thus, as a concerned citizen, I ask for a broader hearing regarding the safety hazards involved pertaining to-the operatirg of the Turkey Point Nuclear Power Plant, and I thank pu for the hearing taking place in t
Miami en M r 10, 1985.
J sincerely, !
i 4k- '
l l
- Rucy Ja
-K.urau
250 g) 1 JUDGE LAZO: And I think if others relying on your 2 statement should come in this afternoon, at a sultable point 3 we might be able to accommodate them.
4 Well then I think the next thing we should do is 5 to get Dr. Edwards' qualifications introduced as p a r t' of the 6 record. Mr. Hodder, you stated that you might have some 7 questions of Dr. Edwards, and perhaps we should swear Dr.
8 Edwards and let you supplement his statement of 9 qualifications, if you feel that is necessary, and then 10 permit the other parties to question Dr. Edwards.
11 MR. HODDER: Let me confer with Dr. Edwards.
12 (Pause.)
13 14 15 16 17 18 19 20 21 22 23 24 25
^
251 gjw 1 1 MR. HODDER: Chairman Lazo, I am ready to offer b 2 the witness, and I am going to ask him a couple of questions.
3 ' JUDGE LAZO: Fine. I wonder, Dr. Edwards, it might
- be convenient if you would take a seat at the witness 5 '_ table.
l 6 MR. EDWARDS: Certainly, Judge Lazo.
7 JUDGE LAZO: Thank you.
8 Whereup'on, 9 GORDON DOUGLAS JOSEPH EDWARDS, 10 was called as a witness, and having first been duly sworn by 11 Judge =Lazo, testifi.ed as follows:
12 .
DIRECT EXA14INATION XX EX 13 BY'MR, H0DDER:
I'l Q Dr. Edwards, could you give us your full name I
a '
15 ' and address?
!! j 16 i A My'name is Gordon Douglas Joseph Edwards, and I l 1
-17 L reside at 53 Dufferon Road, in Hampstead, Quebec. f 18 h Q Could you tell us where you are employed, sir? ll b !
A I am employed as a Professor of Mathematics and 20 h Science at Vanier College in Montreal.
21 Q I am holding here Exhibit A, which'is.a professional 22 qualifications and experience of Gordon D. J. Edwards, which 43 we previously filed accompanying your outline of testimony
- f) 21 MFLJ Reportert. Inc.
in these. proceedings, 25 Have you reviewed that Exhibit A document on your i
3-2-JrcWal 252 1 qualifications, Doctor Edwards?
'~' 2 A Yes, I have.
3 Q And do you adopt that as your testiony in this 4 tproceeding?
l 5 A Yes, I do. It is very brief. I can supplement it 6 if you wish.
7 Q Okay. What I would like you to do by way of I 8 supplementation is tell me to what extent you have acted 9 as a technical interrogator or even a consultant or assistant 10 to an interrogator in prior proceedings in which you have had I
11 experience? l l !
12 -A There were two main occasions in which I was retained i
13 ; as an interrogator, cross-examiner of technical experts.
F 14 ' The first was the Government of Saskatchewan, the l?
1 15 h Board of Inquiry, which was a judicial inquiry into all aspects
< l 16 4 of uranium mining, this inquiry was divided into five phases, j 17 and the fifth phase had to do with all aspects of nuclear power !
18 , beyond the uranium mining stage.
'O This included, for example, reactor safety, and e U
20[ I was retained as a technical person to cross-examine expert ,
21 liwitnesses, including such people as Dr. Norman Rasmusson on j 22 the safety of nuclear reactors. l 23 The seccnd occasion for was approximately five I '; 24 months during a lengthy Royal Commission inquiry into electric uw-Fi_j Reprters, Inc.
25 power planning by the Ontario Government, in which I again was i
m ,
3-3-JocWol.- 253 1
I retained'by the Commission to cross examine technical experts O '2 from the' utility, from the regulatory body, a'nd from the 3
design agency on' reactor safety questions.
I I have also acted as a consultant, not cross-5 l examining, but providing background information during research 6
-and providing testimony'to, for example, the Ontario Legislature ,
7 which has a select committee on Ontario Hydro Affairs, which 8
did a thirteen week investigation of reactor safety following 9
the Three Mile Island Accident here in the United States,
.10 and I was involved throughout that thirteen week period.
II Most recently, I was retained in 1985 by the Auditor I2 General of Canada to participate in a comprehensive audit of 13 the Atomic Energy Control Board, and primarily my involvement M
was limited mostly to questions of reactor safety, and reactor 11 15 'I
- licensing questions, i i
16;- h Q So, even on those occasions when yoti actually didn't I7h conduct the cross examination, there were other occasions when l
l IO '
you prepared questions designed to elicit information. is that 10 1
- true? I 20 A -Yes, that is correct.
'l MR. HODDER: I think that is my complete examination 2l of th.e witness, Mr. Chairman.
73 JUDGE LAZO: -Thank you, counselor. Mr. Bauser or aa moorters, in . Mr. Coll.
25 MR, COLL: Yes, sir. We would like some voir dire 1
.l l
3-4-JccWal 254 I of this witness.
- X I EX 2 VOIR DIRE EXAMINATION 3 EY MR. COLL
4 Q Dr. Edwards, my name is Normal Coll. As I understand l
5 your educational background, it is principally in the discipline 6 of mathematics, is that correct /
7 A My original degree from the University of Toronto 8 was as a Gold Medalist in Mathematics and Physics. I was in 7 a --rogram entitled, Mathematics, Physics' and Chemistry, which l
10 is a triple science maj or. l II My actual, orsginal bachelor's degree from the I2 University of Toronto was in applied mathematics, which is
/~'s !
C, 13 equivalent to theoretical physics. !
I4f ; Q So, you have a master's degree from the University I
15 ! of mathematics in 1962, is toat correct?
I l!
A The university of --
i- i 17 0 Chicago.
Q P
I60, A The University of Chicage, that is correct. I have !
i io ' i 4
.a Master's Begree in Mathemat.ics from the University of :
li i 20 l Chicago, ,
2I And you have a Ph.D Degree in mathematics as well,
, Q 22 I do you not?
23 That is corre ct ,
A r ~. 2;
( ) Q Do you have any other degrees?
wA .J Reporter,, Inc.
'5
^ l A Yes, I have a Master's Degree in English Literature, j l
i 3-5-J'cGWol - 255 i
1 MS, WAGNER: Excuse me for a moment. I am sorry
. ,cy 1
'\-)
2 to interrupt. Judge Lazo, I am having trouble hearing the 3 witness with the noise going on outside in the corridor.
4lCould someone close that door?
l 3 Thank you. Sorry for the interruption. ,
6 BY MR. COLL: (Continuing) 7 Q Do you have any other degrees other than the ones 8 you have told me about? -
'9 A Yes. I have a Master's Degree in English Language 10 and Literature from the University of Chicago.
11 Q Other than that, do you have any other degrees?
12 A No.
I 13 Q' As I understand your occupational experience, it 14l has principally been as a mathematics teacher, is that h
15 . correct?
4 b
l t i 16 A Yes. Mathematics and science. I I teach in an A
17 ? integrated' science program in Montreal. We plan curriculum l
18 l together with physics , biology , chemistry, and mathematics. i
- .! 1 Q Did you teach mathematics at the University of-20 f Toronto?
i 21 A. Yes,.'I did.
72] Q .Did you teach mathematics at the' University of 23 Chicago ?
28 A Yes, I did.
We F Reorters, lec, j
25 Q Did you teach mathematics at the University of
.3-6-JocWai 256 1 Western onatario?
.s -
2 'A Yes, I did.
3 Q Did you teach mathematics at Queens University?
4 i A Yes, I did. By the way, I might mention at 5 Queens, as at Toronto, the students were engineering students.
6 Q Did you teach mathematics at the University of 7 British Columbia?
8 A Yes.
-9 Q Did you teach mathematics at Ottowa University?
10 A Yes.
11 Q And you are presently teaching mathematics at 12 Vanier College, is th'at correct?
f3
(_). 13 , A That is correct.'
14 i. Q- You mentioned other teaching experiences you had I !
15 j in .the sciences, and I beliese you said biology, is that !
l 16
! correct? I 17!; A My teaching in Biology has been very limited,-and 18 in fact has been limited principally to the radiation h L I
. question.
20 Q Where was that, ' sir?
21 A That was at Concordia University.
22 l Q What year was that?
23 A I would have to calculate the year actually. It 24
(~%
p h ] seporm s,Inc.
was-v- just a moment. It was'in the mid-1970s. I can't 25 remember the precise year.
l !
I
"3-7-JodWal 1 .I was an adjunct professor at Concordia University (3)
N- 2 over a period of.three years.
3 Q What was the title of that course?
4 i- A I believe it was social issues in biology.
5 .Q Wasn't it chemistry in society?
6 A No. That was a chemistry course which I taught 7 also. These were, by the way, both biology course and the 8 chemistry course were post-graduate courses.
9 They were at the Master's level.
10 Q What other courses have you taught?
11 A I have taught a great many courses. I am sure '
12 you wouldn't want a rendition of them.
13 Q .What other employment have you had other than as .
14 a mathematics teacher or as a science teacher?
t 15 k MR. HODDER: I think I object to that question 16 Ibecause it is not relevant to *' e proceeding, Chairman'Lazo. }
d v
17 [ We'are here to qualify a witness on his technical 18 !i expertise , not his other employment. If he washed dogs.for I.I.
'OlaLliving, that doesn't have any influence on his ability to 1
20' testify here today.
.21 His other employment, which may be extraneous to
~22 anything such as we are seeking to investigate is not relevant 23 to this inquiry.
O 24 w.hLJ acooners, Inc.
MR. . AZO: Well, Mr. Hodder, the requirements to 25 present testimony as an expert witness depend on the qualifications l
3-8-JacWel -258 l
l
] of the. expert, and those qualifications may be gained not only- 1
(~1 N_/ 2 by educational background, but by practical experience, so 3 we would think'that perhaps the other employment is relevant.
t , MR. H0DDER: May I beg to differ. The question l
5 that Mr. Coll poses:- What other employment?
-6 If somebody asked me that, I would think of all i 7 the things I have ever done from fixing my cars to everything.
8 It is just simply an unfair question. If Mr. Coll has 9 something in mind, he can identify the area of expertise he 10 is seeking to elicit, and he can key in a question about 11 other employment, but just asking the man what his other 12 ! employment is is unfair to the witness, and not relevant to
() 13 h
these proceedings.
14 JUDGE LAZO: We will_ sustain the objection, and F
15 k ask Mr. Coll if he can.present another question.
I 16 BY MR. COLL: (Continuing) i 17 Q Dr. Edwards, I note that you list your qualifications 18 and experience, and it indicates you were employed as the
?
'O President of an orgainzation entitled, The Canadian Coalition n
20 for Nuclear Responsibility, Inc., is that correct?
21 A I am not employed. -This is a voluntary position.
22 Q' How long have you been associated with this 23 organization?.
28 A Since 1975,
' (~
y Fig} Atoorters, Inc.
I 25 Who founded this organization?
Q
3-9-Joolkl 259
- ) A A group of about 30 people, myself i~ncluded.
d 2 Q What is the' purpose of this organization?
3 A The purpose of this organization is to provide i j information on nuclear issues in a form which is understandable 5 have a detailed technical background.
lto people who may not 6 Q How long have you served as the President of
'7 this organization?
8 A' I have served on and off over that period 1975 to 9 1985, several different terms.
10 Q You mentioned that you teach, you work as a 11 consultant,-and that you provide volunteer service to this i 12 Canadian Coalition for Nuclear Responsibility, Inc.
( 13 Can you identify for me what percentage of your 14 time during the last five years that you have devoted to i
15 . teaching, devoted to consulting. .and that you devoted to I! ,
16 0 Canadian Coalition for Nuclear Responsibility, Inc? .I N I A This is highly variable. It depends upon the !
18 l! circumstances. 'l For instance, in the process of doing consulting 20 I work for the Auditor General of Canada, I devoted a considerable 21 amount of time, which is a matter of record with the 22 Auditor General's Office.
23 Going to Ottawa for meetings, preparing documents
') 21 w+[u seconen, Inc.for them, and so on.
25 When I have a contract, I service the contract.
l l.
3-10-JoGWol 260 j When I do.not have a contract, then I am not rm s 2 bound to do that consultation work.
3 So, also from year to year the amount of consultation g is variable. For instance, in 1985, this year, I have had three
!lg 5
l! Primary consulting jobs; one is with the National Film Board 6 of Canada on a film which they produced dealing with nuclear 7 issues.
8 The seco'nd'was with the Auditor General of Canada 9 on the comprehensive audit of the Atomic Energy Control Board, 10 and the third was with the Canadian Broadcasting Corporation.
i End 3. 11 SueW fois. !
12 l
() 13 14 1
15 I '
I6 e
i.
17 '
18 l 1
10 i.
20 2i 22 23
'p 24 Am-FL.O Reporters, Inc.
25
l 261
~$4-1-SueW 1 Q Can you estimate for me, sir, how much of your O
2 time you spent in those three areas during the past five 3 years?
4 A.
The majority of my time is spent -- when you say 5
estimation of time, do you mean percentage
- wise?
6 Q Correct, sir.
7 A Well, I'm also a family man. If you subtract my
-8 ' family time, which I treasure very much, then I would think 9
that it would be approximately fifty percent teaching duties, 10 about twenty-five percent consulting, and the rest of the 11 time would be connected the Canadian Coalition for Nuclear ,
12 Responsibility.
( 13 Q All right, sir. Other than cross-examining witnes-14 ses in a proceeding, have you yourself ever testified as an 15 expert in a proceeding? i l
16 A Yes, I have. t 17 Q What proceedings were those? ,
3 18 A I Those were'the Cluff Lake Board of Inquiry into j oc'- 19 Uranium Mining, the Royal Commission of Inquiry into Electric 20 Power Planning, the Select Committee on Ontario Hydro Affairs 21 hearings on nuclear safety.following the-Three Mile Island 22 accident in Ontario.
23 And there were a few other hearings that I've parti-() 24 we;-.2 r..pormes, inc.
cipated in.
But those three I think are particularly relevant, 25 because the topic of testimony was reactor safety.
262
- 4-2-SueW j Q You say the topic of testimony was reactor safety.
2 Can you identify more particularly what the subject matter 3 of your testimony was --
4 A The subject matter --
5 0 -- during those proceedings?
6 A The subject matter had to do with my area of 7 expertise which is mathematical and which has to do with 8 mathematical analysis, calculation probabilities, the use of mathematical models. l 9 In the case of the Royal Commission on i 10 Electric Power Planning, my testimony consisted in a detailed
-11 critique and counter-argument to certain probability figures 12 which were being advocated by the utility.
() 13 And in the final
- report of the Royal Commission I
1.s on Electric Power Planning they stated in the report that l 15 they found my probability figures more zealistic than the 16 utility's. This was the probability for major reactor ac- '
l 17 i cidents.
18 .
l It involved loss of coolant acci, dents coupled with i 19 loss of emergency core cooling.
I 20 Q You indicate that your area of expertise is mathe- !
{
21 matics; is that correct? !
l
~ 22 A That's correct. !
23 Q Do you have any knowledge, skill, experience, training 24 or education in the field of engineering?
Mie F Reporters, Inc.
i 25 A No, I do not. I
263 I
)4- SueW Q Do you have any knowledge, skill, experience, 2 training or education in the field of nuclear physics?
3 A
~
Only to the extent that I was a theoretical physics 4 student at the University of Toronto.
5 Do you have any knowledge, skill, experience, Q
6 training or education in the field of thermo dynamics?
7 A Yes.
8 Q And'where do you obtain that, sir?
9 A As a student and subsequently. Those.words, 10 they kind of slip by~me, I must say, because mos't of my U experience -- you are saying skill? Would you go through '
12 those-words again?
13 Q Have you ever had any courtes in thero dynamics? --
I# A Skill, training, experience? ,
15 Yes, sir. !
Q i
16 A My experience in the nuclear field has been !
' I7 I cbtained through direct interaction with people in the j l I 18 industry and people in the regulatory body.-
- I Q Do you have any educational background in thermo 20 dynamics? .
j.
21 A Yes. Every science student has an educational 22 background in termo dynamics.
l l 23 Q How many courses did you have in the field of 6., non.n, ne, thermo dynamics?
I
-25
'A At least three. I don't know.
l._
i
[L o
264 ji
- 4 -SueW I Q Undergraduate school?
2 j! A Yes.
l 3l Q Have you had any course work in heat transfer?
I 4 A No.
I 5 Q Have you had any course work in fluid mechanics?
4h A Yes.
/ I! Q Where was that?
6 B A That was at the University of Toronto.
j -
90 Q How many courses did you have?
10[ A Two.
I Q Undergraduate?
II A Yes.
73
\- U Q Do you have any knowledge, or skill, or experience,
'l or training or education in the field of heat transfer from
~3* nuclear fuel to reactor coolant?
5 A Insofar as this is an essential ingredient of nuclear safety considerations, yes, I have had experience i with regard to the inquiries that have been previously mentioned.
'O And what experience is that, sir?
Q M A That is experience is reading official documents, 7 safety analyses, pouring over safety calculations and produc-23 ing critiques of those.
!p/ ,' , ,
Q Have you read the final Safety Analysis Report for Ace F:_.at Atoorters. Inc.
25 the Turkey Point Plant, Units 3 and 4?
l
, 265 l
'l l A No, I have not.
4-5g-u;W' I W 2 L ..
Have you read the. Staff's Safety Evaluation Report I
I Q
3I for the Turk'ev Point Plant, Units 3 and 4?
I 4i A Yes, I have.
5 0 When did you read that?
6 A I read'that approximately.two months ago.
7 Q How many volumes was that_?
i A We are talhing about the_SER?
8ha 9;! Q I'm' ju'st talking about wh'at you read.
i il 10j A If we'are talking about the SER, the Safety Ilf :
Evaluation Report'frcm the Staff for Un'its 3 and 4, it's I2 ) simply one volume. It's one document
(~)/
N- 13 Q Have you ever visited the Turkay Point plant site?
.j
'J A No, I 'have not.
- 5 Q Have you ever designed a heat transfer system?
i 26 ] A. No, certainly not. <
i7 Q Have'you ever been personally ' involved in, or M. . participated in,the conducting of departure from nucleate
. r i
"* ' boiling tests?
2d A No. ,s 2' Q Have you ever been personally involved in, or
?? participated in, the development of a. departure from nucleate 23 boiling correlatidn?
22 I t j A' No. .
W F'LJ 4000rtert, Inc.
25 1 Have you ever personally designed, or participated Q
,i i: .
I-
266 f
44 -SueW l in, the thermal hydraulic design of the core of a nuclear 2 l reactor?
3 A No.
4g Q Have you studied or read the technical specifica-5 tions for the Turkey Point Plant?
6 A No, not all.
7 Q Are you familiar with the Nuclear Regulatory 8i Commission's Standard Review Plan, NUREG 0800?
II -
I 9
I A I'm familiar with some portions of it as they 10 ' apply to the DNBR question.
11 What portions are those?
Q 12 A Those portions which are referenced in the
- ) 13 l l
documents surrounding this particular case.
14 Q Do you recall at this moment what portions they 15 are?
16 1 A Well, for example, the 95/95 criterion is laid 7 down in that document. That's a crucial element of the
- d argument over the acceptability of the DNRB limit of 1.17.
'9 Q You've read that portion of the Standard Review 20 Plan?
21 A Yes.
?2 Q Have you read any other portions of the Standard 23 Review Plan?
24
( _) 1 A I have skimmed. I have not really concentrated on Ace Federd Reporters, Inc. j 25 l any portion except that, because that seems to be the crucial n
o l)
267
.4 -SueW 1 one. ;
-~
i-2 j Q Have you ever designed or used any computer modele 3
to perform thermal hydraulic analysis of,1h eat transfer and ,
4 the fluid flow aspects of a. pressurized water reactor?
.A No, I have not.
5 Q H. ave you ever determined the core operation limits 'i 6
of a pressurized water ' reactor to assure that DNB does not 7
t ,
g occur? ', '
9j A No, i
10 . !
Q Have you ever calculated departure from nucleate
- 31 boiling ratio acceptance limits for a pressurized water
- 7 , reactor? ,
V 73 d- A No.
, 3 ,' MR. COLL: I have no further questions, Your-3-
Honor.
i4 ! JUDGE LAZO: All'right.
e s
7 MR. COLL: ~ i I have;an objection to make after
\ \
3 Staff voir dire' if they have questions. ,
19 [ JUDGE LAZO: Yes. s I
i!
79 - MR. HODDER: As to-this cross, I would like to
- - just pose a couple of brief questions to the witness, Your;
- f 72 Honor, or I can wait after the Staff completes their questions.
23 JUDGE LAZO: It might be^ simpler to do it now and s
(] 24 i AssJeder'A Aeoorters, Inc. !;
have it in the same place in the transcript.
25 MR. HODDER: All right.
\ ,
-a. --. , - . - - - .
l', . \
268 l
- 4-8-SueW' l' REDIRECT EXAMINATION p'
2 i- BY MR. HODDER:~
Y 31 Q Dr. Edwards, you have indicated in responses to t
l 4! Mr. Coll's cross-examination that there are technical studies 7
( ,
5 . such as --
r 6 MS. YOUNG: Judge Lazo, before Mr. Hodder goes any c ;t, 7 further I am uneasy about this procedure. Normally, you have
[4 9 direct from a witness, you have cross-examination from the.
9[ other parties in the proceeding, and then you have redirect.
a 10 ' Now, Mr. Hodder is being asked to have the opportu--
i .
11 j nity to redirect before the other parties have completed A
pr '12 d' cross, could you reconsider your --
^
() 10 - JUDGE LAZO: Well, of course, we want to' hear from b !2 the Staff.
- Basically, this is still voir dire.
?1 ; MS. YOUNG: Right.
Ml JUDGE LAZO: And we've completed the Licensee's
,7 questioning of the witness.
d MS. YOUNG: The Staff also has voir dire.
M
, p (The Board members are conferrin,g. )
-1 20 h JUDGE LAZO: Very well, we have reconsidered based I 'b '2[N on your request. And why don't you proceed? Or, is Ms. Wagner
?I going to do the voir dire?
23 MS. WAGNER: Yes, I would like to do the voir dire.
l$, s. t
.(~)' 2#
Am.F__ a neoo,tm. inc. ,i JUDGE LAZO: Okay.
l 25 ji MS. NAGNER: I think most of my questions have
- i e
a e'
.P-
269 I
.#4-9-SueW. I already been asked. 'Please bear.with me if I end up repeating
-(v) 2 I a couple of questions that have already been asked.
3 CROSS EXAMINATION INDEXX 4 , BY MS. WAGNER:
5 Q Dr. Edwards, have you taken any graduate courses 6 in fluid dynamics or thermo dynamics?
- t 7 A No, I-have not.
8 Q Or heat tr,ansfer or two-phase flow?
9 A " No.
~
10 Q Have you taken any undergraduate courses in i
11 heat transfer or two-phase flow?
12 ! A No.
l'I'
\_ 13 Q You have never taught any courses in fluid dynamics, 1
': c thermo dynamics, heat transfer or two-phase ficw, have you, 15j' sir?
15 , A No, I have not.
i 17 Q Have you written any articles on any of those four
- o ! subjects?
i I A Only insofar as they are related directly to the
'9 q 20 question of nuclear safety. And these articles have not been 21 for publication. They have been for the purpose of my consulta-22 ' tion.
23 Q Could you enumerate the articles in which you 24 wrote on these subjects for me?
Am-F9a neoonm, ene. [l 25 l A Well, there was a summary argument that was.
il
270 1
9 p-SueW. I presented under my authorship to the Royal Commission on 2 Electric Power Planning which dealt with questions specific 3 to the CANDU reactor design. And they did involve accident 4 conditions, transients;and,some of the probability calculations 5 related to failure, cladding and so on.
6 The other document that I produced was a document 7 for the Select Committee on Ontario Hydro Affairs, and this 8 was primarily a critique of a document produced by Professor 9 Rogers of Carleton University on heat transfer and two-phase 10 I flow in the CANDU setting.
11 Q And you authored that article?
12 : A Yes,~I did.
O 13 ,
Q It was not co-authored?
14 / A No, it was not a co-author.
15 ,j Q Did you consult with any experts in heat transfer 86 in preparing that article?
i IT A 'I did not, because the article dealt with mathe-iS matical considerations which I felt confident to deal with.
!9 l Q Were these articles intended to aid a layman in i
20 j understanding general reactor safety issues?
21 A No. These articles were intended to assist an 22 :, inquiry process in determining the truth about the reactor 23 safety characteristics of the CANDU reactor.
24 ii Q Did -- were the results of the inquiry published in Am-Feder:;l Reporters. Inc '
25~ a document made available to the public?
II a
271
- 4- -SueW j A There was a final report, yes. And'also the 2 transcripts were certainly published as well.
3' 0 Nere the members -- what was the composition of 4 the members of the inquiry board, technical expertise?
5 A In the case of the Royal Commission on Electric 6 Power Planning, the Chairman was a Professor of Engineering 7 from the University of Toronto, Dr. Arthur Porter. And there were two other people.who were on the Royal Commission. One was 8l, .
xx 9; Solange Ploarde-Gogrion, who was I believe an economist. And i
10 ! the_other gentleman was also an economist with specific 11 experience in the utility field, in the electric utility 12 field.
I
' i In the case of the Select Committee on Ontario
- 3 ,
- a Hydro Affairs, this was an all-partv committee of the Ontario i
- 3 i
Legislature which had legal staff and which was inquiring
- into the questions of nuclear safety for the benefit of
,7 the Onotario Legislature.
g' Q Was the Royal Commission report prepared for the
,1 29 I .
purpose of cost analysis of reactor safety?
l' 20 A No. The -- that was one consideration. But the 7; Royal Commission on Electric Power Planning was actually'a 77 comprehensive four-year Royal Commission inquiry into all 22 t
aspects of electric power planning.
() 24 Acs Feder:t f:eoorters, Inc. ;,
An important constituent of that was the nuclear, 25 q because Ontario Hydro has a sizeable number of nuclear power d
h e
I 272
~
.1 . plants. And so thev did issue a secarate report in September
- 4<kJ'9-SueW 2 I of'1978 specifically on the nuclear question.
3 And this report did deal with economic cost 4 ~ comparisons but it also did deal with the spectrum of --
5 for example, there was a chapter on' reactor safety, there was -
6 a chapter on waste disposal, nuclear waste disposal, and so 7 on. So, it covered the entire spectrum of issues related to g nuclear power. -
9; Q And your input went to which chapters of the i
ic i report?
11 A My input was fairly broad in the terms that most 12 strongly.it was felt in the chapter on reactor safety and
.O 13 also in the chapter on waste disoosal.
- u. These were the two chapters in which the -- my 15 ,l particular contribution was felt more strongly.
16 Q Have you cross-examined witnesses in the area of i
., fluid. dynamics or thermo dynamics?
y A To some extent.
19 O Could you expand on that, please?
I 20 A Yes.
21 , Q When did you cross-examine these witnesses and 22 9 what was the nature of that?
23 A There was a lengthy period which went on for a O* 24 ' total of about -- I believe it was almost_ fifteen days of u.sw.rsneooneri,ene.;;
251! cross-examination of a succession of technical experts from 4
e
273 t
?#13-SueW'1 Ontario Hydro and from Atomic Energy of Canada and the 2 i
_ Atomic Energy Control Board dealing with accident circumstances 3 in CANDU reactors. And during'this process, there was a good j
7 4 deal of discussion of various transients and various-situa-Si tions where you would have dry out occurring, problems with 6 rewetting of the fuel, problems of pipe breaks which lead
-7 .to, of course, super heated. steam and so on.
3 Q And what did you do to prepare yourself to conduct 9 those cross-examinations in those areas?
O A My main preparation consisted in studying the i
il available documentation which the industry itself was using 12 ' as a basis for its argumentation. And some of those documents
~
Uf were U.S._ documents, because Ontario Hydro, having never I ' h: conducted
- never a major reactor safety study dealing with the r
i: -
. spectrum of large-scale accidents, drew rather heavily upon i the Rasmusen Report from the U.S.
1
.I And consequently I used a great deal of information
.J
..ii from the Rasmusen Report.
T Q Did the reactor safety study include determination
.9! of core safety limits?
[i t
A Would you mind defining what you mean by core i- safety limits? Are you talking about parameter trip values M' . and things of that sort?
) .
d- No. It certainly did not deal with parameter trip
- AmJedefel Aeoorters. W. '
25 ' values. It dealt with questions of probabilitics and
1
.-~. ')
274 )
l
- 6-%3-SueW I consequences of accidents primarily..
' h-) -
2 i Q Have you studied the behavior of a reactor 3: core'or associated systems under normal or transient l
4 -conditions?
l 5 A Yes.
6 Q Have you~ studied the modeling of the behavior of 7 a reactor core ~or associated systems under normal or transient -
a conditions? .
9 A The model.
C Q The model?
I] A *To a limited extent, I have.
12 i Q Is this-informal course you are referring to?
- ! A No. This is not informal course work. This is l'
experience derived from the experiences which I had as a
- r. ; consultant.
I I
L : 0 Mostly in connection with these two --
i i-L A That's correct.
O ish Q -- the Royal Commission and the Ontario Hydro?
. il A
'l That's correct.
c -
H! Q Are you familiar with the term " boiling curve?"
i D
0- A I'm not familiar with that term, " boiling curve,"
no.
I's Q Are you familiar with the term."CHF?"
) 24 A Yes.
w .a m ers iaeoor en, + c..
IS I Q What does that term mean to you?
4
, -- ,, _ ._ c ,_._- ___,._ __ _ __ _ ,,_ _ . m. ._. .-_.,. . . . , . . -
275 i
- 4 4-SueW l' A It means critical heat flux.
2 i Q And could you explain what critical heat flux is?
3, A Critical heat flux refers to a certain value of 4 heat flux which, as the fuel runs hotter the heat flux in-5 creases. You get what is called nucleate boiling, which 6 is particles of steam forming upon the cladding of the fuel 7 which actually enhances the cooling action of the flow.
3 However, as the heat flux continues to increase, 9 the boiling increases and you achieve a certain heat flux 10 which is called the critical heat flux, whereupon these lj individual bubbles coalesce into a film of steam which acts i
12 l as an insulating blanket and thereupon you have a rapid in-
/3, t
('~'I R j, crease in temperature. That is a critical heat flux.
I: A critical heat flux is that heat flux at which d departure from nucleate boiling occurs, lo i 0 Are you familiar with the term " film boiling?"
l
- l
.y '
A Yes, I am familiar with the term " film boiling."
t.
a6
[ Q Could you briefly state --
ip A Well, film boiling is the phenomenon that I des-t 0l cribed. When you have departure from nucleate boiling you i
l' then have film boiling. And this basically creates an in-sulating blanket, if you like, of a film of steam around the
- 2 cladding which inhibits the efficient transfer of heat; and,
/ 2a consequently results in a rapid increase in temperature.
Aa Federb Aeoorte 5,1-c.
25 0 Are you familiar with the term "subchannel analysis?"
c C
i
276 1
4- SueW l A I'm not perfectly clear on that term. These terms, 2 L unfortunately, as you realize, the geometry.of the lake 3l -
water reactors in the United States do not conform to the 1
4 geometry of the CANDU reactor and there are different terms e
5 that are used in the two.
6 Subchannel analysis, I am not perfectly clear on 7 what this means. But I believe that it refers to the analysis 8 of what goes on within a particular bundle. I may be wrong 9 about that. ~I would have to consult my notes.
10- Q You said earlier I believe that you had read the II NRC -- portions of the NRC Standard Review Plan; is that 12 correct?
O 13
- A Yes, I've read portions of it. Yes.
I !
14 =
Including Section 4.4, Thermal and Hydraulic Q
\;
I: l Design?
li h (( A I'm sorry. I can't recognize that title particu-87 larly.
UU Q That is the section that discusses --
" A Perhaps if I saw the section.
20 Q This is the section that discusses the DNBR 2I ' , acceptance criteria?
12 h A Oh, well then I did read it, because I read the --
23 if this is the section which deals with the 95/95 criterion 24 and the derivation of the 95/95 criterion, then, yes, I did y ., - . . ....i,.:
25 read it.
1 l
.l t 4
277 4- SueW 1 Q That is correct. That's the section.
2 i can you tell me what a hot rod is?
A The hot rod is the -- it-really refers to the 3l 1
4 hottest rod in the core. Some of these rods, of course, 5 are going to have more burnup than others. And generally 6l speaking, becauseaaf the geometry, because of the buildup 7 of fission products, because of the amount of fissionable 8 material remaining in a fuel rod, these rods do not all 9 run at the same temperature.
10 The hot rod refers to the hottest rod.
END #4 11 Mary flws I) 13 ,
3 1
14l l
15 !
I 15
- i i7 j is ic 20 21 72 h i
23 !
I
( 24 Aas-Feder3 Reporters. Inc.
25 l
1 v: t.
l I
173
~ '
3 0 Unat 1s the CHF correlation?
2 A The CHF correlation 1s. 11 I may put it in l 5:
3 loose terms, it is a carefully worked out rule of tnumb Ter f 4 est1 mating what the CHF really is. The difficulty is that 5
we do not know exactly how to calculate the precise
.y: J ', 6 cr1tical heat flux. And so a correlatio7 1s a mathematical 7
tool for allowing us to calculate a pred1cted crit 1 cal heat 8
I "** ,
9 This then has to be compared with experimental
.b.
10 results to see how accurate this critical heat flux 2
gg correlation is.
12 Q Can you tell me how DNBR 1s calculated is the' sube annel analysis code?
13 g4 A I am sorry, I can't.
15 0 Do you have some generally knowledge of the g transport of fission prooucts in uranium diox1de?
37 A I have some knowledge of this, yes.
18 0 Can you tell me the basis of that knowledge?
g9 A The basis of tnis knowiecge again, and I have to keep going back to my experience in these various 20 inquirtes, I hac access --
I as privileged to have access 21
^
22 to a great many documents dealing with reactor safety, and 23 among these documents of course, particularly after the 24 Three M11e Island acc1 dent, there was a great deal of 25 discussion of how much fission gases would be present in
! h
\J e
r1 279
(,, i the sap between tne fuel ana the claading. So tnat, for 2 . example,-if you did.have a cladding failure, new mucn of 3 tnese fission. gases would oe available for release.
4 This cecame an issue in Canada Decause one of 5 .our plants, one our older plants, tne Douglas Point Plant, 6 .as w derated to 70 percent of full ocwer because of an
_7 inadequate containment ou11 ding, and the Atomic Energy 8 Control Board, which, plays the role of the Nuclear 9 R.39ulatory Commission in Canada, derated that plant to 70 jo percent partly because, not only does it give a greater ti- margin of safety in terms of various transients, but it 12 also means that there is a lot less fission products in a
_j3 releasable-form in tne event of any -t i n d of failure.
L g4 G Have you every consulted'as an expert In tne 15 areas of fluid dynamics or tnermal :ynam1cs?
l 16 A No, not spec 1f1cally.
l l' 17 C Nor in heat transfer or two-phase flow?
I 18 A No.
I g9 Q What documents nave you reac 1n preparati:n
! '20 for conducting your crcss-examination nere today?
21 A The documents that I have read nave caen the 22 documents wnicn nave oeen provicec by tne licensee 1 .1 terms 23 of trying to nave tne intervention otsmissed, and the 24 affidavits tnat nave'been suomitted, tne SEF, anc I . ave s 25 question'of approximately half of none occuments on my.
i
280
'Ll
(')
j desk over'-there.
2 0 Did you review all of the documents proviced 3
by the licensee in connection witn tnis proceeding?
4 A Well, I am a little~ bit at a disadvantage mere 5
because I can't swear that I did see all of the documents.
I certainly examined all of the documents that were sent
'6 to me by the intervenor, Ms. Joette Lorton. She sent me.
7 8
voluminous documenta, tion upon numerous occasions, and all 9
of the documents that I received, and all the documents which I saw referenced, there were some of the documents 10 3; which I saw referenced which I was unable to find in my package.
12 p 33 Q Which you saw referenced by the licensee?
V g A Yes, tnat is rignt. Set all of the direct documents wntch were forwarded to me by Ms. Lorton I have 15 read.
16
_p Q The documents referenced tnat you were'unaole 18 to find, di' you request of Ms. Lor 1on those dccuments? 01c gg you make a request for those documents?
20 A Some I did and some I got. The d1i11cuity is that -- and I w111 give you an example.
21 22 When you talk, for example, about tne 23 acceptab111ty of tne 15 by 15 OFA fuel assembly, that is 24 the DNBR ratio 11m1t of 1.17, the acceptacility of t.at for 25 the-15 by 15 0FA fuel assembly, there is a letter. .a n d O
281 i accompanying that letter tnere are tour pages whicn are 2 really undated, but which are attached to the letter. Anc 3 following tnose four pages there are four or five 4 references. Now I don't have those four or five references.
5 In fact, I am looking forward to asking some 6 questions about information conta1ned in the four pages, 7 Out the real information 1s in the references, and I don't 8 have those references. So it would certa 1nly be wrong for 9 me to say that I nave availability to all of the documents 10 referenced. I don't.
13 Q Goln3 Dack to the -- 15 it the Rasmusen 12 Report tnat you referenced earlier?
A Yes.
~lh
.- ja Q Ooesn't that report :<imarily deal with the 15 probability of severe accidents?
16 A 'f e s , that is correct. But I might acd, I nac 17 the pr1v11ege of cross-enamining Dr. Rasmusen at the C'uff 18 Lake Doard of Inqu1ry and subsequent!/ v1 siting him in his 39 off1ce at Massachusetts Inst 1tute of Techno'.ogy and ne gave 20 me a good deal of informat1on. He was very nelpful in fa:t.
21 But in the course of tne Rasmusen Re: ort, 22 although the bottom line is severe acc1 cents, as a matter 23 of fact, one of tna interesting find 1ngs of the Rasmusen 24 Report is that small pipe breaks 119ure very or:m1nently in 25 the overall probaD111ty of large a:c1 dents, anc tnese small
282 (j., j
, y pipe breaks in themselves do not necessarily constitute a 2
major accident.
3 For example, the Kemeny Co.mmission on Three 4 Mile-Island found that technically the Three Mile Island 5
accident was a small pipe break accident even though there 6
was no pipe break because the stuck pipe functioned as a 7
small. Pipe break.
So there is a great deal of information in the 8
9 Rasmusen Report.which leads to a greater understanding of 10 what-sees on inside the core of a reactor under various 11 circumstances and not necessarily severe accidents.
12 Q Was'one of the primary focuses of the Rasmusen Report a discussion of DNBR7 13
(-
.s-. y A I do not remember a discussion of DNBR. Now I would n t be at all surprised if there were ir the Rasmusen 15 g Report, which is a voluminous report, a discussion of DNBR.
37 At the time that I read'the Rasmusen Report, 18 that was not one of my concerns, and at that time I j9
.certainly do not recollect a discussion of DNBR in the Rasmusen Report. If there were, it must have been in the 20 appendices because it certainly wasn't in the main text.
21 0 I" *"Y - event, a discussion of DNBR was not a-22 primary c nsideration of the Rasmusen' Report?
23 A Not to my knowledge.
24 Q Just to make sure that I understand your 25
, rm;G GJ
283 R testimony this morning, you do consider yourself an expert
_ i 2 in mathematical analysis and calculations of probability 3 and use of mathematical modelsi is that correct?
4 A Yes.
5 Q And you consider yourself generally well read 6
on general nuclear safety issuesi is that correct?
7 A Yes.
8 Q And you have been an interrogator on general 9 nuclear safety issues?
10 A That is correct.
11 Q But you do not consider yourself an expert in 12 heat transfer, DNBR testing, CHF correlation, determination j3 of operational limits or evaluation of DNBRi is that i L-L-- ja correct?
-15 A That is correct.
16 MS. WAGNER: Thank you.
17 I have no further questions.
18 JUDGE LAZO: Dr. Edwards, just let me ask, in j9 response to a question by Ms. Wagner regarding documents 20 and affidavits that you had read, you stated that you had 21 read documents provided by the licensee.
22 I wonder what about documents p ovided by the 23 NRC staff?
24 THE WITNESS: Oh, yes.
25 JUDGE LAZO: I thought you might have misspoke.
j .i-
'A
284
.f2'i og
_j THE WITNESS: I am sorry. No. I rece1ved 2
doc'umints which had to do with the intervenors' contentions 3
and-affidavits supporting them, I mentioned the SER, and 4
that is the only staff document I mentioned, but indeed I received the affidavits from staff as well, and I received 5
6
'the affidavits that Mr. "C" provided, for example, and also 7
I~ received the Information that the Intervenor herself-has been producing.
8 9
JUDGE LAZO:' We just wanted the record to.show 10 that that was correct.
ij THE WITNESS: That is correct.
JUDGE LAZO: Now, Mr. Hodder, I think we should 12 provide you an opportunity to question your witness if you -
33
's h (J )- gj desire.
MR. HODDER: I have'no questions.
15 g JUDGE LAZO: Mr. Coll, you indicated you-had an
- 7
' objection.
18 MR. COLL: Yes, sir. Judge Lazo, we would j9 object on behalf of Florida Power and Light Company to this 20 witness being used as a technical interrogator in the areas 21 in which he has admittedly admitted he has no expertise.
22 We believe that he has not showed, in fact he' has admitted that he is not qualified by ' scientific 23 training or technical training or experience in the areas 24 of thermal dynamics, heat transfer, fluid mechanics and 25
.q D.
T 285
( g i thermal hydraullc analysis, including areas involving the 2 calculation of DNB and CHF, all of which are the crux of 3 the Board's narrow prequestions in this case which involve 4 procedures and techn19ues used to calculate those values, 5 whether or not certain of those values contain 6 uncertainties or not and whether or not the 95/95 standard 7
of the staff's standard review plan 1s met.
8 We believe that both under the rule, which is 9 10 CFR 2.733, which counsel for the staff has read, and to also under case law, such as the Peach Bottom decisioni 11 which is in ALAB 701, that this witness has not been shown 12 to be qualified in those areas, and we would object to him j3 be1ng a technical interrogator in this proceeding.
14 JUDGE LAZO: Ms. Wagner?
15 MS. WAGNER: The staff believe that if Dr.
16 Edwards' cross-examination is lim 1ted to the areas of his 17 admitted expertise, that is areas relating to mathematics, is including mathematical analysis and calculations of 19 probability and use of mathematical models, we do not 20 object to his conducting cross-examination.
21 We believe that he would meet the three 22 findings of 2733 in terms of qualifications and 23 preparation, et cetera.
24 We do object to any cross-examination beyond 25 those areas because 2733 expressly provides that
.286
)'N j cross-examination pursuant to that section shall be limited
%. ) .
2 to areas within'the expertise of the individual conducting 3 the cross-examination.
4 (Board Conferring.)
5 JUDGE LAZO: The Commission's provision in 6
10 CFR 2.33 is clearly intended to expedite ---
7 MR. HODDER: Your Honor, before'you say 8
anything, I didn't respond to these-motions. I don't know 9 if th,e Board'is prepared to rule at this instance or not, 10 but we ---
n JUDGE LAZO: We would be prepared to rule, but 12 we will provide you with an opportunity to comment on ---
13 MR. HOODER: You got into the huddle so fast
) , 34 .that it took me ---
(Laughter.)
15 16 JUDGE LAZO: That is what I was saying about 37 expediting the proceeding.
18 (Laughter.)
j9 MR. HODDER: Right.
20 (Laughter.)
21 MR. HODDER: I would merely wish to point out.
22 to the Board that all the witness has conceded here this 23 morn 1ng is that he hasn't performed original calculations.
such as the computer models or the other co-relations 24 performed by the vendor or specialists of the staff.
l 25 n
hJ u
287 P He indicates that his expertise is sufficient gf") i V
2 to allow him to make judgments about all of those areas, 3 Including thermal hydraulic analysis and thermal dynamics 4 and even nuclear physics by his training and his 5 experience.
6 He is certainly .more qualified than I am, as I
.7 will attempt to stumble through these questions. And I must 8
say that although I am pleased that the staff agrees that 9
he should be allowed to be a technical interrogator, their to suggestion that there be imposed limits upon the areas of li his cross-examination, which wouldn't be imposed upon me 12 who has less skill, is an unfair one.
13 I think that the motion by the utility compqny LIs ,k 14 is unfounded and frivolous and really ridiculous. I think that the witness has demonstrated that he possesses the 15 16 expertise to address these issues and conduct this 17 technical cross-examination.
ig I think the key to that consideration is 39 recognition of the fact that these mathematical exercises 20 are susceptible to checking by other mathematicians. Today 21 we have one in the form of Dr. Edwards who is eminently 22 qualified to do that interrogationi and I think that it 23 would be a great boon in assistance to this Board as it 24 conducts these hearings to allow Dr. Edwards to cross on
~
75 all the areas identified by the Board.
CD
288 tv ; ) : Thank you.
~wi 2 JUDGE LAZO: Thank you, Mr. Hodder.
3 This provision which has been referred to is 4
clearly intended to expedite the proceedings, and there-are 5
many instances when your expert could sit by your elbow and 6
use the ventriloquist method of providing each question and 7
then having you ask it. That is cumbersome-at best, and we 8
do think that the voir dire examination has demonstrated 9 that Dr. Edwards can certainly assist us all in building a 10 complete record in this proceeding.
g; I think we do agree with you, Mr. Hodder, that 12 it would be very awkward to limit the scope of the y) examination. If we should run into problems there, they
.g .
E.ax _) g, would be general ones of getting beyond the scope of the issue. But we see no reason why Dr. Edwards cannot conduct 15 the cross-examination in all areas, if that is your desire.
16 37 I do think it will contribute to the record, 18 and it will be a worthwhile method of proceeding.
g9 MR. HODDER: Thank you, Mr. Chairman. That is 20 my desire as we approach these technical issues, the three points identifisd by the Board.
21 22 JUDGE LAZO: Well then we will deny the 23 object. ions provided by the parties and state that Dr.
Edwards is qualified to conduct expert examination under 24 the provision that we referred to.
25 p
,5 l t
l
289
([ -
1
-THE WITNESS: Thank you, Mr. Chairman.
2 JUDGE LAZO: You are welcome.
3 THE WITNESS: May I resume my seat?
4 MS. WAGNER: Judge Lazo, excuse me. Are you 5 going to then rule on the areas within Dr. Edwards' 6
expertise? I assume you have made the three findings under 7 2733.
8 JUDGE LAZO: Yes.
9 MS. WAGNER: Are you going to rule as to the io areas within Dr. Edwards' expertise?
11 JUDGE LAZO: I think we are going to wait and 12 see if we develop any problems in that area. We think it will be difficult to divide the scope of the examination. I i3
<) ja would rather start.out letting him examine in all areas 15 which are within the scope of the contention.
16 There may be areas where he is less qualified 17 than others, but Mr. Hodder has indicated that he is 18 certainly more qualified than Mr. Hodder is in these 19 technical areas-and that the proceeding would be slowed if 20 we have to wait for questions to be passed from the expert 21 to coun.sel.
22 Well then let us ask now as to proceeding. We 23 would of course be happy tJ hear the views of the counsel, 24 but I would think that the next thing to do would be to 25 introduce the licensee's direct testimony and then proceed
l~
290 s .
f'S with cross-examination of that testimony.
- Q.
g Dr. Edwards, w'hy-don't you resume your seat at 2
3 the counsel table.
4 THE WITNESS: Thank you very much.
5 JUDGE LAZO: And then we would receive evidence from the NRC staff and have cross-examination on that, and 6
then the intervenors' testimony following the testimony of 7
8 the staff.
9 Does that order of proceeding seem agreeable to the parties?
10 gg MR. COLL: That is fine with the licensee.
12 MR. HOODER: That is fine with the intervenors.
MS. YOUNG: We will proceeding in that order 13
) g4 throughout the hearing?
JUDGE LAZO: That has ---
15 MS. YOUNG: In other words, will staff's 16 gy . questioning always precede the intervenors?
-JUDGE LAZO: You would prefer to go last?
18 MR. HOODER: The intervenors prefer to go last, 39 20 Y'S' MS. YOUNG: Maybe I am confused. I will ask 21 22 again. Do you always prefer that the staff ask questions 23 last~in terms of each witness comes in and testifies?
JUDGE LAZO: We are flexible, Ms. Young. How 24 25 would you prefer to do it?
g 4
291
(, 1 MS. YOUNG: Well, I had antic 1 pated that the 2 staff would go last in asking questions. I understand ---
3 JUDGE LAZO: Generally that is what we do, the 4 staff goes last.
5 MS. YOUNG: But the order of presentation of 6 witnesses w111 be the licensee, the staff and then the 7 intervenorsi is that correct?
8 JUDGE LAZO: That is what we were proposing, 9 unless there is a problem.
to Well, I wonder if this might be an appropriate it time to take a brief mid-morning recess, and then we will 12 come back and proceed w1th the licensee's witness.
i3 Mr. Coll, how many witnesses will the licensee
- 4 have?
15 MR. BAUSER: We will have only one witness, 16 Judge La:o, and we are prepared to proceed now or after a 17 break.
18 JUDGE LAIO: Very well i Mr. Bauser. Thank you.
19 Ms. Young?
20 MS. YOUNG: Before we proceed, is Mr. Hodder 21 going to move his exhib1t into evidence, the exhibit that 22 we had questioning on for the vo1r dire? I don't know if 23 this would be a good time to enter it in the transcript. It 24 really hasn't even been marked for identification.
25 JUDGE LAZO: Well, part of the problem is the D
292
(~N y qualification statement is an exhibit attached to-the
' \_)
2 testimony, and if we try to get' the testimony in now', we 3
may run into another series of objections. ,
4 MS. YOUNG:. Okay. So we are not having a 5
separate exhibit from his testimony theni is that correct?
JUDGE LAZO: Let's keep them together. When the 6
7 proposed testimony is offered, it will contain that Exhibit A. If necessary, we will have to separate them and put 8
9 Exhibit A in all by itself.
10 MS. YOUNG: Thank you.
JUDGE LAZO: Well then, it is 10:45. We will 12 recess until 11 a.m.
13 (Recess taken from 10:45 a.m. till 11:00 a.m.)
. i - [3
- -Q.
34 15 16 17 18 19 20 21 22 23 24 25 ',
<w
'.h
6-1-J;cWcl ll , 293 XX INDEX y .
(11:03)
- 2 JUDG3 LAZO: Would the hearing come to order, 3
please, g Mr. Bauser, are you ready to proceed?
'i' 5
MR. JAUSER: Yes, we arc. I ask Mr. Dzenis to l
6 take the stead.
N 7 Whereupon, ;
XX INDEX 8 FDW(.PD A. DZENIS, s
i 9 was called as a witness, ard having first been duly sworn by 10 Judge Lazo, testified as follows: ,
11 l DIRECT EKAMINATION i s X INDEX 12 ! BY MR. BAUSER: '
T^ ? !
13 [ Q Do you have before you a document entitled Tactimony ,
14 of Edward Z. Dzenis Concerning Contention (d)?
15 A Yes, i'do.
I 16 ' Is this the document that was distributed to the !
Q l,
17 ' Board and partica in this proceeding under my letter dated 18 November 26, 1985'? f
'O A Yes, it is. !
I 20 l Q In this the testimony that you pre. pared as your
~
.i .
I 21 1 testimony in this proceeding?
?2 [ A Yes, it is.
23 Q Do you have any additions or corrections you would,
(~', , 280like to make to this document?
OFi J Reporter $, Inc. I 25 A Yes, I have one' change. On page 5, the first l l
I l ~
3 6-2-JocWol 294
} complete paragraph on that page, Line 6, the section after
'/ 3
\)~ ~2 the. word, Parts, delete II.6, II.7 and II.8, and substitute, 3 II.4 and II.5'.
4 Q Would y'ou read that sentence as corrected?
5 A Procedures and techniques employed in the calculation 6 of DNBR values for this spectrum of normal and anticipated 7 operational occurrences are in accordance with Section 4.4, 8 Parts II.4 and II.5, and the appropriate portions of Section 15 9 of the Standard Review Plan 10 Q Do you have any other additions or. corrections you 11 would like to make?
12 ! A No, I do not.
O).
(_ 13 MR. BAUSER: Judge Lazo, I have handed the reporter 14 copies of Mr. Dzenis' testimony as corrected. I move that his 15 C testimony be made a part of this record as if read, and that F
16 it be bound to the transcript.
17 JUDGE LAZO: Are there any objections?
18 MR. HODDER: Yes, we have an objection. My under- l i
0 standing is that the testimony includes as an exhibit the F
20 ll prof essional qualifications and experience of Edward Dzenis, i'
21 and before we agree to coming in for its truth, we would like 22 to have some voir dire of the~ witness.
23 JUDGE LAZO: That is perfectly acceptable.
(' 24 MS. YOUNG: Judge Lazo, I would also state at this Si_)) Reporters, Inc.
25 time I don't believe Mr. Bauser asked Mr. Dzenis to adopt the L
6-3-JocWal 295 1 document as his testimony.
x 2 MR. BAUSER: I asked the witness if he prepared this
- 3. testimony as his testimony in this proceeding. The witness iianswered, i,
yes.
5 l
By that, I meant to ask if he adopted that testimony 6 as his testimony in this proceeding.
7 JUDGE LAZO: Do you adopt it.; l 8 WITNESS: Yes, I do. '
9 MS. YOUNG: The Staff has no objection to receiving i
10 the testimony, by the way.
11 ! MR. HODDER: Shall I proceed?'
l 12 j JUDGE LAZO: Yes, sir.
~~ !
i XX '~.s; 13 l VOIR DIRE EXAMINATION l 14 '- BY MR. HODDER:
15 Q Mr. Dzenis, I see that you claim expertise in 16 thermo-dynamics, and thermo-hydraulics. Do you also claim l'7 expertise in nuclear physics as it relates to reactor design?
18 A I have some expertiese in the area of' physics in <
'O reactor core design.
I 20 [l' Q Could you tell us what expertise in nuclear physics j-21 it is that you have? Identify your experience or training.
22 A I have taken course work in ine.roduction to nuclear 23 engineering as an undergraduate. Since my responsibilities 24
(" )
4m-Fk,) Reporters, Inc.
as manager of thermal hydraulic design .when we originally 25 submitted this material, I have been responsible for department l
l l
~6-4-JocWel 296 lr j involving the nuclear design analysis of specialized water
. _ i m 2 reactor.
3 Q First of all, that responsibility you just identified ,
4 is that in your employment at Westinghouse?
I 5 A Y e s', it is.
6 Q Okay. Getting back to your training, you indicated y you had an undergraduate -course in introduction to nuclear .
8 'eigneering, is that true?
9 A That is correct.
10 Q Do you know how many credits that course was, or
- 1 1' its duration?- ~[
12 l A It was a three credit course..
i
.() 13 l Q Is that the only undergraduate nuclear study you 14 have been involved in?
15 l A That is correct.
l , i
-16 -Q Have you had anything at the graduate level.that j i
- 17 ; would be in the area of nuclear physics? f i
. . , 18 A In the -- as a graduate student I took two courses !
Il
'!0' in the department of nuclear engineering at Carnegie Mellon E
20 ]' University, - three credit courses.
.( L1 21 Q Could you identify the subj ect matter of 'the
! 22 courses? What their title is?
l--
23 A -Titles, as near as I can remember them were-Thermal 24 dynamic power conversion cycles, and environmental regulatory s>A(_}amortus, Inc.
25 economic aspects of nuclear power.
i t
1
- 44 + ~. - , . , , ,r .<...,-.E.-.4--,....-4 . . . , , .. ew #. -w mee. -e .w . n.r e , e e e era e ew -
16-5-JooWol 297-I Q Those wouldn't address nuclear physics then, would 2 they?'
3 A Only as they relate to particular matters in those 4 particular course areas.
5 l
Q S o ,- there is a passing reference to nuclear matters, 6 Aut'.that wasn't the purpose of the courses then, was it?
I 7 A That is correct.
8 q Therefore, your only training in the field of nuclear 9
physics was that one introductory course that you identified 10 earlier in your testimony?
II A. That is my only academic training, that is correct.
12 Thank you.
Q Do you possess, or claim expertise in' 13 mathematical modeling?
14 A Leg your pardon?
15 f Q Well, let's maybe change the question a li t t 1'e ' b i t . I I,
.16 '! Have you taken any courses in mathematics -- higher mathematics?
17 A Yes. I have taken several courses in mathematics.
I i
18 ,.
V Q Aside-from those that one.gets with the engineering j I'
- degree? I am talking about courses that deal with probabilities.
Il 20 !'or mathematical modeling, that sort of thing?
f 21 A At the undergraduate level, the typical undergraduate 22 engineering student takes courses in calculus and differential 23 equations.
l O kce.Fw Reporters, Inc.
24 I also took a course in mathematical statistics at 25 th.e undergraduate' level, and in addition as part of course work l
! l
6-6-JccWOl 298
] in experimental laboratory technique in experimental laboratory
's) 2 technique, material on the evaluation of statistics of the data 3 from those experiments.
g , Q And you said that was at the undergraduate level?
5 l l A It was at the undergraduate level.
1 6 Q And tha't was in conjunction with your engineering 7
degree? !
8 A That is correct.
9 Q Could you estimate for us the amount of credits 10 involved in those two courses you just identified?
I 11 [ A .The mathematics statistics class was three credits.
I 12 The laboratory techniques program was two terms, of two credits 1 '.I l ) 13 ; each, w/ .
14 Q Okay. Have you ever designed a computer code such 15 as a safety analysis code, such as the WRB-1 correlation for a 16 THINC computer model? l 1:7 A WRB-1 is a correlation, it is not a computer code. !
18 The THINC code was developed many years ago. I have been l 1
to involved in the modifications to the THINC code to incorporate l l
20 l such new correlations as the WRB-1.
' t 21 Q But you never did any original work on a computer 22 j code, is that correct? Is that your testimony?
n 23l A I have developed computer models. I did not do the em !
24 orginal model development of the THINC Code.
Aa.Fii Reporters, Inc.
25 Q When you deal with the THINC code, do you then l !
6-7-JocWcl 299 1 interpret.the work of others; the original work of others?
("s
(_ 2 Your function?
3 A The Code is described in topical reports. Define 4 the;model in the code; define the ranges of applications that 5 we qualify the code to be used'on.
6 That material is reviewed as part. of our training 7 and qualification for performance of the thermal design for 8 Pressurized water reactors.
9 Q Do you claim expertise in the area of statistical 10 analysis' ?
I'l A As they relate to the matters I have testified to. ,
i 12 9 Have you published any papers on statistical l .-
i: I h, 13 L analys'is?
14 A No.
l i
15 L Q In your professional' qualifications, Page 1, in the 16 last'.-paragraph, midway.into'that paragraph, you state: These l i
17 analyses included the determination of core operation units 4 18 ; to ensure ma' r gin f or prevention of departure from nucleate i
l 1'[hoiling. !
! I !
20 j! What do these analyses.that you have identified here l l
L 21 involve? In oth'er words, do they involve physical experimen-22 tation,.-or mathematical calculations?
l
- 23 A These analyses, as I describe in somwhat'more detail 24 , in my testimony, deal with'the << plication of the thermal-L
(^ Reporters, W-FAs} Inc. ,
25 hydraulic analysis code for established conditions, conditions I
l
6-8-JocWcl 300 1 of operations, and set points in the technical specifications.
2 Q Are those codes derived from examination of physical 3 phenomenon? Physical experiements?
4 .. A The analytical codes that we use in this analysis 5 are-derived from those data.
6 .Q Are there any mathematical calculation involved in 7 additional to physical experimentation?
8 A Yes, there is.
9 Q -On the same page, Mr. Dzenis, next to the last 10 sentence, you say you evaluated the results of various postulated i
11'; 0 accidents to' determine whether these core limits met the P
12 d requirements.
. () 13 Could you describe with any more precision what you 14 mean by, ' postulated accidents?'
15 In other words, were they mathematical analyses? ,
f.
16 .A. The group of postulated accidents covers the territory 17 ' ref erred to in the testimony as normal operatio'n anticipated
~
i 18 ' transients.
There is afdefined set of transients which form 20 the' set of accidents that I am referring to. And these accidents 21 are analyzed, and the total system response to those accidents.
22 The results of those accidents at the system level 23 are evaluated by these core thermal hydraulic models, to-determine Z) Fs (N Reporters,
- 24 Inc.
whether the DNB criterion is met.
25 The evaluation of the thermal hydraulic models and
6-9-JocWal 301 1 establishment of the models for those particular applications 2 was my responsibility.
3 Q Are these models calculated, or physically enacted?
4 MR. BAUSER: Excuse me. The area of questioning, 5 it seems to me, has now gotten beyond the witness' qualifications 6 into the substance of matters considered in his testimony, and 7 I would like to object to further questioning along this line. .
8 MR. HODDER: We pose the question because the matters 9 I am inquiring about are contained in the professional 10 ; qualifications, which are an element of voir dire, and it may l l-11 ? be that it is a bit beyond the normal scope of this, but the ',
a t
12 ' witness has offered it in his professional qualifications, and I 13
( _) I felt it was ap p r'op ria t e to proceed. ;
}
14 I was about to finish. I would like an answer to l
15 my question.
16 JUDGE LAZO: Well, we will overrule the objection.
17 Mr. Dzenis has already referred to his testimony on a couple .
18 of occasions in responding to questions on his qualifications.
f
'O We should.though try to not get over into the !
20 testimony itself. Why don't you proceed?
{
21 MR. HODDER: Well, I didn't get an answer to the i 1
22 h question, sir. Could I have that answer.
l 23 WITNESS: Well, with the intervening discussion, I 24 I
/~')
W-FL_.J Reporters, Inc.
have lost tract of the original question.
25 MR. HODDER: Could we have the court reporter re-read l
6-10-JccWcl 302 '
I the question?
(s' 2 JUDGE LAZO: Would you please do that, Mr. Reporter?
3 (Reporter reads back the questionb' 4 ; BY MR. HODDER: (Continuing) 5 ; Q Mr. Dzenis, the question referenced the next to the i
6 last sentence in your professional qualifications on Page 1.
7 And it referred to postulated accidents, and as the court 8 reporter said, I asked were these accidents, as postulated, ;
I 9 calculated or physically enacted?
10 l A These accidents are calculated results based on the 11 i
11 i requirements and conservative assumptions as defined in the i
12 ]istandard review plan. l
( / 13 MR. HODDER: I have no further questions, and we l
14 withdraw our objection to the witness coming in.
l 15 JUDGE LAZO: And the Staff has no objection?
16 MS. YOUNG: That is correct.
i 17 JUDGE LAZO: Therefore, your testimony will bc l 18 received in evidence. I will ask the reporter to incorporate
'C it directly into the transcript as if read. ;
IX INDEX 20 (P refiled testimony of Mr. Edward A. Dzenis follows) l~
r l 21' ll l 22 h,i 23f I r (N 24 ' l Aa-F_J Reporters, Inc.
25 l
( )~ UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION i
BEFORE THE ATOMIC SAFETY AND LICENSING BOARD {
)
In-the Matter of. ) Docket Nos. 50-250 OLA-1
) 50-251 OLA-1 FLORIDA POWER AND LIGHT COMPANY )
)
(Turkey Point Nuclear ) ASLBP No.- 84-496-03 LA Generating Units 3 & 4) ) (Vessel' Flux Reduction)
)
November 26, 1985 Testimony of -
Edward A. Dzenis Concerning Contention (d)
My name is Edward A. Dzenis. I am Manager of Core Operations,
() for the Nuclear Fuel Division of Westinghouse Electric Corporation.
My business address is Westinghouse Electric Corporation, Monroe-ville Mall Office Building, P.O. Box 3912, Pittsburgh, PA 15230.
A resume of my professional qualifications and experience'is j- attached to this testimony'and incorporated herein by reference.
l-In an Order dated August 16, 1985, the Licensing Board, among other things, denied Licensee's motion for summary disposi-tion of Contention (d)*. In connection.with this motion, the Board presented and limited the scope of.this proceeding to
'the following three questions:
! 1. Whether the DNBR of 1.17 which the amendments impose on the OFA fuel in Units 3 and 4 compensates for the.three uncertainties.
outlined'by the Staff in its December 23, 1983 SER on the amendments, at 4.
- 2. Whether, if the.DNBR of 1.17 does
() not compensate for those uncertainties, the SRP's 95/95 standard, or a comparable one, is somehow satisfied. '
- 3. Whether, if that standard is not being satisfied, the reduction in the margin of safety has been significant.
In an Order dated November 8, 1985, the Board denied licensee's September 20, 1985 Motion for Summary Disposition of Contention (d). In a Memorandum dated November 18, 1985, the Board outlined some of its concerns for consideration by the parties in prepara-tion for the hearing as being:
i
- What if any uncertainties are -
included in the ascertainment of the DNBR
, value of 1.177 i.e.,' rod bow, hydraulic flow pattern considerations, ideal condi-tions, etc.
Meaning of the term " design" DNBR
() or "DNBR limit" and/how it is applied to the Turkey Point Plant. How is the " calculated minimum" DNBR related to'this and how is it
, used at. Turkey Point?
Procedures and/or techniques used to ascertain the " calculated minimum" DNBR including any assumptions and uncertainties and how they might affect the validity of the calculation. How much confidence should we have in this~ calculated value?
Concerning the 3 uncertainties l
listed, what are the bases for the estimates and how muck confidence can we place on the estimates? Of particular interest are the
. penalty values discussed in the.NRC's Safety Evaluation related to Amendment 99 at pages 3 and 4.
l -
Is the approacn used for all l Westinghouse safety analyses at Turkey Point (Dzenis, at 3) unique to Turkey Point?
O
, .,_ ..mz ...a- . s 2.. U
. . The purpose of my testimony is to respond, in order , to
.each of the Board's three questions. The testimony will also provide information addressing the Board's stated concerns related to~ Contention (d).
- 1. Whether the DNBR of 1.17 which the amendments impose on the OFA fuel in Units 3 and 4 compensates for the three uncertainties outlined by the Staff in its December 23, 1983 SER on the amendments, at 4.
The answer to this question is: No.
At the outset, it might be helpful to address the Board's concern, as presented in its November 18, 1985 Memorandum, pertinent to nomenclature. The 1.17 DNBR has been referred to a number -
of ways in this proceeding, such as "
DNBR value,'" "' design' DNBR," and "DNBR limit." I believe that it may best be defined
~
as a DNBR acceptance limit. This DNBR acceptance limit of 1.17 is generic to all' Westinghouse plants using Optimized Fuel Assembly (or OFA) fuel. In addition, it should be noted that this DNBR
(
acceptance limit is to be distinguished from what the Board has referred to in its Memorandum as the "' calculated minimum' DNBR," which is calculated on a plant-specific basis and which will be discussed below.
The DNBR acceptan,ce limit of 1.17 for the WRB-1 correlation which is used in connection with the analysis of all Westinghouse OFA fuel, including that at Turkey Point, constitutes, in accordance with the acceptance criterion presented in Section 4.4, Part II.l.a-(at the top of page 4.4-3) of the NRC's Standard Review r
Plan, the 95/95 bounding value for experimental data. Stated O
l
differently, tha 95/95 standard contained in the NRC's Standard Review Plan will be satisfied by assuring that calculated minimum
() DNBR values'for.all normal and anticipated operational occurrences, after accounting for uncertainties, are greater than or equal to the 1.17 DNBR acceptance limit. The method for determining the DNBR acceptance limit is described in detail in paragraphs 6-25 of my August 8, 1984 affidavit, which was submitted in this proceeding as part of Licensee's Motion for Summary Disposi-tion of Intervenors' Contention (d), dated August 10, 1984.
The 1.17 DNBR acceptance limit, however, does not and is not intended to compensate for the three uncertainties referred to in the Board's question, i.e.: rod bow, mixed Low Parasitic (or LOPAR)/OFA fueled core, and 15x15 OFA array fuel. As will be explained in connection with the Board's second question, such uncertainties are considered in the evaluation of normal and anticipated operational occurrences.
- 2. Whether, if the DNBR of 1.17 does not compensate for those uncertainties, the SRP's 95/95 standard, or a com-parable one, is somehow satisfied.
The answer to this question is: Yes; the SRP's 95/95 standard is' satisfied.
l That star.dard is satisfied by assuring that minimum DNBR
~
values calculated for all normal and anticipated operational occurrences, after accounting-for the unc'ertainties referred to above, are greater than or equal to the 1.17 DNBR acceptance i
limit. In the case of the particular amendments under considera-tion here (Amendment No'. 99 to the Turkey Point 3 license and O
~ -
-.e. ,,
.n, - , ,g., ,
AmOndm3nt No. 93 to tha Turksy Point 4 licence), the safety analysis minimum DNBR (referred to as the "' calculated minimum' O DNBR" in the Board's November 18, 1985 Memorandum) is 1.34 using
~
the WRB-1 correlation.
The 1.34 safety analysis minimum DNBR represents the lower bound to the values ~ calculated for the spectrum of normal and I
anticipated operational occurrences for Turkey Point. Procedures and techniques employed in the calculation of DNBR values for this. spectrum of normal and anticipated o'peratimaLoccurrences IE. C AMD E.E areinaccordancewithSection~4.4, Parts (II.6,II.7andII.8 A
and the appropriate portions of Section 15 of~the Standard Review Plan. For example, conservative values of individual system
,, parameters (such as temperature and power level) are used as inputs to these calculations.
These values are consistent with
({} the Limiting Conditions for Operation defined in the Technical
~
Specifications. The safety analysis minimum DNBR is also used in the determination of reactor protection set points which i
are in the Technical Specifications.
1.
The uncertainties referred to in the Board's first question (i.e., rod bow, mixed LOPAR/OFA fueled core, and 15x15 OFA array fuel), reflect specific geometric considerations not explicitly i
modeled in these safety analyses. However, as demonstrated by the equation -
1.34-1.17
= .127 or 12.7%
the 1.34-safety analysis minimum DNBR value -- which is computed using Turkey Point plant-specific reactor parameters -- can O
l accommodate an uncertainty of 12.7% and still meet the DNBR acceptance limit of 1.17 derived from the WRB-1 correlation. '
]}
The three uncertainties referred to total only 10.5%.
~5.5% (for rod bow) 3.0% (for mixed LOPAR/OFA fueled core) 2.0% (maximum, for-15x15 OFA array fuel) 10.5%
Since.12.7% is greater than 10.5%, there is sufficient margin in the l.34 safety analysis minimum DNBR, above the 1.17 DNBR acceptance limit, to compensate ~for uncertainties' associated with rod bow, the mixed LOPAR/OFA fueled core, and 15x15 OFA array fuel.
It is pertinent to note that application of uncertainties -
to results obtained from predictive analysis (in this case, the 1.34 safety-analysis minimum DNBR), rather than to design
- {~)-basislimits.(suchasthe1.17DNBRacceptancelimit), is common
-in.the engineering field. In particular, it is the approach used for all Westinghouse safety analyses, including those for Turkey Point, independent of fuel design or critical heat flux correlations. This approach is also consistent with that in Section 4."4, Part II.2 (at page 4.4-3) of the NRC Standard Review i
Plan. It is not unique to Turkey ~ Point.
Before turning to,the Board's third question, it may be helpful to address two particular points. The first.of these concerns uncertainty values related to a mixed LOPAR/OFA fueled
. core and rod bow.
4 The effect of the mixed core (i.e., of two different fuel assembly. types in the reactor core at the same time) on critical
-O l
. . . . , . _ . . _ . _ . - , . . - _ . _ _ _ . . m - -.._ . . - . __
_. ~ . . . _ _ .,
hast flux'has bsen analyzed.. Spacifically, the differing hydraulic )
1 resistances of the two types of fuel assemblies were modeled explicitly in detailed thermal-hydraulic analyses. The effect of these differing resistances on flow distribution was then calculated for different configurations of fuel assemblies.
Exact reactor core loading patterns cannot be defined in advance for all reactors for every fuel loading because of their dependence on specific plant operating schedules and the specific design requirements of particular refueling cycles. Therefore, these mixed core analyses included various combinations of " checker-board" configurations that could be loaded into actual cores.
These configurations were selected to envelope the expected '
configurations included in reload licensing submittals.
The results of this work, pertinent to transition cores of 15x15 LOPAR and OFA fuel, indicated that a 3.0% DNBR reduction
(]}
for OFA fuel was sufficient to bound all of the calculated effects of transition core geometry. This uncertainty is applied as a penalty'only to the OFA fuel because it has a higher hydraulic resistance than does the LOPAR fuel. No uncertainty is applied to the LOPAR fuel because it always receives at least the reactor coolant flow it would have otherwise experienced.
The value of DNBR uncertainty for rod bow is calculated based on a-correlation of the measured rod bow of irradiated- ,
fuel assemblies. Since the amount of measured rod bow increases with fuel irradiation time, or burnup, the DNBR uncertainty for rod bow also increases with fuel irradiation time, or'burnup, t
l I
The valua of 5.5% DNBR corresponds to the highest burnup at which DNB is a concern because,_at higher burnups, heat genera-tion rates in PWR fuel decrease due to a decrease in the concen-tration of fissionable isotopes and the buildup of fission product inventory. Therefore, the value represents a conservative upper bound to a range of rod bow effects of from 0% to 5.5% DNBR.
Also'concerning uncertainties, the uncertainty properly associated with the application of the WRB-1 correlation to 15x15 OFA fuel is actually 0.0%, not 2.0%. This is because, as noted in paragraph 10 of the Affidavit of Yi-Hsuing Hsii Regarding Contention (d), attached to the NRC Staff response to Licensee's Second Motion for Summary Disposition of Contention (d), dated October 15, 1985, the WRB-1 correlation is applicable to-15x15'OFA fuel with the same 1.'17 DNBR acceptance limit as applies to other designs.
{}
The second point concerns the physical independence of the phenomena of rod bow and variation in hydraulic resistance between LOPAR and OFA fuel. These are separate and independent effects. Each phenomenon is such that the existence or extent of one does not materially affect the existence or extent of the other. Specifically, since for both LOPAR and OFA fuel, rod bow is a local, random variation in the fuel rod position between the grid locations, there is no systematic change which could influence the mixed core uncertainty. Also, the flow redistribution which occurs in the mixed core is not capable of deflecting a fuel rod to any significant degree. Accordingly, the phenomena are not interrelated and the uncertainties asso-(
( ciated with'them can properly be considered independent.
9_
In cummary, as I have explained, in the case of OFA fuel, a DNBR acceptance limit of 1.17 with the WRB-1 correlation meets the 95/95 standard. If the safety analysis minimum DNBR, after accounting for uncertainties, is greater than or equal to the DNBR acceptance limit of 1.17, the SRP's 95/95 standard is satisfied.
As has been discussed, even when penalized for uncertainties consistent-with Section-4.4, Part II.2 of the Standard Review Plan, the 1.34 safety analysis minimum DNBR -- which is the lower 1ound to the values calculated for the spectrum of normal and anticipated operational occurrences for Turkey Point --
more than assures that the allowable DNBR acceptance limit of 1.17 derived from the WRB-1 correlation is satisfied.
- 3. Whether, if that standard is not being satisfied, the reduction in the margin of safety has been significant.
i
(} As I have explained in response to the second Board question, the 95/95 standard is being satisfied. Accordingly, there has been~no reduction in the margin of safety.
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4 November '85 Professional Qualifications and Experience of My name is Edward A. Dzenis and my business address is P. O.
4
~
-Box 3912, Pittsburgh, PA 15230. I am employed by Westinghouse Electric Corporation (" Westinghouse") as Manager of Core Opera-tions in the Nuclear Fuel Division.
I graduated from Lehigh University with a Bachelor of Science
- Degree in Mechanical Engineering in May, 1974. While employed by Westinghouse I gra'duated from Carnegie Mellon University -
l with a Master of Science Degree in Mechanical Engineering in May,'1977. I am currently a Registered Professional Engineer in the Commonwealth of Pennsylvania (certificate number PE-027744-E).
In June, 1974, I joined Westinghouse in the Nuclear Fuel Division of the Water Reactor Divisions as an Associate Engineer. My duties in the Thermal-Hydraulic Design section included the ,
analysis of heat transfer and fluid flow aspects of. reactor fuel assemblies and related components for pressurized water reactors. These analyses included the determination of core operation limits to insure margin for prevention of departure i .
from nucleate boiling (DNB) and other safety criteria. I evaluated the results of various postulated accidents to determine whether these core limits met requirements. I was also responsible for preparing related documentation for submittal to regulatory
() authorities.
! f
_2_
Since that time I have had assignments of increasing responsibility
() in Thermal-Hydraulic Design and was promoted to the position of Engineer in August 1976, and Senior Engineer B in April, 1980. In October, 1981, I was promoted to the position of Manager, Thermal-Hydraulic Design, with responsibility for the efforts l of several engineers and technicians in the thermal-hydraulic analysis of fuel for Westinghouse supplied pressurized water reactors including the Turkey Point units. I am now Manager of Core Operations.
O
]
=
0
o 303
)7-1-SueW - 1 JUDGE LAZO: Mr. Bauser, are you presenting your
' - 2 witness for cross-examination?
3 MR. BAUSER: .Yes. Our witness is now prepared 4L for cross-examination.
i 5 MR. HODDER: Very well. Mr. Chairman, pursuant to 6 the Board's finding, I would like to allow this portion of 7 cross to be conducted by Dr. Gordon Edwards who is sitting 8 with me.
9 JUDGE LAZO: You may proceed, Dr. Edwards.
10 DR. EDWARDS: Thank you.
Il CROSS EXAMINATION 12 BY DR. EDWARDS:
r EDi,)~ If Q Mr. Dzenis, I would like to refer you to Page 3 14 of your testimony. At the bottom of the page, the final 15 paragraph, you discuss the DNBR acceptance limit of 1.17 16 for the WRB-1 correlation which is used in connection with 17 the analysis of the OFA fuel at Turkey Point. I 18 And you mention that this is based on the 95/95 1 i
19 bounding value for experimental data; is that correct?
20 A That's correct.
l 21 Q Would you tell me what that experimental data is ;
i 22 and specifically whether it deals with mixed cores?
23 A That data, the data base which establishes the !
24 w-FL}; (~ Reporters, Inc.
WRB-1/DNB correlation is a data base of rod bundle type 25 geometry with the coolant falling between the fuel rods and. I
, I. ,
_ _ _ _ _ _ _ _ m______ __._
304 J
- 7-2-SueW 1 grid structures like the grids that are in operating reactors.
n l- 2 It includes grid structures of the types employed in both 3 types of assemblies in the reactor core at Turkey Point.
4l So, it' addresses both the previous Inconel L grid 5 _
type of structure and the optimized fuel' zircaloy grid 6 structure.
7 Q Does the experimental data deal ~ with these grid 8 structures in a mixed format?
9 A The critical heat flux testing programs employ 10 only one grid at a time.
11 Q So, would it be correct to say then that the
.12 ' experimental data is derived exclusively on the study of (f 13 homogeneous cores?
14 A The data that is employed in that THINC is based 15 exclusively on single geometry bundles. )
16 O Single geometry bundles?
17 A Yes.
18 Q Now, the sentence following this in your testimony 19 'says, " Stated differently, the 95/95 standard contained in 20 the NRC's Standard Review Plan will be satisfied by assuring 21 that calculated minimum DNBR values for all normal and antici-22 pated' operational occurrences, after accounting for un -
23 . certainties, are greater than or equal to the 1.17.DNBR 24 W-Fg Reporters, Inc.
acceptance limit." l 25 When you say "will be satisfied," are you referring I
3
305
'97-3-SueW 1 to Turkey Po' int among others?
(v) . 2 A Yes. !
3 Q Now, is it not the case that the Turkey Point 4 ,
situation involves'a mixed core where you"have a certain 5 amount of low power' fuel and a certain amount of OFA fuel 6 in the same configuration?
7 A Yes, that's true.
8 Q I have a little difficulty in understanding how 9 in one sentence you can cite experimental data which is based 10 solely on homogeneous cores and in the very next sentence you 11 can apply this to a mixed core.
12 Are you -- are we-to infer from this that in your I) 13 opinion, in your professional judgment, that there is no 14 significant difference between a homogeneous core and a 15 mixed core?
i
,. 16 A No, that's not true. If you -- elsewhere in my l 17 testimony, I cover discussion of.how the effect of the 18 simultaneous existence of the two fuel types in the core at 19 the same time is treated.
20 Q Well, I will come to that. But I'm concerned about 21 this particular juncture of your testimony which is early on, i
22 Page 3 and~4, where you have one sentence that says that the 23 DNBR acceptance limit of 1.17 is in accordance with the 24 accepted criterion, the 95/95 bounding value, for experimental
} .w(~}n n.n. u.
t 25 data.
l
306
- 7-4-SueW 1 A That's correct.
2 l Q Then, you say: Stated differently, that 95/95 3 standard will be satisfied. And you've confirmed that you 4 are referring to Turkey Point here.
5 A That's correct.
6 Q What I have difficulty in understanding is how t
7 you can say that these two statements are the same.
8 MR. BAUSER: Excuse me. The --
9 DR. EDWARDS: Could you clarify that? I'm sorry, 10 I didn't phrase it as a question.
11 MR. BAUSER: You anticipated part of my objection.
12 I object to the form of the question and also the fact that
() 13 the entire sentence, the second sentence, to which is being 14 referred has been omitted.
15 I would like that considered all at once.
16 DR. EDWARDS: Well, I would be glad -- would you.
17 like me to read the' entire sentence? l 18 MR. BAUSER: That would be helpful.
19 BY DR. EDWARDS: (Continuing) 20 Q Would you like me to read the entire two sentences?
21 A If necessary, yes. j 22 I Q Let me read the two sentences. The first.
23 sentence is: "The DNBR acceptance limit of 1.17 for the I
24 Am-Fjggtearsn,tu. WRB-1 correlation which is used in connection with the !
i i 25 analysis of all Westinghouse OFA fuel, including that at j
. 1 307
- 7-5-SueW. 1 Turkey Point, constitutes, in accordance with the acceptance 2 -criterion presented in Section 4.4, Part II.l.a:(at the top 3 of page 4.4-3) of the NRC's Standard Review Plan, the 95/95 4' .
bounding.value for experimental data." End'of sentence.
5 Now, parenthetically, you have confirmed that the 6 experimental data is for homogeneous cores'.
7 Continuing with the next sentence: " Stated 8 different, the 95/95 standard contained in the NRC's 9 Standard Review Plan will be satisfied by assuring that 10 calculated minimum DNBR values for all normal and anticipated
'll operational cccurrences, after accountinf for uncertainties, 12 .are greater than or equal to the 1.17 DNBR acceptance limit."-
() 13 End of sentence.
14 A That's correct.
15 Q Now, Mr. Dzenis, is it your intention in the ;
I 16 testimony to imply that these two statements have the same 17 content?
18 When you say " stated differently," does that
' 19 mean that the two statements are intended to have the same 20 content?
21 A Yes, with respect to addressing the question that j 22 this material is presented in response to.
23 Q So, in other words, you see no problem in taking 24 f'
m-F;;_-)cuemn. ine. experimental data which is based exclusively on homogeneous l 25
. cores and.saying that that'has the same content as applying I-i
308
- 7- 6-SueW 1 that criterion to a mixed core?
'I 2 i A That's true, because the data for each bundle type 3
is appropriate used in the subsequent safety analysis of l 4! , either type of bundle, both in the transition or the homo-5 geneous core zones.
6 The method I described later in my testimony totall 7 accounts for the effects of the two types of fuel in the core ,
8 at the same time. That's an independent discussion on the 9
material in response to Question 1 of my testimony.
I 10 Q Is it not the case that there was some question II of acceptability of the DNBR limit of 1.17 for 15 byil5 OFA ;
12 fuel assemblies at an earlier point in time?
13
) A In my technical judgment, we have presented '
Id information which supported the use of the 1.17 for the 15 by 15 15 optimized fuel geometry.
i 16 So, in my technical judgment, there was no question.
17 There was in the process of the review, as ordinarily occurs ,
18 during the review of any material that we submit, questions l t
19 and answers on many aspects of the material. These are re-20 sponded to and resolved in the process of our licensing 21 review of the docketed materials. ;
22 0 Would you agree that there are important differences:
23 in terms of DNBR considerations between homogeneous cores and
-~
24 i W-F Reporters, Inc.
25 <
A I can't agree with the word "important." There is .
l
. .- . . - - . - . - . . ..~ . . - -. .
309' i
47-7-SueW 1 a consideration of the two types of fuel in the core which 2 ,
is explicitly analyzed and included in the mixed core un-3 certainty that I describe later in my testimony.- t
- 4. Q Could you explain briefly why the situation with 5 the mixed core _is not immediately applicable to data based 6 on a homogeneous coue?
7 In other words, why is.it that you could not take
, 8 the data-which was derived experimentally from the study of 9 _ homogeneous cores and simply apply it without any penalty 4 10 to a mixed core?-
11 A I think I need to get a better understanding as 12 to what you mean by data from homogeneous cores.
, () 13 Q I'm referring to your use of the experimental 14 data here in your sentence.
15 A Okay. The experimental data base that is used ;
16 to_ derive the -- any DNB correlation for that matter, is a i
! 17 relatively -- it's a subsection of an assembly, one fuel 18 assembly, ordinarily twenty-five or thirty-six rods,'of 5 by 5 19 or 6 by 6 bundle.- The reason for that is because the amount f
20 of power it takes to actually take a rod to DNB is very high
-21 and it's not possible to test a bundle larger than that.
I 22 So, this small subsection of the bundle is-used c 23 as determining the data base for the determination of L
24 correlations and establishing the appropriate acceptance y-Fi n con.n, Inc.
l 25 limit DNBR. 1 F
2 c -
m',y: 310 m .
g
- 7-8-SueW 1 .Even in a homogeneous core, a full core of any
-- g. <
homogeneous fuel assembly, there is flow redistribution 2
3 between.the assemblies and that core because of the differing 4L power levels and the different assemblies. Iberefore,this -
l 5 is why there are tools which are used to analyze the flow 6 distribution,in a core-wide imulation or model of the core.
'7 It is to that application -- and once the lo' cal 8 thermal con'ditions are developed fr(m that core-wide cal'cula- I t
9 tion -- that the DNB correlation is ap^olfed and whether the N
% 4 10 calculated DEBR exceeds the acceptance limit. That's the I s ' ' s. ,,
11 homogeneous core-wide analysis.
12 In the case of a tr'ansition core the sa'r.e.,
.g analvtical
~
O ia too1= wou1a 99 1 r, due iir'ai= c e we aeri=ea e" i=cre e=*=1 14 bounding uncertainty for the effect of the two types of
(
15 assemblies that could be $e~ corded at the same time. '
>. t (
16 Q In short, are you saying that the' difference kith '
!v 17 a mixed core is a more complicate'd hydraulic flow due $o the
^
i l
18 differenceinhydrabicrssistance?
19 Would you say that that i the reason why?
20 A More complicated than what? q -
s, 21 Q More complicated than in a case of'a homogeneous N ,
~'
22 core?
23 A _.It's a different flow distribution.; In other I
24 words, in a homogeneous core calculation'we supplied the \
f W.F Reporters, Inc. . f j
25 power distribution that occyrs across the core as an input to s,
( N, '
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311
- 7-9-SueW l the analysis. In the case of transition core, we supply the 2 same input but we define the differing hydraulic resistances 3 of each assembly. .This is well within the capabilities of
- 4. the model to calculate the flow distribution and the changes 5 in the conditions, flow and temperature conditions in the 6 core which occur because of the transition core geometry.
7 And then the effect of those changes is -- results 8 in a change-to the calculated DNBR which establishes the 9 transition core uncertainty.
- 10 Q It is a fact chough that the experimental data 11 upon which the 1.17.DNBR limit was originally derived was 12 based on the study of homogeneous cores; and, also is it not
() 13 a fact that the, plant specific safety analysis for Turkey 14 Point, Units 3 and 4, also proceeded from the assumption of 15 homogeneous cores; is that correct? !
16 A No. Both of those statements are not correct.
17 The first statement is correct. The second statement, which 18 deals with the safety-analysis of the Turkey Point cores, 19 did not assume homogeneous cores.
20 Certain calculations as a subset of that safety 21 analysis assume a homogeneous core. However, the other calcula-22 tions which are provided as the basis of the transition core i 23 uncertainty are done in full recognition of the fact that g-) 24 there will be a transition core operating in the Turkey Point ko-F(_) Rmomrs. Inc. J l 25 Plant.
- l I
I N !
o _ . _ . -_ --
- 1 312 *
)7-l'0-SueW' I Q In connection with DNBR, the calculated DNBR, O- 2 using.the THINC computer model -- '
[
'3 A Yes.
4 Q -- is it not the e,ase,that the calculated, DNBR 5 limit of 1.34, which is referred t!o.in your testimony, was' ,
6 derived proceeding from the assumptiongof a homogeneous core?,
7 -A That particular calculated value does, yes.
8 Q On Page 7 of your testimony, you refer to thermal-3 9 hydraulic analyses. This is at the top of Page 7. "Specifi- i 10 cally, the diffe ing hydraulic resistances of the two,. types 11 of fuel assemblies were modeled~ explicitly in detailed thermal-12 hydraulic analyses." ,
13 My question is:' Were these actual experiments? t 14 Or, were these mathematical' calculations? p s o )
15 A We have experiments on the hydsaulic 'characteris-1 y
i 16 tics of the two types of fuel assemblies.1 ,In thisxcase, a
- i. 17 ' full-fuel assembly of either type of fuel. And we also have 18 hydraulicLexperimental data on the case o'f two assemblies 19 side-by side in an actual test loop. t 20 It is that experimental data which is used to 21 determine the hydraulic resistances that I refer to in this 22 sentence. a 23 Q So you have e erimental data dealing only with / I O.
La Reportm, Inc.
24 two bundles side by side in a test loop? #
L 25 Is that the extent of the experimental data?
~... .
i l.
d 313
~
- 7-ll-SueW1 A We have --
2 q
Q That's for a mixed configuration?
3 A For the hydraulic characteristics,'that's
, 4 j. . correct.
5 Q Okay. So I presume then that these thermal 6 hydraulic analyses which we are discussing on Page 7 are 7 mathematical calculations, not physical experiments?
8 .A They are mathematical calculations.
9 j Q Thank you. Did these calculations specifically 10 involve 15 by'15 OFA fuel assemblies?
<y 11 A Yes.
12 Q When we come to-the middle of the page, you say, NL- 13 "The~~results of this work, pertinent to transition cores 14 of 15x15 LOPAR and OFA fuel, indicated that a 3.0% DNBR
{# 15 reduction ~for OFA fuel was sufficient to bound all of the
./
16 calculated effe~ cts of transition core geometry."
17 Now, this 3.0 DNBR reduction, this is the penalty 18 that'you were referring to earlier?
19 A That's correct.
20 Q Does the three percent penalty involve only the 21 mean.value of the probability distribution? i 22 Or, does it also affect the standard deviation 23 of the probability distribution?
t 24 A It reflects neither of those, ke-FL Reporters, Inc.
25 Q You are saying, are you not, that if you take the y .-*+---.p,. .myr2p--, , - - - - ,- , ,,-yy----- ,y ,,m._.-. v rrv .m ww m- m -rw, -ww9-- y-T r -- r.-yv' = r-i == e-
314
- 7-12-SueW l DNBR limit of 1.17 and assign a penalty of three percent, i fl/
2 that this will compensate for the variability in hydraulic 3 resistance caused by the mixed core?
- 4. Is that what you are saying here?
5 A No. What~I'm referring to is the -- the 1.34 6 safety analysis DNBR represents a lower bound to a spectrum f
7 of individually calculated DNBRs.
8 None of those DNBRs would have to be reduced by 9 more than three percent to accommodate the effect of the 10 transition core for that particular safety analysis calcula-11 tion.
12 Q And that is based on the mathematical calculations
(_ ) 13 which are referred to above as the thermal-hydraulic analyses?
14 A That's correct.
15 Q Thank you. Now, at the bottom of this page -- I ,
16 would like to ask you whether the exact core _ loading patterns 17 can be determined in advance?
18 A No, they can't.
19 Q Well, if they cannot be, then aren't you essentially 20 reduced to a certain amount of guesswork in terms of what 21 the precise configuration of the fuel is going to be? ;
22 A Let me define advance more explicitly. We performed 23 this type of safety analysis to be -- to serve as the basis j
(') 24 Mz-F;._a' Resmners, Inc.
forachangetotheoperatinglicenseatthatparticularunit.l 25 Once it has become a portion of the operating ;
l i
i
315
- 7-13-SueW l license it's not intended to change because of particular l C)
' / 2 choices of a specific reload core loading pattern. There-3 fore, at the point in time that this sort of a. submittal is I
4; made, which is in advance of the two to three cycles of I
5 transition geometry and the subsequent operation of a full 6 optimized fuel core, we do not know all the specific loading 7 patterns for those particular cycles of operation.
8 That is why we choose a methodology of this type 9 which uses bounding values for the particular uncertainties 10 to cover the effects of any conceivable specific loading 11 pattern which may be established during those intervening 12 cycles.
() 13 Q Are you saying then that your thermal-hydraulic 14 analyses include all possible core loading patterns?
15 A They include a consideration. There is a finite 16 number of conceivable geometries in this checkerboard. I i
17 Q Yes. And do you study all of those possible l 18 configurations?
I 19 A We -- as I've stated, the checkerboard only refers 20 to the way the fuel assemblies are loaded in the core. There 21 are defined locations to feed the new fuel versus the fuel !
22 which is being carried over from a previous cycle. l l
23 We don't --we. define the fuel within the particular ;
24 assembly of their position in the core, not the change during us-F;(]j Rnomn, Inc.
j 25 the reload. Therefore, it was only given two types of assemblies
! l l I i !
I
316
- 7-14-SueW 1 that go into the varying positions'on the checkerboard.
(k 1
2 .There is only a finite number -- you can't do alternating, l l
3 diagonals or --
'4 ,
O I fully understand that there is only a finite number of possible arrangements.
~
4 5 What I'm asking you is 6 whether you looked at all possible arrangements?
7 I don't see that your answer addresses the question.
8 A I' don't remember exactly which configurations, 9 Ebut we studied several configurations --
10 0- ~Several?
[ 11 A -- and found the pipes configurations.
12 Q =Would you care to give a numerical estimate _of
-( ) 13 the number of possible configurations that might exist?
14 A The number of possible configurations? For the 15 purposes of this type of calculation, there is probably only 16 three that could really-occur.
17 In other words, for the purposes of calculating-18 the effect'of the thermal conditions in a particular assembly, 19 what's going on with its neighbors.
20 Q So the few has become three? You studied three 21 core configurations?
- 22 A There is three particular that I can identify.
23 Q Okay. So, what we have~is a situation then-where 24 p the calculation is done for a homogeneous core but the applica-bF() Reporters, Inc.
25 tion takes place for a non-homogeneous core; is that correct?
t
re 317 !
- 7-15-SueW 1 A That's correct.
2 Q And in order to get some understanding of what 3 happens with the non-homogeneous core, you conduct these 4 mathematical thermal-hydraulic analyses; is that correct, 5 .in order to understand how the flow pattern changes?
6 Is- that - my understanding, is that a correct 7 understanding'of what you are saying?
8 A That's correct.
~
9 Q Is it also correct that since you cannot know 10 the exact pattern, the checkerboard, of the core loading that 11 you are forced to look at a finite number of possibilities 12 and base.your' conclusions on those finite number of possibilities;
() 13 is that' correct?-
14 A That's true.
-15 Q Is there any particular reason why the analyses :
16 do not proceed immediately from the assumption of a mixed
~17 core?.
18 A. The reason for that stems directly from what I 19 mention in my testimony, that the exact reactor core loading 20 patterns ~cannot be defined in advance.
l 21 What we are' submitting here is the basis for~a 22 change to the license of a particular unit, such that regardless
.23 that what particular loading is inserted at the refueling its E
[ 24 effects are bounded by the results of the analysis that we've W: n corws, inc.
i 25 previously provided.
, s
.._ . _ _____.__..__ - . . . ~ _
318
- 7-16-SueW '1 To that end, that defines the methodology'that
(')
\d- 2 we use in the submitted safety analysis.
3 Q Now, you've used the word " bounded" several times.
4 Would you explain -- is.your understanding of-the word-
'5 " bounded" that this is an absolute maximum or minimum beyond
'6 which no fuel configuration would give you values either 7 above this bound or below this bound?
8 Is that what you mean by bounded?
9 A In the case of the 15 by 15 standard optimized to fuel configuration, we looked at a variety of cases with 11 respect to the geometry which you've already asked me about.
12 We: looked at a range ~of thermal conditions like
(]) 13 pressure, temperature, flow, part-power, full-power conditions 14 offer the range of which the DNB criterion is required to be 15 applied. Each of the cases results in a-calculated change 16 to'DNBR.
1:7 All of those calculated changes.in DNBR are less 18 than three percent.
19 Q Uh-huh.
20 A So that we've covered the spectrum of the 21 geometries, the spectrum of the thermal conditions. And no 22 calculated value exceeds three percent.
END #7 23 Mary flws w as n porters, i .
25 i
319
-p- O So if I understand you then, what you are N.)
say1ng 1s that you have examined a fin 1te subset, a rather small finite' subset of the possible core configurations of.
-core geometries for the mixed fuel and you have found that 4
in that subset none of the penalties exceeded three percenti is that correct?
A 'I wouldn't that is a small subset. We tested a wide -- we calculated a wide range of conditions.
0 I am-sorry. I thought that when we were 9
discussing the geometry that you had recalled three and that.those were the three that you had recollected. Am I misunderstanding you?
A The calculations set of the effects of the 13
- k[dh' 14 transition cores included three sets of conceivable.
checkerboard geometric conditions. It also included 15 different sets with respect to power distribution in 16 thermal conditions, which results in many more cases than the Phase III I referred to.
18 Q I understand. But in terms of the geometry of 19 t he 'cor e , you would agree that there are roughly three
-20 checkerboard patterns that you examined?
21 A Yes.
22 Q. And the examination existed not of physical 23 experiments, but of thermal hydraulic analyses of a 24 mathematical naturel is that correct?
25
320 m
s.
() 1 2
A That is not true. As I stated previously, we do have a full-s1:ed hydraulic experiment to determine the 3 resistance characteristics of both assembly types 4 individually and in a direct side-by-side configuration.
5 That gives us direct experimental information about the 6 resistance characteristics of each assembly and the 7 resistance characteristics in the transitional mixed 8 configuration.
9 That information is used in the thermal 10 hydraulic calculation of the effect of the transition cores 11 on the core safety analysis.
12 Q So the basis then of the experimental data 13 upon which these calculations are based is confined to Y
14 single fuel assemblies and side-by-side fuel assembliesi is 15 that correct?
16 A That is true.
17 Q Now on page 7 you refer also to also to 18 another penalty, and this has to do with something called' 19 rod bow, which is -- am I correct in understanding that 20 this is bending of the fuel rods as a result of their 21 experience in the core of the reactor?
22 A That is a reasonable explanation.
23 Q And could you, first of all, just explain 24 briefly why rod bowing affects ONBR calculations?
25 A That would require a slightly more detailed n
l 321
/m.
i
). I description of the nature of the fuel rod bowing. In a 2 reactor fue1 bundle the rods are supported at several axial 3 elevations by'the grids that form the structure, the 4 mechanical structure of the fuel assembly.
-5 These grids are located every two feet or so 6 axially up and down the fuel bundle. Therefore, at those 7 positions the fuel rod's position in the-lattice in the 8 core is defined.
9 In between those axial positions there can be
- 10 some horizontal displacement of the fuel rod from its 11 nominal position.to make it closer to its neighbor. This 12 displacement effect results in some rods being closer to 13 each other than what the nominal geometry was.
'(.h)
- ds- . 14 Because of that change in the displacement, 15 'that can affect the. local thermal hydraulic conditions and 16 therefore result in an effect-on the critical heat flux.
17 Q So are you saying then that because of the 18 closer. proximity of the fuel rods there is a' greater 19 likelihood that you m1sht get film boiling?
20 A- Not necessarily. If the core is properle 21 analyzed and protected, we can determine a basis for 22 quantifying the DNBR effect of rod bow, which when applied 23 to.the safety analysis calculated DNBR's, it still assures 24 the same 95 by 95 probability of not having DNB on the lead 25 rod.
322 1 Q Is it not the case, however, that the critical 2 heat ~ flux ~in the' case of rods which are severely bowed 3 would be lower?
4 A That is correct, and that is the basis for the 5 calculated DNBR's ce'rtainly for rod bow that I have 6 . described in my testimony.
7 Q -And this translates into an operational 8 criterion that you would not have as high a limit in heat ~
9 flux, or in temperature for that matter of operation ---
10 A No, there is no operational limit. The effect II of the rod bowing is accounted for by the uncertainty.that 12 I described in my testimony. There is no independent j
13 operational limit or criterion on that effect.
.ye
'14 Q Now is this rod bow penalty applied to the 15 LOPAR fuel? i l
16 A Yes, it is.
37 'Q Because there is no mention of it in your 18 testimony and I was just wondering about that.
!~ 19 With regard to LOPAR fuel, and I am not 20 familiar with the' refueling schedules'for LOPAR fuel, but i
L 21 is LOPAR fuel intended to be refueled at 12-month intervals 22 for the most part?
23 A' Well, I am not familiar with the particular 24 refueling schedules.on this unit, but either fuel is 25 refuelable. The refueling schedule is set by the utility
- ._________m_ . _ _ _ _ _ . _ _ . _
323
, ej s _) I who basically operates the plant and not by the fuel type.
l x>
2 Q Is there any way of determining what is the
! 3 maximum possible rod bow under normal operation l
circumstances and at what burn-up level does it occur?
~
4 5 A Well, we have taken -- as I describe in my 6 . testimony, we have taken measurements on a wide range of 7 burn-up, in other'words, time and history conditions of 1 8 fuel assemblies from operating reactors and have used that 9 as the basis for establishing the rod bow penalty.
10 Q So in other words, you are saying that the rod 11 bow penalty is based on direct experimental evidence?
12 A Yes.
13 Q. Is there a. reason why the rod bow phenomenon
([r')
\/ 14 'is not based just on mathematical calculations? Is it 15 possible to predict rod bow phenomena just mathematically?
16 A Well, we originally, you know, we have seen 17 rod bow in fuel assemblies and then by the observation of 18 that rod bow determined the amount of rod bow that we 19 account for in the safety analysis. So the original sources 20 .many years ago of the consideration of rod bow is based on 21 its observation in reactor fuel.
22 Q So the rod bow penalty is based on direct 23 experimental evidence and not primarily on mathematical 24 calculations, although mathematical calculations do figurei 25 is that correct?
Q)
I -
324 1 A That is correct.
b._ ll 2 Q Is the rod penalty accounted for in the 1.17 3 bounding value that you referred to earlier, the DNBR 4 bounding value?
5 A No, it is not.
6 Q Is it not the case that whether you have a 7 homogeneous core or a mixed core that rod bow would be a 8 consideration in DNBR calculations?
9 A It is a consideration and it is accounted for 10 by assuring that all of the calculated safety analysis II DNBR's allow sufficient space to the 117 acceptance limit 12 such that even when reduced by the appropriate rod bow 13 penalty, all the DNBR's are still greater than the 14 acceptance limit.
15 Q So is it correct then to say that the 1.17 16 acceptance limit is predicated on a core which does not 17 have rod bowing in it?
18 A No. The 117 acceptance limit is predicated on 19 the experimental data used to define the correlation. It 20 goes with the particular correlation and data set. The use 21 of the rod bowing and transition geometry, since they are 22 geometric characteristics of the plant and the fuel in the 23 plant, are applied to the plant safety analysis.
24 O So that in operational terms, is it true that 25 the 1.17 DNBR limit was derived from experimental data with
325 f.,,5 fuel which does not show rod bowing?
.;l
): 1 2 A That is correct.
3 Q Now on page 8 of your testimony you say that 4 i'The unce r t a i n t y associated with the application of WRB-1 5 -- the WRB-1 critical heat flux correlation -- to 15 by 15 6 OFA fuel is 0.0 percent." Is that correct?
7 A' Yes.
8 Q Do'you really mean 0.0 percent? In your
-9 opinion there is absolutely no uncertainty at all?
10 A In my technical-opinion, that is correct.
11 Q Do you know of any critical heat flux test 12 data involving the 15 by 15 OFA fuel assemblies?
13 A Not that specific geometry.
14 Q Have you personally reviewed the statistical 15- analyses which were conducted in order to justify the 16 application of WRB-1 to the 15 by 15 OFA fuel assemblies?
17 A I am familiar with that analysis. I did not 18- perform that analysis myself.
19 Q Are you convinced that those analyses indicate 20 a 0.0 percent uncertainty?
21 A Yes.
22 Q Is there a document in which the 0.0 percent 23 figure is cited?
24 A In the material we supplied to the NRC in 25 justification of the application of 1.17 DNBR for the 15 by
.a
326 r'
i.. L 1 15 optimized suel assembly. There was supplied analysis and 2 justification of no incremental uncertainty for the use of 3 the 15 by 15 OFA geometry.
4 Q Was there in that document the figure 0.0
$ percent?
6 A I don't know that.
7 Q Or in any documents which you have seen from 8 the staff or from Westinghouse?
9 A I don't know that.
10 Q So this 0.0 percent figure is your figure and 11 it reflects your judgmenti is that it?
12 A This figure is presented in relation to the 13 question from the Board addressing the uncertainties. We 14 were work'ing from the question of what is the basis of the 15 two percent uncertainty, and we would propose that at this 16 point in time a proper combination of the uncertainties 17 would include a zero value instead of two percent.
18 Q Further on page 8 you refer to the question of 19 the physical independence of the phenomena of rod bow and 20 variation in hydraulic resistance between low power and OFA 21 fuel.
22 Could you cite any experimental data which in 23 fact confirms that they are independent?
24 A Well, the hydraulic experiment that I referred 25 to previously provides the basis for the hydraulic
- 9 F f k
327
. n. - ,
(j ,' ) I characteristics and the different resistances of each (L '2 assembly.
E 3 'O Did those experiments involve rod bow fuel 4 displaying rod bow?
5 A No, but they involved as-built actual fuel 6 geometries.
7 Q So the experimental data which determines the 8 hydraulic resistance which consists of two bundles side by 9 side did not involve any fuel with rod bowing?
-10 A That is correct.
11 Q Then could you cite any experimental data 12 . which in. fact confirms that rod bowing and hydraulic
,, 13 -resistance are independent?
hJ'") i A If you take the geometry of a fuel assembly, a 14 15 fuel rod is over 12 feet long and it is supported at a two 16 foot elevation, at least 26-inch elevations by these grids.
17 Q Excuse me. I don't like to interrupt, but are-18 you addressing the question of whether you could cite any
.19 experimental data, or are you answering a different 20 question?
21 A I am' citing the basis of my response that the' 22 rod bowing does not affect the hydraulic resistance of the 23 fuel assemblies.
24 Q What I specifically would like to know is 25 whether you can cite any experimental data, physical sa
328
_h I experimental data which confirms that they are independent?
2 A I can cite no experimental data.
3 0 Okay. So you say there is no systematic change 4 which could influence the mixed core uncertainty. Excuse 5 me. Yes, here we are. This is on page 8, 6 lines from the 6 bottom. "There is no systematic change which could 7 influence the mixed core uncertainty."
8 Now are you saying that this is an 9 impossibility, or are you simply saying that it is 10 improbable or are you saying that it is just your judgment?
11 A It is my judgment that it is an impossibility.
12 Q It is your judgment that it is an 13 impossibility.
L.
14 A Yes.
15 Q So you don't think there is any even a 16 probability?
17 A No.
18 0 Can you refer us to any scientific study which 19 justifies this conclusion?
20 A As I-was previously stating, the length of a 21 fuel if over 12 feet long, and it is supported every 26 22 inches by a grid structure which serves as the structural 23 element of the fuel assembly.
24 The distances between adjacent fuel rods are 25 only on the order of about an eighth of an inch. In other
329 qq s.J ) I words, if you take the total checkerboard area, if you 2 will, of the fuel assembly, the vast majority of that area 3 is occupied by.the fuel rods.
4 .The deflections that occur with rod bowing are 5 in'most' cases much less obviously than that 1/8th of an 6 inch dealing with only a few hundredreths of an inch over 7 an axial distance of two feet in that the total localized 8 change in the area is very smooth and very, very small.
9 The total flow area o'f the fuel assambly is 10 unchanged because we are only dealing with displacements of 11 fuel rods from their n om '; n a l positions. There are numerous 12 engineering studies on the issue of effects of changes in
. , 13 area. This-change in area is far too smooth and far too
'- 14 insignificant to result in any hydraulic characteristic 15 change to that fuel assembly.
16 Q Are you saying that there are scientific 17 studies which quantifies this improbability which basically 18 state that this probability is negligibly small?
19 A The effect-of the area of changes on hydraulic 20 resistance is a common area of flue mechanic study and 21 experimentation. The area of changes that we are dealing 22 with here is nowhere near the type that induce incremental 23 hydraulic resistances of any sort.
24 Q So you are saying that based on your 25 e'xperience you consider it an impossibility?
330 bh 1 A That is correct.
I 2 Q But there are no test data to support that?
3 A I can't cite the specif1c experiments in 4 hydraulic effects of area change;.
(
i 5 Q Again'you say the flow redistribution, and 6 this is continuing on on the same page, page 8, "The flow 7 redistribution which occurs in the mixed core is not 8
capable of peflecting a fuel rod to any significant 9 degree." Does this statement fall in the same category and 10 are there any test data which determine this or which 11 demonstrate this?
12 A When we run the experiment, I define two 13 assemblies in the hydraulic test loop side by side. We run 14 that experiment for purposes other than just obtaining the 15 hydraulic data.
16 One of the other issues that our mechanical 17 design experts are interested in assuring is that there is 18 not any unacceptable v1bration or mechanical interaction 19 effect between the two assemblies. The fuel rods in the 20 experiment are instrumented to determine their vibrational 21 characteristics; in other words, the amplitudes and 22 frequencies of their motion during the hydraulic 23 characteristic typical of a reactor.
24 That is experimental evidence in addition to 25 our technical and analytical judgment that there
\
i l
c'
- 331 s
e
~'
- 1' 's'ignificant deflection.of a fuel rod in this type of an in - 2 environment.
3
.4 '
5 6
7
.8 9
- 10 11
~ 12
. 13
(:;p)
~% ' 1a 15 16 9
17 18
. 19 20 21 22 23
~24 25 sh .
C!: ' ,
9-1-JacW21 332 1 Q But this test data does not involve fuel which is 2 already bowed, or which is about to bow, is that correct?
3 A It is the same fuel assembly. It is a production-4 ,
i type fuel assembly with respect to the rods and the grids, but 5 it is the same mechanical structure that is loaded into the 6 Core.
7 Q Is it not the case that rod bowing occurs as a !
8 function of burnup, and is not expected in fresh fuel?
9' A That is true. You asked me whether the fuel in the 10 test was of the same type that would experience rod bow.
I 11 They are as-built --
as-built new fuel assemblies 12 l do not have rod bow in ther . i i
13 Q And were these new fuel assemblies which were used ,
l 14 in the test, or were these fuel assemblies which were old and ;
15 . subject to rod bowing?
16 A These are new fuel assemblies.
1 l'7 Q So,-in other words, there have been no tests done ;
18 involving fuel which is either bowed or which is about to bow l
10 because of its long burnup, is that correct, which would ,
l 20 involve the interaction between rod bowing and hydraulic l
., I 21 j .resis tence , is that correct? Do you have any test data on l l
I l 22 I that?
23 A I have answered previously that there is no ;
i 24
-') specific data on that geometry. '
% > F i J a m uters. w . j 25 And that would apply also to the statement the j Q
l l' j i
i l >
9-2-J0cWO1 333 1 flow redistribution is not capable o.f deflecting a fuel rod
(~)
A/ 2 to any significant degree. Would you apply that statment also 3 to a fuel rod which is about to bow, or getting to the point 4 where it is going to start bowing?
5 Is there any test data to confirm which fuel which 6 is about to experience bowing might not be affected systematically 7 by the flow redistribution? '
8 A It is my technical j udgment that the previous 9 axperimental data I have cited to demonstrate the non-deflection 10 of fuel rods in the transition core also provides experimental 11 evidences of non-deflection of bowed fuel rods. !'
4 i
12 ! Q So, the physical independence then of the phenomena 1
() 13 ' of rod bow and variation of hydraulic resistence between low 14 power and 0FA fuel is then a matter of your professional 15 opinion, but it is not based on any test data, is that correct?
/ i 16 A I think I cited the experimental data in the f
17 transition hydraulic test that furnishes some evidence of !
I I
18 Independence.
l 10 i Q Now, Mr. Dzenis, if I might return to the two 20 j! penalties that were discussed earlier. The penalty of three 21 percent --
22 A Which penalties? I am sorry.
23 Q These are the penalties of three percent DNBR 24
(~)
hFiJ Reporters, loc.
reducti'on to compensate for the variability and hydraulic
( 25 resistence between the two types of fuel in the core, and
n 9-3-Jr:Wal 334 1 the second penalty is the rod bow penalty, which is 5.5 percent, l' ,
2 which is attributable to the bowing of the rods.
3 Is it not the case that the second penalty, the a rod bow penalty, is based directly on experimental data, whereas 5lthe first penalty, the hydraulic resistence penalty, is based 6 ;j not on physical experimental data of an actual mixed core, but 1
7 rather of mathematical calculations?
e MR. BAUSER: I object to the question. It has been 9 asked and answered. !
10 ' MR. EDWARDS: Is it clear to the Board what the t
11 answer to that question is?
12 JUDGE LAZO: I think, perhaps, the witness should ,
/m
) 13 answer the question again.
- 14 WITNESS: The three percent transition core effect l 15 is based on the results of analyses of configurations of mixed 16 core geometries.
17 Included in that analysis are results of experiments 18 on the hydraulic characteristics of each fuel type. The rod 20 bow penalty is determined on the basis of observations of 20 ' f uel rod bowing in irradiated fuel samples.
11 .
21!! ,
Each effect has some experimental data which serves i i
0 l 22 j as the basis for elements of the calculation of a penalty. Each{
l 23 effect is calculated in following certain calculations. !
~
/N 24 Q One final question. If fuel operates according to us-L J Reponen, inc. j 25 the 1,17 limit attached to the WRB-1 correlation, derived from !
l
9-4-JocWal 335 1 the WRB-1 correlation, compared with the 1.3 DNBR limit associated 2 with the older W-3 correlation, does this mean that in the 3 former case the fuel would be allowed to run somewhat hatter 4 hefore violating the 95/95 standard, than in the latter case?
5 A The 117 or the 130, deal only with the probabilities 6 I of DNR not occurring at that particular evaluated condition.
7 They do not deal with what the temperature of the 8 fuel is at that condition. The temperature of the fuel is a 9 function of many other considerations well beyond any i
10 ! consideration of DNB --
li t
11 - Q Let me rephrase the question. Under the 1.17 DNBR '
i 12 limit derived from the WRB-1 correlation, is the permissible j i
.) 13 actual heat flux, the limit of the permissible actual heat flux,!
14 higher than under the 1.30 limit associated with the W-3 ,
i 15 correlation?
16 A The Turkey Point amendment, the peaking factor limits; i
17 on both fuel types are the same. In other words, each fuel i
18 type is allowed the same radio peaking factor, and the same i i
'O total peaking factor, so that in this case each fuel type, l
20 although analyzed in its respective correlation and compared :
21 to its respective acceptance level, each fuel operates at the 22 q same peaking f actor limitation of technical specification.
23 lN Q Well, Mr. Dzenis, what are the consequences of not l
24 granting the~ amendment? In other words, what difference does i
('l l ArFi_j Reponen, Inc.
1 25 it make whether you use one DNBR limit or another? l:
i i
l
9-5-JocWal 336 1 A The amendment which was filed to allow for.an
( 2 increase in the' peaking factor limits for both fuel types in
~
3 the Turkey Point reactor, was the -- the increase of these 4 peaking factors allows for more design and operat1onal 5 flexibility in'the setting of the reload cores for those 6 ! Operating plants.
7 9 If in fact, it doesn't make any difference in 8 ' terms of the actual heat flux which is permitted, how does 9 it give you greater flexibility?
10 A- When we design a reload core for any operating i
i 11 i plant, the utility provides us with .he design constraints
- 12 in. terms of the operation of the plant.
I
(') 13 They tell us when they expect the " plant to shutdown ,
14 for the refueling, and how long we have to design the next 15 cycle to operate.
l 16 We set a fuel loading -- in other words, the number i 17 of new fuel assemblies that will be supplied. The enrichment 18 they will contain, and the actual specific loading pattern l 10 based on those assumptions.
20 g Each fuel loading must meet all the constraints of 21 ' the license in terms of peaking factor limitations. With 22 an increased limit on the peaking factors, there is more 23 l flexibility in the te rms of the -- when the plant can shut
(^3 24 down and the plant can operate the next cycle with the same wnJ n.ponen. anc.
25 f.nel loading.
r 337
'9-6-JocW51 I Q I am sorry. I had difficulty.following the-latter
- (^}.
\- 2 part of your answer.
3 You talke'd about increased peaking factors?
4 .A That is what we requested. What was requested in 5 the amendment was an increase in the radio and total peaking 6 factor of the Turkey Point reactor. Reactors, excuse me.
i 7' Q- My question, though, has to do with the practical 8 consequences of this particular amendment, not another amend-9 ment having to do with peaking factors.
10 This particular amendment, I would like to know what 11 would be the practical consequences -- suppose, for example, that l
12 ' you were faced with the prospect of. designing a reloading r
-( ) 13 of fuel, and it was decided that you had to go with.the 3.0 14 DNBR limit for all the fuel.
15 What would be the practical consequences of that?
16 MR. BAUSER: Excuse me. Objection. The question has .
17 ' been asked and answered. The only amendments to which the 18 witness was referring were those that'are at issue in this t
10 proceeding.
20 Those amendments had to do with peaking factors.
!! ~
214 In addition, there was reference to a 3.0, I believe, DNBR.
22 qI am not sure where that came from.
23l JUDGE LAZO The request of the amendement was p 24 f rom 2.30 to 2.32?
Am F_,,_) Reporters, Inc.
25 WITNESS: Radio peaking factor from 1.5 to 1.62.
d
,-e, , . . . _ _ . .-
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- v. ? 338 9-7-JecWol -
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1 MR. EDWARDS: My probl'em>xMr. Chairman,;ts.that j( ) 2 I didn't recall asking a question about the peaking factgrs, s
3 I was wondering what were the pract(cal consequencies of this g.
4 particular amendment. not some other amendment. i j.
5 I understand there were other' amendments th t 6 related to peaking factors. ,
7 JUDGE LAZO: You understand that t'h'i s particular A
'8 amendment also relates to a" peaking, factor. '
s ,
9 MR .- EDWARDS: Contention (d). '\' w[, '
i s
't 10 . JUDGE LAZO: Oh, you said,tLis pary.idular amendment, j ,
- l. ,
s s 1 11 , MR. EDWARDS: Oh, I am sorry, excuse me.
I wasfs' i 12 limiting myself to Contention (d), which I understood was the
(")
U 13 sole subject.of this hearing.
14- JUDGE LAZO: All right. I ,think that is where our N
15 ' misunders tanding arose.
s ,,
.+
16 MR. EDWARDS: My apologies. I.used confusing '
17 terminology. May I' rephrase my, question? - N l s
s j' N,
18 BY MR. EDWARDS. (Continuing) '
z ,
to Q With regard to Contention (d), which deals solely !
}
20 with the DNBR ratio e the DNBR limit of -- what would be the 21 practical consequences if, instead of Amendment (d) bei.ng !
1
\
22 !, Sranted, whereupon 1,17 would be the DNBR-limit, you were 23 faced with a situation where 1.3 as under the old W-3 corr' elation -
i 24 l were the DNER limit.
Am F Reporters, Inc. '
25 7tR. BAUSER: Excuse-me. I have to object again.
l i' 1 0
{
339 9-8-Jc Wal 1 This amendment does not in any way address the 1.17 DNBR 2 acceptance limit.
3 This amendment has only to do with certain technical 4 specifications involved primarily with peaking factors, 5 and therefore the question is as stated, without basis.
6, JUDGE LAZO: Well, again it is the use of the word, !
7 ' amendment.'
8 MR. EDWARDS: I an afraid I have done it again. l 9; Perhaps I could get some guidance from the Board on this.
I 10 I am addressing the questions which the Board posed.
11 The first question is whether the DNBR of 1.17 which the ,
12 amendments impose on the OFA fuel in Units 3 and 4 compensate
- s. !
)* 13 ' for the three uncertainties outlined by the Staff, and this I 14 think we have a clear answer to.
I 15 The second question is whether if the DNBR of 1.17 16 does not compensate for those uncertainties, the SRP's 95/95 17 standard or a comparable one is somehow satisfied, and the I
18 third is whether if that standard is not being satisfied, the
'O reduction in the margin of safety has been significant. ,
l 20 Now, particularly in connection with the latter j i
21 1 question, I would like to know what would be the practical i 0 l 22 li consequences in terms of the fuel loading regime of the plant, 23 if li this contention were -- if the contention -- if the Board 24
'~'l were to rule that the contention was valid, and that the 1.17 ka-Fi j Reporters, Inc.
25 DNBR ratio is not applicable to the OFA fuel in Units 3 and 4, l
I ll, I
9-9-JcCWol i
1 that is really the question that I should have been asking.
(}
\/ I am afraid I did a poor job of it. Is that 2
3 understandahle, 4; JUDGE LAZO: That seems quite clear to us. Do you 5 understand the question?
6 WITNESS: I would interpret that to mean that we 7 supplied all this DNB related material in addition to a lot 8 of other material in a lot of other areas in support of an 9 operating license amendment for the increase in peaking factors 10 : at the Turkey Point units.
I 11 S We furnished this information in' order to demonstrate!!
12 f that we had met all the appropriate safety criterion, the 95 by
() 13 95 DNBR acceptance limit. ,
14 Then, taking into account the geaking factor limits 15 - that have been requested in this amendment.' khat is why I have I l
16 taken -- why I have answered the question with what the practical i
17 implications of the lack of the change in the peaking factor {
i
^
18 limits would be. l l
10 BY MR. EPWARDS: (Continuing) l 20 .l Q Was the request for the increase in peaking factors 21 related'to the new DNBR -- the application of a new DNBR limit 22 l. of 1.17 ?
23 A The use of the DNBR limit of 1.17 was established
-(~y 24 on a previous amendement dealing with the induction of the Am FU Reporters, Inc.
25 optimized assembly. That assembly was --
the acceptance of
- - - _ - - - _- ., - - _ . _ ~
10-JooWo1 341 i
j that assembly and its appropriate correlation.and its limit,
(^h 1 s_/ 2 was addressed in detail in a prior amendment to this one. !
3 The question arose because we have to continue to 4 , address all the fuel types which will exist in that core during 5 th* period with the increased peaking factors that were requesta i 4 in this particular amendment, and that is why we have presented 7 the material -- the material --'the safety analysis being 8 heyond the DNB acceptance limit, because that is what 9 establishes the basis for. acceptability of the amendment.
10 Q Now, how does the peaking factor affect DNB?
11 A' The -- an increase in peaking factor tends to l
12 !. decrease the DNBR.
4 (). 13 ' Q And consequently, is.it safe to conclude, therefore' ,
14 that an increase in peaking factor would naturally involve a 15 collelary decrease in the DNBR limit?
16 A Well, the - 'more.recently the material in support 17- o.f th.e increase in peaking factors, we demonstrated that 18 including analysis with the increased peaking factors in it, I.
!c 'and accounting.for whatever sensitivities there are, the r .20 effect of the-change in the peaking factor on DNBR, we still
.y 1 the same 95 by 95 criterion when used with the appropriate 21l meet 22 correlation for that particular fuel type, so that the analysis, t i
23 including _the effects of the 162 radio peaking factor result fS 24 in the 134 values, which when reduced by the appropriate am-Fd ft pori .. inc.
25 uncertainties, are still greater-than the DNB acceptance limit
9-ll-JccWal 342 1 of 1.17. -
2 Q I would like to come back to this DNBR question 3 which is central to Contention (d), and it has to do with this 4 y question of the reduction in the margin of safety, and also the 5 question of the 95/95 standard.
6i .Is it not true that the 95/95 standard is intended 7l to he quite a precise statistical measurement of acceptability /
- 8) A That is true.
i 9 Q And is it not true that that 95/95 standard was 10 derived from experimental data using only one fuel type at a Il time?
12 A The 95/95 standard is a fundamental assumption and !
,- j O 13 the data for each particular fuel type is analyzed, and an l Id appropriate DNBR acceptance level is derived consistent with the 15 95 by 95 assumption.
16 Now, I understand that there were experimental data Q
17 involving homogeneous ~ fuel arrays, in which the 1.17 DNBR limit 18 was derived for OFA fuel, in conformity with the 95/95 acceptance criterion, is that correct?
20 A That is correct.
21 ] Q Where is the specific evidence that in the case of 22 Turkey Point that this limit is acceptable for the Turkey Point 23 lplant?
i
() 24 l As-F_ - J Reporters, Inc. l
.A It is included in the calculation of the uncertainty 25 for transition cores, which is applied to the safety analysis i
l
9-12-JccWc1 343
] DNBR's.
(.) 2 Q So you are saying that the calculation of the 3 complicated geometry of the transition cores is sufficiently g, precise to guarantee 95/95; 95 percent probability within a c
5 95 percent confidence interval, that DNBR will not exceed or 6,fgobelow that DNBR?
I 7 l A Yes. ,
I i 8, Q Where does the decision of these mathematical I
9l analyses, these thermo-hydraulic analyses, where does the i
i I
I 10 '! Precision, the 95/95 precision reside in those analyses? l 11 A The most fundamental 95 by 95 decision, if you will, 12 is in the assurance that we have a sufficient data base, i
,~ .
( ) 13 sufficient methodology and sufficient correlation to properly 14 model the data that was the basis of the correlation.
15 Q Okay. Could I take those one at a time, please?
16 A Yes.
17 Q You say a sufficient data base?
4 18 A Yes. '
m Q What is the sufficient data bcse for the Turkey Point' i
i 20 -core? ,
21 , A The -- you have asked me for the basis of the 95 l i
22 m by 95, now you have jumped to the Turkey Point core.
23 q I am sorry. What I am discussing is specifically i
,3 24 ! the applicability of the 1.17 DNBR limit to Turkey Point.
AnF j Reporters, Inc.
25 A The data base upon which the WRB-1 correlation was l
344 9-13-JocWol 1 derived, and the range of geometric parameters under which
() 2 the correlation is applicable includes the characteristics 3 of the Turkey Point 15 by 15 optimized fuel, in particular 4 the rod diameter, the pitches, the grid spacings, all the 5 hydraulic diameters, and the grid structure of the zircoid 6 CYPe which is included in the optimized fuel assembly.
7 All of those features are included in tests which 8 are in the data base of the WRB-1 correlation development, and 9 I would say the ranges of applicability of that correlation.
10 ] Q In the case of a mixed core, where you have hydraulic 11 ' resistance variations, and consequently flow diverted through the 12 low power field that might otherwise go through the OFA field,
() 13. do you know if 95 probability and 95 percent confidence what 14 the actual flow through each 0FA fuel' bundle is going to be? l l
15 MR , BAUSER: Excuse me. I object to the question.
16 ' I t has been asked and answered. The witness has explained I 17 tests that ~were done in connection with the effect of having l 1
18 assemblies in the core of different hydraulic resistances, as !
10 expected, how bounding values were selected based on that
- 20 ' inf o rmation , and then applied to the~different possibilities.
21 ; MR.. EDWARDS: Mr. Chairman, I know en'ough about u
22 j geometry, fluid flow, to realize that the complexity of the i'
23l geometry is a very important determining factor in fluid flow 24 problems, and what I would like Mr. Dzenis to explain is how
("}
Am FV Rept rters. Inc.
25 one can, by studying two bundles side by side, arrive at a l !
9-14-JooWal 345
- I decision which would effect all possible field configurations l
2 in a core such as Turkey Point.
3 JUDGE LAZO: That is a somewhat different question.
- End-9. 4 l
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346
- 10-1-SueW I MR. EDWARDS: And how one can have the desired ,
> ~
2 degree of confidence, the 95/95 criterion.
3 JUDGE LAZO: Mr. Dzenis, can you answer that 4> question?
5 WITNESS: The tools that we use to analyze the 6 flow conditions in the reactor cores in conjunction with the 7 subchannel analytical methodology are derived in order to 8 properly model the flow and coolant temperature distribution 9 in the types of geometries to employ here. -
10 They are reviewed and verified against that type II of geometry. I would beg to differ with what you term as 12 the complexity of the transition geometry relative to a (j 13 homogeneous core geometry.
I4 BY DR. EDWARDS: (Continuing) 15 Q Is it not the case that in deriving the WRB-1.
I 16 correlation for critical heat flux estimation, is it not the !
l 17 case that in order to verify the 95/95 criterion was satisfied 18 there had to be a painstaking process of calculating and check-19 ing against physical experiments to see whether the critical 20 heat flux that was predicted corresponded to the critical heat 21 flux that was experienced?
l 22 Is that -- am I correct in understanding that that's!
23 the process by which the 95/95 criterion was established?
24
/ A That's correct.
W#Q],! Reporters, Inc.
l 25 I Q What I'm getting at is, where is the comparable l
l i
347 J-
- 10-2-SueW 1 check with experience in the case of a non-homogeneous core?
2 How can you take your calculations and check them against 3 physical experiments in order to see whether in fact the 4 theoretically predicted hydraulic patterns correspond with 5 95/95 percent confidence to the actual experienced hydraulic 6 ~ patterns _in the core?
7 A I think I'm going to have to explain again that 8 there is a defined data base which establishes the base for 9 the WRB correlation and establishes the DNBR acceptance 10 limit of the 1.17 on a.95 by 95 basis.
11 Inherent with that is the subchannel analysis ,
12 analytical methodology which is used to analyze the-data
- () 13 from'the experiment and the range of conditions. There is 14 a geometry of the fuel bundles, the nature of the grids, 15- the range of thermal-dynamic conditions over which the 16 correlation is applicable.
17 The limit, the DNBR acceptance limit, only applies
[
18 .within those clearly defined ranges of applicability.
'19 Now, the question of, does a particular plant 20 analysis, safety analysis, meet that requirement deals with
, 21 the question of, well, is the probability of the DNB on the 22 hot rod in that core higher - . probability'of'DNB.not-occurring, 23 being higher than the 95 by 95, which is to say it's a
[p- 24 requirement.
FL Reportees, Inc.
i 25 It is in the analysis of that hot rod configuration I
348
'#10-3-SueW I that any question of what's the environment within the core, i pd 2 which is the changes in geometry, what is the amount of fuel 3 rod bowing, all of those factors are accommodated for by'the 4j conservative uncertainties that apply to the safety analysis.
5 Also, it is verified that the ranges of conditions 6 and the ranges of thermal-dynamic conditions, the geometry 7 of the fuel and the nature of the grid striking is consistent 8 within the ranges of defined applicability for the DNB cor-9 relation. 4 10 It is on that basis that since the -- all the 11 ' physical parameters which are to be evaluated are within 12 the range of conditions of applicability of the correlation, 13 then that correlation can be used to evaluate the conditions Id- in the safety analysis.
15 Q So with regard-to the data base, which is used 16 in assigning the penalty for hydraulic resistance, I think 17 you've made it perfectly clear what the data base consists 18 of, two side by side bundles.
19 And I've asked the question, and I think you've 20 answered it, about where we can get 95/95 confidence about
~
21 what is actually going on in the OFA fuel in the core. l 22 I would like to turn to the second point that 23 you raised a moment ago, and that is the methodology. You O 24 no-e w a. corm s,inc.
referred to the data base and the methodology.
25 JUDGE LAZO: I wonder, Dr. Edwards, if we are going l l
349
- 10-4-SueW 1 on into a somewhat different area if this might not be a
( )- 2 convenient time to take a luncheon recess?
3- DR. EDWARDS: Certainly.- No objection.
4, JUDGE LAZO: Very well, then. We will recess 5 until 2 o' clock.
6 (Whereupon, the hearing is recessed at 12:34 p.m.,
7 to-reconvene at 2:01 p.m., this same date.)
8
'9 10 11 12
() 13 14
?
15 16
-17 18 19 20 21 22 .
23
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h-Fi v 24 Reporters, Inc.
25 l
350
- 10-5-SueW l A_ F_ T_ E_ R N_ O_ O,N P_ R_ O_ C_ E_ E_ D,I_ N_ G S, 2 (2:01 p.m.)
3 JUDGE LAZO: Will the hearing come to order?
4 .
Dr. Edwards, are you ready to proceed?
5 DR. EDWARDS: Yes, Mr. Chairman, I am. Thank you, 6 Mr. Cha~irman.
7 Whereupon, !
L 8 EDWARD A. DZENIS t
9 resumed the witness stand on behalf of the Licensee and,
~
10 'having previously been duly sworn, was furthir examined and 11 testified as follows:
4 12 CROSS EXAMINATION 13 BY DR. EDWARDS: (Resuming) i- 14 Q- Mr. Dzenis, I would like to return to a point 15 which you were mentioning over which there was a little bit i
16 of confusion on my part I think prior to breaking for lunch. I 17 And this had to do with my poor use of terms in ;
18 . terms of the contention versus the-amendment. And you had 19 mentioned-the increasing of the peaking factors, that's the 20 F Delta H and FQ.
21 A That's correct, yes.
22 l Q Is'it the case that the increase.of those peaking j L ~
i 23 limits allows-the Turkey Point reactors to run the fuel l
24 hotter than they would have had the peaking limits been' set W-F1 Reporters, Inc.
! .25 -lower?
t s
I
- _.- c.- m - ~ , . . . - -. _ . . . , . . , , . . . . . . . . . . . . , , . . . ...._:.. ._,,. - , , _ . . - . . - . - . .
c 351
- 10-6-SueW 1 A The increase -- the peaking factor limit change !
k- 2 represents a change in the allowable limits. It is not )
3 possible to design cores which are exactly at those 4 limits.
5 Q Right.
6 A~ .In the assumed case of two cases, all else equal, 7 one operating at a higher peaking factor, the other there is i 8 a small change, less'than a few degrees, and the difference 9 Ein the operating temperatures is the fuel in those two 10 cases. .
11 Q But ycu would agree that it does allow more 12 margin at the upward end of the temperature scale'in terms
() 13 o'f planning for operational contingencies, expected transients 14 and so on?
15 A No. -The temperature I'm referring to is localized 1-6 fuel cladding or cooling temperature. The. operating tempera-17 ture of the unit as a whole and the operating power level of 18 the unit as a whole is unchanged by this amendment.
19 Q So, this would effect primarily the hot fuel rod?
20 The hottest fuel---
21 A That's right.
22 Q -- rod in the core?
23 A The hot spot location, that's correct.
(3 24 Q And consequently this is where the peaking factor .
4FLJ Reporters, Inc, l 25 relates directly to the DNB argument?
l l
I' l
352 i
- 10-7-SueW l A That's correct. i
~
1-) 2 Q So, it is true then that running the fuel hotter l
3 just -- if you were to run the fuel hotter, this does take 4; you closer to DNB in absolute terms?
5 A No. The DNB criterion is evaluated based on the 6 conditions which feed to the DNB correlation, not the cladding 7 temperature itself per se.
8 Q No. Excuse me. I'm not talking about the DNBR.
9 A Yes.
10 Q But the phenomenon of departure.from nucleate 11 boiling. If you -- what I'm saying is that if the hot fuel 12 pin in the core --
() 13 A Yes. i 14 Q -- is to run at a hotter temperature, then it is 15 that much closer to the possibility of departure from nucleate i
16 boiling; is that correct? !
I i
17 A The only -- the limitation with respect to DMB l i
18 phenomenon, the DNB phenomenon is effectively precluded by 19 the hot -- in terms of the definition of hot with respect to 20 the DNB criteria -- meeting its DNBR exceeding the DNBR 21 acceptance limit of 1.17. i 22 That -- the meausre of DNB occurring or not occur- '
23 ring is not directly relatable to the mere temperature at !
(~' , 24 which the rod operates.
w-F; _/ Reponers, lnc, j 25 Q I'm a little confused. I understood that DNB was l
i i
l
353
- 10-8-SueW 1 .a. physical phenomenon which was related to a critical heat n
2 flux and that that. critical flux was a physical phen menon 3 -which one can either get closer to the critical heat flux or 4; less close to the critical heat flux.
.5 Is it not the case?
6 A '- For therurposes of explaining the phenomenon, for
.7 example, in didactics fashion, it's appropriate to refer to 8 the physical phenomena in that type of a relationship.
9 I use those types of terms in explaining the
.10 fundamental aspects of the DNB physical phenomenon. When 11 relating to the' matter as to whether a particular reactor 12 core is likely to undergo a phenomenon or not, the specific
() '13 analytical tools and procedures and correlations come into 14 play.
15 And it is not possible on the basis of the descrip ;
16 tion of a particular temperature to determine whether the DNB 17 phenomenon will or will not occur.
I 18 Q Would you agree that there is an uncertainty 19' -involved in the occurrence or nonoccurrence of DNB when you 20 .are at the upper end of the temperature scale?
.21 Within the acceptable limit? Within the DNBR
-limit?
22 l
23 A I don't understand that question'.
24 Q~ Do you believe -- you said a moment ago that pF Reporters, Inc.
- i. 25 you believes that as long as the fuel, let's say, the hot fuel
! i 1
354 i
- 10-9-SueW I pin is operating at a temperature, at a heat flux, which is l
' ' ' 2 within the DNBR limit that it effectively precludes the 1
3 phenomenon of DNB?
4 Is that your understanding?
5 A I said that if calculated DNBRs are in exctss of 6 the DNBR acceptance limit, then DNB is precluded.
t 7 Q Is precluded? Would you say that that's a zero 8 point zero percent probability, then?
9 A I -- if you wish me to go to the equivab3nt l
10 statistical statement that the 1.17 DNBR acceptance limit II when analyzed with the NRB-1 correlation assures that DNB 12 will not occur at that condition at 95 percent probability n
(). 13 with 95 percent confidence.
Id Q So, what would be the converse of tP_at statement?
15 Suppose it were operating precisely at the DNBR limit,' what .
I 16 would be the probability that it would occur? That departure !
I7 from nucleate boiling would occur?
18 A I don't think I understand your question.
19 Q Is there a probability that departure from 20 nucleate boiling could occur if the fuel were operating at I
2l the limit, the DNBR limit? l 22 A The statistical -- it would seem to me that there l l
23 is a statistical inverse to a phenomenon not occurring , that !
24 f- the phenomenon would occur. !
4.-Fi]jnconm.ine. ;
25 And can you assign any probability to that?
Q Yes.
1 0 !
355
- 10-10-SueW1 A I presume that one hundred minus ninety-five 2 percent would be five percent.
4 3 Q Five percent?
4 A Yes.
5 Q Okay. So, five p'ercent probability is what 6 you mean by precluding? That the possibility of depature l
7 from nucleate boiling is precluded?
8 A No, that's not what I said. I said when the 9 event is precluded at the 95 by 95 level on the hot rod 10 of the core when defined by the methods that I describe in 11 my testimony, that effectively precludes the phenomenon 12 of DNB in that core, which is the question you asked me.
() 13 0 .So, you are saying that even if the core were 14 operating at the DNBR limit the phenomenon of departure from 15 nucleate boiling within that core, including the hottest ;
16 fuel pin, is precluded?
17 MR. BAUSER: Excuse me. I object to the question.
18 I do not see how the question, Mr. Chairman, relates to the 19 scope of this proceeding or in is within the scope of this L 20 proceeding as defined by the Board in its three questi'ons.
21 It seems to me the question is actually raising !
22 an issue that the Board has identified as not being proper l 23 for dispute in this proceeding.
I
('s ks-FQj Roorms, lm.
24 DR. EDWARDS: Mr. Chairman, if I might call vour l 25 attention to the three questions which the Board identified I
l-
. . . , , , . . . . . . . . . . . . . . . , , _ , . , . . . . . . , . . . . . . . . . . . . , _ _ _ , , _ , . . . , . , . . . . . , , , , , . . , , , , , , _ _ _ m . , ,,, , , . . , , . _ , .
356
- 10-ll-SueW 1 on the question of Contention D, I believe that'the third
_ e-l s_/ 2 question, or'is it the second -- the second question relates 3 specifically to the.95/95 criterion, and part of the intent 4 l, of my question is to see whether Mr. Dzenis' interpretation
.5 of the 95 percent probability, the criterion which is involved 6 in acceptability of DNBR, is~ commonly understood between Mr.
7 Dzenis and myself. !
8 JUDGE LAZO: Yes, we agree. You may answer the 9 question.
10 Do you want it repeated, sir?
11 WITNESS: I think Dr. Edwards and I understand 12 equivalently what 95 percent probability at'95 percent
.(). 13 confidence means, if that's the, question.
14 BY DR. EDWARDS: (Continuing) 15 Q Well, this is where I'm having some difficulty, '
16 because I would not use the word " preclude departure from.
17 nucleate boiling" when there is a five percent probability i
18 that it might not be precluded.
19 So'-I'm just asking.you really to explain your 20 . terminology I guess.
21 A The questioning began asking me about what I 22 felt the effect of the amendment on DNB at Turkey Point was.
23 And my response there was that DNB is effectively precluded
'24 at Turkey Point either way, either before or after the w}e]neoorem m inc.
25 -amendment.
l- l.
357
- 10-12-SueW1 The basis of my statement on that is contained in ;
x l
-' 2 my testimony which explains the basis of the particular DNBR 3 acceptance limit and the DNB correlation which is based 4 .on its appropriate data base and its derivation to meet the 5 commonly accepted and approved 95 by 95 basis for the purpose 6 of setting that particular DNBR ' acceptance limit.
7 It also contains a' description of the bounding 8 ; conservative approach taken to the use of all the input 9 values used in.the evalu'ation of the plant safety analysis 10 and the determination of the particular uncertainties that 11 are in the Board's questioning, in that that material relates ;
12 directly to the question of what's a hot rod in that reactor.
-( ) 13 Now, the evaluation of~that hot rod is done on a ,
14 very conservative bounding parameters basis. When the DNBR 15 for'that bounding value's conservative basis hot rod exceeds 16 the 95 by 95 correlation acceptance limit value, I would i
17 consider that to be effectively precluded in that particular l 18 reactor application.
19 Q So you are using the word " precluded" to basically I
20 mean 95 percent probability?
21 A No, that's not what I said. !
l
- 22 0 Okay. Let me move to another question, then, !
23 related to this subject. And that is, you have said that i
24 the Turkey Point Units 3 and 4 could run acceptably under l Es-FG_ Reorwrs, lm.
l 25 either the amendment or if the amendment were in force, as it is i
358 l
- 10-13-SueW i now, or if the amendment were not in force --
I) 2 l A That's correct.
3 Q -- the Turkey Point Units 3 and 4 would run at 4L acceptable levels.
1; 5 Would it require changes in the operational
'6 Parameters of the plant if this amendment v: .e to be denied 7 at this point? '
8 A If the amendment were deleted, if one would 9 hypothesize -- I still have heard no technical basis for 10 its deletion, but if it were to be deleted, the conditions 11 of the license would revert to their condition prior to the 12 amendment request. '
() 13 And that was a condition that they operated under 14 Prior to that amendment.
-15 Q Is it your understanding that one of the purposes 16 of the amendment was to' avoid a power limiting reduction in 17 the FQ. limit?
18 A No, that's not true.
19 DR. EDWARDS: Mr. Chairman, I'm a bit at a loss as
- 20 to the precise procedure here. I'm not a lawyer.
21 I have a letter in my hands on the letterhead of 22 Florida Power and Light company addressed to Mr. Eisenhut, 23 Director of the Division of Licensing of the Nuclear Regulatory i
24 Commission, which contains a paragraph that I think is perti- !
6..FMlh amorws. iu. I 25 nent to this most immediate question and answer.
m_ _ . . _ _ _ _ _ . _ _ _ . _ _ _ _ . . _ _ . _ _ _ _ _ _ _ _ _ _ . __ _ _ . _ _ _ _ _ . . _ __ _ ._ _ _ _ _ _ . . _ _ _ _ _ . _.
l 359 l
'#10-14-SueW1 Is there -- can I identify this letter and perhaps l 2 file it as an exhibit and ask a question based upon that?
3 JUDGE LAZO: You can identify it. You may ask 4 the witness if he has ever seen it --
5 DR. EDWARDS: Okay.
6 JUDGE LAZO:. -- is familiar with it.
7 DR. EDWARDS: 'Very good.
8 JUDGE LAZO: I think you should show it to him --
9 DR. EDWARDS: Sure.
10 JUDGE LAZO: -- and then if you are going to ask 11 questions regarding'it, he should have an opportunity to 12 study it.
( )- 13 DR. EDWARDS: Fine. Thank you.
14 MS. YOUNG: I request that.he show it to opposing 15 counsel also.
16 JUDGE LAZO: Yes. Why don't you right now show 17 it to counsel for the other parties?
18 DR. EDWARDS: 'All right.
19 (Dr. Edwards is showing the letter to all counsel.)
kND#10 20 Mary,f1ws 21-
,. 22 ,
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E 360 MR. EDWARDS: Mr. Chairman, shall I show it to
, "l x .
v) the Board as-well, or shall I read the relevant paragraph?
JUDGE LAZO: Well, I don't believe there is any
~
need to show it to the Board at this time. It has not been 4
determined whether is-going to come into the record or be part of the record.
6 MR. EDWARDS: Fine. I apologize for not knowing 7
the proper procedure, but if I might just ask ---
8 JUDGE LAZO: You ar, doing just fine.
9 MR. EDWARDS: --- if i could just ask Mr.
10 Dzenis whether he is familiar with unis letter 11 JUDGE LAZO: Does he now have a copy of it?
MR. EDWARDS: Yes, he does. He has a copy in 13 w.,
'/ ) front of him. --- whether he is familiar or whether he has
%s 14 previously seen this letter dated July 6th, 1983 directed 15 to Mr. Eisenhut.
16 BY MR. EDWARDS:
17 Q Mr. Drenis, have you seen this letter previously?
19 A No,-I have not seen this letter previously.
20 Q Well, I would call your attention to the 21 second paragraph under point one, Turkey Point Unit 3 Cycle 22 9 Licensing, which reads as follows:
23 "As discussed in Reference 1, an increase in 24' the F Delta H limit of the technical specification is 25
,. ~p
361 1 required to achieve the desired flux reduction in Turkey 2 Point Unit 3 Cycle Nine. In addition, to achieve the higher 3 F Delta H, without a power limiting reduction in the FQ 4 limit, approval of the improved LOCA reflood model BART was 5 requested."
6 A That is right. It . indicates that in order to 7 design a core which utilizes the increased F Delta H for 8 the purposes identified in the letter, a corresponding 9 increase in FQ must also be included. Now by the deletion 10 of the amendment, both F Delta H and FQ would be 11 constrained.
12 Q And judging from the text of this letter, the 13
. author would have concluded that this would be a power I
14 limiting reduction?
15 A I wouldn't agree with that conclusion.
16 0 You would not agree with that?
17 A No. The way I read this is if the increase in 18 the F Delta H limit is requir'ed to achieve the flux 19 structure, and if that is implemented and you design a core 20 to take advantage of that F Delta H increase, you need the 21 corresponding FQ increase to go with it.
22 And, as aaid before, the amendment requested 23 the change of both parameters. I think this discussion 24 Indicates the relationship of the rationale as to why both 25 parameters are being changed at the same time.
t i
.362 Q So it is'your belief'then'that if the
) 1 2 amendment were not denied', this would'have no implications 3 for the power of the Turkey Point Units 3 and 4 reactors?
f-4 A If.the amendment were not denied or ---
^
5 Q If this amendment were denied that this would 6 have zero implications for the power ou'put t of Turkey Point 7 Units 3 and 47
-8 A The' rated thermal output of the plant could be 9 achieved without this amendment.
10 Q The rated thermal output ---
11 A Yes.
12 Q --- but I am talking about actual.
13 A Well, t'he plant operates at its rated thermal
.14 output.. A' condition of the license is the rated thermal 15 output of the_ plant. That is defined as a term of the 3 license. The plant operates at the rated thermal power.
p 17 Q What is the r a t e'd thermal power?
18 .A I believe it is 2200 megawatts thermal.
19 Q And what is the capacity. factor?
20 A I don't know what the capacity factor is.-
21 JUDGE LAZO: Mr. D:enis,' the recipient of the.
22 letter and the date has been identified, but I don't think 23 anyone mentioned the author.
24 THE WITNESS: The author is Robert E. Uhr_ig, 25 Vice President of Advanced Systems and Technology of
>J
363
( ) i Florida Power and Light Company.
2 JUDGE LAZO: Would you read into the record the 3 relevant paragraph that we have been referring to here, the 4 paragraph regarding. power limitation.
5 THE WITNESS: This is the third paragraph on 6 the first page of the letter. It says "As discussed in 7 Reference 1 -- and Reference 1 is Florida Power and Light 8 . letter L-83-180 R. E. Uhrig to S. A. Varga, NRC, Turkey 9 Point Units 3 and 4, Docket Numbers 50-250 and 50-251, 10 Pressurized Thermal Shock, dated March 25th, 1983.
11 It says "As discussed in Reference 1, an 12 increase in the F Delta H limit of the technical 13 specification is required to achieve the desired flux f: '
14 reduction in Turkey Point Unit 3 Cycle Nine.
15 "In addition, to achieve the higher F Delta H 16 without a power limiting reduction in the FQ limit, 17 approval of the improved LOCA reflood model BART was 18 requested. FPL will be submitting for NRC approval 19 technical specifications and supporting safety analyses to 20 raise the F Delta H technical specification limit and a new 21 LOCA analysis for both Turkey Point units."
22 There is more to the paragraph, but I think 23 that is the section they were referring to.
24 JUDGE LAZO: Thank you, sir.
25 MR. EDWARDS: Thank you, Mr. Dzenis and thank
r 364 you, Mr. Cnairman.
'%} }
2 BY MR. EDWARDS:
3 Q MR. Dzenis, I would like to return to the 4 meaning of the DNBR limit of 1.19 through the WRB-1 5 correlation that we discussed earlier.
6 JUDGE COLE: Do you mean 1.17?
7 MR. EDWARDS: I am sorry, 1.17, yes. Thank you.
8 BY MR. EDWARDS:
9 Q Am I correct in saying that_the DNBR limit of to ~1.17 is to be applied with the appropriate penalties added 11 afterwards-to the calculated DNBR limit?
12 In other words, you have a calculated DNBR 13 through onsite, plant-specific analysis, you have DNBR
(~7%x_/ 14 limit of 1.17 which does not include consideration of such 15 things as rod bowing phenomena, and the way one reconciles 16 these two is^to take the calculated DNBR and subtract 17 cert'ain penaltiess and see if the answer is still greater 18 -than the 1.17 limiti is that correct?
~
19 A That last summary is correct.
20 Q That last summary is correct. Thank you.
21 Now-in principle if the limit were something 22 other than 1.17, would the same philosophy apply? In other 23 words, suppose that experimental data had shown a 1.20 DNBR 24 limit as the appropriate limiti would one then follow the 25 same procedure,-i.e., do onsite, site-specific calculations s3 l
eg
365 (m g 1 of the DNBR and then subtract penalties and see if the 2 result if greater than the 1.20?
3 A Let me clarify-two points. First of alli the 4 methodology that you have used is fine, in otner words, 5 with three steps to it.
6 First of all, the definition of an appropriate 7 DNBR acceptance limit to define an experimental base with a 8 particular correlation for a particular fuel type is an 9 appropriate first step. That methodology applies to a broad 10 range of fuel designs and for different correlations and 11 with different numerical values across the industry.
12 The second case of the calculated DNBR's, that 13 calculation does not occur at the power plant site. Those 14 calculations are performed in advance of the operation of 15 the olant and are of evidence of meeting the safety limit 16 prior to the acceptance of the amendment to the operating 17 license.
18 But with that correction, moving on from here, 19 that analysis is performed, as I have described in my 20 testimony, with conservative input assumptions consistent 21 with the technical specifications and consistent with the 22 other rules defined -- or consistent with the standard 23 review plan methodology.
24 And then that calculated DNBR is reduced by 25 the appropriate uncertainties or penalties, whichever you
(
366 m
' care to refer to them as, and then compared to the j I V;
2 DNBR pcceptance limit determined in step one.1 3 Q Now if one goes by the W-3 correlation, the 4 appropriate. figure would be a.DNBR limit of 1.3; is this 5 correct?
6 A Yes, that is correct.
7 O And if one followed the same procedure with a 8 DNBR' limit of 1.5. would it not be the case that the 9 presently calculated minimum DNBR for Turkey Point Units 3 10 and 4, i.e., 1.34 would already be in violation of the 11 acceptable DNBR limit?
12 A The calculation of the DNBR must be done with j3 the1same correlation that you are comparing as the limit
-((:,) q -
-N_j 14 for. You can't cross-compare apples and oranges, if you 15 will, from W-3 calculations and WRB-1 calculations.
16 Q So, in other words, in the THINC' program, 17 which is used to simulate the actual core operation of the
~
18 plant for the purpose of calculating the minimum DNBR, you 19 would have to use the appropriate correlation, the critical 20 heat flux correlation in that computer modeli is that 21 correct?
22 A That is correct.
23 Q Has that been done using the W-3 correlation 24 Of I- ---
25- A Yes, the W-3 correlation is used -- both I'
N)
_- - -. .a. . . . _ . . . ,
367 V ^ \.
L .. ; I correlations are programmed in the code. One selects the 2 appropriate correlation for the purposes of the analysis of 3 the particular fuel type in question.
4 Q Could you tell us what is the calculated 5 minimum DNBR if the analysis reflects the W-3 correlation 6 route?
7 MR. BAUSER: Excuse mei objection. Mr.
8 Chairman, I fail to see how this question again falls 9 within the ambit of this proceeding as defined in the scope 10 of the Board's three questions, which refer to the 1.17 11 DNBR as it relates to the OFA fuel.
12 MS. YOUNG: The staff joins in that objection, 13 Judge Lazo.
y wt h-(, ! 14 JUDGE LAZO: For the same reason, Ms. Young?
15 MS. YOUNGi Basically the contention was 16 challenging the acceptance limit of 1.17 for OFA fuel and 17 did not challenge the 1.3 W-3 correlation limit.
18 JUDGE LAZO: Well, perhaps Dr. Edwards will 19 tell us where he is going with this, if this is an 20 introductory question.
21 MR. EDWARDS: Yes. I am thinking of the third 22 question that the Board posed, which has to do with 23 the margin of safety, Mr. Chairman.
24 One thing that I an unclear'about is whether 25 -- I am still grappling with the question of what are the
[
_p V)
r- q 368 i
l 4
t ~~T 1 practical consequences of this DNBR limit of 1.17 as l V
-2 compared with the DNBR limit of 1.3 using the W-3 3 correlation, and it is in that context that I was 4 attempting to ask this question.
5 I was hoping that Mr. Dzenis' answer would 6 cast some light on whether there has been a significant 7 reduction in the margin of safety between the two 8 correlations.
9 MR. BAUSER: Mr. Chairman, first, the third 10 question is based on the predicate that the standard, the 11 95/95 standard is not being met when the-1.17 DNBR 12 acceptance limit is used. There is no predicate for that
_ 13 question'that I am aware of that has been established.
[ f%.,)
s ja And, secondly, the question that Dr. Edwards 15 has indicated he would like to.have answered I believe has 16 already been asked and answered. I would have no objection 17 to him asking that question again, however.
.18 MR. EDWARDS: I am sorry. I haven't heard the 19 answer before. Perhaps I missed it.
20 JUDGE LAZO: Well, why don't you ask the 21 question again, Dr. Edwards.
22 MR. EDWARDS: Okay.
23 'BY MR. EDWARDS:
24 Q The question, Mr. Dzenis, is what is the 25 minimum calculated DNBR using a W-3 correlation for the 3
%.g I
. . . . , , , , .. ,. . , . - - ,-w . - - , . - . - . .
369 R_ I entire mixed core?
)
2 MR. BAUSER: The question to which I was 3 referring that had already been asked and answered was what 4 was the effect of the grant of this amendment that was in 5 question, and I believe that has been delved into. I 6 continue to object to the question that Dr. Edwards has 7 just asked. There is no relationship to that question and 8 the three questions the Board has asked.
9 JUDGE LAZO: We are going to overrule the 10 objection. You may answer the question.
11 THE WITNESS: From which fuel tank?
12 SY MR. EDWARDS:
13 Q For a mixed core of exactly the same type that
\ ; 'K ,
L-L 14 you were considering. In your affidavit you -- or, pardon 15 me, in your prefiled testimony you discuss a minimum 16 calculated DNBR o'f 1.34; is that correct?
17 A Yes.
18 Q Under exactly those s a i. circumstances, but 19 using the W-3~ correlation, what is the answer?
20 A I don't know the answer to that. We selected 21 the WRB-1 option for the optimized fuel, and all the 22 calculations are performed based on that assumption. We 23 not do the second set of analyses. We used the correlation 24 which is appropriate for that fuel t a i: k .
25 0 So at no time was it ---
.si l
370
'.x,
/~h i A I might also add here that this amendment did LJ 2 ~ not. introduce the use of -- it was not the first 3
introduction of the WRB-1 correlation on the 15 by 15 OFA 4 ' fuel at the Turkey Point reactor. That was submitted and 5 approved on a' previous amendment.
6 All we are doing here is indicating that we 7
continue to meet that'same 95 by 95 criterion with the 8
revised input which occurs because of the increase in the 9 peaking f actor -limit .
10 Q At the present time this analysis which 11 results in the 1.34 minimum-DNBR calculated, are you using 12 the same correlation for-the LOPAR and OFA fuel?
13 A No. The LOPAR analysis is covered by its.
!.?^g Ms _/ 14 previously existing submitted analysis which meets the 1.3 15 limit using the W-3 correlation-appropriate to that fuel 16 type.
17 Q 'And you anticipate that that will remain the 18 case for the foreseeable future?
19 A Well, that type' fuel will be phased out of 20 operation in the reactor, and at some point in time it will 21 no longer be used in the Turkey Point reactors.
22 Q And at that point the fuel core will be 23 homogeneous?
24 A Yes. -
25 'O And it will be homogeneous OFA fuel?
- 7%m M
371
(
L;lg 1 A OFA, yes.
2 Q Fifteen by 15?
3 A Correct.
4 Q Of the other Westinghouse plants which have 5 suffered from pressure vessel embrittlement problems, 6 Robinson and San Onofre, are these plants also using 15 by 7 15 OFA fuel?
8 MR. BAUSER: I object to the question. It 9 assumes a matter that has not been established in this 10 proceeding. In addition, I do not know the relationship 11 between the other plants that Dr. Edwards has mentioned
, 12 after Turkey Point.
13 MR. EDWARDS: May I rephrase the question, Mr.
14 Chairman?
15 JUDGE LAZO: Yes. We will have to sustain the 16 objection. You may want to lead into it in another fashion. i 17 MR. EDWARDS: May I rephrase the question.
18 BY MR. EDWARDS:
19 0 Are there any other plants, to your knowledge, 20 which are using a mixed LOPAR/OFA fuel with a 15 by 15 21 configuration?
22 A Yes.
23 Q May I ask what they are?
24 A They are in the Zion reactors, Commonwealth 25 Edison.- D. C. Cook Unit 1 employs the 15 by 15 OFA design.
372 tW)
Vfv 't However,' it does not have the LOPAR reference. It has fuel 2 supplied by another vendor ^as the predecessor fuel type, 3 but the same methodology of transition core analysis is
>4 employed there.
5 I believe Indian Point Unit 3 may have the 6 transition, but I don't know if they have actually loaded 7 the new fuel tank or have licensed the methodology.
8 Q And are some of these reactors that you 9 mentioned also undergoing a flux reduction program?
to A I don't know.
11 MR. BAUSER: I object to the question, 12 Mr. Chairman. I~believe that question is beyond the scope of i3 .the proceeding and it is irrelevant.
(?~h Well, the question has been AA )- 14 JUDGE LAZO:
15 answered.
16 MR. BAUSER: May I ask what the answer was?
17 THE WITNESS: I said I don't know that.
18 BY MR. EDWARDS:
19 Q Mr. Dzenis, I would like to go back to the 20 . mixed core data base which we had discussed prior to the 21 lunch break.-
22 Is my understanding correct that the hydraulic 23 calculations which have been performed-for the mixed core 24 for Turkey Point Units 3 and 4 have been based upon 25 mathematical modeling which in turn is based upon hydraulic ey Q ,1 1
-. .- ~ . . , . . - ,,.,e. .n. -. - . - - , , . . , , - = , . , , -. -- - - ,
373 g-l
(( 1 resistance factors derived from the two-bundle experiment 2 that you described earlier?
3 A That is correct.
4 .Q Now since these calculations result in a 5 penalty, a numerical penalty of three percent for the 6 variability in hydraulic resistance in the mixed core, my 7 question is how do you know that these calculations satisfy 8 the 95/95 criterion that is the basis of the DNBR 9 acceptability criterion?
10 A The calculations did not result in three 11 percent. The calculations resulted in a range of 12 calculated effects, all of which were less than three 13 percent. Three percent was selected as an upper bound value ic v 14 to the range of conditions studied in the analytical 15 Program and for consideration of what the likely 16 configurations in the reactor core might be. Three percent 17 represents an absolute upper bound.
18 Q And as you explained earlier, this is an 19 absolute upper bound on those configurations which you had 20 specifically modeled?
21 A That is corract.
22 Q Which were three in number?
23 A There are three specific geometric 24 possibilities.
25 Q Now is it true to say that the reason for
374 lM
.ca ) 1 . assigning a penalty to the 1.34 calculated minimum DNBR is v
2 because of the fact that the THINC program which produced 3 the 1.34 figure was working at that time on the assumption 4 of a homogeneous core? In other words, the 1.34 figure 5 itself was calculated using the THINC program assuming a 6 homogeneous core?
7 A That is correct.
8 Q Why? Why wouldn't the THINC program be 9 programmed to treat a non-homogeneous-core?
10 A- There are'two reasons for that. The model can
_11 be prepared which simulates the transition core. Two 12 reasons for that.are, first of all, as I explained-in my i3' testimony, you do not know the specific core-wide loading h(- 14 pattern configurations of which assemblies will be in which 15 positions throughout the transition period. Therefore, it-16 is not possible to set up the specific core-wide models 17 with the OFA and standard positions known.
18 The other reason is that the intent of the 19 transition period is-to get to-the full core optimized fuel 20 configuration, and that the analysis that we performed'with
! 21 the THINC code with the 1.34 safety analysis value is 22 intended to be the new reference analysis for the optimized 23 fuel in eventuality.
24 And in order to prevent resubmittal on a 25 cycle-specific basis of additional data, it is possible to
- 3Y~%
':Lk.)
L
l 375 1 bound the effects through the transition period in this 2 fashion.
3 4
5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25
'12-1-JocWal 376 1 Q' I see, So, in oth.er words the analysis was done
\ I
/~'/
(_ 2 with a view-towards establishing a criterion for the homogeneous i
3 core, which is ultimately going to cover both.
4 A I wouldn't use those terms. We did the analysis 5 of the transients, knowing where the plant was going to be 6 loaded to eventually. The criterion was still -- the DNBR
=7 limit criterion is.still the 95/95 hased 1.17.
8 Q Is it not true that the hot rod, fuel rod, fuel 9 Pin, would be' located -- would be expected to be located in the 10 I 0FA section of the core?
11 ' A The hotest power rods tend to occur in the. feed ,
12 fuel. That is not necessarily always the case. We covered --
.() 13 .the analysis that we submitted covers the possibility of the I 14 hot rod occurring in either the OFA or the low power fuel.
15 Q Just one final question'on this point, _and that is a
16 with regard to the calculations that you get from the THINC !
l 17 program, what experimental basis is there for the degree of l t
18 confidence. l to I am thinking again of this 95/95 criterion. How 20 : can you quantify the confidence to be placed and the numerical j 21 results from a THINC analysis?
l!'
r 22 h A This is inherent in the qualification program of the 23j use of the code for reactor applications, 7"] 24
-FU Reporters, Inc.
First of all, the thermal hydraulic methodology, the 25 suh-channel thermal' hydraulic methodology which THINC is one p r - .na,-, , -s .-~g.--, .,.y,.m ,,,n,.,--- , , , - ~ . , . , , . , , -,w c-- - - , - - , , , , - ,-rn . , ~ - - , ..-,e ,
12-2-JoGWal , 377 I
i y of the most well known examples in the eactor d_ n d u s t ry , is used l
(- 2 in the analysis of the test data so that the analysis of the 3 data,.the critical heat flux testing program, since it is 4 a reactor bundled type configuration in a sub-channel mode, 5 was used to reduce the data.
6 Therefore, this qualifies the use of the code with ,
t 7 the appropriate limit'DNBR with the appropriate correlation. ,
l 8 In addition, as part of the qualification program 9 for any reactor code, and code using reactor analysis, we 10 l provide analysis of this code and various experiments of our 11 core wide type nature.
12 ! In other words, experiments have been performed in I
~s ;
{) 13 - t his industry where coolant temperatures are instrumented 14 within the core, and the performance of the core with respect 15 to the prediction of those temperatures is valid, and is part 16 of the qualification.
17 Q Do they use a criterion such as 95/95 criterion to 18 determine the acceptability of those estimates? i m A In terms of how well the code predicts phenomena, it l l
20 has to predict better than a 96 by 95 percent bounding value. i 21 } In other words, you are qualifying the code for the d .
22 purposes of evaluating a particular physical configuration. l
, i 23 j The issue is not how well does it model the 95/95, but how does j . ,
24 ' it model the data, and the data and the model of it is supplied
(]
Wr-Fi j Reporters, Inc.
25 as the qualification program for the use of computer code in l l
1223-JocWol- 378 ;
I this' application.
(O j 2 I might add that the same code is included in the 3 reference list of the standard review plan as an example of the 4 type of-code-to use the sub-channel analysta methodology of 5 Pressurized water reactors.
6 Q In your-affidavit of August 1984; August.8, 1984, 7 which I believe you have included by reference in your testimony' 8
9 MR. BAUSER: Do you'hve a citation to where that 10 I is included by reference?
11 l- WITNESS: It is on page 4, the top paragraph. I
_12 l MR. BAUSER: Well, that is a reference to another
( )~ 13 Idocument'. I don't believe that that document is included by 14 reference in' this test 1 mony.
15 ' JUDGE LAZO: Well, Mr. Dzenis mentioned it for a >
i 16 particular purpose, and I guess that would be to incorporate 17'it in some fashion.
{
18 g MR. EDWARDS: His exact words are: The method for 10 : d e t e rmining the DNBR acceptance limit is described in detail '
e 20 hon paragraph 6 to 25 of my August 8, 1984 affidavit, Mr.
21 !(Chairman .
l' 22 JUDGE LAZO; Yes, I see that, and for what purpose F
23 {are'you going into this at-this time? Do you wish to refer to 24 the-earlier affidavit?
ham-Fi Reporters, Inc.
25 MR. EDWARDS: Yes.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ . . _ . _ . _ _ . . . . . .. ._ - , , _,- u =m --
12-4-JccW:21 379 1 JUDGE LAZO: Very well, 2 BY MR. EDWARDS: (Continuing) 3 Q Now, I --
4 , JUDGE LAZO: Does counsel for Licensee have a copy o
5 of tha't affidavit to provide the witness?
f 6 .MR, BAUSER: Yes, we do, Mr. Chairman.
7 CMr . Hauser passes document to witness.)
I 8l BY MR. EDWARDS: (Con tinuing) 9 Q Mr. Dzenis, I would like to refer you to paragraph 10 h 27 on page 10.
l 11 A Paragraph 27?
12 Q Yes. ;
, ~N i
( ~
) 13 A It is not within 6 to 25, referred to in my 14 affidavit.
15 Q The paragraph in question states: It should be 16 noted that although F Delta H does have a direct impact on i
i7 calculated DNBR values, the changes of the F Delta H Amendment 18 do not reduce DNBR values to a point where they are below the l
10 acceptance limit.
i 20 Previous DNB analyses prior to the F Delta H l l
21 Amendment, show that the minimum DNBR values of both transient
- i 22l: and normal operation not only met the'DNB acceptance limit, but 23: in actuality were greater than the acceptance limit by an amount a
(3 24 !! which will be called the DNBR available for design flexibility.
Amd_ _ ) Reportcrs, Inc.
25 That is the paragraph. j l
d2-5-JcoWol
^
l l
'2 That is the paragraph.I wo'uld like to ask you about.
3 Now, when you talk about they do not reduce the DNBR values to 4 ,a point where they are below the acceptance limit, are you 5 ref errin's here to the 1.17 acceptance limi ?
6 A That is correct.
7 Q Then you go on to say previous DNB analyses prior
~8 to the F' Delta H Amendment show that the minimum DNBR values 9 of both transient and normal operation not only met the DNB l I
10 , acceptance limit -- again, this is the 1.17 acceptance limit?
l 11 ' 'A That is correct. !
k.
But' in actuality were greater than the acceptance 12 'Q j 13 ,. limit . Mr. Dzenis, were these calculations done with a 1.3 14 acceptance limit?
t 15.l' A No.
16' Q -Back an Page 9, paragraph 23, we read: As noted 4
17 - above ', the WRB correlation is based strictly.on data from Rob ,
18 ~Bunglepeste, whereas the old grid correlation is based on single 10 tube data. i 4
20 , The fact that the WRB-1 correlation is a better j 21,; predictor of DNB for actual nuclear reactor geometries is shown N
22 y hy the result that a minimum DNBR acceptance limit required with b
23 f th.e use of the WRB-1 correlation is one 1.17.
q J24 Now, Mr. Dzenis, this question of the WRB-1 correlation Re-FLJ Reporters, snc.
25 is e. hetter predictor of DNB for actual nuclear reactor geometries
12-6-JooWol. 381 1 is that not pertaining to homogeneous geometries?
2 A. It pertains to the nature of the geometries in the 3 DNB test situation.
4 ,. First the rods with reactor-type grids,.with the t
5 flow on the outside of the rods, as separate and distinct from 6 the interior flow heated tube-type experiments which had l
- 7. Preceded that. configuration.
8 Q So this question about actual nuclear reactor 9 geometries was not intended to relate to the question of 10 homogeneous.versus non-homogeneous cores, but simply to the i
11 fact that you are dealing with a complete fuel bundling. l I
12 'i A That is correct. l
-O(_/ 13 MR. EDWARDS: Thank you. I have no further 14 questions, Mr. Chairman.
15 JUDGE LAZO: Thank you. Does Staff have cross I
16 '
examination of this witness? ,
i 17 MS. YOUNG: Yes, we do. j II INDEX 18 CROSS-EXAMINATION BY MS. YOUNG j 20 , Q .Mr. Dzenis, in your testimony you conclude that
?
- p. 21) a calculated minimum DNBR of 1.34 for.OFA fuel' meets the 95/
h 22 li 25 criteria set forth in the Staff's Standard Review Plan, is u
.f 231' tha t correct?
. f') 24 '
Am-FL.J Reporters, Inc.
A Ihat is correct, 25 q Can a calculated minimum DNBR for mixed core be
12-7-JooWal 382 1 determined either by using a homogeneous core model, which is f 1
'i penalized for the effects of mixed core, or by modeling the 2
3 actual transitional core configuration?
4p A Either method would give the desired result.
1 5 Q Is using the method that gives you the lower DNBR 6 conservative? ,
l 7 A Yes.
i 8 Q On Page 7 of your testimony, you state that a 9 detailed thermal analysis was used to derive the three percent :
10 bounding value for the effects of transitional core geometry.
11 How was that done?
12 I A That was performed consistent with the former of
. I.
i 13 the two methods you described. In other words, there is a -
1 14 homogeneous core reference calculation. There are detailed 15 calculations of the transition core effect and the penalty --
16 the bounding penalty derived for the transitional core 17 configurations.
l 18 Q How did you determine the bounding penalty? i
" A What we looked at was a series of cases dealing with ;
20 different geometries, and a range of thermal conditions across i l 21 the corresponding range of thermal conditions involved in the 22 . normal operation and anticipated transients.
a i
23 P There was a range of computed values three percent l L' '; 24 jgreater than any of the individual calculated values.
w F_ _ i Reporters, Inc. '
25 Q Among the configurations you looked at, did you
! l i
i
12-8-JoOW31 383 1 determine if there was a. worse-core configuration, --
!qV- 2 A No.
3 Q In your opinion, is a situation where OFA fuel 4 assembly is. completely surrounded by LOPAR fuel, the worse
'~
configuration in terms of the effects of a mis-core?
_5 jcore 6 A That tends to be the worse configuration, yes.
7 Q Was that configuration looked at in determining 8 the bounding value?
9 A Yes.
10
.Q . In analyzing the effects of various core configu-111 rations, would you deem that a sensitivity study, basically?
1 12 i A Yes.
( 13 Q .And is a sensitivity study a valid means to determine 14 the effect of geometric differences between LOPAR and local 15 fuel? j i
16 A In my opinion, yes. !
I 17 Q Dr. Edwards asked you a series of questions regardingj i
18 the number of potential fuel loading arrangements in a mixed-I
'9 core.
20 Isn't it true that you do not have to analyze whole
' 21 -n (d
core configuration, whether mixed or homogeneous for the purposes
~ 22 0 of determining DNBR7 23 'A That'is correct.
.e.
24 !- And why is that true?
(~}
Ass-Fm,s Reporters, Inc.
Q 25 A As I explained previously, the fuel assemblies are
12-9-JcGWal 384
) loaded as entire assemblies into a defined, checherboard r, -
x_/
2 location in the core, We can't change the position of the 3 assemblies relative to each other.
4, The three configurations I described -- one is the i
I 5 case of the optimized fuel surrounded by the other -- the case 6 you asked me about in the previous question.
7l The second case would be going back to our checker-l 8 board again,'and if you have the white and the black squares ,
'l 9 l the whites are one type of fuel, and the black is the other 4 -
10 ;
i type of fuel assembly in the core, so we studied that configura-j' 11 tion.
12 The third possible configuration is a row of one l em
) 13 ' assemblies adj acen t to a row of the other. All or combinations 14 or subsets of one of those three.
15 Q And isn't it true that only the conditions 16 immediately surrounding the fuel bundle with the hot rod, are 17 important for determination of DNBR7 18 A That is correct.
10 So the amount of permutations of possible fuel Q
i 20 bundle arrangements isn't important for determining DNBR?
t.
21 'A That is correct. !
i 22 ; Q What assurance do you have that if you model the 23 !: actual mtred core you would not calculate a DNBR lower than a 1
(~') 24 j DNBR f or a homogeneous core, which is penalized for the Aos-F_ j Reporters, Inc. ;
25 effects in a-mixed core?
l 1
12-10-JacWol 303 1 A That is based on my physical understanding of the
, i j 2 l phenomena that relate to flow distribution within the transition
'3 core.
4 We have modeled a case which is more extreme which 1
5l that which you experience in your reactors.
6 Q Could you explain that a little further? ,
7 A In particular, if you take the case you asked me i
8 about, which dealt with the single assembly, surrounded by 1
9j your eight immediate neighbors, that are the other fuel type.
i 10 'l The amount of flow which would leave the higher resistance l i
11 optimized fuel tank assembly in that type of environment would ,
i 12 be related to the fraction of the analysis, if you will -- the 3
' i 13 fraction of the assemblies in the analysis, which are one l 14 fuel assembly type relative to the fraction of assemblies in the i
15 other.
16 Therefore, that type of specific analysis contains 17 only one-ninth of the fuel of the high resistance, therefore, 18 the percentage of its flow that it would lose to its neighbors l 10 would be represented by that kind of configuration. .
i i
20 Actual fuel loadings involve, generally speaking, i
21'at least in the order of a third of the core is replaced. So, o
22 ' each loading of the optimized fuels constitute at least a third 23 which is a higher fraction of the core than the one-ninth that 24 is modeled in a specific case you described.
- )
Aa-F- _j Reporters, Inc. l 25 Q Let me move on for a moment to the 5.5 percent 1
12-ll-JocWal 386 1 rod how penalty you mentioned in your testimony,
( )
k' 2 Is this penalty based on several sort of conservative 3 assumptions?
4, A Yes, it is, li 5l Q Could you explain those?
6 A Well, it is --
as I described in my testimony, the 7 . amount of rod bowing observed in reactors increases with the i
8 operating lifetime or burnup of the particular fuel assembly.
t 9 This particular penalty is calculated at a high 10 l near end of life burnup type value, and applied across the l! l 11 J Board to all of the fuel in the core, regardless of burnup.
s .
12 As I was asked previously in the questioning, the rm
() 13 limiting peaking factors tend to occur near the beginning of l
i 14 life of the fuel. ;
15 In other words, at very low burnups, when the rod 16 how penalty is considerably less than the 5.5 percent at near 17 end of life conditions.
18 In addition, we take data from all positions within the fuel assembly, and we assign the maximum observed regardless 3 20 of actual position, to the limiting positicn calculated in j si l 211DNBR L
calculation. 1 22 ? Therefore, we presume maximum amounts of rod-bow; 3
23 bregardless of actual positions.
1 It is applied to our positions f) 24! equally.
As-F_J Aeporters, Inc.
End 12, 25 SunW fols.
l l,
I
387 f#13-1-SueW I Q Okay. JYou described in answering the preceding
'() 2 question that you used end-of'-life fuel burnup.
3 Could you describe a little more what that means
.4 and why you don't have to use a higher level?
5 A The - when-the fuel.is loaded, when new fuel is 6 loaded, in-the reactor core it has enriched fuel material and 7 no fission products accumulated. Over its lifetime, the
~8 .fissonable isotopes decrease with time.because of the burn-9 cut of the fuel.-
10 In addition, there is a fission product inventory II which accumulates over~ time. This serves to -- near the end 12 of the lifetime of the fuel to reduce its power-producing
() 13 capability to be potentially limiting, in that it still pro-14 duces some net power but.at a much lower than average level.
15 The value that the rod bow penalty is determined l l
16 at is determined at a value, at a maximum level to which one l 17 could even conceivably expect a fuel rod to be capable of a 18 near-limiting peaking factor.
19 After that point in time, it's intuitively physical-20 ly obvious that the rod is not capable of a limiting' peaking 21 -factor.
[
22 O So, is assuming this high burnup a conservative 1.
23 consumption?
"}'
hes-F-( J Reporters,24 Inc.
A Yes.
L25 g .And does fuel rod bowing increase with burnup?
l
388 I
l
- 13-2-SueW 1 A Yes, it does. l I'"\ I 2 Q And would, therefore, a rod bow penalty increase 3 with burnup?
4 A Yes, it does.
5 Q So the 5.5 percent rod bow penalty bounds the 6 effects of high burnup?
7 A In terms of its effect on DNBR.
8 Q So, therefore, it's a maximum value that protects 9 the burnup?
10 A That's correct.
11 Q Does rod bow magnitude for an OFA assembly in 12 a homogeneous core differ-from the bow magnitude of an OFA n
() 13 assembly.in a core containing both LOPAR and'OFA fuel? !
14 A No.
15 Q Is there -- well, could you explain that?
l 16 A As I describe in my testimony, the existence of !
I 17 theotherfuelassembliesinthecoreeffectstheflowdistri-l i
18 bution within the core but not to the amount that it could l
i 10 effect the magnitude of the rod deflection of the -- in the l f
20 other a: ambly. l I
21 Q And is this flow distribution due to a difference l
.22 in grid design between the 15 by 15 LOPAR and OFA fuel?
23 A That's correct.
24 Q And does this difference cause higher flow resistance f~')
Am-F__J Rmorters, lm. ,
25 and reduced flow rate within an OFA assembly in a mixed core? !
i l
389
- 13-3-SueW 1 A That's correct.
- a. I l
2 Q So, would the 5.5 percent rod bow penalty remain 3 bounding in a mixed core, given the effects of these declin-4 ing flow rates?
5 A Yes.
6 Q Could you explain why?
7 A The effects of rod bowing are dominated by the ,
8 magnitude of the bow and the heat flux and the pressure which 9 are uneffected by the transient, the transition core condition.
10 Therefore, for the -- each assembly type the change 11 in the localized flow rate distribution accommodated in the i 12 transition core penalty does not effect the magnitude of
<s
( ,)
13 the calculated rod bow penalty.
14 Q So that's the reason why these two penalties 15 are independent?
16 A That's correct. f I
17 0 Okay. I would like to ask you questions concerning l j; E 18 the WRB-1 correlation, or CF correlations in general.
19 Is the CHF correlations for OFA fuel applicable ,
l 20 to a mixed core'
}
21 A Yes , it is.
22 Q And, isn't it true that a CHF correlation can !
23 only be used within the range of applicability of its para-(~N 24 meters such as, for example, mass flow?
Am Fb_.j Resmners, lnc. ;
25 A That's correct.
l
390
@3-4-SueW 1 Q Is the WRB-1 correlation for OFA fuel based on .
,~ \
_ '2 a range of flow conditions or flow rates?
3 A Yes.
4 Q And is that range of applicability of the correla-5 tion the same for a fuel assembly either in a mixed core or 6 a homogeneous core?
7 A The range of applicability is the same for that 8 OFA fuel in either configuration.
9 Q Okay. Earlier, you testified that there is a 10 reduction of flow through an OFA assembly in a mixed core 11 versus the homogeneous core.
12 Does this flow reduction result in a flow rate
()
c I 13 outside of the range of applicability of the WRB-1 correla- !
14 tion?
~15 A No.
16 Q So, basically that flow rate was introduced ,
l 17 into the experiments to derive the correlation, the flow !
i 18 rate you would see in a mixed core configuration? l 19 A Well, the range of flow, it's an experiment as j l
20 defined by the range that the correlation is to be bowed i 21 over. We also reviewed the calculated flow rates in the l
22 safety analysis and the transition core calculations to j 23 assure that they are within the range of applicability of l 24 that correlation. ;
r]
W-Fi _j Reporters, Inc.
25 Q But the flow rate in the mixed core is encompassed l
i
i 391 l
.913-5-SueW. 1 by the flow rates, the range of flow rates, to which the
\
("#
'- 2 'WRB-1 is applicable?
3 A Yes.
4 Q Thus, the penalty associated with the. applicability 5 of the WRB-1 correlation is not effected by the penalty for 6 the effects of the mixed core?
7 A I be'g your pardon? -Again?
8 Q I asked you whether the mixed core penalty --
9 A Okay.
10 Q -- has any effect on the penalty for the appli .
11 cability of the WRB-1 correlation?
12 A No.
' ('i 1- () 13 Q And is this in part because some of the flow rates 14 are within the range of applicability for the WRB-1 correla-15 tion? :
'~
16 A- That's right. As I stated before, all transition --
17 all the calculations we did to determine a transition core 18 penalty and all the calculations in'the homogeneous core 10 configuration are within the range of applicability to'WRB-1 20 and DNB correlation.
21 Q All right. Moving on to a different subject, in 22 your testimony on Pages 5 and 6, you state that there is a 23 12.7 percent margin between the calculated minimum DNBR of L24 1.34 and the DNBR acceptance limit of 1.17.
(~}
Am-F6J Reporters. Inc.
25 Is the calculated minimum DNBR included in the ,
392 i
& SueW l safety limits for the operation of Turkey Point?
I
/
2 A Yes, they are.
3 Q And how are they incorporated?
4 A They are incorporated in the determination of the 5 core, DNB core limits, which are listed in the technical 6 specifications and those core limits are eventually converted 7 to over temperature Delta T setpoints also included in the 8 technical specifications of the unit.
9 Q Is the limitation on the combination of thermal 10 power, pressurizer pressure, and coolant temperature based 11 on a calculated minimum DNBR of 1.34 OFA fuel?
12 A Those conditions are uneffected by this minimum.
) 13 We retain the same condition on power, pressure and tempera-14 ture.
I 15 l The values listed in the technical specifications i
16 are the conservative bounds included in the safety analysis, i 17 !! Q But my question to you is, is the limiting curve .
t L
h l
18 in the tech specs regarding -- j i
19 i A I'm sorry. I 20 Q -
the operation of reactor --
21 A I misunderstood your question. l l
22 0 -- based on the calculated minimum DNBR of 1.34 23 for OFA fuel? ;
p 24 A Yes.
As-Fi .J Reporters, Inc.
25 Q So, under the plant's current technical specifications
393 f ' l3 15-7-SueW' I the reactor cannot be operated in a manner that the limits
,- ~li 2 based on the 1.34 calculated minimum DNBR would be exceeded 3 during normal operation or anticipated operation on transients; 4 is that correct? ,
5 A That's correct.
6 Q And the safety limits based on the calculated 7 minimum DNBR of 1.34 compensate for the three uncertainties 8 associated with rod bow and mixed core and the applicability l
[ 9 of WRB-l?
I 10 A That's correct.
11 Q Suppose for the moment that you were to increase l
12 the DNBR limit of 1,17 to' account for these three penalties totaling 10.5 percent in this case, would the value be !
, greater than the limits based on the 1.34 calculated minimum DNBR used in the tech specs?
A No. l 16 I i-Q l 17 l Would the safety limit ~ curve remain conservative?
i ;
18 A Yes.
l 4
g Q Why is that? !
A All you've done is there is change the arithmetic --
you have two numerical values which are a distance apart, wheth r you start from one and approach the other.or the other and approach -- if you start from One and approach Two, and if you
~ start from Two-and approach one, it's the same distance and !
P "*'"*-
25 Nse am&deddem. i I
394 G
j#26-8-SueW1 .Q So, for the present technical specifications for j
2 'the plant, it doesn't matter whether you raise the DNBR 3 limit of 1.17 to account for uncertainties?
l
, , .4 A Well, the 1.17 is still the~ fundamental acceptance
{
5 limit'for the-correlation in use for that fuel. The un-6 certainties are derived and applied to -- they are derived 7 based on the geometric. considerations of the plant.
8 Q And I'm asking in terms of the practical conse-9 quences on the curve limiting the operation for the plant?
10 A Arithmetically they are equivalent.
I 11 Q And that the curve based on the 1.34, which would 12 ~ still be greater.than the 1.17. compensating for the penalty
() 13 is not changed?
- 14 A It would still.be the same curve.
15 MS. YOUNG: I have no further questions.
16 JUDGE LAZO: Redirect, Mr. Bauser?
17 l
MR. BAUSER: Mr. Chairman, the witness has been 18 on the stand for over an hour. May I suggest that we take a M short break?
20 JUDGE LAZO: Does anyone have any objections?
21 (No' reply.)
22 Very well. Why don't we take a fifteen minute mid-23 afternoon recess?
24 MR. BAUSER: Thank you.
Me# R eorms, ine, f 25
~
JUDGE LAZO: Could we go back on the record for jus 6
395 13
- TS-9'-SueW 1 -one minute? In discussing the matter further at lunch time,
'2 I think the Licensing Board has determined that it would be 3 more orderly to have the Intervenors direct' case go next 4
after the Licensee and then let the Staff finish up after 5 the Intervenor's case has been heard.
~
6' I think that's the way we should proceed. So, when we have completed with Licensee's witnesses,the redirect
~
7 8 examination, there will be a few Board questions and then 9 the next order of business would be the Intervenor's case.
10 Thank you. We will recess for fifteen minutes.
11 (Whereupon, a recess is taken at 3:12 p.m., to 12 reconvene at 3:31 p.m., this same date.)
() 13 JUDGE LAZO: Could we come to order, please? Jhr. .
14 .Bauser, are you ready to proceed?
15 MR. BAUSER: Yes, we are, Mr. Chairman. We have 16 no redirect.
17 i JUDGE LAZO: No redirect.
18 MR. HODDER: Mr. Chairman --
10 JUDGE LAZO: Yes, Mr. Hodder.
20 MR. HODDER: Earlier this morning, the Board 21 indicated that they might take some Limited Appearance 22 statements, and I believe that we have a lady here that 23 wants to give at least two such statements. And she has 24 been waiting awhile, and she is sitting in the audience. !
(~}
Am F~ J Rmorters. Inc. f 25 I just wanted to direct the Board's attention to l 1
i
396 13
- b5-10-SueW that. I am not advocating any - I have another matter though !
(~N i
2 I would like to address to the Board.
3 JUDGE LAZO: We are at a break point. May I 4 ask, does the young lady have a statement to present for 5 another as well as your own statement?
6 MRS. MEYERSON: Yes, I do.
7 JUDGE LAZO: Do you have written statements, ma'am, 8 or do you want to make an oral statement?
9 MRS. MEYERSON: Well, I would like'to make an 10 oral statement. And I have three, also very brief, written 11 statements from people who head groups that I represent. !
i 12 JUDGE LAZO: May I ask, ma'am, approximately how
.) 13 l long do you think you would like to have to make your oral l
14 l statement?
I 15 MRS. MEYERSON: I just want about two minutes 16 for.my own oral statement.- And I think the whole thing, 171l reading the statements from these other groups, is probably I i 18 a total of about ten minutes. I I
I IC JUDGE LAZO: Well, you are most welcome. Would !
20 you come forward? ,
21 Perhaps, Mr. Dzenis, we are not dismissing you 22 because the Board has some questions, but if you would let 23 this young woman have the microphone we would appreciate ;
i 24
(~)
Nz Fi_j Rmorters. Inc.
that. j 25 WITNESS: Yes.
I 397 l I
- 13-ll-SueW1 JUDGE LAZO: We are making a verbatim transcript k_) 2 of the' proceeding. If you would, give your name, please.
3 MRS. MEYERSON: I certainly will.
4 LIMITED APPEARANCE STATEMENT 5 OF BRENDA MEYERSON 6 MRS. MEYERSON: My name is Brena Meyerson. And I'm 7 the Chair Person of the South Florida Peace Coalition, which 8 is an anti-war, anti-nuc coalition of more than forty groups 9 in the South-Florida area.
10 And as part of my Coalition, there are somewhere i
11 between three and five groups who, in addition to working on i
12 the problem of nuclear weapons, are also very much concerned l l
-() 13 with nuclear power. And those are the statements that I have.I 14 I personally would like to say that I have known 15 Joette Lorion since I have been working on this issue for.
16 about three years now. And during that' time, she has been l i l 17ll working on the nuclear power issue and has kept me abreast !
In i, 18 of what she has been doing as it has evolved. i
'O And I know her to be a very sincere and dedicated q l
20 person and somebody who grapoles with an enormous amount of i 21 detail. She is not somebody who wants to muck up the works. .
I l
22 She is somebody who is sincerely concerned with the safety .
23 of ourselves and our entire community.
( x, 24 And I am very grateful that you have come here j wee ~ l aemnm. ine. i 25 and that you have given her a fair hearing, because in the l
398
- 13-12-SueW1 final analysis this'is not about which side wins, this is g
kJ 2 about the safety of the entire community.
3 I'm going to read a statement first from the 4 Physicians for Social Responsibility, which is a member group 5 of my Coalition, of which my husband is.the President. And 6 this is to the Atomic Safety and Licensing Board, U.S. Nuclear 7 Regulatory Commission. ,
8 Ladies and Gentlemen: Physicians for Social 9 Responsibility is a national organization of physicians, 10 health professionals and supporters who are concerned about 11 the risk to public health and safety casued by the building i
12 of nuclear weapons and the use of nuclear power for the ex i
(_) 13 generation of electricity. !
14 Serious safety issues have been raised by the 15 Center for Nuclear Responsibility concerning the embrittle-16 ment of the reactors at Turkey Point. The hearing being held i,
17 p on December 10 in Miami, Florida concerns one aspect of an !
I 18 0 attempt to adapt to the deterioration of the pressure vessels. 1 i
l
'O ! It is unacceptable that the proposed solution will reduce j 20 safety margins when the very problem itself seems to mandate l 21 closer attention to reactor safety. The terrible and irreversi-22 ble. consequences of a pressure vessel failure are so catastro-23 phic that it is vital that the public be protected to the
(-] 24 w Fn~J anmnen. ine.
full extent of the safety margins originally designed into 25 the plant. ,
i l
399
- 13 13-SueW j It is the recommendation of the Miami Chapter of n
\_/ Physicans for Social Responsibility that the scope of the 2
3 Present inquiry be broadened to look at all aspects of the 1
4 4
embrittlement problem at the Turkey Point reactor and that 5
whatever measures are necessary be taken to assure public 6 safety. This is the charge of your Board and the public has 7
a right to expect no less.
8 Thank you for your concern about this matter.
9 Sincerely, Steven J. Meyerson, M.D., President of Miami 10 Physicians for Social Responsibility.
11 I have another letter here from a couple written 12 by the woman. Her name is Rae Newman, and she has worked !
(') 13 with the ConShell Alliance which is another organization in 14 South Florida concerned about nuclear power.
15 To Whom It May Concern: Please accept this note 16 in lieu of.our appearance today. Our family is deeply con- l l
17 cerned about the potentially dangerous condition at the i
18 Turkey Point Nuclear Power Plant.
to We feel the situation must be thoroughly looked I
20 into and corrected immediately. This threat to many lives !
21 15 Preventable. Let's not wait until the worst happens.
22 We request a broader review of the hazards of 23 the plant than exists'now, than this hearing affords.
Thank you. And that's from Dr. Mike and Rae
(~} 24 Am FM 7.ep0fttts, IOC.
25 Newman. ,
i
l 400 1
- 13-14-Su eW1 I have one more letter from a woman named Rucy
( ~' , 1
\--
2 Jason-Kurau, who is the Secretary of the Coalition that I head.
3 Gentlemen: Although the present hearing regarding 4 the safety of the fuel core redesign at the Turkey Point Nuclear 5 Power Plant is to be commended, I would like to state that I 6 believe this is not enough.
7 The Plant's last test ten years ago determined that 8 the weld metal in Reactor Number 4 was thirty percent more 9 brittle than Reactor Number 3. Yet, Florida Power and Light.
10 Company has managed to receive an extension period for the re-11 testing of Reactor Number 4 of fourteen additional years. ;
12 Twenty-four years will have passed between the
( ,) 13 testing of Reactor Number 4 if it is not tested this year as 14 originally planned. Not to test this particular reactor 15 I as originally intended showes blatent disregard for public i
16 safety. I find this appalling. t i
17 l When originally built, I understand that it was l 18 determined that the reactors should be tested every ten l 1
" years for safety reasons. Many problems the Nuclear Regulatory l
20 Commission is facing today, such as the Three Mile Island l
i incident, meltdowns and pressure vessel embrittlement, here
~
21 22 never considered since these were problems no one could fore-23 see.
/~l 24 Am#n_j aeporms, ine.
I believe what you are accomplishing with this l l
25 hearing today is only the very beginning of an extremely l
l
l 401
'#13-15-Sud needed full hearing on the safety and operations of the 2 individual nuclear reactors at Turkey Point.
3 I would demand that the testing of the weld metal 4 in Reactor Number 4 be done within the next six months and 5 that a citizen's right to participate in discussions, 6 information and regulation.of a nuclear utility that serves 7 the public and can threaten life itself, within the large 8 surrounding area of nuclear energy plants, is an undeniable 9 democratic right of'every American citizen.
10 Thus, as a concerned citizen, I ask for a broader 11 hearing regarding the safety hazards' involved pertaining to 12 the operating of Turkey Point Nuclear Power Plant. And I
() 13 thank you for the hearing taking place in Miami on December 14 10, 1985.
I 15 l Most sincerely, Rucy Jason-Jurau.
l 16 JUDGE LAZO:- Mrs. Meyerson, thank you .very much.
I And we appreciate the fact that you were willing to come !
17 j
!' l 18 out and present these statements. l 10 MRS. MEYERSON: Thank you.
20 JUDGE LAZO: Thank you. Mrs. Meyerson, do you 21 want to leave the letters? They have been transcribed. But 22 if we keep the original letters, we will make them a part 23 of the file.
24 MRS. MEYERSON: Yes. f We F(L Reorters, loc.
25 JUDGE LAZO: Thank you. !
402
- 13-16-SueW1 (The documents follow.)
(Ns) 2 MRS. MEYERSON: Yes, sir. Thank you.
3 (The witness stood aside.)
END #13 4 Mary f1ws 5
6 7
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6830 S. W. 127th Place ,
Miarti, Florida 33183 Tel. (305) 385- 3714 M r 9, 1985 Statement Directed to:
~ The Atanic Safety and Licensing Board of the U.S. Nuclear Regulatory Ccmnission Re: 12/10/85 Hearing on the Safety of Ebel Core- . .
Redesign at Turkey Point Nu.riear Power Plant Gentlemen:
- Although the present hearing regarding the safety of the fuel core .
redesign at the Turkey Point Nuclear Power Plant is to be m..ded, I would 1dJce to state I believe this is not enough.
The Plant's last test 10 years ago, determined that the . weld metal in Reactor No. 4 was 30% more brittle than Reactor No. 3. Yet, Florida '
Power & Light Conpany has managed to receive an extension period for the re-testing of Reactor No. 4, of 14 additional years!
1*' 24 years will have passed between the, testing of Reactor No. 4, if it is not tested this year as originally planned. Not to test this particular reactor as originally intended, shows blatent disregard for public safety. . .
I find this appalling. . .
When originally built, I understand that it was determined that. the reactors should be tested every 10 years for safety reasons. Many problems the Nuclear Regulatory Ccanission is facing today, (such as the Three Mile Island incident, meltdowns and pressure vessel embrittlement) were never considered, since these were problems - no one could foresee.
I believe what you are acccmplishing with this hearing today is only the very beginning of an extremely = needed full' hearing on the safety and operations of the individual nuclear reactors at Turkey Point.
I.would demand that the testing of the weld metal in Reactor No. 4 be done within the next 6 nonths and that a citizen's right to participate ,
in discussions, . information, and regulation of a nuclear utility that serves the public and can threaten life itself, within the large *i surrounding area of nuclear energy plants, is an undeniable democratic right of every American citizen. l Thus, as a concerned citizen, I ask for a broader hearing regarding the safety hazards involved pertaining to the operatirq of the Turkey Point
-"- ' Nuclear - Power Plant, and I thank you for the hearing takirq place in
' Miami on December 10, 1965.
%]) . sincerely, GuuW
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PHYSICIANS FOR SOCIAL RESPONSIBILITY 7701 S.W. 76 Avenue, Miami, florida 33143 (305) 662-1521 L
MIAMI CliAPTER December 9,1985 Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, D.C.
Gentlemen and Ladies:
Physicians for Social Responsibility is a national organization of physicians, health ' professionals and supporters who are concerned about the risk to public health and safety caused by the building of nuclear weapons and the use of nuclear power for the generation of electricity.
Serious safety-Issues have been raised by the Center for Nuclear Responsibility concerning the embrittlement of the reactors at Turkey Point. The hearing being held on December 10 in Miami, Florida concerns one aspect of an attempt to adapt to the deterioration of the pressure vessels. It is unacceptable that the proposed solution will reduce safety margins when the very problem itself seems to mandate closer attention to d reactor safety., The terrible and irreversible consequences of a pressure vessel failure are so catastrophic that it is vital that the public be protected to the full extent of the safety margins originally designed into the plant.
It is the recommendation of the Miami Chapter of Physicians for Social Responsibility that the scope of the present inquiry be broadened to look at all aspects of the embrittlement problem at Turkey Point and that whatever measures are necessary be taken to assure public safety.
This is the charge of your Board and the public has a right to expect no less.
Thank you for your concern about this matter.
Sincerely, h) f,g v-Steven J. Heyersod, M.D.
President, PSR/ Miaml
(*
k'O -
403 JUDGE COLE: I have just a few questions, Mr.
1 D ""is' 2
3 more. I have one more request.
4
- ^ ' '
- 5 MR. HOODER: I am a little slow on the uptake today as you have not1ced, but just when the Board 7
announced the change in schedule and you adjourned, every time that happens I am not fast enough.
9 10 yy discussed with my client, and not only is he fatigued from g arriving late last n19ht and staying up late in preparing questions, but also his extensive cross-examination, which y,
will if he were to testify today or any time soon, would exceed the fatigue exposures of Witness Zenix.
(Laughter.)
g So that 1s one basis, but the other and more im rtant basis to us is surprise. We did plan to be last 18 and we would still like to be last in our presentation of 9
our testimony. And We are asking the Board to reconsider that, and specifically because were we to have to go forward today, we allocated our time differently at lunch on other latters and we didn't really prepare for the leading questions that might be necessary to introduce his test 1 mony.
404
'P~ '
I am asking the-Board to consider the fact
'(
that they did announce the schedule that allowed us to go last and we would like the Board to adhere to that because 3
we relied upon it. And we have asked the staf1 counsel if 4
they have any objection to our request, and they have stated that they have none.
I didn't pose that question to counsel for the 7
utility company.
8 JUDGE LAZO: You'did not?
9 .
MR. HODDER: I did not. When I as talking here 10 they weren't there, and then I didn't ask them.
MS. YOUNG: Judge Lazo, let me clear up the record for a moment. Mr. Hodder and I spoke informally, and 13 I( f )
I didn't express any strong preference for the order of presentation of witensses in this proceeding. But after 15 consulting with co-counsel and the witness himself, the 16 staff would prefer to go last.
It would give I think the Board an opportunity to get a tying up of the testimony presented in a 19 proceeding by intervenors' witness and the licensee.
JUDGE LAZO: Ms. Young, that was our thought, 21 too, and that is why we thought it would be a more orderly procedure. But let us suggest that perhaps we can 23 accommodate everyone. The Board has some questions of Mr.
I Drenis and then we should be getting -- I keep looking at l 25 l
v .
L 405
(
1 that silly clock which stopped ---
( (Laughter.)
2
--- we should be getting near the end of the 3
day and perhaps overnight, Mr. Hodder, your witness would 4
have a chance to prepare.
5 t
MR. HODOER: Well, of course, that is true. We would be -prepared to present testimony in the morning. The 7
f only point is we would still prefer to go last. We are like
\ 8 p the staff in that respect, 9
l We feel that this witness will be of greater j g benefit to the Board if he is able to make his presentation which will evaluate and take into consideration everything that the proponents of the amendment have voiced.
And we feel that specifically the testimony of Dr. Edwards will-be of greater value to the hearing process and this Board if it goes last. Therefore, we will abide by
~
g the decision of the Board, but we just wish to point out to
- * * ** * **" *" * * "" *8' **
- 18 perceive it to the Board in the process.
JUDGE LAIO: Well, again, perhaps we could 20 accommodate you and still satisfy the staff's desire.
We might as well stay on the record. I am just thinking out loud that it appears very likely that we can complete this proceeding tomorrow. Dr. Edwards has come a long way, anc I am not sure when he wants to get away. But
406
,s 11 you went .first tomerr:w, an: nen the staf' would
) I v
follow, if you felt that there st 11 were some matters you wou?d like to put into the record througn ycar witness, it
- 3 m1gnt be appropriate to do that through rebuttal testim:ny.
4 So tnat in effect, assuming that there are 5
some areas where rebuttal would be proper, you would have 6
- ne cenefit or in the sense naving the last word.
7 MS. (OUNG: Judge La:o, woulc that opportunity 8
for rebattal also be proviced to the staff?
9 JUOGE LAZO: Yes, out it gets narrower anc 10 ,
narrower.
11 (Laugnter.1 12 MR. H000ER: Chairman La:c, I tnink tnat is 13 m
certainly a air offer anc we will ao1ce ey tne cecision et
(_) 14 ne Scarc.
15 JUDGE LAZO: Thank you, sir.
16 MC. 5AUSER: Mr. Chairman, we certainly will 17 acide by the cecision ci :ne 5carc also, sut we woul: like to pernaps wnen anc 11 the time :cmas, cons 1cer tna:
19 further. But, as I uncerstanc it, tne s:ne:ule 1s sa:n that 20 tne intervenors will be going next anc present1ng tne1r 21 witness and the staff will follow at least as far as that 22 is concernea.
23 JUDGE LAZO: Well, good. I hope none of you are 24 too c1ssatisfie: with tnat.
25
, yy N_j
407 Let's set en w :, tne Boarc ues:1cns so we j
can release Mr. 0:ents.
2 e04 O XAM ANTIQN 3
SY JU GE COLE:
4 5
talked about a lot of different things, anc I a rr. soing to have to go over certain of those again ?cr my own 7
8 On page 3 of your filed testimony in :nis case 9
you indicate in the middle of the page there the DN3R g acceptance limit of 1.17 is generic to all Westinghouse plants using optimized fuel assembly fuel, OFA fuel.
Is there a generic DNBR acceptance limit for LOPAR fuel?
A For the 15 by 15 LOPAR fuel that we used, the 1.3 limit DNBR across the Board with the W-3 L-grid g correlation.
0 All right, sir. So the 1.17 applies to OFA 18 fuel and the 1.3 is LOPAR fuel, and they are in the same 9
"* 'I # ** * "9
- 9*"*' " *""* """*
- 20 A That is correct.
Q The 1.3 value for LOPAR fuel that was in use prior to this proceeding or these modifications, was that used the same way you are proposing to use the 1.17 in that you made a calculation of a ONBR and then added or 4
r 403
]eq subtrar.ted penalties to the appropriate number, depending x;
upon which direction you are going?
A The 1.3 value corresponds to the 1.17 acceptance limit value. In other words, the 1.3, the W-3 correlation for the LOPAR type fuel is as the 1.17 is to the WRB-1 correlation for'the optimized fuel.
6 ,
Q All r.i g h t , sir. You have already answered this 7
question, and I am not sure I understood the answer. So 8
please bear with me.
9 Does the WRS-1 correlation apply to LOPAR fuel?
A We have applied it to the 17 by 17 type LOPAR fuel. We have not reperformed the analysis. As I said 13 previously, the standard LOPAR type fuel assembly in the
( )n/ )
u Turkey Point type reactor is being phased out of operation.
15 Therefore, we are not doing any reanalysis of that fuel 16 type with the new correlation.
Q My concern is not so much with what happens after we have the full core change to OFA fuel, but in the 19 transition stage. And considering what you have just told me of the generic limit of 1.3 for the LOPAR and a generic 21 limit of 1.17 for the OFA fuel, I guess I do not understand 22 why the calculation was made as it was.
23 If the transition is going to take place over 24 a period of two or three years, and I don't know exactly 25 g
Q ,)
409 l
P- 1 what it is, I assume it is going to be a third at a time or
.lk s mething close to that. After the first exchange we will 2
have two-thirds LOPAR fuel and one-third OFA fuel, but you 3
are proposing that we use a ONBR of 1.17 plus penalties.
4 Why don't we use 1.3 plus penalties because it is mostly 3
LOPAR fue1?
A I think I understand your question.
7 Q All right, sir.
A If you go back in time to the point prior to 9
the introduction of the optimized fuel to the core, the g
y, Turkey Point units had a reference anlaysis of all these normal operation anticipated transients. That set of analyses existed analyzed wi*h the L-grid W-3 based
- g --
g correlation to the 1.3 limit. Now that applies to that particular reactor bundle geometry type.
We then submitted an amendment which proposed the incorporation of the new fuel type into that core. Now g
the first question would be, does by the introduction of 18 the new fuel type, does that in any way invalidate the 20 The answer to that is no. As I described in the transition core methodology, since the new fuel type g
has a higher resistance, the older, lower resistance fuel type receives more coolant flow than it would have under its previous operating condition.
, . . - .. -. . . - - - - . - - - . . - . = _ - - -
l 410
) 1 Therefore, the previous existing homogeneous 2 core LOPAR fuel geometry W-3 correlation based anlaysis is
, 3 still valid.
4 Q Are you then saying, sir, that you are not ,
5 changing anything except the simple insertion of'another 6 fuel?
7 A That is right. We are inserting another fuel. ;
8 For each fuel type the geustion is does the insertion or 9 the existence of the other affect the calculated DNBR's.
10 In the case of the previous LOPAR fuel, the II answer is no, it doesn't affect the calculated DNBR's. In 12 the case of the new fuel, we are introducing the newer 13 technology correintion with its corresponding 1.17 limit.
14 And since we are going to a point where eventually there '
15 would be a full core of the new fuel homogeneously, that 16 that was the analysis we were working toward. I D Now the question is does the existence of the 18 residual LOPAR fuel affect that calculated DNBR, and the 19 answer to that is yes, and that was that the intent of the 20 transition core DNBR penalty was. That was the three 21 percent penalty bounds the effect on that homogeneous 22 optimized fuel core analysis with respect to the existence c
23 of the residual LOPAR fuel in the core.
24 Does that clarify the point? I 25 Q Well, I believe ! understand what you are 3 i L
+
411 I 'saying,. sir, but let n.e try to make sure I understand.
2 A Perhaps I can add another clarifying point.
3 The license, if you will, contains with it inherent some 4
' reference analysis. That reference analysis is not 5 resubmitted for each refueling of the reactor. And what'we 6 do in the refueling is assure that each reload core meets 7 all the same key parameter input assumptions to that safety 8 analysis, and the safety analysis is not completely 9 reperformed and resubmitted for recycle.
10 In order to be able to perform that type of a H reload' licensing methodology, we performed the reference 12 calculations that I described earlier in my response to 13 your question.
" I4 Q So are you_saying, sir, that your only concern 15 now with the change from LOPAR fuel into OFA fuel is any 16 safety considerations associated with the OFA fuel because i
II previous anlayses that have been verified and checked and I8 approved on the LOPAR fuel are already in the record and do I9 not have to be redone?
20 A That is correct.
21 Q And why are you so sure that that is not going 22 to be a problem again?
23 A Because the LOPAR fuel is the lower resistance 24 assembly. The same total flow goes through the core. The
- 25 pain of nigner resistance, i.e., tne newer fusi gets iess
412
) I flow than what it gets relative to its low resistance 2 neighbor, the LOPAR fuel. Therefore the previous anlaysis 3 is conservative in that more flow is going to be going 4 through the LOPAR assemblies than what was attributed to it 5 in the previous reference analysis.
6 Q All right, sir. But now you are proposing that 7 a lower value be used. Can you go to that lower value while 8 you still have LOPAR fuel in there which has a generic DN8R 9 value of 1.37 10 A Oy lower value you referring to the 1.17 Il value?
12 Q yes.
,s 13 A Okay. Any change to the operating license i T
'\ > 14 parameters must be evaluated with respect to all fuel types 15 in the core at that time. Therefore, any change to any 16 operating parameter would have to be re-evaluated for both 17 fuel types.
18 The 1.17 value is based on the experimental 19 data base that was used to derive the correlation and does 20 not directly relate to the particular, you know, operating 21 technical specification parameters.
22 The 1.34 safety analysis ONBR value that 1 23 have described in my testimony is related and is based 24 directly on the conservative parameters defined in the 25 technical specifications and with the methods described in
.l g Q/
413 I the final safety analy51s reoort.
h 2 Therefore, any change to those parameters 3
would have to be resubmitted and rever1fied in terms as to 4 whether the change in that ONBR is still above the 5 acceptance limit 1.17 value.
6 All right, sir.
Q 7 I want to talk about rod bow now. On page 2 of 8 your testimony, Item 4 wherein you repeat the concerns that 9 we put in our November 18th memorandum, reference is made 10 to the NRC's safety evaluation to Amendment 92, and I II believe it was --
Amendment 99. That actually was to 12 Amendments 99 and 93.
13 We refer you to pages 3 and 4. Do you have
("
Id that, sir?
15 A I don't have that here, no.
16 JUDGE COLE: Is there an extra copy of that in 17 If you could let him see that.
the room?
18 (A document was handed to the witness by 19 counsel.)
20 BY JUDGE COLE:
21 Q I am particularly interested in page 3.d 22 A Yes, sir.
23 Q And the next to the last paragraph on that 24 page, which is the third paragraph.
25 A Yes.
414 I 1 Q Are you familiar with that, sir?
2 A Yes, I am.
3 Q All right, sir. Well, here is my concern with 4 that. In that paragraph you indicate that using an older 5 method you calculated a rod bow penalty of 14.9 percent, 6 and a new procedure was developed, a more recently approved 7 method topical report, WCAP 8691, Revision 1.
8 In that section it states that the method 9 applies statistical convolution to the test data and the 10 test data -- the interfuel rod gap closure data and the CHF 11 test data.
12 You used a penalty value of 5.5 percent, which
_ 13 is the rod bow penalty associated with a burn-up rate of r .
14 33,000.
15 A That is correct.
16 Q Now how did we get from 14.9 to 5.5 percent? i 17 Where did we get the 14.9 earlier and how come it is down 18 to 5.5 percent?
19 A Between the original England method as 20 referred to in the NRC document and the final method, we 21 had obtained more data on the extent of rod bowing in our 22 1rradiated fuel assemblies, and we had also obtained more 23 information with respect to the effects of that rod bowing 24 on the ONOR.
25 This additional material was included in the
.m i
415 l
l I topical report referred to here as the " Final."
h As the 2
Commission refers to it there was an interim method. We 3 compiled all of the relevant information into the final 4 report on the subject W-CAP-A 691, Revision 1, and th'at 5
method was all the available data that resulted in a 6 reduced penalty.
7 All right, sir. So you have got two Q
l ,
8 considerations working against each other. Increased 9 burn-up would tend to increase the rod bow, and then i
10 fission product buildup and insulation and possibly other II L factors would go against that and would tend to reduce the
- l. 12 penalty.
13 Now do you have enough information over the !
^
Id history of the bunn-up in a rod to say at what point you !
l 15 would get the maximum rod bow? l i
- 16 A You are referring to the first of the two 17 competing effects? ;
! 18 0 Well, I am looking at a rod, and you picked 19 33,000 burn-up as the figure and 5.5 percent. Do you have 20 enough information to state that that is the maximum rod 21 bow that will be associated throughout the history of a 22 fuel rod?
I 23 A As ! stated before, the rod bow increases wiht i
24 increasing burn-up. Therefore, the amount of rod bowing l 25 does increase beyond the 33,000 value. !
i i
l
416 1 However, for the purposes of determining what
)
2 is the effect of that rod bow on the DBNR analysis that I 3 have' described previously, 33,000 is selected as a cut-off 4 value.
5 The peaking factors that -- the power 6 production capability of the rod begins to. drop off even 7 before 33,000, but the 33,000 value was selected as being a 8 value which is for the nature of our pressurized water 9 reactor designs is physically incapable of achieving a near 10 limiting peaking factor.
11 12 13 l( y 14 15 16 17 18 19 20 1
T 21
[ 22 23 l
! 24 25 i
15-1-JcOW31 417 q Do you know if rod bow would be at any greater 3
A I) s 2 value prior to achieving that burnup rate?
A Not. It is monotonically increasing.
3 It is monotonically increasing. All right, sir.
4 Q r
l 3 So, the original estimates for rod bow penalty of 14.9, what i
6 were they based on?
7 A They were based on earlier preliminary estimates 8
on the hoving based on smaller data samples for the amount 9 of how, and also an incomplete set of calculation methods.
[
10 These methods were all completed and submitted in
- 11 h the reference, approved topical report.
12 l Q Does the 5.5 percent fit the 95/95 statistical
() 13 btandard test that you referred to earlier?
r 14 A For that particular phenomenon, rod bow phenomenon.
i It is a bounding calculation to the data. The function that 15 16 we used to define amount of bow bears the observed data of l 17 rod bowing on that basis. l JUDGE COLE: All right, sir. Thank you. That is 18 I ,
J 10 all I have. Thank you.
Jul INDEX 20 BOARD EXAMINATION I I BY JUDGE LAZO:
21 f 22 I Q Mr. Dzonis, I just have two or three questions.
t l
l 23 I I note that it is almost two years since the Commission issued l
24 l the amendments to Licensee's license, amended the tech specs.
l (~)
wra repoet.... ine f 25 Do you know if the Licensee is complying with those tech spec
9 15-2-JocWo1 418 1 changes, and is operating under those new conditions?
- N 2 A To the best of my understanding, yes.
3 Q That they are? Well, now, some of the amendments 4 ,that were authorized involved an increase in the hot channel n
5 factor limit. Of course, an increase in the total. peaking 6 factor limit that we have been talking about.
7 I think it was mentioned there would be changes in 8 the overpower set points.in terms of hydraulic limit -- in 9 order to accomplish these changes, did_the licensee have to 10 make any equipment changes? Did it involve different components I.
11 or mechanical changes involved? Any hardware changes, do you ,
12 Iknow?
() 13 A To the best of-my understanding, no. They shoul I -
i 14 be accomplishable by changing settings or instructions to the 15 operators.
16 I am really not sure of the equipment necessary to i
l 17 do that. There are just changes in the values. '
j 18 So that the answer to the next question is self-Q l 10 '
explanatory, namely if this licensing board should decide that !
20 the licensee should go back and operate the reactors under the
~
21l previous conditions and the previous limits of the technical i
h 22 h apecifications, the licensee would have to do very little to E
^
23 accomplish that? p 24 l A I don't.really know that. I do not implement changes l
('Reporters. Inc.
be Fw) 25 at the plant site. I'am not sure as to what is involved in that.
15-4-JcoWal 419 1 JUDGE LAZO: Maybe we will ask the Staff that l )
-> 2 tomorrow.
3 I have no further questions.
XXX INDEX 4. BOARD EXAMINATION t
5 BY JUDGE LUEBKE:
6 Q Mr. Dzenis, I have a question here which is rather 7 general, and for purposes of background orientetion and 8 clarification, and I don't mean to broaden the scope of the l 9 hearing beyond its understood limits, I want to explore a i
10 hypothetical situation.
l s h 11 ' Maybe a little oddballing, and that is that some j i : I i
- 12 errors have been made in your testing procedure, or your sm' f
13 hydraulic effects, and so on, and in the calculations and i1
(/ l
'4 'n i the interpretation of results, and after all of this fuel
,15 gets shipped to Turkey Point and installed, which everybody 16 things is characterized by this DNBR of 1.17 that we have been 67 ' talking about all day, and actually it is 0.9.
18 My question then is: In this hypothetical, outragious I
situatian, if the reactor is run up to full rated power, what 20 : would happen? What is expected to happen?
Y 21 'l!g A 3
I think -- to answer your question, I think I have l
d 22 ll to go back through a set of steps that occurs in the evaluation I
23 h of ; the -- as to whether a particular reactor operation meets the j '
(') 24!!DNE limitation requirements.
Aa Fi J Reponen, lm. ! i 25l The first step is the critical heat flux experiments f s
n 1
'15-5-JocWal 420 1 -- the associated analysis of those experimental data. The 2 statistical analysis measured to predict critical heat flux 3 values, and the determination of an appropriate limit DNBR 4 j for.that correlation, and that the second step of the operation 5 ; deals with the analysis of the plant. operating restrictions
'6 as defined in technical specifications, and the analysis of 7 anticipated transients, assuming the reactor protection set 8 points and all other requirements in the technical specifications.
9 I think each of these pieces receives careful review 10 and verification, and I am aware of no error in these at this 11 point in time.
I 12 There are procedures'in place --
I am sur'e the
./~) 13 J owners and_the cperators of the plant to bring to proper I
(_/ --
14 ' attention that there were -- in the possibility of errors that 15 'they-are properly evaluated.
16 If there were an error in any piece of the analysis, f' l
17 - it would require re-evaluation of the full set of the' cases to !
r i
18 . a s's u r e that, indeed, the proper DNB limit were not exceeded. !
I might also add that in defining. normal reactor
[
l- !
20 operation, I believe at the end of your question yoj asked me 21 if it is running at full power, the reactor does not operate 22 lat a DNBR which approaches these limiting values we have been 23l discussing here today.
[j 24 l In fact, the DNBR itself is not a control parameter, Ace-Fu Reporters, Inc.
25 if you will. Your independent controls and definitions on all l
1
15-6-JooWol 421
) the operating parameters, s uch as pressure, temperature, power,
() 2 flow, peaking factors, which affect the DNB analysis, and that 3 those operating restrictions on normal operations combined to 4,, form a far more restrictive set of conditions with respect to t
5 DNBR than the spectrum of conceivable set point conditions.
6 So, even if there were some sort of error in 7 analysis somewhere,.an error in the consideration of how a set !
8 Point is there for a protection system, that does not necessarily 9 have any implication with respect to the normal operation, 10 operating condition.
11 nl In other words, the DNBR, that condition is al' ready 12 f much higher than the -- either the safety analysis, DNBR, the
() 13 1.34, or the correlat' ion of the 1.17.
14 Q Are you'saying it is not.likely to happen. Is that 15 sort of the gist of your explanation?
l i
16 A It is not likely to happen. An error -- a hypothetical 17 error to induce a normal operation situation that would --- ;
i 18 Q I was looking for a simple answer; and a simple answe(
10 was that as they ran this plant p to full power, someplace 20 i !along the way, because it was the wrong fuel, it would start
. I overheating you you voold get some failire of fuel, because 21 li 22 of the fact you had shipped the wrong fuel.
i 23 .! A There are controls in monitoring procedures for such 1
(~3 24 l situations --
Aap-FU Reporters, Inc. '
25 Q And one of those in my mind might be that you would I
15-7-JooWo1 422 I
] start seeing radioactivity in the primary coolant.
,3
(_) 2 A Well short of radioactivity in the coolant, there 3 is monitoring on the temperature --
gg Q You would claim you wouldn't even get to the point
.I 5 'i f seeing radioactivity.
i 6 You see I am trying to assess --
I read this 7 w rd, 'significant' increase of hazard, and I am trying in 8'.
rather -- without any significant figures --
9 - A I understand what you are asking. For example, there i
jo ' is monitoring on flow rate, pressure, temperature, and power i,
11 5 distribution within the coolant, both within core --
12 Q Operators know what is going on? They might get
/ 13 up to eighty percent power, and say: Hey, there is something
)
i 14 wrong. j 15 A That is correct. There are start procedures 16 that define --
17 Q And they are not going to get to the situation where 18 the primary coolant has radioactivity in it?
'O A No. They should see this well before that ;
20 development.
21 Q And the same answer would apply if I gave you the ;
22 j hypothetical situation where the error in shipment of fuel was i
23 ] one characteristic of a DNBR of 1.0 instead of 0.9/
i 24 ij A The case of a misloaded assembly , a mis-constructed WFQ Reporters, foc.
25 i assembly explicitly dealt with that in the final safety i
i
i
>15-8-JooWal 423
] analysis report.
- O~
l 2 In that case, we indicate what systems there are r i
3 in place to monitor.that' type of an event. .I mentioned 4 , explicitly the in core ~ monitoring systeu for determining 5
jpeaking facors and fission distrib'ution within the core.
6~ If-there were a misload assembly of some sort, 7 such-that it requires a change in the peaking factor to cal 1 8 into question the DNBR, that woul d be visible on the in core 9 power distribution monitoring system.
10 Q Because I have been intrigued all along in this l 11 " case by.three significant figures; 1.17 and 1.34.. It really 12 doesn't matter, because if you ship fuel that was 1.20 that l.
13 - would be okay. If you ship fuel that was 1.15, we maybe you 14 could run to 99 percent power before these factors thatyyou i
15 have mentioned come into play, and the operating crew begins ;
l-
-16 to sense that gee,.it isn't just right. .l l-t 17 A Now, I think you made a very important. That one l 18 cannot base the total question of safety on any particular s
10 -individual criterion,1or any individual, particular operating 20 [ procedure.
~
~21 It'is the entirety of the system of analysis and i
22 kaonitoring equipment, training procedures, which provides the i
23,l!overallsafety.
I 24 p Fw( )d Reporten. Itte.
i .
~
My belief'is the types of errors, or misloadings of 25 fuel that you describe, that there are sufficient systems to l
15-9-JeaWel 424 1
i determine that at the -- in the reactor prior to the DNB
('"' )
2 phenomenon occurring.
Q So this thing we talked about today in such bitter 3l 4; detail really isn't all that hazardous and a one-shot concept.
5 l In other words, there are many protective measures?
4 6 A That is correct.
7 JUDGE LUEBKE: Thank you, s i r'. ;
I I
8 ' JUDGE LAZO: Now, before excusing Mr. Dzenis, I 9 want to ask if any of the parties have any follow-on questions ;
I 10 j regarding questions which have been raised by the Board.
't II '
If so, I think you should have an opportunity to ,
i i 12 ] ask them now bef ore we excuse the witness.
,/ 13 Dr. Edwards? i 14 MR. EDWARDS: Yes, I do have some follow-on questions, 15 if I may, Mr. Chairman.
16 JUDGE LAZO: Surely. !
i QG INDEX 57 RECROSS EXAMINATION i 18 BY RR. EDWARDS: I Q Mr. Dzenis, in connection with the last sequence i
20 ..of questions, I was intrigued by your comment about various i I
2I j monitoring capabilities which exist prior to an actual onslaught !
h o l 22 [ of the critical heat flux, even assurang that the critical heat
- t. ,
23 flux were wrongly estimated.
( ) Are you referring to such things as in-core temperature ke F1_j Aeroners. Inc. ,j l
'S i
} monitors that would be capable of actually identifying the statuk I l l !
15-10-JecWcl 425 j of the hot fuel pin the core, or --
-)' 2 l A I was referring to th a in-cc.! a fission detectors.
3 There is a detector which would sense the neutron flux at i
4j,different portions within the core.
i 5 And from those measurements can be verified the i
6 Power distribution is consistent with tech spec limitations 7 on radio peaking factor, and total peaking factor like the i
l g, type that we were discussing with respect to this amendnent, j I I 9! and if there were hypothetically induced deviation in that j l '
10 j value,-that would be sensed by that type of a monitoring 11 equipment.
12 That is why there is an independent technical I
j 13 specification on these values of peaking factors to assure a la proper monitoring of that behavior.
15 Q But I am not sure if I understand you. If the 16 calculation of critical heat flux was incorrect for one reason ;
I 17 or another, which I think was the premise of the question, so 18 that you were in fact -- the conitors were looking for a W critical heat flux which was actually quite a bit higher than 20 the true critical heat flux, how would these in-core monitors 21 be able to detect this unanticipated lower critical heat flux? ;
I it
?
22 A Let me make a couple of things very clear here.
23 9 First of all, I am aware that no such error, or no such technical i, !
(~) 24 y basis for such a deviation as was hypothesized to clarify, l i
s=L__j neporim, rne. lif f 25 j' -- to impact with this particular criterion. To overall reactor l.
l l
l l l
l
If PS-11-J ocWo1 426 I i
i safety.
(_/ 2 And, --
but we do not rely on this detection to 3 meet the DNB criteria. The f undamental reliance Lis the overall al,' technical specification requirements, reactor protection
'I 5 set points which are defined in those. tech'uical specifications, 6 and then the previous review and analysis of the calculations 7 we have'done.
8 'l Q Yes, I understand that. But isn't it the case that 9 if departure from nucleate boiling ever does happen, it is 10 almost by definition an unexpected an unwanted result, and 11 'therefore must result-from a combination of errors of some ,
12 sort? j
() 13 -What I am wondering is under the hypothesis which f
14 I agree you don't see where this hypothesis comes from, but 15 under the hypothesis the critical heat flux is significantly 16 lower than that which is predicated in the analysis, I have !
1 17 difficulty mderstanding what kind of monitor would be able !
I
-18 to identify that the hot pin was approaching a critical heat i
10 flux which you don't even know is the critical heat flux, i l
20 because your analysis predicates a higher one. -
i 21 ' Could you enlighten me on that subject?
22 !. A I guess I am having a very difficult time dealing 23, with this hypotheticality, because -- at least in our experience 1
(- 24 as a Westinghouse vendor of nuclear reactor fuel -- we have never
% FL} Reporters. Inc. i I 25 ! observed failures attributable to DNB mechanism.
I
IS tJ-12-JecWal 427 1 Therefore, it is difficult for me to presume a
> ; I
-/ 2 ' hypothetical error in the formulation of the critical heat 3 flux that would not have already manifested itself in a visible 4Ifailure by that mechanism in auy of a host of reactor locations
$ which are supplied without fuel, and analyzed with this 6 technology.
7 I guess that is why I have a very difficult time 8 relating the capability of a particular sample -- example monitor-9 ing system, and I cited two --
this hypothetical gross error 10 in the prediction of this phenomenon.
11 b Q So, what you are saying then is that you are confident I l 12 , tha t the various monitoring procedures would decect something, i
() 13 although you can't put your fingure'on what they would be? ;
I 14 A In aggregate, that is correct, yes.
f 15 Q Another follow-up question that has to be --
you made 16 it quite clear that because of the higher hydraulic resistance, il the greater hydraulic resistance of the OFA fuel, that the 18 LOPAR fuel in the core would get more than its share, if you ,
'O like, of the flow.
20 A That is correct.
l 21 . , Q And consequently, the previous analysis applied s
22 Ito the 10 PAR fuel would be conservative for that reason?
23 A Yes.
("; 24 y i
q voesn't it follow by the same token that the new I
I w FL__J Heporters, Inc.
25 ganalysis for the OFA fuel would tend to be non-conservative 1
ll i
15-13-JocWel 428 1 for that very reason.
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2 i A And that is why we have quantified the transition 3 core penalty I described in my testimony, and was questioned i
- about earlier today. It is because of that specific physical o
l 5 ' f act that we determined the transition core effect, and applied 6 it to the optimized fuel analysis.
End 15. 7 Su W fois.
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429 Ol6-1-SueW 1 Q So there is an explicit recognition in that J 2 procedure that the new analysis based upon the new DNBR limit 3 of 1.17 is because of some starvation of flow, you might say --
4 A Uh-huh.
5 0 -- diverted to the LOPAR fuel, but in itself it 6 will tend to be non-conservative and, therefore, has to be i
7 accounted for quantitatively by a numerical penalty.
8 A I guess in the aggregate, if you take the homo-9 geneous optimized fuel-safety analysis and the transition 10 core penalty, that the net effect of those is conservative.
I 11 And without one or the other, the analysis is ,
12' l incomplete. But the two together form a conservative ex l
( ,) 13 analytical technique. i i
14 f Q Even though not confirmed by any direct physical !
I 15 i measurements of the actual hydraulic flows through the core?
u i
i 16 4 A No. The flow rate in the core is confirmed. There !
a .
17 ti is -- !
18 Q The aggregate flow rate in the core is confirmed. ,
n 10 A Yes.
o b i 20 l Q But the precise ratio of flow through the OFA fuel i
21 as compared with the LOPAR fuel, is that confirmed by a 22 specific --
23 A That's confirmed in the hydraulic test program
(~) 24 I referred to earlier today. !
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Is that a calculation, or is that a physical l 25 0 i I
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l I
430
- 16-2-SueW 1 measurement?
2 A It's a physical measurement of the flow rate 3 through the assemblies.
4 Q Through the assemblies?
5 A Yes.
6 Q And would you refresh my memory, please?
7 A That was the experiment to determine the hydraulic 8 resistance characteristics of each fuel soundly. We run 9 first each assembly individually, then two assemblies, one 10 assembly of each' type side by side in a hydraulic test loop.
11 Q So, you are saying that -- again, we are going back i I
I 12 to the same data base. The data base consists of two fuel i j 13 assemblies side by side. ;
O l 14 ll A Correct. i 15 l Q And that this is where the test resides. And from 16 that you extrapolate to the entire more complicated geometry 4 i 17 d of the full core configuration, so that you feel confident 18 that you know accurately exactly what.the flow is through the '
0 i t
'9 !, OFA fuel in the existing core geometry? l
!l I J i 20 ! A It really isn't anymore complicated. There are ;
21 just several sets of adjacent pairs, actually many sets of ,
1 22 adjacent pairs of standard and optimized fuel assemblies ;
i 23 within the core.
(^) 24 Like I said, the assemblies are loaded checker- !
.w t _ ; n, port.<i.ine. :
25 roard fashion. We cannot have alternate configurations. Theyf l
431
- 16-3-SueW1 have to line up box by box. Therefore, the only alignment 2 that is adjacent to each other is a one-to-one lineup.
3 Q In other words, you have to have adjacent to every 4 OFA fuel bundle a LOPAR fuel bundle?
5 A No. It's either another optimized fuel bundle 6 or a LOPAR fuel bundle.
7 Q Well, yes, that goes without saying. Now, with 8 regard to the entire amendment as a whole, does not this 9 entire amendment relate to the vessel flux reduction program 10 'for Turkey Point 3 and 4?
i 11 A I believe that's the -- in the title, this is 12 one of the elements of it.
() -13 Q Now, all thing being equal, doesn't the vessel 14 flux reduction program result in effectively a physically 15 smaller or more compact core?
l 16 [ A No. The core is the same size with either fuel 17 i type or with either peaking factor. l 4
0 18 'i O What I mean here is that is it not the case -- l l
10,' perhaps I misunderstand. But is it not the case that there 20 has to be a detection for the walls of the pressure vessel i
i 21 by means of rods which are essentially neutron absorbers but 22 not giving off the ful neutron flux that would normally be 23 anticipated?
24 MR. BAUSER: Excuse me. Mr. Chairman, I am going WFL Reporters. Inc.
25 to object to that question. I fail to see how -- ,
I i
432 416-4-SueW 1 JUDGE LAZO: And we are going to sustain the
'e ' 2 objection.
3 MR.BAUSER: Okay.
1j JUDGE LAZO: I think we are getting beyond the 5 area of answers that were provided for the Board questions, 6 Dr. Edwards.
7 DR. EDWARDS: Fine. Thank you.
8 JUDGE LAZO: I didn't mean to cut you off. But, 9 you have no more questions?
10 DR. EDWARDS: I have no more questions.
11 JUDGE LAZO: Thank you, sir. Do any of the other ,
12 parties have any questions?
(,) 13 MS. YOUNG: Yes, Judge Lazo. I have one follow- l
+
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14 up question regarding a question of Judge Cole's.
i 1 15 0 RECROSS EXAMINATION 16 ,l BY MS. YOUNG: f i
INDEXX 17 c Q He was discussing with you whether there was a 18 i change in the DNBR limit applied to the core at Turkey Point j i
'o '
as a result of the introduction of OFA fuel. And he asked ;
d i 20 a series of questions basically inquiring whether 1.3 was
! i 21 any good for LOPAR fuel anymore now that you have an OFA '
22 assembly adjacent to the fuel.
23 I would like to ask you, what DNBR limits apply
) 24 4e F__J Ceporters, Inc.
to Turkey Point core, the transitional core?
25 A The 1.17, WRB-1 based acceptance limit applies to l
433
- 16--S-SueW I. the optimized fuel in that core. The 1.3 based,W-3 based O 2 acceptance limit applies to the LOPAR fuel.
3 Q So, basically you have one core with two limits 4 for --
5 A That's correct.
6 Q All right. Now, is this use of two limits also 7 implemented in establishing the operational limits for the 8 reactor, and how is that done?
9 A Well, there is only one set of operational limits 10 to the plant as a whole. So it's possible for one -- for 11 , example, the allowable temperature at a particular point to i I .
12 ! be more constrained by the one fuel versus the other. Which-13 ever fuel is more constraining establishes the limit for 14 the plant as a whole.
15l! O Okay. And where would this' limit be for the 16 plant? !
- 1) A For example, the DNB core limits which are 18 g supplied in the technical specifications. I i
li
'O Q Okay. And is -- the curve effecting pressurizer l h I I
20 i pressure and cooler temperature, does that incorporate both 21 the DNBR limit of 1.3 for LOPAR fuel and 1.34 OFA fuel?
22 A Yes, that's correct.
I 23 Q I believe that in response to a question from
~
24 Judge Luebke you stated that even at full power the hot rod
()
w FQj Ceporters, Inc. l 25 would not experience departure from nucleate boiling?
l
434 fl6-6-SueW I A That's correct.
2 Q Isn't it true that the core would only experience 3 DNB if there was a transient while the reactor was at high 4 power and the transient was severe?
5 A You have to add, in my opinion, considerable 6 numbers of such compounding hypotheticalities to achieve 7 DNB in a reactor, yes. '
8 Q And so if you miscalculated the critical heat flux 9 of the DNBR ratio and if the hot rod does experience DNB for 10 a period of time before a trip following a transient, would 11 you expect severe damage in the core?
I 12 A No. DNB does not necessarily result in fuel
() 13 cladding failure.
I 14 ,- O Could you explain that a little further?
15 ll A The DNB is to be avoided or precluded by intent i
16 l in the safety analysis, because DNB can result in rapid ,
17 temperature changes in the cladding if the DNB phenomenon 18 i occurs. !'
i
'O l Just because a temperature increase occurs does m
20 not mean that the cladding would necessarily fail. There l 21 have been some research programs studying under what conditions 22 might one begin to consider the likelihood of failure after 23 DNB and it can take long time periods or very high tempera-24
(~)
Aers_J r.worters, ine.
tures to result, or the combination thereof, to result in the 25 actual failure. !
435
- 16-7-SueW 1 ,
So, therefore, the existence of DNB would not
{}
2 necessarily result in a direct failure.
3 Q So, is it true that there are other reactor core 4 ~ protective systems that would come into play to prevent --
5 A ~ Well, there are other reactor core protection 6 systems. There is this physical characteristics I have 7 described. And in addition to after all this, as Judge Luebke 8 has mentioned, there are technical specifications on the 9 cooling activity as monitored regularly.
10 Q So, DNB or DNBR is just one line of defense 11 against --
12 A That's correct.
)- 13 Q -- core damage?
14 A That's correct. A very conservative inner-o '
15 0 bound. .
16 1 .MS. YOUNG: Thank you. l I !
f.
17' JUDGE LAZO: Mr. Bauser, do you have any questions?
18 MR. BAUSER: No.
I i
~
'O J BOARD EXAMINATION INDEXX 20 BY JUDGE COLE:
21 Q Just one point. Ms. Young mentioned the 1.34 22 , safety analysis on DNBR for OFA fuel and 1.30 for LOPAR fuel.
23 I guess my confusion is this.
,p 24 If 1.30 is the generic limit for LOPAR and 1.17 wrc J ceponen, inc.
25 is the generic limit for OFA, wouldn't you have to add any
436 516-8-SueW I penalties to the 1.30? For example, a bending rod bow G 2 penalty of 4.7?
3 A As I define in my testimony between -- if I start 4 with the optimized fuel assembly which has been the dominant 5 portion of discussion, there is a 12.7 percent difference 6 between the 1.34 safety analysis value and the 1.17 limit.
7 In the case of the LOPAR fuel, there is an 8 equivalent margin credit of 11.1 parcent which corresponds 9 to 12.7 percent in the optimized fuel.
10 In the case of LOPAR fuel, there is obviously II no OFA geometry question. There is no transition core !
12 ' penalty. There is a rod bow penalty effect on LOPAR fuel, 13 as I identified previously in my testimony, and that is ,
Id j! accommodated by that 11.1 percent equivalent margin credit.
d 15 d Q I guess I still don't understand if both of them ,
I i 16 are their generic limits, why wouldn't you add the 4.7 rod l c i 17 I bow penalty to the 1.30, as you added the 5.5 percent to the i i
18 OFA limit? !
, A Well, the -- those -- I'm certain these are l p !
20 I subtracted from the plant safety analysis DNBRs, not added i 21 to the design limit.
i 22 Well, the safety analysis minimum DNBR for LOPAR !
Q 23 fuel is what, sir? Do you know?
n
'i 24 A It's equivalent to 11.1 percent, in excess of 1.3.
42 FL.I Ceporters Inc.
25 I don't know what that value is, i
437 16-9-SueW I Q Oh, in excess of 1.3?
2 A That's correct.
3 JUDGE COLE: All right, sir. That's all.
4 JUDGE LAZO: Well, one last chance. I'm sure 5 Mr. Dzenis ic ready to relax.
6 Are there any further questions from anyone?
7 MS. YOUNG: Could we have a moment to confer?
8 JUDGE LAZO: Yes, of course.
9 (Pause.)
10 MS. YOUNG: I'm sorry for the confusion and delay.
11 RECROSS EXAMINATION ,
i 12 BY MS. YOUNG:
I bXX 13 Q Does the 1.30 limit for the W-3 correlation have i
14 ! any room for uncertainty?
9 o
15 ! In other words, hasn't over a period of years the 0
16 l W-3 correlation been improved and that number is actually :
!i i 17 1' lower? l 18 j A I have defined this equivalent margin discussion
" l, as the basis of the margin on the previous L-grid analysis.
t 20 j The W-3 based correlation limit is 1.3.
21 There have been other intermediate correlations 22 which accounted for other factors which had demonstrated 23 that that is conservative, that the value could be lower 24 than that. These are intermediate to the -- if I'm going
[
As F._ ~) Ceoorters, Inc.
25 from a W-3 to the eventual WRB-1, 1.17 value. j I
438 16-10-SueW l O But isn't it Westinghouse's present position 2 that the W-3 limit of 1.3 actually has margin over what 3 that limit is actually understood to be based on further 4 tests?
5 A That is correct.
6 MS. YOUNG: No further questions.
7 JUDGE LAZO: Dr. Edwards, you have the look of 8 someone who would like to ask a question.
9 DR. EDWARDS: Yes, Mr. Chairman.
10 RECROSS EXAMINATION l
11 BY DR. EDWARDS: !
l INDEXX 12 Q Following along with Judge Cole's question, could l
- s i
) 13 I, Mr. Dzenis, ask for a clarification because now I'm not '
l 14 I sure that I understand again?
15 The 1.34 calculated minimum DNBR, is that specific i l 16 3 for OFA fuel? i l i 4
17 [ A That is correct.
I 18 0 That is specific for OFA fuel? What is the l 4
corresponding figure for the LOPAR fuel?
20 l A The corresponding figure would be the equivalent i i
21 of 11.1 cercent margin above 1.30. I don't know what that 22 value is offhand because the Board's questions dealt with 23 the OFA.
- 24 Q Now, just again I'm not quite sure. Do you mean l Amt ; ceporters. Inc.
j 25 11.1 percent plus -- 1.3 plus 11.1 percent of 1.37 i i
439
- 16-ll-SueWl A That's correct.
[~D
\> 2 Q That's what you are referring to?
3 A Yes.
l O Okay. So that comes to 1.443; is that -- does 3 that ring a bell?
6 A Not offhand. But if that's what the calcula'ed t 7 value is, yes.
8 Q So, in other words, if what you've told me is 9 correct, then I would gather that the 1.34 figure would be
'10 the minimum calculated DNBR for the OFA fuel --
l l 11 A Yes.
, l t
12 ! Q -- and 1.44 would be the minimum calculated DNBR I
() 13 i
for the LOPAR fuel?
14 ;. A It's equivalent to that, yes.
1 l 15 [6 Q ' Equivalent?
, i l 16 l A Yes, s
17 i Q It's not that?
i 18 A Well, 11.1 percent above 1.3.
Q Oh, that's what it is?
]b 20 ' A- Yes.
21 DR. EDWARDS: Okay. Thank you. No further 22 questions, Mr. Chairman.
l 23 JUDGE LAZO: Well, then I guess we have come to 24 MFQ)[~ Ceporters, it'c. a reasonable stopping point for this afternoon.
25 Mr. Dzenis, you may be excused. Thank you very l
t L
440
- 16-12-SueW much for your testimony.
.m 2 WITNESS: Thank you.
3 (The witness stood aside.)
4 j JUDGE LAZO: Is 9:30 in the morning convenient 5 for everyone to resume?
6 Do you want to go earlier or later?
7 (No reply.)
8 None of the above? Then, let's recess until 9:30 9 in the morning when we will start off with Dr. Edwards' 10 direct testimony and follow with cross-examination.
11 Thank you.
I I
12 ! (Whereupon, the hearing is adjourned at 4:43 p.m.,
() 13 Tuesday *, December 10, 1985, to reconvene at 9:30 a.m.,
14 i Wednesday, December 11, 1985.)
l 15 [ ************
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. 1 CERTIFICATE OF OFFICIAL REPORTER NO PAGE NUMBER O
This- is to certify that the attached proceedings before ,.-
the UNITED STATES NUCLEAR REGULATORY COMMISSION in the matter of: .
NAME OF PROCEEDING: FLORIDA POWER AND LIGHT, COMPANY (Turkey Point Nuclear Generating Plant, Units 3 and 4)
DOCKET NO.: 50-250 OLA-1; 50-251 OLA-1 PLACE: MIAMI, FLORIDA DATE: TUESDAY, DECEMBER 10, 1985 were held as herein appears, and that this is the original transcript thereof for the file of the United States Nuclear Regulatory Commission.
(sigt) AM -
(TYPED) GARR T [J. WALSH, JR.
Official Reporter Official Reporter Reporter's Affiliation ACE Federal Reporters (siqt)T % 8.2, k. h'f47 8Q (TYPED) M Y PIT L E H . WALSH Ofpicial Reporter -
o ( sigt &
)./- e'O * % M/ h /t' N (TYPED) MARY C,f /SIMONS Official Reporter
.