ML15335A538
| ML15335A538 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 11/17/2015 |
| From: | Atomic Safety and Licensing Board Panel |
| To: | |
| SECY RAS | |
| References | |
| 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01, NRC-2016, RAS 28598 | |
| Download: ML15335A538 (321) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Entergy Nuclear Operations, Inc.
Indian Point Nuclear Generating Station Docket Number: 50-247-LR and 50-286-LR ASLBP Number: 07-858-03-LR-BD01 Location: Tarrytown, New York Date: Tuesday, November 17, 2015 Work Order No.: NRC-2016 Pages 5002-5321 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.
Washington, D.C. 20005 (202) 234-4433
5002 1 UNITED STATES OF AMERICA 2 U.S. NUCLEAR REGULATORY COMMISSION 3 + + + + +
4 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 5 + + + + +
6 ________________________________
7 In the Matter of: : Docket No.
8 ENTERGY NUCLEAR OPERATIONS, INC. : 50-247-LR 9 (Indian Point Nuclear Generating : 50-286-LR 10 Station, Units 2 and 3) : ASLBP No.
11 ________________________________ : 07-858-03-LR-BD01 12 Tuesday, November 17, 2015 13 14 Doubletree Tarrytown 15 Westchester Ballroom 16 455 South Broadway 17 Tarrytown, New York 18 19 20 21 BEFORE:
22 LAWRENCE G. MCDADE, Chairman 23 MICHAEL F. KENNEDY, Administrative Judge 24 RICHARD E. WARDWELL, Administrative Judge 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5003 1 APPEARANCES:
2 On Behalf of the U.S. Nuclear Regulatory 3 Commission:
4 DAVID E. ROTH, ESQ.
5 SHERWIN E. TURK, ESQ.
6 BRIAN HARRIS, ESQ.
7 U.S. Nuclear Regulatory Commission 8 Office of General Counsel 9 Mail Stop 15 D21 10 Washington, D.C. 20555 11 david.roth@nrc.gov 12 sherwin.turk@nrc.gov 13 brian.harris@nrc.gov 14 301-415-2749 (Roth) 15 301-415-1533 (Turk) 16 301-415-1392 (Harris) 17 18 On Behalf of Entergy Nuclear Operations, Inc.:
19 KATHRYN M. SUTTON, ESQ.
20 PAUL M. BESSETTE, ESQ.
21 RAPHAEL "RAY" KUYLER, ESQ.
22 Morgan, Lewis & Bockius, LLP 23 1111 Pennsylvania Avenue, NW 24 Washington, DC 20004 25 202-739-5738 (Sutton)
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5004 1 202-739-5796 (Bessette) 2 202-739-5146 (Kuyler) 3 ksutton@morganlewis.com 4 pbessette@morganlewis.com 5 rkuyler@morganlewis.com 6
7 On Behalf of the State of New York:
8 JOHN J. SIPOS, ESQ.
9 LISA S. KWONG, ESQ.
10 MIHIR A. DESAI, ESQ.
11 of: New York State 12 Office of the Attorney General 13 Environmental Protection Bureau 14 The Capitol 15 Albany, NY 12224 16 brian.lusignan@ag.ny.gov 17 18 On Behalf of Riverkeeper Inc.:
19 DEBORAH BRANCATO, ESQ.
20 Riverkeeper, Inc.
21 20 Secor Road 22 Ossining, New York 10562 23 800-21-RIVER 24 info@riverkeeper.org 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5005 1 On Behalf of Westchester County:
2 CHRISTOPHER INZERO, ESQ.
3 Assistant County Attorney 4 Westchester County Government 5 148 Martine Avenue 6 Room 600 7 White Plains, New York 10601 8 914-995-2000 9
10 On Behalf of Westinghouse Electric Company:
11 RICHARD J. COLDREN, ESQ.
12 Westinghouse Electric Company 13 1000 Westinghouse Drive 14 Cranberry Township, Pennsylvania 16066 15 412-374-6645 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5006 1 P R O C E E D I N G S 2 8:31 a.m.
3 CHAIRMAN MCDADE: This hearing will come to 4 order. We will continue with discussion on Contention 5 Number 25. Before we get started however, there was 6 one administrative matter that I forgot from yesterday 7 that I'm embarrassed that I forgot. Looking at the 8 Statement of Position of Entergy, this is Exhibit 615, 9 Page 33, third paragraph down. First word of the 10 paragraph is stricken from the record.
11 MR. KUYLER: Your honor, could you say 12 again the page?
13 CHAIRMAN MCDADE: Page 33, third paragraph.
14 MR. KUYLER: Yes, your honor.
15 CHAIRMAN MCDADE: We are striking from the 16 record as grossly inappropriate the first word of that 17 paragraph. Are you ready to proceed?
18 ADMIN. JUDGE WARDWELL: What about the 19 homework assignments? Should we start with Dr. Lahey 20 first and then we'll assume Entergy has something in 21 regards to the data? If not, we can give you more 22 time if you need it. But let's start with Dr. Lahey.
23 DR. LAHEY: So, he wanted these.
24 ADMIN. JUDGE KENNEDY: Well, we don't what 25 these are, Dr. Lahey.
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5007 1 CHAIRMAN MCDADE: Yes.
2 DR. LAHEY: Oh, they're -- you asked me to 3 --
4 CHAIRMAN MCDADE: Dr. Lahey, excuse me. If 5 you're going to talk, you've got to be sitting with 6 the microphone. Otherwise, it's going to get lost for 7 the record.
8 DR. LAHEY: Okay.
9 CHAIRMAN MCDADE: And we can get someone --
10 DR. LAHEY: Richard Lahey. Your honor, you 11 asked me to give you the references that I had cited 12 yesterday. So I have copies of them for you.
13 CHAIRMAN MCDADE: Well, could you just tell 14 us what the cites are?
15 DR. LAHEY: What the references are?
16 CHAIRMAN MCDADE: Yes.
17 DR. LAHEY: Okay. There's three of them.
18 One of them is Kanaski, the other one is Arai, and the 19 other one is Korth, et al. And these had to do with 20 low cycle fatigue versus high cycle fatigue and the 21 effect of this on failure for testing components that 22 were irradiated.
23 ADMIN. JUDGE WARDWELL: And were these 24 previous exhibits that you had submitted?
25 DR. LAHEY: Yes. They were in my NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5008 1 testimony, but Judge McDade asked me if I would -- I 2 thought it was my homework assignment if I would make 3 him a copy of them, so I did.
4 MR. SIPOS: And Judge, this is John Sipos, 5 we have the exhibit numbers if you would like.
6 ADMIN. JUDGE WARDWELL: And that's all we 7 needed.
8 MR. SIPOS: Would that be helpful?
9 CHAIRMAN MCDADE: Yes.
10 ADMIN. JUDGE WARDWELL: That's what I think 11 we were trying to imply is we just needed the cites 12 for those, we didn't need the cites.
13 MR. SIPOS: So the first one, New York 14 State 564 is Arai, A-R-A-I. Another one is 15 Riverkeeper 152 and that's Korth, K-O-R-T-H. And the 16 third is NRC 177 and the first author is Kanasaki, K-17 A-N-A-S-A-K-I.
18 CHAIRMAN MCDADE: Okay. Thank you, Mr.
19 Sipos.
20 ADMIN. JUDGE WARDWELL: And thank you, Dr.
21 Lahey, for digging out those specific ones. We only 22 really needed the cites though. Sorry for the 23 misunderstanding, we appreciate your effort.
24 DR. LAHEY: Okay.
25 ADMIN. JUDGE WARDWELL: Entergy, is there NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5009 1 anything you'd like to offer in regards to the data 2 that demonstrates the effect of, I believe it was 3 looking at the effects of embrittlement on fatigue in 4 some fashion or fracture toughness.
5 DR. LOTT: Yes. My understanding was there 6 were two questions that came up yesterday related to 7 data. That being one of them, the irradiation effects 8 on fatigue. And there was another question about high 9 fluence properties in general and the survivability of 10 materials at high fluences. As to the question of the 11 effects of radiation on fatigue, I must admit, I've 12 been reliant on NRC NUREG/CR-6909, which in Section 13 1.32 has a discussion of irradiation embrittlement and 14 fatigue.
15 Let me say that I have a difficult time 16 with the word synergism when we talk about these 17 particular relationships. I understand that all 18 irradiated materials are embrittled and the question 19 to me is whether or not irradiation also has effect on 20 the strain life of the material or on the fatigue 21 resistance of the material. I don't see that as a 22 synergism, that's just a question of irradiation 23 embrittlement and irradiation fatigue life.
24 I can give you a summary of what I think 25 it says in NRC 6909. It does reference the papers by NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5010 1 Korth and Harper. I'll point out that, that data in 2 that paper is basically generated for fast meter 3 conditions, it's not at a temperature that we would 4 generally think would be applicable for PWR 5 interactions. The paper by Arai was actually 6 reporting on irradiation embrittlement work that was 7 done at Westinghouse. I'm not sure that there's any 8 fatigue data in that paper at all, I don't recall any.
9 And the paper by Kanasaki, I'll point out 10 I'm a co-author on that paper, so we're certainly 11 aware of it. It is the one paper that, and we can put 12 the data -- do you want to discuss the data or just 13 note that the data is there? That is the paper that 14 does show that at low strain amplitudes, below I 15 believe it's 0.6, the data that was tested of PWR 16 relevant conditions all showed an increase in fatigue 17 life with irradiation. We're not to discussing the 18 CUFs in Contention 26 yet, but when we do, I think 19 we'll find that there's, A, a limited number of 20 materials of the internals which have CUF values 21 calculated are irradiated. And those materials, I do 22 not expect to have a large strain amplitude. So I 23 think the data in the Kanasaki paper is directly 24 relevant to the conditions we're talking about.
25 ADMIN. JUDGE WARDWELL: Okay. When you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5011 1 first started off referencing a NUREG was it?
2 DR. LOTT: It's CR-6909 and I'm sorry, I'll 3 find you the --
4 ADMIN. JUDGE WARDWELL: And is there an 5 exhibit number. If you just --
6 DR. LOTT: Yes, there is.
7 ADMIN. JUDGE WARDWELL: -- have an exhibit 8 number, that would suffice because then we can find it 9 through the --
10 MR. STEVENS: Your honor, Jerry Stevens of 11 the Staff. That's New York State 490 Alpha.
12 ADMIN. JUDGE WARDWELL: Okay.
13 DR. LOTT: Yes.
14 ADMIN. JUDGE WARDWELL: Thank you.
15 DR. LOTT: Of which Mr. Stevens is an 16 author.
17 ADMIN. JUDGE WARDWELL: Okay. Is there 18 anything else you'd like to offer in regards to 19 sources of data backing your claims from yesterday?
20 DR. LOTT: Well, there is, in addition, the 21 issue of high fluence properties. There was some 22 question in our mind whether the question was high 23 fluence data on fracture toughness or high fluence 24 data on the yield stress of materials. My observation 25 again would be that, when we come to the very highly NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5012 1 irradiated components in the plant, particularly the 2 baffle former bolts, and you look at strategies for 3 dealing with the baffle former bolts, anything that we 4 observe with a crack, we assume has failed. So we 5 never actually use a fracture toughness value for very 6 highly irradiated materials because we have this 7 failure assumption in our acceptance criteria.
8 So if you want to know about those 9 materials at very high fluences, I would recommend to 10 you and I'm going to find it here, it's Exhibit 11 Entergy 000646, MRP 210, has in there in Figure 1-4 a 12 plot of yield strength versus neutron fluence that 13 goes out to fluences data certainly greater than 90 14 dpa, actually one data point closer to 120 some dpa.
15 Which would be well beyond what we would expect the 16 fluence on the vast majority of the reactor internals 17 ever to see.
18 ADMIN. JUDGE WARDWELL: Thank you. And 19 again, is there a -- did you reference an exhibit 20 number in your --
21 DR. LOTT: Yes, I believe I did. ENT 646.
22 ADMIN. JUDGE WARDWELL: Oh, ENT, okay.
23 Thank you.
24 MS. BRANCATO: Your honors, this is Debra 25 Brancato from Riverkeeper. Dr. Hopenfeld has NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5013 1 indicated to me he has some helpful input to this 2 discussion as well if you care to entertain some 3 testimony from him at this time.
4 CHAIRMAN MCDADE: Well, probably not at 5 this time, but before we leave the topic.
6 MS. BRANCATO: Okay.
7 ADMIN. JUDGE WARDWELL: Okay. So let's 8 continue on with some questioning and dealing with 9 Contention 25 --
10 CHAIRMAN MCDADE: But before you go on, Dr.
11 Hopenfeld is, as I said yesterday, sometimes things 12 move and we hear an awful lot of things. Take a card, 13 write down the comment to remind yourself, and you 14 will be testifying later. Thank you.
15 ADMIN. JUDGE WARDWELL: And I'll refer 16 again to Entergy's Exhibit 616, their testimony, 17 Answer 174, Pages 113 to 114. During its technical 18 review of MRP 227, the NRC Staff specifically 19 requested additional information on how the program 20 accounts for synergistic effects. And I guess I'll 21 ask again for Entergy, how does your AMP look at and 22 include actual analyses which have addressed the 23 change in fatigue strength as a function of varying 24 degrees of embrittlement of the specimen that occurs 25 with time?
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5014 1 DR. LOTT: Well, again, to me there are two 2 questions in mind there. One is when we talk about 3 embrittlement, if we're talking about the loss of 4 fracture toughness, we calculate or look at lower 5 bound fracture toughness curves based on the dose of 6 the material. So we're using what we believe to be 7 bounding fracture toughness curves and then analyze 8 the crack that has grown. So there's certainly an 9 interaction there between the tolerance of the 10 material and the embrittlement and the fatigue crack 11 growth, they're tied into the same calculation.
12 ADMIN. JUDGE WARDWELL: But that's a 13 proposal of what you would do if a crack was observed, 14 right? And until a crack is observed and as part of 15 the development of your AMP, you haven't done any 16 analysis in regards to that, is that correct?
17 DR. LOTT: Well, we've certainly done 18 analysis to show that we do not expect to see cracks 19 occur in these materials whatsoever. Again, the CUF 20 would be a good example of that. We've looked at 21 extensive analysis of IASCC, particularly in the 22 highly rated components we generated a fairly complex 23 model of the reactor internals to look at the aging of 24 those reactor internals and predict where IASCC might 25 occur. The only place that we actually predicted it NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5015 1 would occur was in the baffle former bolts, which we 2 discussed. There's a fairly detailed process for 3 that.
4 In our processing and recommendations for 5 inspections, we certainly looked at susceptibility to 6 multiple degradation mechanisms as part of that and we 7 based our inspection program on determination of the 8 effect of those mechanisms. So as far as we're 9 concerned, we're obviously inspecting for irradiation 10 fatigue, IASCC, or SCC, all three of which are 11 cracking mechanisms and the inspection program doesn't 12 care which one of those caused that, it's just simply 13 looking for that. So I think in our prioritization of 14 inspections, we certainly have looked at that. We've 15 looked at that in the design of our program in 16 general.
17 ADMIN. JUDGE WARDWELL: I'm just trying to 18 think of whether I'll wait until 26 to discuss this 19 further or not, but I'll bring it up again now, I 20 guess, because we did talk about it yesterday and I 21 want to fix again, I guess with you, Dr. Lott. I'm 22 trying to get a grasp on the handle between the peak 23 strength and fatigue durability. As I heard our 24 discussion yesterday, I can understand why someone may 25 say embrittlement won't come into effect really until NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5016 1 there's a crack due to excessive peak strength.
2 You've got to get past the peak strength and then 3 embrittlement comes into play to a certain degree.
4 But fatigue durability doesn't rely on loads that 5 exceed the peak strength. It's a repetitive loading 6 that causes the types of cracks, is that correct?
7 DR. LOTT: Yes.
8 ADMIN. JUDGE WARDWELL: And so, again, I'm 9 back to the question of where is anyone looked at 10 whether or not that fatigue durability is influenced 11 by embrittlement and to what degree is it?
12 DR. LOTT: Well, again, this is, at least 13 from my perspective, I don't necessarily -- to me, 14 embrittlement and fatigue life, well, again, 15 embrittlement is not a property in and of itself. The 16 properties are yield stress, ultimate stress, 17 ductility measurements, such as total elongation, 18 fracture toughness. And I would see, again, as 19 another issue, would be the question of what's the 20 impact on the S-N curve, the number of --
21 ADMIN. JUDGE WARDWELL: Well, can't we use 22 fracture toughness as an indication of embrittlement?
23 DR. LOTT: It's an indication that the 24 material has been irradiated and irradiated materials 25 have a decrease in that. But it does not necessarily NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5017 1 indicate how the material will survive under fatigue 2 conditions. So basically, we have a test for that, 3 just like we have a test to determine yield stress.
4 That is the fatigue life test to generation of the S-N 5 curve. We actually take a specimen, strain it at some 6 strain amplitude, we put it on repeated cycles, and 7 count the number of the cycles it takes to fail the 8 specimen or to have a load drop in the specimen 9 actually. So those fatigue life curves, again, are 10 behind basically the calculation of the CUF factors 11 based on a design curve, which is bounding to all of 12 the measurements of fatigue life.
13 ADMIN. JUDGE WARDWELL: Let me ask you 14 this, if I did a fatigue test, whatever the fatigue 15 test might be, if I try to fatigue it, as a specimen 16 that's not irradiated whatsoever, and I repeated that 17 test under different degrees of radiation exposure, 18 what would you expect the results of the fatigue test 19 to do? Remain the same, improve, or degrade?
20 DR. LOTT: That's exactly what we did in 21 the Kanasaki paper that, again, I'm a co-author on.
22 ADMIN. JUDGE WARDWELL: Okay.
23 DR. LOTT: We looked at that. We looked at 24 that at some fairly high fluences. So, if you'll --
25 the fluences are in two different papers reported in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5018 1 different units. If you'll take my word for it, in 2 the Kanasaki paper the data is about 20 dpa. At 20 3 dpa, if you look at the yield stress or the fracture 4 toughness curves, you'll see that there's a sharp 5 decrease in the fracture toughness of the material, 6 there's a large increase in the yield stress and, yet, 7 the fatigue life in those specimens that we tested for 8 Kanasaki, the fatigue life of that material got 9 longer. So the fatigue life improved at the same time 10 that the fluences were such that the yield stress 11 would increase and the fracture toughness decreased.
12 ADMIN. JUDGE WARDWELL: Thank you. Dr.
13 Lahey, do you have any other types of cites or 14 evidence in your testimony that differs from what they 15 have just expounded upon and/or what would you believe 16 would be the change in that fatigue property as the 17 materials are irradiated?
18 DR. LAHEY: Right. This is Richard Lahey 19 for New York State. I don't disagree with anything he 20 said except he didn't say everything. If you look at 21 the Korth paper, which admittedly was for high 22 pressure or higher temperature conditions for the 23 reactor, what it says is for high cycle fatigue where 24 you have larger amplitudes, you can have a reduction 25 by a factor of one half of the cycles to failure for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5019 1 irradiated materials, irradiated up to 1.1 times ten 2 to the 22 neutrons per centimeter square.
3 It is inconclusive. The reason people say 4 it's inconclusive is there's other data that shows if 5 you have high cycle fatigue or have low amplitude, 6 then it strengthens and you're in the region where 7 things get better. Here, things get significantly 8 worse. So all the real experts, the people who do 9 research on this, say the same thing. We don't really 10 have any good data for light water reactor conditions 11 and it's sorely needed. And this is why in the 12 sustainability program they're doing those kind of 13 tests. So I don't disagree with that at all.
14 What I am concerned with is there's 15 evidence, admittedly it's not perfect, but there's 16 evidence it can have a significant degrading effect.
17 And then the question is, what do you do in the 18 meantime until you can definitively tell and quantify 19 the effect? Do you just ignore it and keep looking 20 until you get a crack? Or do you put on some sort of 21 factor to account -- put a cushion in to account --
22 ADMIN. JUDGE WARDWELL: In your cite that 23 you just gave us, would that temperature have a big 24 difference and is that temperature representative of 25 what might be experienced in a PWR at IPEC?
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5020 1 DR. LAHEY: It's somewhat higher than the 2 temperature that you would have in a light water 3 reactor for sure. Therefore, it's not a perfect data 4 set.
5 ADMIN. JUDGE WARDWELL: And would that 6 likely have an influence on the data results, the 7 extra temperature?
8 DR. LAHEY: It's hard to say.
9 ADMIN. JUDGE KENNEDY: Dr. Lahey, this is 10 Judge Kennedy. I'm curious now that I've heard from 11 Dr. Lott, we've got two conflicting views of the data, 12 one for slightly different conditions. It appears 13 that Dr. Lott's paper addressed the right conditions 14 and addressed different levels of irradiation. How 15 would you challenge his paper? You've offered up the 16 higher temperature data, but what would you say to the 17 data that he's presented? I mean, I recognize it's 18 one of the three references you provided.
19 DR. LAHEY: Right. Well, my understanding 20 -- I mean, he's the author, so --
21 ADMIN. JUDGE KENNEDY: But I think you're 22 --
23 DR. LAHEY: -- he has a little advantage 24 there, but --
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5021 1 offering his same paper for a different conclusion.
2 DR. LAHEY: Yes. I mean, it is one of the 3 few papers that addressed fatigue and irradiation.
4 And as I understood the focus of his paper was more 5 into what happens once you get a crack and how does it 6 propagate and the initiation of the crack, rather than 7 the cycles to failure.
8 ADMIN. JUDGE KENNEDY: But isn't that what 9 he just said? That the number of cycles to failure 10 increases with the irradiation of the sample material?
11 DR. LAHEY: Well, that's what --
12 ADMIN. JUDGE KENNEDY: That the fatigue 13 life went up?
14 DR. LAHEY: It depends entirely on the 15 amplitude. It depends on the amplitude of the fatigue 16 cycle.
17 ADMIN. JUDGE KENNEDY: Are you suggesting 18 that the amplitude in his test data or in his 19 calculations using the test data are not the right 20 characterization of the amplitudes that would be 21 present at Indian Point Units 2 or 3?
22 DR. LAHEY: I mean, I don't -- I mean, he 23 should say what the purpose of his test was. But as 24 I understand it, it wasn't to specifically address 25 that.
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5022 1 ADMIN. JUDGE KENNEDY: And maybe this is a 2 more general question about the term high cycle/low 3 amplitude and low cycle/high amplitude has been pushed 4 around here. I'm not sure sitting up here, are both 5 applicable to the operating conditions at Indian 6 Point? Is one grouping, I don't know if it was -- is 7 high cycle/low amplitude more applicable to Indian 8 Point? Or is low cycle/high amplitude more 9 applicable? Or are they both? And is there a hole in 10 the data as Dr. Lahey is suggesting?
11 DR. LAHEY: Well, as I under -- this is 12 Richard Lahey again. As I understand high cycle, it's 13 things like flow induced vibration, turbulence 14 induced, flow induced vibration. This is not a great 15 concern. It is in the steam generators and what that 16 might do to fretting and things like that, but it's 17 not in the primary side. It's more low cycle and many 18 of the kind of transients they have would in my 19 opinion give a larger amplitude. So it's more like 20 low cycle, larger amplitude fatigue.
21 CHAIRMAN MCDADE: Okay. Dr. Lahey, let me 22 interrupt for a second here. The document that you 23 referenced was the Korth paper, K-O-R-T-H.
24 DR. LAHEY: Right.
25 CHAIRMAN MCDADE: And that was Riverkeeper NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5023 1 152. That paper was originally presented back in June 2 of 1974. Is that data still valid? I mean, isn't 3 there something more recent that you can address us to 4 as far -- you're talking about the absence of data 5 here and that particular study is more than 40 years 6 old.
7 DR. LAHEY: Well, that's correct. But 8 nobody that we could find in a literature search has 9 really systematically done that and as a consequence 10 that's why they took it on in the Light Water Reactor 11 Sustainability Program. I agree there's data needed.
12 CHAIRMAN MCDADE: Okay. So that particular 13 study, although it's more than 40 years old, it's your 14 position that since then there has been no significant 15 work that has generated more informative data?
16 DR. LAHEY: I haven't been able to find it 17 if there has been. And I don't know anybody else that 18 has. All the -- this has been discussed by the NRC 19 because they had input, like I'm giving you, from some 20 of their experts at Argonne and they took all that 21 input in, looked at this data, the other data, and 22 then decided it's inconclusive and so we'll do our 23 inspection program, but not do anything separate in 24 terms of putting a penalty factor in. But that's 25 their job, in my view, as regulators. That's their NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5024 1 opinion.
2 ADMIN. JUDGE KENNEDY: I guess -- this is 3 Judge Kennedy. I'm still curious about low cycle/high 4 amplitude and high cycle/low amplitude. Is this worth 5 discussing in more detail? Is it relevant to the 6 metal fatigue for reactor vessel internals?
7 DR. LOTT: Well, I think probably what's 8 relevant to this particular discussion is the limits 9 that are in the Kanasaki paper in general, which is 10 this 0.6 percent strain amplitude. We believe those 11 conditions were chosen such that they would be 12 relevant to reactor internals, that was the intent of 13 the testing in the first place. And, again, as I 14 indicated, did a quick check of the CUF values that 15 are reported for the reactor internals and those that 16 are in components that also see irradiation, which is 17 only a fraction of the total. And having surveyed 18 those, I believe that we'll find that all of the 19 strain amplitudes are within the limits that are in 20 the Kanasaki paper. If you look at the number of 21 cycles to failure in Figure 8 of that paper, you'll 22 see that in some cases they are as few as 1,000 23 cycles. So it's not like this is thousands of cycles 24 per year kind of numbers that we're talking about.
25 ADMIN. JUDGE KENNEDY: All right. Thank NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5025 1 you.
2 ADMIN. JUDGE WARDWELL: Moving on from 3 that, Entergy's testimony, Exhibit 616, Answer 144, 4 Pages 93 to 94, present eight age relating degradation 5 mechanisms. And they include stress corrosion 6 cracking, irradiation assisted stress corrosion 7 cracking, wear, fatigue, thermal aging embrittlement, 8 irradiation embrittlement, void swelling and 9 irradiation growth, thermal and irradiation enhanced 10 stress relaxation, or irradiation enhanced creep.
11 They then go on to say that for each of 12 these eight mechanism, MRP 227 identifies the 13 resulting aging effect, which will then be managed 14 through inspections under MRP 227-A guidelines.
15 Notably, in most cases, the key effects are cracking, 16 dimensional changes, or wear, but in all cases, as 17 explained below, the inspections specified in MRP 227-18 A are designed to detect potential aging effects 19 applicable to each reactor vessel internal component 20 regardless of the underlying mechanism.
21 And I guess I'd address this to Entergy, 22 whoever would like to answer, if you say that in most 23 cases the key effects are cracking, dimensional 24 changes, or wear, but given that the effects are 25 managed, not the mechanism, what are the key effects NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5026 1 in the other minority cases and are they detectable by 2 your AMP? I'm addressing your throw away statement in 3 this testimony that in most cases the key effects are 4 cracking, dimensional changes, et cetera. What I'm 5 asking is, what about in those other cases that aren't 6 included in that, what are those key effects because 7 it is true you are claiming to be monitoring for 8 effects and not mechanisms?
9 DR. LOTT: I may have to take a minute to 10 think about this one. I suspect that, that is, as you 11 suggested, a throw away sentence that we were probably 12 just overly cautious. I don't --
13 MR. DOLANSKY: Dr. Lott?
14 DR. LOTT: Yes.
15 MR. DOLANSKY: This is Bob Dolansky with 16 Entergy. Perhaps one example would be the internals 17 hold down spring.
18 DR. LOTT: Okay.
19 MR. DOLANSKY: We're actually taking 20 measurements of the hold down spring, we're not 21 looking for cracking or wear. We're going to take 22 dimensional measurements of the hold down spring and 23 use that to determine whether that is acceptable. Is 24 that --
25 DR. LOTT: Yes.
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5027 1 MR. DOLANSKY: Would that be one example?
2 DR. LOTT: That would be a good idea.
3 ADMIN. JUDGE WARDWELL: Okay. So in that 4 particular case then for that spring, you're doing 5 some actual measurement rather than just observing for 6 a crack?
7 MR. DOLANSKY: Correct.
8 ADMIN. JUDGE WARDWELL: Do you have any 9 other examples of this?
10 MR. DOLANSKY: Give me one --
11 ADMIN. JUDGE WARDWELL: And let's just 12 assign that as a homework assignment so you don't 13 break out in beads of sweat panicking --
14 DR. LOTT: Right.
15 ADMIN. JUDGE WARDWELL: -- trying to find 16 everything you possibly can at this moment, I know the 17 feeling all too well being on that side of the table 18 often. Let's just -- when you feel comfortable, get 19 back with a response if there's any other examples you 20 can give in regards to those. Because I think it is 21 somewhat important because, again, you are monitoring 22 for the effects and I want to make sure we don't have 23 some giant hole of something else that's out there 24 that we would like to be able to track.
25 DR. LOTT: May I say that we did go through NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5028 1 this FMECA process which looked at each component and 2 asked ourselves what are the appropriate failure modes 3 and consequences? So we certainly did look at that 4 and asked ourselves the questions in the process.
5 ADMIN. JUDGE WARDWELL: So it sounds like 6 your reference to most cases, you were referring to in 7 most of the internals, we are looking for some type of 8 strain and deformation or a crack or something in that 9 neighborhood.
10 DR. LOTT: Yes. Well, we --
11 ADMIN. JUDGE WARDWELL: And in some of your 12 -- and then other cases, you are doing something else 13 besides looking for some sort of strain.
14 DR. LOTT: Yes. We tried to identify in 15 each case what the effect was that we thought we were 16 looking for in the prescribed inspection. So if it 17 was a wear inspection, we would be looking for loss of 18 material or evidence of wear on the surfaces. If it 19 was something that was subject to a cracking 20 mechanism, we would say, we're looking for cracks.
21 And this is where the question of irradiation 22 embrittlement would come up as well.
23 ADMIN. JUDGE WARDWELL: Yes. And I kind of 24 view wear as a strain, it's not really a strain, but 25 it's a loss of the dimension if nothing else.
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5029 1 DR. LOTT: Loss of --
2 ADMIN. JUDGE WARDWELL: A change in a 3 dimension.
4 DR. LOTT: Yes. Opening of a gap, 5 displacement of a component one with respect to the 6 other in some small way. We tried to find the places 7 where that was most evidence where we could see it.
8 ADMIN. JUDGE WARDWELL: Is it not true 9 though that your AMP is based on the fact of 10 monitoring for effects and not trying to deal with the 11 mechanism that caused those effects?
12 DR. LOTT: Yes. I think that was the 13 instruction of the Aging Management Program in 14 general.
15 ADMIN. JUDGE WARDWELL: And, NRC, would you 16 agree that, that's the motive behind the inspection 17 program for the reactor vessel internals AMP?
18 DR. HISER: This is Allen Hiser with the 19 NRC. Yes, that's correct. It's monitoring, 20 inspecting for aging effects. Mechanisms create 21 effects and they're important to understand what aging 22 effects you need to manage, but the mechanisms 23 themselves are not managed.
24 MR. STROSNIDER: This is Jack Strosnider 25 for Entergy. If I could just add there, if you want NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5030 1 a citation on that, you can go to the Statements of 2 Consideration that were issued when the License 3 Renewal Part 54 was issued in 1995. And it explicitly 4 discusses this notion of managing effects rather than 5 mechanisms. And I think part of the reason for that 6 is to recognize that the effect doesn't care what 7 synergisms, if there are any, that are happening. If 8 you see a crack, you see a crack and whatever 9 contributed to it, contributed to it and then you need 10 to take the right corrective actions. But I just 11 wanted to get you that citation in case you want to 12 look at that.
13 ADMIN. JUDGE WARDWELL: How does one 14 observe embrittlement?
15 MR. STROSNIDER: This is Jack Strosnider of 16 Entergy. You asked -- is that question directed at 17 me?
18 ADMIN. JUDGE WARDWELL: Yes.
19 MR. STROSNIDER: Okay. So the discussion 20 in the AMP is that embrittlement is not directly 21 observed, but it is managed through the detection of 22 cracks. And, as we've been discussing, if you find a 23 crack, you then need to assess it considering the 24 material properties that would be associated with 25 whatever level of embrittlement that component has NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5031 1 experienced. So it's an indirect approach. There is 2 not an embrittlement meter, if you will, that you can 3 go in and look. You have to do this indirectly and 4 that's what's laid out in the program.
5 ADMIN. JUDGE WARDWELL: Thank you.
6 DR. LOTT: May I add? This is Randy Lott.
7 In MRP 227, when there were components that were 8 subject to irradiation or thermal embrittlement, we 9 tried to note them in that way. There was effects 10 listed and then there would be the conditional note 11 that says, this effect should be -- aging management 12 for irradiation embrittlement, thermal embrittlement 13 would be required in this component. So we would mark 14 those places where this would be a concern.
15 ADMIN. JUDGE WARDWELL: Okay. Thank you.
16 Dr. Lahey --
17 CHAIRMAN MCDADE: Okay. If I could, just 18 -- and, Dr. Lahey, as I understand your position is 19 that given the fact that you can't directly observe or 20 monitor embrittlement, that there is a significant 21 change, even in the absence of cracks, that could set 22 up the reactor vessel internal for failure in the 23 event of a shock load.
24 DR. LAHEY: The reactor vessel internals, 25 is that what you're talking about? Yes, I agree with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5032 1 what they said that, that's what they're monitoring.
2 They're not looking at incipient failure, they're 3 looking at failures that have occurred. And I'm 4 concerned with the mechanism of microcracks, there are 5 plenty of cracks, but they're microcracks, and how 6 this weakens the material in the event that you have 7 an impulsive load on this material.
8 CHAIRMAN MCDADE: Okay. And it's your view 9 that there's no specific inspection technique 10 currently available that would be able to identify the 11 effects of embrittlement prior to failure and that, in 12 your view, the only way to adequately manage the 13 effect of aging is to have a reasonable replacement or 14 repair system. Am I correct in summarizing --
15 DR. LAHEY: Yes. I think you are correct.
16 I mean, the problem is, if we had all the data that we 17 really need to have, we wouldn't be having this 18 discussion. We would know what to do, the NRC would 19 be requiring it, Entergy would be complying with it, 20 and everything would be fine. But we don't. We have 21 some fragmentary data, which indicates concerns, and 22 so how do you deal with that? To me, the easiest way 23 to deal with it, for things like bolts, they're 24 relatively easy to replace, would be just replace 25 them. Get rid of the problem rather than try to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5033 1 calculate how many bolts can be failed and still keep 2 running and that sort of thing. I think that's a very 3 dangerous game, you're playing with fire when you do 4 that.
5 CHAIRMAN MCDADE: Okay. And let me address 6 this in turn to Dr. Hiser and then to Dr. Lott, it's 7 your view, Dr. Hiser, that in the absence of cracking, 8 there is no reason to believe that you are on the 9 verge of failure, even in the event of a shock load, 10 with these reactor vessel internals and that, 11 therefore, the inspections that are currently 12 available are adequate to monitor the aging of these 13 reactor vessel internals? Is that correct?
14 DR. HISER: This is Allen Hiser. Yes, that 15 is correct.
16 CHAIRMAN MCDADE: Okay. And, Dr. Lott, do 17 you agree with what Dr. Hiser just affirmed?
18 DR. LOTT: Yes, I do.
19 CHAIRMAN MCDADE: Okay. So that's 20 basically the difference of opinion here between the 21 NRC Staff, Entergy, and the position of New York's 22 witness, Dr. Lahey, as you see it Dr. Hiser?
23 DR. HISER: If I could just -- this is 24 Allen Hiser. If I could just elaborate a little bit 25 because --
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5034 1 CHAIRMAN MCDADE: Please.
2 DR. HISER: -- some of the discussion 3 yesterday related to ductility of the materials at 4 high fluence and a couple of the exhibits, New York 5 487 shows data, they're up to about 70 dpa that have 6 measurable fracture toughness, which is indicative of 7 ductility, NRC 209 has data that are up to about 12 8 dpa, that again show reasonable fracture toughness.
9 So there is still ductility in the material at these 10 fluence areas of interest. And, again, there also is, 11 my understanding, there was no exhibits provided to, 12 no data that we've seen that would indicate that 13 fatigue weakens a material or a component in the 14 absence of cracks.
15 ADMIN. JUDGE WARDWELL: And what types of 16 fluences do we expect after 60 years of operation?
17 DR. HISER: My understanding is for the 18 internals on the upwards of 75 dpa for the maximum for 19 the baffle bolts. If you go beyond that area, go 20 beyond the baffle assembly, they're much lower. So 21 the baffle bolts --
22 ADMIN. JUDGE WARDWELL: But the ones you 23 just quoted were well below that, weren't they? You 24 say it was over --
25 DR. HISER: The data in New York State 487 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5035 1 is upwards of 60, 70 dpa.
2 ADMIN. JUDGE WARDWELL: Okay. Thank you.
3 CHAIRMAN MCDADE: Okay. And, Dr. Hiser, 4 we're talking perhaps about different things. We've 5 been talking about cracks. Dr. Lahey mentioned 6 microcracks, cracks that are there, but are not 7 observable given current inspection techniques. Do 8 you agree that there's probably microcracks in many of 9 these reactor vessel internals, such as the baffle 10 former bolts?
11 DR. HISER: This is Allen Hiser. As the 12 CUF gets much closer to one, I think the likelihood 13 increases that you could have microcracks. But I 14 think the impact of those microcracks on the fracture 15 response of the component is negligible. I think 16 that's been demonstrated through many tests.
17 CHAIRMAN MCDADE: So the degradation in 18 fracture toughness would be minimal in your view?
19 DR. HISER: Well, I think the effect of the 20 degradation of fracture toughness would not be 21 significant. There may be reductions in fracture 22 toughness, but the impact of that in the presence of 23 even microcracks is not significant.
24 CHAIRMAN MCDADE: In your view, would not 25 be of consequence?
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5036 1 DR. HISER: Yes, that's correct.
2 CHAIRMAN MCDADE: Dr. Lott, do you agree 3 with that?
4 DR. LOTT: Yes, and I believe that the data 5 that Dr. Hiser just cited basically shows you that 6 these components can survive with actual cracks, 7 macrocracks, measurable cracks, not just the 8 microcracks that are suggested here. Certainly if we 9 can withstand cracking and we can demonstrate 10 stability of the component with a crack in it, concern 11 about microcracks does not, to my view, seem to be 12 important.
13 CHAIRMAN MCDADE: Okay. And this is the 14 data in New York State 487 and NRC 209, correct, Dr.
15 Hiser?
16 DR. HISER: Yes, that's correct.
17 CHAIRMAN MCDADE: Okay. Dr. Lahey?
18 DR. LAHEY: All right. I want to try to 19 clear up something we talked about yesterday and here.
20 Dr. Kennedy brought up the question, have I invented 21 new loads, do I need new loads to show that these 22 things can fail or not? And my answer was, no, the 23 existing type of accidents and seismic events are 24 sufficient. What's happened, if you go back in 25 history, is there was a point in time when we were NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5037 1 real concerned about design basis LOCA event and the 2 decompression wave that went in and what that would do 3 to the structures. Because in a particular instant, 4 you can have atmospheric, essentially atmospheric 5 pressure on one side and several thousand PSI on the 6 other and that would give you a very large impulsive 7 load.
8 So codes were written to address that, 9 method of characteristic kind of codes, like the WHAM 10 code that Stan Fabic wrote in Westinghouse, and 11 detailed analysis was done that showed for ductile 12 structures, they can withstand it. And we confirmed 13 that in the Loft experiment, which was an experiment 14 which we ran a simulated loss of coolant accident.
15 And so that concern was mitigated and as a 16 consequence, now people do the analyses using codes 17 such as RELAP and TRAC, those kind of codes, and they 18 really smear out this type of shock. They don't give 19 you the kind of shock loads that you would get if you 20 tracked the wave, the rarefaction waves, throughout 21 the vessel.
22 So I'm concerned with the real shock 23 loads. It's time, in my view, to go back and take a 24 look at this again with degraded materials. These are 25 weakened materials, you can't have lots of microcracks NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5038 1 in there and not have it weakened. It depends on how 2 much weaker it is, that's just due to fatigue. And 3 then irradiation makes it weaker yet, makes it more 4 brittle, more subject to failure. So that's the real 5 difference. That's why when I say shock loads, I may 6 mean something quite different than what they're 7 talking about with shock loads because they're talking 8 about the normal safety analyses using these big 9 system codes, which really are intended to look at the 10 inventory of the liquid and the coolability and that 11 sort of stuff. They're not very good at giving 12 instantaneous loads, either the thermal or pressure 13 loads.
14 CHAIRMAN MCDADE: Okay. But according to 15 Dr. Hiser and Dr. Lot, even in the situation where you 16 have macrocracks, observable cracks as opposed --
17 there still would not be sufficient degradation in 18 order to create a real risk, a significant risk of 19 failure. And they rely on, I believe it was, what was 20 it, New York 487, which was an Argonne Lab study from 21 2010. Somewhat more recent data than the one that you 22 had cited. Does that not alleviate your fears that 23 this material would continue to be sufficiently 24 robust?
25 DR. LAHEY: That's not what the Argonne NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5039 1 experts are saying. When they give input, they're 2 saying very similar things to what I'm saying. They 3 have the same kind of concerns about the lack of data 4 and the effect of it and what it may imply. And 5 they're, of course, hoping to get funding to run more 6 experiments. And I agree, more experiments are 7 needed.
8 CHAIRMAN MCDADE: Okay. So you don't 9 disagree with the study, it's just what you take away 10 from that Argonne Laboratory study is different from 11 what Dr. Hiser and Dr. Lott take away from it.
12 DR. LAHEY: Apparently. I mean, I don't 13 agree with what he said. I think a degraded structure 14 is inherently weak and more subject to failure.
15 CHAIRMAN MCDADE: Okay. But what we're 16 dealing with here is basically a professional 17 disagreement. We're looking at the same data and 18 you're interpreting it, it creates more concern in 19 your mind as to the potential for failure than what 20 was expressed by Dr. Hiser and Dr. Lott. Is that 21 correct?
22 DR. LAHEY: I suppose.
23 CHAIRMAN MCDADE: Well, I'm not trying to 24 -- this is new material for me and I'm trying to make 25 sure that I understand the relative positions here and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5040 1 what the basis of the disagreement is. It seems to me 2 that it is an interpretation of the data, the 3 significance of the data that's available and the data 4 that isn't available. And that you seem to be very 5 concerned with an absence of data that leaves 6 questions in your mind. Am I correct?
7 DR. LAHEY: Well, not entirely. I'll tell 8 you why I feel the way I do is I'm on a science 9 council for a program called CASL, which is based in 10 Oak Ridge. It's a very large program, which is funded 11 by DoE, to develop computational capability for 12 nuclear reactors. It involves many national labs and 13 many universities. One of the members of the board 14 was also an executive or the person in charge of the 15 Light Water Reactor Sustainability Program, so I've 16 had ample opportunity to talk to the people who are 17 working in that program. And I know from the comments 18 that I have there's people who would like to think, 19 that's just for operating out beyond where we're 20 talking about. We're talking about 60 years, they're 21 going on to 80 years.
22 But in nobody's mind is there a sharp 23 demarcation. Those are the same concerns, they have 24 the same concerns that I do, and they're working on 25 it. And we're now looking at how to take the code NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5041 1 that we've developed under CASL, which is a three 2 dimensional neutronics, thermohydraulics, fuel, water 3 chemistry, crud deposition, everything, code, and 4 apply it to these kind of issues for relicensing of 5 nuclear reactors. So there are certainly things you 6 can do, like the water chemistry, what is the water 7 chemistry during transience and how does that affect 8 FN?
9 So these are things we'll do in the next 10 -- I think we're going to get to in the next issue.
11 But it's not -- I haven't just made this stuff up. I 12 mean, these are valid concerns by people who are 13 working in it, some of them are even working under NRC 14 funding. So, I think it's an honest professional 15 disagreement and nobody has the perfect data set right 16 now to say, here's the answer. But there's 17 significant concerns about these type of things.
18 CHAIRMAN MCDADE: Okay. Thank you, Dr.
19 Lahey.
20 ADMIN. JUDGE KENNEDY: Dr. Lahey, since you 21 responded to my design basis question -- this is Judge 22 Kennedy. I'm just curious, it seems like the scenario 23 that you painted in terms of analysis methodologies 24 would be just as applicable to non-irradiated material 25 shock loading analysis as irradiated material shock NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5042 1 loading analysis. Are you -- am I correct in that?
2 Do you have the same concern with non-irradiated 3 material analysis, shock loading, as you do with the 4 irradiated material shock loading analysis?
5 DR. LAHEY: I didn't explain it very well.
6 Originally, that was the concern and so they were more 7 ductile materials. These were newer reactors and so 8 the concern was, will they maintain a coolable 9 geometry during these type of events? And so, I think 10 that question has been settled and now the issue is, 11 given an intact geometry, can you cool them with 12 emergency core cooling engineered safety systems? The 13 new thing is, now you have degraded structures, which 14 you never had to deal with before, highly degraded 15 both from fatigue and irradiation, and it's time to 16 now relook at that. Because they can be, not only 17 deformed, they can be failed and relocated and then 18 the concern is core coolability.
19 ADMIN. JUDGE KENNEDY: I guess I thought I 20 understood you to say that there was a problem with 21 the analysis techniques that are being utilized to 22 analyze the shock loadings, that they were deficient.
23 Is that not the case? Is that not what you were 24 trying to say?
25 DR. LAHEY: The tools that were developed NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5043 1 and used previously were focused on tracking 2 depressurization fronts and their effect in terms of 3 loading. They're rather expensive to run method of 4 characteristic type codes, but they give you good 5 answers. Since then, people have concluded that, at 6 least for ductile geometry, the core will stay intact, 7 so given this geometry, the geometry is assumed, now 8 we can use codes that do not do that, they're more 9 control volume kind of codes, like RELAP and TRAC. So 10 when you do the analyses for plants like Indian Point, 11 you don't look at the deformation of the core, you 12 have a certain geometry. What you look at is where's 13 the liquid, how's the cooling, what's the peak clad 14 temperature, that sort of thing, to see if you're in 15 the compliance with the safety regulations.
16 ADMIN. JUDGE KENNEDY: All right. Thank 17 you.
18 ADMIN. JUDGE WARDWELL: Getting back to the 19 inspections where we were, do you have any criticisms, 20 Dr. Lahey, of their approach of trying to monitor for 21 effects rather than mechanisms? Strictly this issue, 22 not in regards to your overall issue, but as an expert 23 witness before us, do you agree that monitoring for 24 effects is a good approach rather than trying to 25 monitor for any mechanism?
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5044 1 DR. LAHEY: The answer to that, your honor 2 -- this is Richard Lahey for New York State. As I 3 tried to say yesterday, I think the program that they 4 have in place, which is inspection based, is a pretty 5 good program. So I'm all for it. I just don't think 6 it addresses all the real concerns.
7 ADMIN. JUDGE WARDWELL: You've said that, 8 thank you. In your testimony, 482, Page 17, Lines 6 9 through 17, you talk about the rather complex and 10 interacting metal degradation mechanisms associated 11 with fatigue, irradiation, and corrosion interact is 12 still an area of active research. And you point to a 13 DoE, USNR, in conjunction with various other national 14 laboratories that have recently embarked on a program 15 to understand and resolve issues related to these 16 interacting and synergistic effects. And you 17 reference NUREG/CR-7153, which is entitled Expanded 18 Materials Degradation Assessment, or EMDA --
19 DR. LAHEY: Right.
20 ADMIN. JUDGE WARDWELL: -- Aging of Core 21 Internals. And in that particular NUREG, do you know 22 the age of the study of how far out they were looking?
23 Was it within the license renewal period or was it 24 beyond that, to your knowledge?
25 DR. LAHEY: That report is part of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5045 1 Light Water Reactor Sustainability Program. So 2 they're looking beyond the design life of 40 years, 3 they're going out farther. They're actually charged 4 with seeing if you can get out to 80 years. But the 5 phenomena has no sharp demarcation at any point in 6 time. So they're looking at it all the way out.
7 ADMIN. JUDGE WARDWELL: But any conclusions 8 they reach, if they're making them in regards to how 9 it is out to 80, that would be certainly different 10 than if it was out to 60, wouldn't it?
11 DR. LAHEY: Yes, but if they find phenomena 12 that is of concern at 50, they will communicate that 13 and it will be relevant to what we're talking about 14 here.
15 ADMIN. JUDGE WARDWELL: And did they have 16 any conclusions from that report that related to out 17 to 50 or 60 years that demonstrate or support your 18 positions?
19 DR. LAHEY: Dr. Busby is one of the authors 20 of that report and one of the things that I recall 21 from that report was he was greatly concerned about 22 irradiated assisted stress corrosion cracking and what 23 the impact of that may be.
24 ADMIN. JUDGE WARDWELL: Okay, thank you.
25 You go on, in fact on that same page and extending NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5046 1 over to Page 18 with lines 17 through 22 and then 1 2 through 2 on the top, that the federal government has 3 also embarked on a fairly large research program. And 4 this is what you've termed -- or not you -- yes, you 5 call this Light Water Reactor Sustainability Program, 6 which includes research into whether the different 7 materials and light water reactor components can 8 continue to perform their intended functions during 9 the period of operations. This report that you cite, 10 which is Exhibit New York State 485, was in August of 11 2014. And, again, what were the periods of years that 12 they were looking at under that particular document?
13 DR. LAHEY: From now until operating out to 14 80 years.
15 ADMIN. JUDGE WARDWELL: Okay, thank you.
16 DR. LAHEY: I mean, it's being done at 17 various national labs and they put out monthly 18 newsletters on where they stand on various things.
19 ADMIN. JUDGE WARDWELL: Okay. Thank you.
20 NRC's testimony, 197, Answer 124, Page 75, states 21 that, "Entergy has also implemented a low leakage core 22 design for IP2 and IP3 prior to 30 calendar years of 23 operation, which reduces the potential for irradiation 24 driven aging mechanisms, such as the IASCC, the IE 25 void swelling, and ISR." I guess I'll go to Staff NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5047 1 considering this was your exhibit. I knew what all 2 those other acronyms were for, but what about the ISR?
3 MR. POEHLER: This is Jeffrey Poehler of 4 the Staff. So ISR is irradiation assisted stress 5 relaxation or irradiation stress relaxation.
6 ADMIN. JUDGE WARDWELL: Irradiation stress 7 relaxation, is that what the ISR is for?
8 MR. POEHLER: Correct.
9 ADMIN. JUDGE WARDWELL: Okay, thank you.
10 How is this low leakage core design achieved?
11 MR. POEHLER: This is Jeffrey Poehler of 12 the Staff. So, basically the core design is such that 13 you have fuel assemblies that have higher, I guess, 14 levels of burn up or depletion or placed around the 15 outside of the periphery of the core so that the newer 16 fuel assemblies are concentrated more towards the 17 middle of the core.
18 ADMIN. JUDGE WARDWELL: And what does this 19 do for you supposedly?
20 MR. POEHLER: It reduces the, I guess, the 21 leakage of -- it reduces the fluence levels.
22 ADMIN. JUDGE WARDWELL: That's what I was 23 wondering. Okay, thank you.
24 CHAIRMAN MCDADE: I'm sorry. It releases 25 the -- I just didn't hear.
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5048 1 ADMIN. JUDGE WARDWELL: It reduces the 2 fluence --
3 CHAIRMAN MCDADE: Fluence.
4 ADMIN. JUDGE WARDWELL -- or the flux 5 actually, so over time the fluence will be less.
6 ADMIN. JUDGE WARDWELL: Dr. Lahey, do you 7 have any comments on the low leakage core design?
8 DR. LAHEY: No. No, I understood what they 9 were doing and why. I mean, it's certainly helpful to 10 the concern we talked about yesterday with the core 11 plates and the pressure vessel to try to reduce the 12 fluence.
13 ADMIN. JUDGE WARDWELL: Okay. Thank you.
14 Entergy's Exhibit 616, testimony, Answer 182 on Page 15 120, failure of a component without a pre-existing 16 crack is governed by the mechanical properties of the 17 material, the yield strength, the ultimate strength in 18 particular. Irradiation increases the yield and 19 ultimate strengths, and we talked about this. I guess 20 my question for Entergy, does this increase in 21 strength with irradiation occur without any bound? I 22 mean, will it continue on forever, the longer you 23 irradiate it, will it continue to gain strength --
24 DR. LOTT: Okay. This is Randy --
25 ADMIN. JUDGE WARDWELL: -- keep on going NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5049 1 unlimited?
2 DR. LOTT: This is Randy Lott for Entergy.
3 I think it might be clearer if you examine the exhibit 4 I offered this morning about yield stress in material 5 versus fluence. In general, most of the changes that 6 happen to the mechanical properties of the material 7 happen within the first five to ten dpa. So they 8 actually happen early in life and saturate such that 9 there are much less changes. That's true both of the 10 increase in yield stress and the ultimate stress, the 11 decrease in ductility. Most of the action happens, I 12 would say, at less than ten dpa. Which, for your 13 highly irradiated -- ten dpa may be the end of life 14 fluence for some components, for others it's as much 15 as 60, for others it's one. It just depends on where 16 the component is.
17 ADMIN. JUDGE WARDWELL: And are you 18 referring to all components or just the reactor vessel 19 internals?
20 DR. LOTT: The internals. The internals 21 are the only ones that are going to see enough neutron 22 exposure to exceed one dpa. We talk about milli-dpa 23 when get out to the reactor pressure vessel.
24 ADMIN. JUDGE WARDWELL: Okay. Thank you.
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5050 1 are you claiming those still aren't caused by 2 irradiation or where would those microcracks be 3 occurring from?
4 DR. LOTT: The microcracks, as I understand 5 the argument, are one of the precursors to crack 6 formation in the fatigue specimens. So, as where it 7 was offered into evidence, and I think this is also 8 explained in NRC NUREG-6909, at least in the draft 9 version, they're sort of stages in the process of 10 forming an observable crack and one of those early 11 stages is the microcracking stage. So, again, yes, 12 it's a form of -- I mean, obviously, if there is a 13 limiting number of cycles, then something must be 14 changing over the course of time. It's been a 15 struggle to identify those things in these materials, 16 but this microcracking and some of these other small 17 micro-structural changes are just the evidence of 18 accumulating fatigue in material.
19 ADMIN. JUDGE WARDWELL: Okay, thank you.
20 Does irradiation decrease the resistance to the crack 21 propagation?
22 DR. LOTT: Okay. Let me ask, I want to 23 clarify here. When we talk about crack propagation, 24 are we talking about crack propagation due to loading 25 or crack propagation due to corrosion cracking or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5051 1 crack propagation due to fatigue? There are multiple 2 kinds of ways that, that term might be used.
3 ADMIN. JUDGE WARDWELL: Answer for each 4 situation.
5 DR. LOTT: Okay. So, I've created my own 6 question now, didn't I?
7 (Laughter.)
8 DR. LOTT: In terms of mechanical 9 properties, we talked and I think Dr. Hiser talked 10 earlier about the fracture mechanics mechanisms that 11 we look at or fracture toughness values that we use.
12 And in general, those toughness values are measured in 13 terms of J-resistance curves. That would tell you 14 effectively how much work it takes to advance a crack 15 in the material mechanically. And the slope of that 16 J-resistance curve gives you an idea of what the, 17 again, the resistance the material is to crack 18 advance. And in general, as the irradiation goes up, 19 that number goes down. So there is some decrease in 20 the resistance. The good news, however, is if you're 21 even measuring resistance, you're into ductile 22 failure, not into the brittle failures we had before.
23 Fatigue, I think we talked about the 24 effect of radiation on fatigue earlier and it's our 25 contention that where we have relevant data, that data NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5052 1 says that the resistance to fatigue initiation is 2 decreased. Fatigue propagation, I think, follows 3 similar kinds of rules, but it is a different 4 mechanism. I think we'd have to go back -- I'd have 5 to do a little more research if you wanted to know 6 about that.
7 ADMIN. JUDGE WARDWELL: But by different 8 rules, I mean, does irradiation decrease that 9 resistance to crack propagation? Does a crack 10 propagate faster when it's been irradiated under 11 fatigue cracking?
12 DR. LOTT: Not necessarily. I think we'll 13 get back to the same kind of data that we discussed in 14 terms of the initiation. And I'd have to go back and 15 review that data for you in detail. And, again, we 16 talked about -- so there's fatigue stress corrosion 17 cracking, again, that effectively is irradiation 18 assisted stress corrosion cracking, that's the concern 19 we have is that there will be crack formation and 20 growth due to irradiation.
21 ADMIN. JUDGE WARDWELL: Okay. Thank you.
22 Dr. Lahey, do you believe that irradiation decreases 23 the resistance to crack propagation once a crack forms 24 from whatever mechanism?
25 DR. LAHEY: Yes, definitely the crack will NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5053 1 propagate faster.
2 ADMIN. JUDGE WARDWELL: Thank you.
3 Entergy's Exhibit 616, Answer 114, Page 71, says RVIs 4 have no pressure retaining function. A pressurized 5 thermal shock, PTS, transient, therefore, does not 6 subject the RVI components to the sustained membrane 7 stresses characteristics of the effects of a PTS event 8 on a reactor pressure vessel itself. I guess my 9 question to Entergy is, why wouldn't the internals 10 still feel the pressure wave from a PTS transient if 11 one occurred?
12 DR. LOTT: Well, and again, I'm not sure if 13 someone else from Entergy panel wants to -- I'll step 14 forward first, I guess, and they can help me out as I 15 go along. Effectively, the pressurized thermal shock 16 is a repressurization, it's not necessarily -- it's a 17 long time in developing. It's not, I don't believe, 18 the same kind of process we're talking about here.
19 And one of the key elements of it is that you've 20 cooled down the vessel at a time when the internal 21 pressure on the vessel, the membrane stresses, remain 22 high because you've not seen the depressurization of 23 the system. So, the pressure itself in the vessel 24 creates the baseline high loads that the thermal shock 25 challenges. I'm not sure I've -- perhaps somebody can NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5054 1 help me out with my words.
2 MR. AZEVEDO: Yes, this is Nelson Azevedo 3 from Entergy. What Dr. Lott said was correct, the 4 repressurization during a PTS event is not an 5 instantaneous event like a pipe break, we have a 6 fraction wave traveling through the system. This is 7 a repressurization that takes some time to 8 repressurize the system.
9 ADMIN. JUDGE WARDWELL: But even so, 10 wouldn't the vessel internals feel however small 11 gradual change there is regardless? And I'm trying to 12 understand your statement that, as I assume you're 13 trying to imply on this Answer 114, is that we don't 14 have to worry about PTS because it's not a pressure 15 retaining -- these aren't pressure retaining 16 components, but they're in and amongst the pressurized 17 area and so why wouldn't it still feel that change, 18 whatever shock that does occur?
19 MR. AZEVEDO: This is Nelson Azevedo for 20 Entergy. And, yes, that's correct, but the way I 21 visualize at least, for a pressure boundary component, 22 like the reactor vessel, you have 2,200 pounds on the 23 inside and essentially zero on the outside. So you 24 have that whole delta P across the component. When 25 you're talking about a reactor vessel internal, like NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5055 1 a column, it's pressurized the same amount all the way 2 around, so there's no pressure differential. That's 3 why they're differentiated between pressure boundary 4 components and vessel internals.
5 ADMIN. JUDGE WARDWELL: Thank you, that's 6 helpful. And, Dr. Lahey, do you agree with those 7 statements? Any disagreement?
8 DR. LAHEY: I was somewhat surprised when 9 I got comments on my concerns about thermal shock, 10 significant thermal shocks, because I was not worried 11 about pressurized thermal shocks, that's a pressure 12 vessel phenomena. I was worried about cold water, for 13 example, coming in and some of the internal structures 14 would then be suddenly changed in temperature, it 15 would hit, it would shock the surface, it would try to 16 contract, it would crack, it could fail. That was my 17 concern.
18 ADMIN. JUDGE WARDWELL: Sure. And I just 19 want to verify that you don't have any disagreement 20 with the pressurized thermal shock specifically that 21 we just discussed now.
22 DR. LAHEY: No, I have no disagreement with 23 it.
24 ADMIN. JUDGE WARDWELL: Okay, great.
25 DR. LAHEY: I wasn't concerned with it.
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5056 1 ADMIN. JUDGE WARDWELL: Just want to make 2 sure they're not trying to spoof me over here with 3 some voodoo.
4 DR. LAHEY: No.
5 ADMIN. JUDGE WARDWELL: That's why I'm 6 going to you to see if you can agree with those small 7 points. That's why I go back and forth. But I 8 understand your comments about the other thermal 9 shock.
10 DR. LAHEY: Okay.
11 ADMIN. JUDGE WARDWELL: That was a 12 different issue. Thank you. Entergy's Exhibit 616, 13 Question and Answer 185, Page 122. The Question 185 14 says, do accident loads need to be considered as a 15 contributor to the effects of aging on reactor vessel 16 internals? And the Answer in 185 says, no, aging is 17 a gradual, long-term degradation of a component 18 resulting from sustained environmental conditions, 19 that is applied loads and residual stresses. And then 20 goes on to talk about the ASME codes and Staff 21 Guidance.
22 While I can understand why the loads 23 aren't a contributor to aging, I just want to make 24 sure that it's clear that I haven't confused this by 25 saying that still those loads, those design basis NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5057 1 loads, are considered in your AMP in regards to any 2 evaluation that you might be doing to look at any 3 changes in strength or whatever else occurs during the 4 aging process. Is that correct? And I'll address 5 this to Entergy, I guess, because it was your 6 testimony.
7 DR. LOTT: This is Randy Lott for Entergy.
8 Yes, what we were trying to say, I think, is exactly 9 what you have said that it does not contribute to the 10 condition that you would observe at any time of the 11 components, but if we're looking at the ability of the 12 components to withstand an accident mode, that's a 13 different question.
14 ADMIN. JUDGE WARDWELL: You're still --
15 DR. LOTT: It's not defined as the aging.
16 ADMIN. JUDGE WARDWELL: You still will 17 consider those loads --
18 DR. LOTT: Right.
19 ADMIN. JUDGE WARDWELL: -- design basis 20 loads. And aren't the design basis loads -- do not 21 the design basis loads also include LOCAs?
22 DR. LOTT: Yes.
23 ADMIN. JUDGE WARDWELL: Thank you.
24 Entergy's testimony, 616, Answer 185, Page 122, the 25 ASME Code Section 3 compares accident loads such as NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5058 1 large break LOCAs and large main steam line breaks to 2 the stress allowables to ensure that the affected 3 components remain capable of performing their intended 4 safety function during and after the event. And so I 5 want to just confirm again with you that doesn't the 6 degraded age strength of the PEO determine these 7 stress allowables in your analysis?
8 DR. LOTT: I -- go ahead.
9 MR. AZEVEDO: Yes, this is Nelson Azevedo 10 for Entergy. The stress allowables are directly 11 obtained from the ASME code. So based on what 12 material the component is made out of, the ASME code 13 specifies what the allowables are, we don't get to 14 choose those.
15 DR. LOTT: Yes, and I think that what 16 you're talking about is basically the design section 17 of the code, right? So it's --
18 MR. AZEVEDO: Right. So this would be 19 discussed in ASME Section 3, the original analysis, 20 and the analysis has been revised since then.
21 ADMIN. JUDGE WARDWELL: So isn't it taking 22 advantage of the full virginal strength of this 23 material if it's part of your -- or, I guess not, if 24 it was an allowable.
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5059 1 for Entergy. You're right. The original Section 3 2 design process uses ASME minimum yield strength and 3 code allowables in evaluating the margins that are 4 intended to be there. What we've described 5 previously, those tensile strengths and yield 6 strengths may actually increase due to the irradiation 7 and aging process. That's not taken into account, you 8 still use the code required properties of the virgin 9 material when you do those analyses. They're still 10 valid, in fact.
11 ADMIN. JUDGE WARDWELL: But if you observed 12 a crack and now you're trying to evaluate where you go 13 from there, does this statement not apply to those 14 types of analyses that you may or may not perform as 15 part of your corrective measure for the observation of 16 that crack?
17 MR. AZEVEDO: Yes, this is Nelson Azevedo 18 for Entergy. The design process, the ASME Section 3 19 specifically that we're talking about right now, does 20 not allow cracks. So cracks are not allowed during 21 the design phase. If you find cracks, then you 22 evaluate them either under ASME Section 11, which is 23 the operating version of the code, or another NRC 24 approved methodology. But the original design that 25 we're talking about, stress calculations do not allow NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5060 1 cracks.
2 ADMIN. JUDGE WARDWELL: Thank you, that's 3 helpful. Let's move now on to talk a little bit more 4 about these inspections. Yes, Dr. Lahey, would you 5 like to -- I hadn't gotten back to you after telling 6 you why I was getting back to you all the time.
7 DR. LAHEY: This is Richard Lahey again.
8 I think that was a great question because it really 9 captures a fundamental difference in our view, or my 10 view and the view that they have expressed. I'm 11 concerned with having degraded properties and the 12 ability if you have high enough strain to fail those 13 properties, those components in core. And I don't 14 believe it's adequate just to do a safety analysis 15 using ductile materials. I don't believe it's 16 adequate at all. That's the fundamental difference in 17 our view.
18 ADMIN. JUDGE WARDWELL: Okay. Thank you.
19 CHAIRMAN MCDADE: Excuse me. Judge 20 Wardwell, before we move on, I've got a question, 21 perhaps going back a little bit, that I need some 22 clarification on. Maybe a little bit off point, but 23 Dr. Lahey had raised an issue at Pages 45 and 46 of 24 his testimony, about issues at Davis Besse. And in 25 the response, in Answer Number 102, by Entergy, you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5061 1 indicated that -- dismissed it by saying, it does not 2 appear that conditions similar to those which led to 3 the degradation at Davis Besse are present at Indian 4 Point. But there isn't any explanation as to why the 5 conditions are different or how the conditions are 6 different. This is your Answer 102 on Page 59 of your 7 testimony. Anybody from Entergy can address what the 8 differences are between the conditions at Davis Besse 9 and Indian Point and why you believe the concern of 10 Dr. Lahey is unwarranted?
11 MR. AZEVEDO: Yes, this is Nelson Azevedo 12 for Entergy. The events at Davis Besse, I'm assuming 13 you're talking about the corrosion that the reactor 14 vessel had --
15 CHAIRMAN MCDADE: Yes.
16 MR. AZEVEDO: -- at Davis Besse.
17 CHAIRMAN MCDADE: And I realize that's 18 covered by a different AMP, but --
19 MR. AZEVEDO: Yes, so the events at Davis 20 Besse occurred because the reactor vessel had 21 penetrations, which are LI600, cracked and they were 22 undetected for a period of time. And that caused 23 leakage onto the reactor vessel head, which eventually 24 corroded the base metal itself. Indian Point does not 25 have this issue and one of the reasons is because we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5062 1 do inspections every outage. We inspect those 2 penetrations every outage and we have not found any 3 cracks to date. So we know there is no leakage going 4 on similar to what happened at Davis Besse.
5 MR. DOLANSKY: This is Bob Dolansky with 6 Entergy. Additionally, we not only inspect the 7 penetrations for cracking, but we also do what's 8 called a bare metal visual inspection on the top of 9 the head, on the outside surface of the head, where we 10 go around every penetration and look visually and make 11 sure that there's no evidence of any corrosion. So we 12 actually look for where cracking could start inside 13 the head and then we also verify by a bare metal 14 visual inspection that there is in fact no corrosion 15 going on like there was at Davis Besse.
16 CHAIRMAN MCDADE: Okay. But, Dr. Lahey, am 17 I correct, what your concern was, was the mechanism 18 that caused the cracking as opposed to the failure of 19 inspecting?
20 DR. LAHEY: Well, I'm concerned about what 21 happens internally and what happens if it's a through-22 crack. So it's heartening to hear that they do those 23 types of inspections. I think that's very 24 responsible. But there are some welds in the inside 25 that they can't do full inspection of and if they NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5063 1 weaken for various degradation mechanisms, like stress 2 corrosion cracking or whatever happens, you can have 3 a concern, particularly for some of the stub tubes 4 associated with the control rod drives. So this is 5 one of the concerns that we had.
6 CHAIRMAN MCDADE: Okay, but are we 7 concerned about the mechanism that caused the 8 cracking? As I understand it, at Davis Besse, there 9 was an issue that many of these locations were either 10 inaccessible to inspection or very difficult to 11 inspect and that they didn't inspect and that's why 12 the problem was able to reach the level that it had.
13 Here, Entergy is indicating that they do have an 14 inspection program that identifies these potential 15 problems and are able to ensure that there has not 16 been cracking. What I'm concerned with is the 17 mechanism that would have caused the cracking in the 18 first place, not the ability to identify it, but the 19 mechanism that would cause the cracking. And would 20 there be any difference in the situation leading to 21 that mechanism between Davis Besse and Indian Point?
22 DR. LAHEY: Well, one of -- this is Dr.
23 Lahey again. One of the things in the conclusions in 24 the document that you asked about before, which was 25 authored by Dr. Busby, was that irradiated assisted NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5064 1 stress corrosion cracking was still one of the 2 greatest concerns that we have. And in fact, EPRI had 3 put that out as a statement about a year before as 4 well. So it is an issue, I mean, that's the mechanism 5 as I understand it that's of most concern and since 6 you can't do full inspection inside, it remains a 7 concern.
8 CHAIRMAN MCDADE: Okay. When --
9 MR. STROSNIDER: This is Jack Strosnider 10 for Entergy, if I could just comment on this. First 11 of all, there could be differences in the 12 susceptibility to cracking based on who manufactured 13 the vessel head and the specific configuration. I 14 can't speak to that without a lot of details, but 15 there could be differences. But I think the important 16 thing to recognize is that GALL says that the 17 potential for cracking does exist and that's why these 18 inspections are done.
19 So, if you look at the overall framework, 20 the GALL report doesn't say you're not going to see 21 that kind of cracking, it says, it's a potential and 22 you need to go look for it and they're doing two 23 different types of inspections. And, by the way, at 24 Davis Besse, they had several outages of opportunity 25 to identify the problem that was occurring there, but NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5065 1 the inspections weren't being done that needed to be 2 done.
3 CHAIRMAN MCDADE: And all I'm trying to 4 find out is just to make sure I understand, when you 5 use the term, does not appear to have conditions 6 similar, are we talking about only the fact that you 7 have an effective inspection program, and specifically 8 your reactor vessel head penetration inspection AMP, 9 that it is effective versus the one at Davis Besse?
10 Or is there something different about the reactor 11 itself where when you use the term conditions similar 12 which would result in cracking as opposed to just your 13 ability to identify it?
14 MR. AZEVEDO: Yes, sir. It's Nelson 15 Azevedo for Entergy. There are differences between 16 the way the reactor vessel heads were fabricated for 17 Davis Besse versus Indian Point. Davis Besse's 18 reactor vessel head was made by B&W. Our heads were 19 made by Combustion Engineering. So the process they 20 used to install the penetrations and the way they 21 strength the penetrations, the way the welding was 22 done, was different. Also, the materials were 23 different. They're both LI600, but the Davis Besse 24 materials were B&W tubular products, ours are 25 Huntington alloy materials.
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5066 1 And if you look at the OE, at the 2 operating history for the leaks, essentially all the 3 leaks occurred in heads made by B&W. If you look at 4 Oconee, Davis Besse, most of the other leaks were by 5 B&W heads. Again, ours is different. So they're both 6 LI600, so they're both material susceptible, but I 7 feel that our material is much less susceptible than 8 the Davis Besse heads. Also, Oconee --
9 CHAIRMAN MCDADE: Excuse me. Why and is 10 there anything here relevant that carries over to the 11 reactor vessel internals, which we're focusing on?
12 MR. AZEVEDO: Yes, that's what it was going 13 to say. As far as the irradiation on the upper head, 14 there's no -- the fluence is very, very low, less than 15 one times ten to 17. And, no, I don't think there's 16 anything that carries over to the reactor vessel 17 internals.
18 CHAIRMAN MCDADE: Okay. Dr. Lahey?
19 DR. LAHEY: Yes, I misspoke, I should have 20 said primary water stress corrosion cracking. He's 21 absolutely correct.
22 CHAIRMAN MCDADE: Okay, thank you. Dr.
23 Wardwell?
24 ADMIN. JUDGE WARDWELL: Entergy's 25 testimony, 616, Answer 139, Page 86, states at Table NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5067 1 5-2 of Attachment 2 to NL12-037, and that Attachment 2 2037 is the Inspection Plan, so this is Table 2 of the 3 Inspection Plan, specifies a required timing of the 4 first inspections and subsequent intervals for the 5 primary components in the reactor vessel internals 6 AMP. For most components, the first planned 7 inspection at Indian Point are scheduled for two 8 refueling outages from the beginning of the PEO, i.e.,
9 the Spring of 2016 for IP2 and the Spring of 2019 for 10 IP3.
11 You go on to state in Answer 142 of Page 12 92 that the NRC Staff, in a safety evaluation for MRP 13 227-A, acknowledged the justification for the timing 14 of the initial PEO and subsequent inspections and 15 found the inspection intervals acceptable, and 16 referencing the SE that Staff put out for MRP 227, 17 which I believe is Entergy's Exhibit 230. And I guess 18 I'll start with Entergy. What's your technical basis 19 for justifying not performing the first inspections 20 during the first refueling outage?
21 DR. LOTT: This is Randy Lott for Entergy.
22 As I think someone said earlier in the day, we did a 23 lot of evaluations for these components and we did not 24 identify any component where degradation, shall I say, 25 fell off the edge of the table. That it was a gradual NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5068 1 process, it was a process that seemed to be 2 appropriately managed. We wanted to see that it was 3 integrated into -- that there were baseline conditions 4 set up within the program. And so we thought within 5 the first two years of PEO was appropriate.
6 It also gave a chance to coordinate some 7 of these examinations with the ASME Section 11 8 examinations, gave some flexibility to that schedule, 9 which I think is important to the implementation of 10 these exams. It's proven to be -- and we've already 11 done a number of baseline exams and I think we've 12 shown that our number of findings have been extremely 13 low and it seems an appropriate response.
14 ADMIN. JUDGE WARDWELL: But I guess I don't 15 understand the timing need. I mean, the application 16 was submitted in 2007, so they knew this was coming 17 up. Why did they need more time to get ready for the 18 first inspection besides the first refueling outage 19 after the PEO started?
20 MR. DOLANSKY: This is Bob Dolansky with 21 Entergy. To perform the MRP 227-A inspections, we'll 22 remove the component called the core barrel. We also 23 do that when we do the ten year ISI inspections. So 24 we wanted to do both of those together. And the ten 25 year ISI inspection, which inspects the reactor vessel NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5069 1 itself, the welds that make up the reactor vessel, 2 that's required on a ten year frequency for ASME 3 Section 11. So the reason we're doing it during the 4 second outage of the PEO is to allow us to do both of 5 those together so we only have to remove the core 6 barrel one time.
7 ADMIN. JUDGE WARDWELL: NRC Staff, would 8 you like to comment on why you are convinced through 9 their justification that the second refueling outage 10 was adequate?
11 DR. HISER: This is Allen Hiser. The Staff 12 -- I think some of the bases relate to the 13 expectations of degradation and the desire to allow 14 for a higher likelihood that degradation would be 15 detectable. Some of the mechanisms, the analyses by 16 Westinghouse, it led to MRP 227, the degradation that 17 they -- the levels that they were using relate to 60 18 year fluences and things like that. At 40 years, the 19 fluences aren't to that level. So I think we just 20 thought that it was reasonable to delay the baseline 21 inspections into the PEO.
22 As Dr. Lott mentioned, plants have done 23 inspections so far, so far there have been very 24 limited, if any, indications of degradation identified 25 at all. So I think that reinforces -- if the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5070 1 inspections that had been done at other plants 2 indicated problems, then I think the NRC would have 3 likely pushed to accelerate the inspections at Indian 4 Point and similar plants.
5 ADMIN. JUDGE WARDWELL: Back to Entergy.
6 I believe your Table 5-2 and 5-3 of your Inspection 7 Plan shows inspection intervals of ten years for many 8 components. Were there any other shorter intervals 9 incorporated into your Inspection Plans beside the ten 10 year cycle that you can remember?
11 MR. DOLANSKY: No.
12 ADMIN. JUDGE WARDWELL: And what's the 13 justification for what seems like a pretty long 14 interval between inspections considering the 15 importance of which you're placing on these 16 inspections?
17 DR. LOTT: This is Randy Lott for Entergy.
18 Again, there were two issues. One was coordination of 19 schedules because we felt that, that was actually 20 important, it's a fairly difficult operation to remove 21 the core barrel and it's not something people want to 22 do lightly. And I have to point out that it was our 23 feeling that, particularly as an industry, a single 24 inspection gives us information about a wide variety, 25 for instance, the 800 baffle former bolts in the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5071 1 plant, there's factors of three and four in the 2 fluences, so we see lead components, we see a whole 3 range of fluences in a single inspection. As we 4 gather data, that gives us a much broader data base.
5 And I'll say that we looked at in 6 particular one of the components that had us most 7 concerned were the baffle former bolts themselves.
8 That, as Dr. Hiser has kind of alluded to here, that 9 drove a lot of our inspection schedule decisions. As 10 well, we found because of the assumptions in our 11 analysis and the way that the analysis was put 12 together that the rate of degradation of the baffle 13 former bolts was actually slower in the last part of 14 our irradiations and that it took at least, I think it 15 was 25 effective full power years of operation before 16 we had a reasonable number of predicted failures such 17 that we thought there would be actually something to 18 see. We have experience on baffle former bolt 19 inspections from the 1990s when things were done at 20 lower fluences and very little is seen in any of those 21 plants. So, again, we just felt that it gave us the 22 best opportunity to collect data and that we would get 23 a robust data base from the industry.
24 MR. STROSNIDER: This is Jack Strosnider 25 for Entergy. I'd also like to add to this and what NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5072 1 was mentioned a minute ago that I think it's helpful 2 if you look at this from a fleet perspective. There's 3 a lot of plants doing these inspections at different 4 times and they have similar designs, similar 5 environments. So if something shows up in the 6 operating experience, in accordance with the Program 7 Element 8, Operating Experience portion, that's going 8 to be reviewed and evaluated and if it required a 9 change in the inspection frequency, then they could do 10 that. So there's a lot more data than is coming just 11 from the inspections at this point.
12 CHAIRMAN MCDADE: Okay. If I can clarify 13 in my own mind here, so the difference between the 14 initiation and propagation of a crack, do you have 15 sufficient data to determine, for example you inspect 16 today, tomorrow a crack is initiated, it's another ten 17 years before you inspect again. At the rate of 18 propagation of that crack, do you have data that would 19 indicate how long it would take after initiation for 20 a crack in the bolt to become problematic?
21 DR. LOTT: Well, again, our presumption in 22 the bolts was that any bolt that had a crack would 23 fail. So we basically did not do a calculation of 24 crack propagation in the bolts. We basically said, if 25 it's cracked, it's gone. We do have --
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5073 1 CHAIRMAN MCDADE: Well, what I'm getting 2 at, Dr. Lott, is you do the inspection today, you 3 don't see any cracks in a baffle former bolt. You now 4 don't inspect it for another ten years. You don't 5 know when during that period of time a crack may 6 initiate. If it initiates nine years nine months 7 after the inspection, there's no reason to believe 8 there will be significant propagation by the time you 9 inspect the next time.
10 My question is, is there any way of 11 knowing, for example, if the crack initiates a day, a 12 week, a month after the inspection of whether it is 13 subject to failure within that ten year period before 14 it's inspected again? Is there data that would lead 15 you to believe that the propagation would be at a rate 16 slow enough that it would not be problematic prior to 17 the time of the next inspection? Do you understand 18 where my --
19 ADMIN. JUDGE WARDWELL: And if it helps 20 you, because this was my question I was going to ask, 21 don't relate it to the bolts, relate it to anything.
22 What's to say that a crack that is just ready to be 23 initiated when you inspected it, but hadn't occurred 24 yet, or at least wasn't large enough to be visible, 25 and the next day, it became visible, you don't inspect NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5074 1 it now for ten more years, how do we know there won't 2 be some catastrophic type of result associated with 3 that crack over that ten year interval? What gives 4 you confidence that, that interval is sufficient 5 enough to still be within the range of it being able 6 to maintain its intended function?
7 DR. LOTT: I guess in response to that, I 8 would point out that in preparation for these 9 examinations, we're working with Entergy to develop 10 inspection acceptance criteria. Those inspection 11 acceptance criteria have built into them, again, how 12 large a crack would be allowable and that includes an 13 allowance for crack growth. So that would be starting 14 with a crack, that means there is an allowable size.
15 I would suggest to you that if a crack were to 16 initiate now, it would be less than that allowable 17 size because there's ten years of growth in the 18 acceptance criteria. So as long as --
19 ADMIN. JUDGE WARDWELL: But what is the 20 allowable size in your acceptance criteria? I thought 21 it as soon as a crack appears, you have to take some 22 -- that is --
23 DR. LOTT: Well, we need to --
24 ADMIN. JUDGE WARDWELL: -- that isn't an 25 acceptance criteria.
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5075 1 DR. LOTT: Well, that would trigger an 2 engineering evaluation. What I'm talking about is the 3 engineering evaluation, the basis for the engineering 4 evaluation.
5 ADMIN. JUDGE WARDWELL: Yes, but the ten 6 years is passed, you're not doing that evaluation, 7 you're not doing anything on it, but the crack is 8 already there and off and running. And what's to lead 9 you to believe that it will maintain its intended 10 function for ten years as this crack propagates?
11 DR. LOTT: Well, again, within the 12 evaluation, we started with a finite crack length and 13 allowed it to grow ten years and showed at the end of 14 ten years it would still be acceptable. That's part 15 of the engineering evaluation.
16 ADMIN. JUDGE WARDWELL: And you know that 17 for all the RVIs that they will still be able to 18 maintain their function as a crack initiates, assuming 19 that the criteria for evaluating it occurs the day 20 after you inspect it, when you didn't see anything, 21 and you do nothing during that period, do you have 22 enough data to comfort yourself that it won't reach 23 some -- it will still be able to maintain its intended 24 function ten years later?
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5076 1 Dolansky with Entergy. I don't think I can say right 2 now that I've reviewed and looked at that for every 3 component. I mean, I know that -- what Dr. Lott is 4 saying, Westinghouse is doing work for Entergy right 5 now coming up with acceptance criteria for the 6 inspection. I don't know for every single component 7 what that acceptance criteria is. I do remember for 8 some of the components where there is an acceptance 9 criteria, where there's an acceptable crack length, 10 that it says crack below this length are acceptable.
11 That means that they have done the calculation out to 12 ten more years before we would inspect it again and 13 determined that, that's acceptable. But I can't say 14 right now if -- I'd have to go back and -- I don't 15 think we have -- we don't have the final acceptance 16 criteria at this time. I'd have to go ask 17 Westinghouse if they've --
18 ADMIN. JUDGE WARDWELL: You've got 19 acceptance criteria in your AMP, don't you?
20 MR. DOLANSKY: The methodology and -- yes.
21 Acceptance criteria, but not the detailed plant 22 specific where we look at the plant specific loads, 23 the plant specific licensing basis, where they 24 actually did the engineering evaluation. You'd have 25 to ask --
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5077 1 ADMIN. JUDGE WARDWELL: Well, we've got 2 more questions, we'll get into the acceptance criteria 3 in more detail a little bit later. But I guess I'd 4 like to turn to Staff in regards to, what did you 5 review that comforts you that you could accept this 6 program knowing that there's a potential for a crack 7 to occur the day after an inspection was finished and 8 it wouldn't be looked at again for ten more years that 9 it would still maintain its intended function?
10 DR. HISER: This is Allen Hiser with the 11 NRC. I was not involved in the specific review of MRP 12 227 and their program is based on 227. In general 13 though -- and there's nothing in the SCR that really 14 describes the basis for acceptability, the ten year 15 reinspection interval. In general, I think there's an 16 expectation that given the knowledge that we have of 17 crack growth rates and things like that, that a flaw 18 that could initiate the day after the inspection as 19 you mentioned, that it would not have sufficient time 20 to propagate to where it could cause a failure of that 21 piece, maybe not even the assembly, but of that 22 individual piece, before the next inspection.
23 ADMIN. JUDGE WARDWELL: And there's data to 24 support that or is that just based on, because that's 25 what we've doing all along?
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5078 1 DR. HISER: I don't believe it was, that's 2 what we've been doing all along. I believe there was 3 more consideration to it. Fundamentally though, it 4 came, my guess is that it came down to engineering 5 judgment. There is nothing in the SCR that really 6 provides a roadmap to that, but I think given our 7 knowledge of crack growth rates, evaluations that have 8 been done for similar components, for example in BWR 9 internals, that, that was found to be a reasonable 10 reinspection interval to provide reasonable assurance 11 of the integrity of the RVI.
12 ADMIN. JUDGE WARDWELL: Some of these 13 inspections have been going on as part of your current 14 licensing basis, haven't these, for some of these 15 internals?
16 DR. HISER: That's correct.
17 ADMIN. JUDGE WARDWELL: And what's the 18 frequency for those?
19 DR. HISER: Those would be every ten years.
20 ADMIN. JUDGE WARDWELL: That's what I 21 thought. With the ten year interval, wouldn't it mean 22 that these inspections basically are going to occur 23 just once over the PEO?
24 DR. HISER: Well, they would occur twice.
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5079 1 53. So they would occur twice, every ten years during 2 the PEO.
3 ADMIN. JUDGE WARDWELL: Well, you've got a 4 20 year PEO period. That's going to be the end of the 5 --
6 DR. HISER: Right. So --
7 ADMIN. JUDGE WARDWELL: -- 53 is going to 8 be the end, you're going to shut down then. You're 9 going to inspect them as you shut down and tear it 10 down?
11 DR. HISER: Year 43 would be the first 12 inspection under this program. Year 53, 13 approximately, would be the second inspection. So it 14 would be two times during the PEO.
15 MR. DOLANSKY: This is Bob Dolansky with 16 Entergy. Just to clarify, the first inspection would 17 be within two outages of the beginning of the PEO.
18 ADMIN. JUDGE WARDWELL: Right.
19 MR. DOLANSKY: That's the first. And then 20 the second one --
21 ADMIN. JUDGE WARDWELL: Oh, that's what 22 you're -- I'm calling the first -- that's what I call 23 a baseline inspection. The first subsequent interval 24 inspection will occur --
25 MR. DOLANSKY: Ten years.
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5080 1 ADMIN. JUDGE WARDWELL: -- 53 and then, so 2 you have only have one interval inspection is what I'm 3 saying beyond this baseline inspection.
4 MR. DOLANSKY: Okay. We count the baseline 5 as the first. We say that we're doing the MRP 227-A 6 exam in Spring of 2016 at Indian Point 2.
7 ADMIN. JUDGE WARDWELL: That's your 8 baseline, correct?
9 MR. DOLANSKY: Yes, but we --
10 ADMIN. JUDGE WARDWELL: Right.
11 MR. DOLANSKY: We treat that as our first 12 227-A exam.
13 ADMIN. JUDGE WARDWELL: Fine. Semantics 14 and that's -- I understand the difference between our 15 discussion. Okay.
16 CHAIRMAN MCDADE: Okay. And I understand 17 the, I think, genesis for the ten year period of the 18 difficulties and what is involved in the inspection.
19 And trying to become sanguine that the ten year 20 inspection is adequate. And I was using the term 21 crack propagation, Dr. Hiser, you used the term crack 22 growth rates. Where do you look to determine the 23 anticipated crack growth rates for these kinds of 24 materials? To satisfy you that a ten year period is 25 adequate?
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5081 1 MR. POEHLER: This is Jeffrey Poehler of 2 the Staff. So MRP 227-A has guidance on crack growth 3 rates to be used for engineering evaluations and 4 that's in Chapter 6 of that report. Also, the 5 industry plans to use WCAP-17096, which is under 6 review by the Staff and we mentioned that yesterday.
7 But that provides methodologies for engineering 8 evaluations when degradation is found. And that 9 report gives guidance on which crack growth rates to 10 use. But they are referencing industry accepted crack 11 growth rates that have been developed for -- well, I 12 probably shouldn't say because it's proprietary at 13 this point.
14 CHAIRMAN MCDADE: Okay. But basically, as 15 I understand what you're saying, is that you look 16 right now to MRP 227, the crack growth rates that are 17 projected there. That your review suggests those are 18 valid and you're willing to rely on those crack growth 19 rates --
20 MR. POEHLER: Right.
21 CHAIRMAN MCDADE: -- and your review of the 22 validity of those crack growth rates in determining 23 that a ten year period is sufficient to ensure that 24 these reactor vessel internals will maintain their 25 intended function.
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5082 1 MR. POEHLER: The Staff believes or it's 2 our opinion that the crack growth rates that they're 3 recommending would be conservative for pressurized 4 water reactors, given the --
5 CHAIRMAN MCDADE: Would be or are?
6 MR. POEHLER: Are.
7 CHAIRMAN MCDADE: Okay. Okay, Dr. Lahey, 8 do you have some input here with regard to crack 9 growth rates and MRP 227 and the validity of the ten 10 year period?
11 DR. LAHEY: Yes. Your honor, we have the 12 same concern as you expressed in your questions. Two 13 inputs that I would give to this discussion is there's 14 been publications which indicate the baffle former 15 bolts, when you do ultrasound inspections, you're 16 unable to detect up to or below 30 percent through-17 crack. So that means you could already start out with 18 a significant weakened bolt at that point. And I want 19 to remind you that on top of all of this, this is an 20 inspection based program and I agree with the concerns 21 that you have raised, but on top of this, at any time 22 during this event, if you have some of these highly 23 degraded structures subject to shock loads, you can 24 fail them. That's the concern that I have. Because, 25 as you know, I'm focused, the bottom line is on the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5083 1 safety of the plant. And the safety of the plant 2 means to be me core coolability.
3 CHAIRMAN MCDADE: It means core --
4 DR. LAHEY: Coolability.
5 CHAIRMAN MCDADE: -- coolability?
6 Coolability, you're saying?
7 DR. LAHEY: Yes. Once you lose an intact 8 geometry, all bets are off as to core coolability.
9 And by far the most vulnerable reactor vessel 10 internals are these baffle bolts.
11 CHAIRMAN MCDADE: Which of course Entergy 12 and the Staff represent can have a 50 percent failure 13 rate without impacting the integrity.
14 DR. LAHEY: Well, my understanding of how 15 that conclusion was drawn was really based on the kind 16 of loads you get during steady state operation and the 17 redundancy to hold them in place. But during accident 18 loads, if you have significant loads, you can unzip 19 the rest of your bolts. That's what I'm worried 20 about.
21 CHAIRMAN MCDADE: Okay.
22 MR. STROSNIDER: This is Jack Strosnider 23 for Entergy. I just should comment that those bolting 24 patterns that were analyzed in the WCAP, and I don't 25 remember the exact number right now, that they're NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5084 1 based on accident loads.
2 DR. HISER: This is Allen Hiser with the 3 Staff.
4 ADMIN. JUDGE WARDWELL: What was the --
5 DR. HISER: I can fill in. Entergy 654 and 6 655 are exhibits that are, at least in one case, is an 7 NRC approved report that looks at the bolting patterns 8 under accident loads.
9 ADMIN. JUDGE WARDWELL: Okay. Well, we'll 10 talk about baffle former bolts in more detail this 11 afternoon too. But before that, I do want to ask the 12 Staff in regards to what's their understanding of the 13 current acceptance criteria that's in the AMP as it 14 stands now?
15 DR. HISER: This is Allen Hiser with the 16 Staff. The acceptance criteria are provided in the 17 RVI Inspection Plan for Indian Point. I think there 18 maybe was some confusion with some of the discussion 19 earlier. The acceptance criteria that are under 20 development by Westinghouse would be an engineering 21 analysis option under corrective actions. So it's --
22 maybe if we say inspection acceptance criteria are in 23 the Inspection Plan and they are definitive.
24 ADMIN. JUDGE WARDWELL: And what do they 25 state there?
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5085 1 DR. HISER: I think in general for 2 cracking, it's no cracking. Any identified cracking 3 is subject to corrective actions.
4 ADMIN. JUDGE WARDWELL: Thank you.
5 CHAIRMAN MCDADE: Okay. Let me -- Dr.
6 Hiser, mentioning that any identifiable cracks, we've 7 talked quite a bit over this morning and yesterday 8 about microcracks. The existence of microcracks that 9 could be present even during an inspection, but not 10 identified. Does the possibility of those microcracks 11 affect your consideration of this, of the frequency of 12 inspection?
13 DR. HISER: No, I don't believe so.
14 Because a microcrack would be subsumed under the 15 inspection -- or depending on the size of the 16 microcrack. I mean, microcracks, if they're below the 17 inspectability limit of the NDE method would not --
18 clearly you would not be able to detect those. And I 19 don't believe that microcracks would have a 20 significant impact on the integrity of the RVI 21 components, sort of as a starting point. So with the 22 analyses --
23 CHAIRMAN MCDADE: But isn't the microcrack 24 basically the crack initiation and then you have the 25 propagation. So the question is, being able to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5086 1 identify where you are on that spectrum at the time of 2 the inspection. The clearer picture you have, the 3 better way to follow it. Now, you made reference to 4 the crack growth rates in MRP 227, does that take into 5 consideration that at the time of the inspection there 6 may well be significant, and perhaps maybe not use the 7 word significant, but microcracks that are approaching 8 visibility, but not yet visible?
9 DR. HISER: Yes. This is Allen Hiser of 10 the Staff. When the report was reviewed for 11 acceptability, my expectation is that the 12 consideration was along the lines of what kinds of 13 flaws could be missed by the inspection given the 14 knowledge that we have of crack growth rates. Was it 15 likely that there would be a challenge to 16 functionality at the end of the ten year reinspection 17 interval? Based on that analysis, be it -- and my 18 guess is this may have been an engineering judgment 19 based that there was not thought to be a significant 20 concern. There was reasonable assurance that the RVI 21 would still maintain their functionality.
22 CHAIRMAN MCDADE: Okay. Thank you, Dr.
23 Hiser.
24 ADMIN. JUDGE WARDWELL: Entergy's 25 testimony, 616, Answer 209, Page 140. The results of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5087 1 the IPEC inspections will be available to the NRC 2 Staff for onsite inspection. Are these results also 3 publically available?
4 MR. DOLANSKY: This is Bob Dolansky for 5 Entergy. These inspection results are put into a 6 report and supplied to EPRI for the whole industry and 7 EPRI rolls that up. I don't know if that's publically 8 available. I don't think it's publically available, 9 but it's available to the industry for sure.
10 ADMIN. JUDGE WARDWELL: So publically it's 11 not available?
12 MR. DOLANSKY: I don't believe so.
13 ADMIN. JUDGE WARDWELL: Okay, thank you.
14 MR. POEHLER: And, your honor, I would just 15 like to add to that. And that report is also 16 submitted to the NRC Staff for our information. So, 17 that's in the MRP 227 implementing process that they 18 will do that. So we will get a chance to review what 19 the operating experience has been and if there's been 20 any trends in failures, then we'll know about it.
21 ADMIN. JUDGE WARDWELL: And is that 22 available to the public or is it considered 23 proprietary information?
24 MR. POEHLER: I don't know the answer to 25 that offhand, I can check.
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5088 1 ADMIN. JUDGE WARDWELL: Do that, would you 2 please? And I think now would be a good time to take 3 a break.
4 CHAIRMAN MCDADE: Okay. Would anybody need 5 more than ten minutes? Okay, it's now -- why don't we 6 come back at 10:40.
7 (Whereupon, the above-entitled matter went 8 off the record at 10:26 a.m. and resumed at 10:44 9 a.m.)
10 CHAIRMAN MCDADE: We're back on the record.
11 ADMIN. JUDGE WARDWELL: Let's turn to some 12 inspection actions now. NRC's Exhibit 197, testimony, 13 Answer 80, Page 51, says inspection techniques include 14 ultrasonic UT testing, EVT1 enhanced visual 15 examinations, and VT3 visual examinations. And so, 16 considering it's NRC's testimony I'm referring to, 17 I'll let them answer. Could you explain each type of 18 test and its applicability for the various reactor 19 vessel internals?
20 MR. POEHLER: Okay. This is Jeffrey 21 Poehler of the Staff. So, just to clarify the 22 question, you want us to explain the applicability of 23 each type of inspection technique?
24 ADMIN. JUDGE WARDWELL: I want you to 25 explain what's a UT test, what's an EVT1 test, and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5089 1 what's the VT3 test.
2 MRP: So UT test is an ultrasonic 3 examination. And it's basically using sound to detect 4 cracks in a material. Do you want a detailed 5 explanation or --
6 ADMIN. JUDGE WARDWELL: How do they show 7 up? How does a crack show up? So, you put an --
8 MR. POEHLER: So you have a transducer that 9 generates ultrasonic sound. It has to be in contact 10 with the material and it puts sound into the material 11 and you get echos back from the material basically 12 that are detected by either the same transducer or a 13 separate transducer. And those are processed 14 electronically such that you get a signal. So if you 15 have a discontinuity, like let's say in a bolt, you 16 have a partially cracked bolt, that's going to reflect 17 the sound back and be detected and it'll be processed 18 by the electronics such that you can determine the 19 location of that discontinuity.
20 And with certain ultrasonic techniques, 21 you can create images, you can image the 22 discontinuity. So ultrasonic is used throughout the 23 nuclear industry for piping weld exams to detect 24 cracking and also in vessels, in large structural 25 welds. So it's a very established technique for welds NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5090 1 and it's also for bolting, such as baffle former 2 bolts, that's been used -- it was first used back in 3 the 1990s when baffle former bolt cracking was first 4 detected.
5 ADMIN. JUDGE WARDWELL: So that's a type of 6 test that's used for the baffle former bolts?
7 MR. POEHLER: It's a pretty well 8 established technique for bolts.
9 ADMIN. JUDGE WARDWELL: So is it used for 10 the clevis bolts that we'll talk about later also?
11 MR. POEHLER: It is not specified for the 12 clevis bolts at this time. It could be, but it's not 13 --
14 ADMIN. JUDGE WARDWELL: Okay.
15 MR. POEHLER: -- has been determined not to 16 be necessary.
17 ADMIN. JUDGE WARDWELL: So is the baffle 18 former bolts the only one it's used for in regards to 19 reactor vessel internals?
20 MR. POEHLER: The only component in a 21 Westinghouse design reactor internals that it's 22 currently used for is the baffle former bolts.
23 ADMIN. JUDGE WARDWELL: Okay.
24 MR. POEHLER: You also asked about visual 25 examination.
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5091 1 ADMIN. JUDGE WARDWELL: How about the 2 enhanced visual examination, EVT1, what is that 3 composed of?
4 MR. POEHLER: Well, so enhanced --
5 ADMIN. JUDGE WARDWELL: Well, let me ask 6 one other question about the UT. So is this a sensor 7 that you put in at the time you do the inspection or 8 are these permanently mounted so you can turn it on?
9 Or how are the mechanics of this, the logistics of 10 this achieved?
11 MR. POEHLER: Right, it's a sensor that has 12 to be put in. It's put in for the time of the 13 inspection and there's special tooling to access the 14 bolting. And so it's not permanent.
15 ADMIN. JUDGE WARDWELL: Okay, thank you.
16 CHAIRMAN MCDADE: What percentage of the 17 baffle former bolts are accessible to the UT 18 inspection?
19 MR. POEHLER: For the UT inspection, 20 essentially 100 percent of them are accessible. There 21 are minimum coverage requirements in MRP 227-A, so you 22 can credit 100 percent -- you can take credit for an 23 examination if you exam 75 percent of a population in 24 general. And that would include both the accessible 25 and inaccessible members of the population. But, to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5092 1 my knowledge, I don't think there are any major 2 obstructions that would prevent you from accessing all 3 the baffle former bolts.
4 CHAIRMAN MCDADE: To Entergy, are 100 5 percent of the baffle former bolts accessible to the 6 ultrasound inspection?
7 MR. DOLANSKY: Yes. This is Bob Dolansky 8 with Entergy. We believe 100 percent will be 9 accessible. We won't know for sure until we actually 10 get out there and do the exam, but based on all of our 11 drawing reviews and so forth, we expect to get 100 12 percent of all the bolts and we expect, there's 832 of 13 them, we expect all 832 to be accessible.
14 CHAIRMAN MCDADE: So unless somebody 15 changed the design since you last looked at it?
16 MR. DOLANSKY: Well, there could -- in 17 theory, there could be -- let me give a little bit of 18 background. The technique to perform this inspection 19 is a submarine that goes down underwater, it's done 20 underwater. It basically docks up against the baffle 21 plate and there's a special head that goes in and 22 inspects the bolt. So although all the bolts should 23 be accessible, we're not 100 percent sure we can reach 24 every one with this tooling technique. So that's why 25 there's a possibility that we wouldn't get every one.
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5093 1 But other Westinghouse plants have got 100 percent, we 2 expect to get 100 percent, but there could be some 3 interference that would limit us, but only a very 4 small number.
5 CHAIRMAN MCDADE: Okay. Thank you.
6 ADMIN. JUDGE WARDWELL: Enhanced visual 7 examinations?
8 MR. POEHLER: Yes. This is Jeffery Poehler 9 of the Staff. So enhanced visual examinations are 10 specified for examining welds for cracking. Or 11 anytime you're looking for cracking specifically in 12 MRP 227-A. For example, the core valve girth welds in 13 a Westinghouse design, reactor internals, you use 14 enhanced visual testing. And that's a visual 15 examination that has a fairly stringent detection or 16 resolution requirement. So you have to be able to --
17 the way that they test this in situ is that they have 18 to be able to identify a character, like a letter, 19 letter C or A or O, and they have to identify that 20 letter and the letter has a 0.04 inch size on the 21 card. So it's a pretty small letter. So that's how 22 they determine the resolution is adequate in the 23 environment that you're going to do the testing.
24 ADMIN. JUDGE WARDWELL: And what's this 25 card?
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5094 1 MR. POEHLER: It's like a plaque --
2 ADMIN. JUDGE WARDWELL: A plaque, a little 3 plaque?
4 MR. POEHLER: Yes, it's a --
5 ADMIN. JUDGE WARDWELL: Okay.
6 MR. POEHLER: It's something that -- yes.
7 ADMIN. JUDGE WARDWELL: Is it on each 8 internal or is it just a test plaque for you to start 9 off to verify that you've got enhanced --
10 MR. POEHLER: It's just a test plaque. You 11 could -- analogous to calibration standard.
12 ADMIN. JUDGE WARDWELL: Calibration coupon 13 --
14 MR. POEHLER: Right.
15 ADMIN. JUDGE WARDWELL: -- shall we say?
16 MR. POEHLER: Right. And so then you have 17 to -- so that's basically how that's qualified. And 18 you would do that before you start the examination.
19 ADMIN. JUDGE WARDWELL: This is a camera 20 that you put --
21 MR. POEHLER: Yes, remote --
22 ADMIN. JUDGE WARDWELL: -- in there, it's 23 not someone's eyeball that you're calibrating.
24 MR. POEHLER: I think they used to -- the 25 submarine type delivery system, or I think for the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5095 1 core valve girth weld, they have special tooling to go 2 up under the thermal shield to deliver the camera. So 3 those are used for when you're specifically looking 4 for cracks such as stress corrosion cracks, IASCC, 5 irradiation assisted stress corrosion cracks, in welds 6 or other components. Other than bolting, it would not 7 be used for bolting because of the location. Where 8 the cracks are would be at the head, the shank area in 9 the bolt, which you're not going to be able to look at 10 side-on. So that's why ultrasonic is used for those 11 bolts. So you also asked about --
12 ADMIN. JUDGE WARDWELL: VT3, visual 13 examination.
14 MR. POEHLER: So VT3 is another visual 15 examination technique. The main difference between 16 VT3 and EVT1 is that it is -- VT3 has a slightly lower 17 resolution requirement and it's used for general 18 mechanical and structural conditions. So you're 19 looking for gross failure, like broken components, 20 broken bolts, and other distortion in structures such 21 as it's used for the baffle former assembly to look 22 for effects of void swelling. But the VT3 visual is 23 only specified in MRP 227-A when it's for either a 24 redundant population of components or components that 25 have been changed to be highly flaw tolerant, such NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5096 1 that they could tolerate cracks. So that's VT3.
2 ADMIN. JUDGE WARDWELL: And VT3 implies 3 there's a VT1 and a VT2 and where did they go and are 4 they ever used at Indian Point?
5 MR. POEHLER: Yes. VT1 is not used for, 6 and Entergy may correct me, but I don't believe it's 7 used for any of the Westinghouse RVI exams. But in 8 some other designs it's used for looking for gaps.
9 But VT1 is also a more -- would have a higher 10 resolution requirement than a VT3. But it's an 11 examination that's called out in ASME code, so it's 12 defined in there. But it's not used very much for the 13 internals. And VT2 is another type of visual 14 examination where you specifically look for leakage of 15 pressure boundary components. So that's not 16 applicable and they don't use that for reactor 17 internals.
18 ADMIN. JUDGE WARDWELL: In New York State's 19 testimony on 482 at 62, lines 3 through 8, Dr. Lahey 20 criticizes the use of VT3 in visual inspections as 21 inadequate for use in inspections for cracking, 22 stating that there are significant shortcomings of 23 this technique to detect material cracking, 24 degradation, or wear prior to failure, as illustrated 25 by the visual detection of only seven out of 29 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5097 1 fractured clevis insert bolts at the Westinghouse PWR 2 in 2010. Why I don't I turn to -- well, I'll stay 3 with you. So, why do we have any confidence in VT3 4 for any components and what components beside the 5 clevis bolts is this technique used and what's it's 6 track record for the other components?
7 MR. POEHLER: Well, for example, it's used 8 for the baffle former assembly, for the general 9 examination for the assembly for distortion due to 10 void swelling. So that wouldn't be something you 11 would use -- it's the most appropriate for looking for 12 that type of aging deformation.
13 CHAIRMAN MCDADE: Okay. Well, let me jump 14 in here a second and to Entergy, in your Exhibit 616 15 at Page 87, your Table 1, you list the various items 16 to be inspected and how you're going to be inspecting 17 them. Is there any difference in the, for lack of a 18 better phrase, degree of difficulty in the EVT1 and 19 the VT3? I mean, it appears that the EVT1 gives you 20 greater resolution, more information. Is there any 21 reason why you use the VT3 for certain items, like 22 baffle edge bolts, as opposed to the EVT1?
23 MR. DOLANSKY: This is Bob Dolansky with 24 Entergy. An EVT1 is looking for cracking. A VT3 is 25 not looking for cracking. So, for example, the baffle NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5098 1 edge bolts, if the cracking is down in the bolt, a 2 visual inspection is not going to see it. So if you 3 do an EVT1 on the baffle edge bolts, it's not going to 4 give you any more information than a VT3 would and it 5 is in fact a more difficult exam to do because you 6 have -- the character requirements are more stringent 7 and --
8 CHAIRMAN MCDADE: I'm sorry, I just didn't 9 hear. The what requirements?
10 MR. DOLANSKY: The character card 11 requirements are more stringent and you have to 12 inspect at a certain speed. You typically use video 13 enhancement to do that inspection. So, for the baffle 14 edge bolts, since you can't see the area where they 15 would be cracking anyway, an EVT1 doesn't buy you 16 anything. What a VT3 tells you about baffle edge 17 bolts is that if you look at all the edge bolts, if 18 you saw that there was -- the two plates where they go 19 together are shifted or moved, a VT3 is very good for 20 that. You're not --
21 ADMIN. JUDGE WARDWELL: But I thought you 22 used the UT for baffle bolts.
23 MR. DOLANSKY: For baffle former bolts, a 24 UT is used.
25 ADMIN. JUDGE WARDWELL: Yes.
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5099 1 MR. DOLANSKY: But a baffle edge bolt --
2 ADMIN. JUDGE WARDWELL: Oh, okay, sorry.
3 MR. DOLANSKY: -- it's not.
4 ADMIN. JUDGE WARDWELL: Sorry.
5 MR. DOLANSKY: A VT3 is used for the baffle 6 edge bolts. Did that answer your questions, your 7 honor --
8 ADMIN. JUDGE WARDWELL: Yes.
9 MR. DOLANSKY: -- about the differences?
10 MR. AZEVEDO: Your honor, this is Nelson 11 Azevedo. Maybe I can add a little bit to that. A VT1 12 is done up close. So, for example, if you're looking 13 for deformation of a component, see if a component is 14 bent, if you're up real close to a component, you may 15 not be able to see if the component is bent. So you 16 actually, if you step back, you can be further from 17 the component, in a lot of cases, you actually get a 18 better assessment of what the component is 19 experiencing rather than be within a couple inches of 20 the component. So that's another reason why sometimes 21 VT3 are used versus VT1.
22 CHAIRMAN MCDADE: Sort of why I take my 23 glasses off to thread a needle?
24 MR. AZEVEDO: If that's the example you 25 want to use.
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5100 1 CHAIRMAN MCDADE: Do you thread needles?
2 (Laughter.)
3 MR. AZEVEDO: Occasionally.
4 CHAIRMAN MCDADE: Sorry. Staff, do you 5 thread needles?
6 (Laughter.)
7 ADMIN. JUDGE WARDWELL: Back to the clevis 8 insert bolts, and I'll direct this to, I guess I'll 9 stay with Entergy, it doesn't seem like the seven out 10 of 29 fractured clevis insert bolts that were detected 11 by this process is a very high percentage of success.
12 Well, let me rephrase that, you look confused. New 13 York State claimed that only seven out of 29 fractured 14 clevis insert bolts were detected in 2010 at a 15 Westinghouse PWR. Are you aware of that and, if so, 16 why is VT3 a successful inspection technique for seven 17 out of 29?
18 MR. DOLANSKY: This is Bob Dolansky with 19 Entergy. We definitely were aware of that. There was 20 a Technical Bulletin that was issued as a result of 21 the clevis insert bolting. I would say that the VT3 22 is what initially found it at the other plant. I 23 think it was sufficient, they did find it, they didn't 24 find every one, but it showed them that there was 25 something going on and then they did additional exams NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5101 1 to determine that. To me, with the clevis insert 2 bolts, and Dr. Lott can step in and help me with this, 3 but the clevis insert bolts don't have any safety 4 function. So there's no -- the fact that they cracked 5 and that they detected that with VT3 shows that, as 6 part of the Section 11 program, shows that, that is in 7 fact working in my estimation. Do you want to add 8 anything about that?
9 DR. LOTT: I don't think we ever believed 10 or contended that the clevis insert bolts, the VT3 11 inspection would identify failed clevis insert bolts.
12 What we were worried about, as Mr. Dolansky said, was 13 that the location and the securing of the clevis in 14 the log on the reactor pressure vessel -- and the 15 Technical Bulletin that he referred to, which is 16 Westinghouse Technical Bulletin 14-5, is ENT Exhibit 17 656, just for the record.
18 And we looked very closely at the clevis 19 insert bolts in that case and made additional 20 recommendations about inspections that would help us 21 determine what we thought was the key feature, which 22 was the clevis being still in place. Once the reactor 23 is loaded, the lower core plate is in, the clevis is 24 effectively locked in place and the safety function of 25 the clevis, as long as it's there when you start up, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5102 1 there's no place for it to go when you're basically 2 running the reactor. We can go through this in more 3 detail if you want to, but we'd probably have to pull 4 up some of the diagrams.
5 ADMIN. JUDGE WARDWELL: Yes. I think I at 6 least have the picture of that. How did you determine 7 that eventually there was 29 -- or how did 8 Westinghouse determine that there were 29 fractured 9 bolts, not just the seven that were detected by the 10 VT3?
11 DR. LOTT: Well, Westinghouse's involvement 12 in this program, we had advised the utility that, when 13 they observed the first, or you told me the number 14 seven, broken bolts, advised them that they should 15 think about replacing the bolts, not because of a 16 safety concern, but our concern was that if that were 17 to become dislodged in the refueling cycle, it would 18 be extremely expensive and difficult to put it back in 19 place, it's a precision fit part. So therefore --
20 ADMIN. JUDGE WARDWELL: This is the clevis 21 itself, the insert itself?
22 DR. LOTT: The clevis itself. The clevis 23 insert is specifically machined to match the keys in 24 the core barrel.
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5103 1 it correctly, these bolts merely hold that in place 2 until that superstructure or whatever you call it, the 3 --
4 DR. LOTT: The core barrel.
5 ADMIN. JUDGE WARDWELL: -- the core barrel 6 coming in and --
7 DR. LOTT: Yes, once the core barrel is 8 engaged --
9 ADMIN. JUDGE WARDWELL: -- resting on that, 10 which --
11 DR. LOTT: -- in the key, it's effectively 12 restrained by multiple factors that we can go into if 13 you want to. And so, therefore, we determined it 14 wasn't a safety concern, but that it might be in the 15 best interest of the utility to replace those bolts 16 because if it were to come loose, it would be 17 incredibly difficult to replace.
18 ADMIN. JUDGE WARDWELL: Are either the --
19 any of these bolts, whether it's the clevis or the 20 baffle edge or the baffle former bolts, if they fail, 21 do they become dislodged and have a potential to 22 impact the geometry, if you will, and the coolability 23 of the core itself? Or a function of any of the other 24 vessel internals?
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5104 1 of the advantages of the VT3 examination is that it 2 allows you to determine whether the lock bars, which 3 hold the heads of all the bolts in place, lock bars 4 that are locked across the heads, are in place. And 5 as long as those lock bars are in place, there's no 6 way for the bolt to escape. In particular, the clevis 7 insert bolts are, again, geometrically constrained.
8 Once they're in place, there's no way for that head to 9 escape from this very small tight dimensions of the 10 object.
11 ADMIN. JUDGE WARDWELL: Well, couldn't the 12 bolt itself crack and fall away from the head and go 13 down below?
14 DR. LOTT: Well, it's threaded in and it 15 can't -- there's nowhere for it to go except for out 16 past the head.
17 ADMIN. JUDGE WARDWELL: Okay, thank you.
18 While we're on it, Dr. Lahey, would you agree with the 19 statements that were made in regards to the clevis 20 bolt inserts and do you have comments about the seven 21 out of 29?
22 DR. LAHEY: Well, our concern was it was 23 not a very effective non-destructive testing technique 24 and the accuracy of it wasn't very effective. I'm not 25 very concerned about the safety aspects of the clevis NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5105 1 bolts, but the technique itself had some deficiencies.
2 ADMIN. JUDGE WARDWELL: Thank you.
3 CHAIRMAN MCDADE: Okay. If I could, just 4 to clarify. We've talked about these clevis bolts and 5 not having a safety consideration. If you could 6 elaborate for me, as I understand from that, Dr. Lott, 7 you referred to the Westinghouse Technical Bulletin, 8 which was Entergy Exhibit 656, and it talks about 9 maintenance risk due to the complexity and cost of 10 repair and the required level of contingency planning.
11 That inability to remove the core barrel or need to 12 replace the insert to reestablish customized design 13 gaps. Could you explain to me, what really --
14 elaborate a bit on the implications of a failure of 15 these clevis bolts?
16 DR. LOTT: Sure. The clevis inserts 17 themselves are actually, they're positioned -- they 18 locate these keys and they're relatively high 19 tolerance components, such that if it -- when the 20 reactor is actually built, they're precision machined 21 to match. And then they're shrunk fit into the lugs 22 in the component, so they're frozen, put in -- so 23 they're a tight fit. That's one of the reason they 24 don't come out is because they're shrunk fit in place.
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5106 1 in the plant and relocate it into the exact location 2 where it was in the first place if the bolts were to 3 come loose and you wanted to even put the same one 4 back in, much less replacement, is a very difficult 5 job. So if we were to lose it, it would be hard to 6 replace. Is that --
7 MR. DOLANSKY: Dr. Lott, can I just add 8 something?
9 CHAIRMAN MCDADE: Hard to or impossible and 10 then --
11 DR. LOTT: I don't think I would say 12 impossible to replace, but, again, this was a 13 maintenance concern, not a safety concern. So it was 14 really a matter of what did it make sense for the 15 utility to do at that point?
16 MR. DOLANSKY: This is Bob Dolansky with 17 Entergy. Just to try to clarify a little bit, when 18 you refuel the reactor vessel, basically you do 19 everything remotely, so you're not unbolting things.
20 Basically, everything slides together and slides 21 apart. So when you're removing the core barrel, the 22 clevis insert basically is like a key and there's a 23 key on the core barrel that just slides in there.
24 Once it's in there, there's no safety function, it has 25 no safety function.
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5107 1 But if I as the owner am planning to 2 refuel the reactor vessel and remove the core barrel 3 and that clevis insert moves, then I cannot get that 4 back in. That's very expensive for me to then try to 5 get that fixed, just so I can get it back in, it has 6 no safety function, but just to be able to put it back 7 together. So that's why it's a commercial issue and 8 not a safety issue. And that's why what Dr. Lott is 9 saying is, if three out of six bolts or four out of 10 six bolts failed, but two are still holding it, I can 11 get those four that are failed out and put new ones in 12 without that thing shifting. Once it shifts, to try 13 to get it lined up exactly again, underwater, would be 14 very, very difficult. It could be done, absolutely 15 could be done and would be done, but it just would be 16 much more difficult. So for us, it makes much more 17 sense to try to get those things replaced before it 18 causes us a commercial issue. Does that help?
19 CHAIRMAN MCDADE: Yes. So basically what 20 you're saying is, it's not an inability to remove the 21 core barrel, it just makes it a much -- the degree of 22 difficulty increases significantly?
23 MR. DOLANSKY: Getting it out or putting it 24 back together. Putting it back together, if it's 25 cocked, it wouldn't slide, it's a very close machine NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5108 1 tolerance. The core barrel may not slide back into 2 the key if it were totally -- if it had moved.
3 CHAIRMAN MCDADE: Okay. Thank you.
4 ADMIN. JUDGE WARDWELL: Entergy's 5 testimony, 616, Answer 188, Page 124, states that the 6 basis for the adequacy of these inspection techniques 7 is described in the companion document to MRP 227-A, 8 which is MRP 228. This document describes the 9 standards to be met by each specific examination 10 method. The standards in MRP 228 reflect the latest 11 information and regulatory documents, such as 12 NUREG/CR-6943, which addresses visual examinations, 13 including remote visual examinations, and describes 14 the characteristics of flaws to be detected in nuclear 15 reactor components. In particular, such critical 16 characteristics as the crack opening displacement.
17 And I guess I just want to verify, is 228 an exhibit 18 in this proceeding, Entergy, as far as you know?
19 DR. LOTT: I believe it is.
20 MR. DOLANSKY: I believe it is.
21 ADMIN. JUDGE WARDWELL: Okay, thank you.
22 MR. KUYLER: Your honor? MRP 228 is 23 Exhibit Entergy 645.
24 ADMIN. JUDGE WARDWELL: Thank you. I guess 25 I'll ask Dr. Lahey, have you provided any evidence NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5109 1 challenging these details on the basis for the 2 development of Entergy's Inspection Plan or that 3 Entergy's Inspection Plan does not meet all the 4 requirements of MRP 227?
5 DR. LAHEY: Are you talking about other 6 than the inability to detect the clevis bolts?
7 ADMIN. JUDGE WARDWELL: Right. Just --
8 DR. LAHEY: I would have to look back, I 9 don't know off hand.
10 ADMIN. JUDGE WARDWELL: Okay. Thank you.
11 CHAIRMAN MCDADE: Excuse me one second, and 12 perhaps I just got confused. Did you say, MRP 227?
13 ADMIN. JUDGE WARDWELL: Yes.
14 CHAIRMAN MCDADE: Okay. Because 645 is 15 EPRI 228.
16 ADMIN. JUDGE WARDWELL: Right. That was my 17 first question.
18 CHAIRMAN MCDADE: Thank you. I'm just --
19 thank you.
20 ADMIN. JUDGE WARDWELL: Okay. We switched 21 MRPs on you. NRC 197, testimony, Answer 183, Page 22 105, states that we disagree with Mr. Lahey's concern 23 regarding synergistic effects because, one, primary 24 components are to be inspected under Entergy's program 25 are those components which are most likely to be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5110 1 effected by synergistic effects, if they exist. Two, 2 that Entergy's AMP inspects for and will detect 3 cracking whether a single aging mechanism or multiple 4 or synergistic aging mechanisms contribute to 5 cracking.
6 Three, embrittlement alone will not cause 7 failure without the presence of a crack and the 8 inspections performed by Entergy's AMP are sufficient 9 to detect cracking. Four, Dr. Lahey has not 10 identified any tests or operating experience which 11 demonstrates that synergistic effects are significant 12 for PWR RVIs and existing laboratory test data on 13 synergistic effects is inconclusive. And, five, the 14 industry reactor vessels internal program in which 15 Entergy is participating is a living program which 16 shares operating experience among all PWR licensees 17 and, thus, any occurrence of an unexplained or 18 accelerated degradation due to synergistic effects 19 will be identified and adjustments to the industry 20 guidance and the Entergy AMP will be made to ensure 21 continued integrity of the RVI across the fleet.
22 I think with this statement, I'd like to 23 start off fixing one other point again. And this is 24 this statement that embrittlement will not occur 25 without a crack. And I'll address this to Entergy.
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5111 1 Does this statement mean that embrittlement won't have 2 any affect until cracking is caused by some other 3 mechanism or is this a statement that says that 4 embrittlement won't alone cause failure without a 5 crack? Meaning that embrittlement can cause a crack, 6 but it won't result in failure unless there is a crack 7 demonstrated from the embrittlement?
8 DR. LOTT: This is Randy Lott for Entergy.
9 The statement, I believe, is saying that if a material 10 is embrittled, that fact alone will not result in the 11 failure of the component, it needs to be combined with 12 a crack and a load that would challenge the stability 13 of the component.
14 ADMIN. JUDGE WARDWELL: With embrittlement, 15 can embrittlement eventually cause a crack in and of 16 itself?
17 DR. LOTT: No. It's not identified as one 18 of the crack causing mechanisms in the component.
19 ADMIN. JUDGE WARDWELL: What is the effect 20 of embrittlement then?
21 DR. LOTT: What is the effect? The change 22 in the -- for me, the discussion of embrittlement is 23 a discussion of the change in properties that happen 24 when the material is irradiated. And we've listed 25 them several times before, it's the increase in yield NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5112 1 stress, the decrease in ductility, the decrease in 2 fracture toughness, those combined effects.
3 ADMIN. JUDGE WARDWELL: Is the resulting 4 effect?
5 DR. LOTT: Those are -- yes. If I had a 6 test specimen, those are measurable, objective things 7 I could measure.
8 ADMIN. JUDGE WARDWELL: So regardless --
9 DR. LOTT: I can't do that test non-10 destructively.
11 ADMIN. JUDGE WARDWELL: So regardless of 12 how long or how much fluence there is at a location, 13 a piece of metal there will not exhibit a crack due to 14 just embrittlement causes? Let's say you have a test 15 coupon and --
16 DR. LOTT: If I had a test coupon and I 17 were to just -- we certainly irradiated lots of 18 specimens to the fluences that we've talked about 19 lots. We've irradiated specimens to the fluences that 20 we've talked about here, the 60, 70 dpa, and they do 21 not just spontaneously crack.
22 ADMIN. JUDGE WARDWELL: Okay.
23 DR. HISER: And, your honor, this is Allen 24 Hiser of the Staff. I just want to make sure, the 25 Staff did not imply that embrittlement can cause NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5113 1 cracking. The statement that you read just says 2 embrittlement cannot cause failure unless there is a 3 crack.
4 ADMIN. JUDGE WARDWELL: Correct.
5 DR. HISER: So the Staff did not imply 6 embrittlement causes cracks.
7 ADMIN. JUDGE WARDWELL: Right.
8 DR. HISER: Okay. Just to make sure.
9 ADMIN. JUDGE WARDWELL: Yes. I didn't mean 10 to imply that you implied that.
11 DR. HISER: Okay.
12 CHAIRMAN MCDADE: Okay. And --
13 DR. HISER: Maybe I inferred that from you.
14 CHAIRMAN MCDADE: And either Dr. Hiser or 15 Dr. Lahey, let me -- and, again, I want to make sure 16 I understand this before we go off to try to make a 17 decision here. Embrittlement is a mechanism, an 18 effect of embrittlement is a decrease in ductility.
19 And it's Dr. Lahey's position that, that decrease in 20 ductility makes the item more susceptible to cracking.
21 Dr. Lahey, is that correct?
22 DR. LAHEY: The effect of embrittlement is 23 -- I mean, fundamentally, the difference is no crack 24 no problem. I mean, there are people who believe no 25 crack no problem, to me that was the quotable quote NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5114 1 from yesterday. I don't believe that. I believe with 2 or without a crack, if you have an embrittled 3 structure and you hit it with a significant impulsive 4 load, you can fail the structure. And that's the 5 concern.
6 CHAIRMAN MCDADE: But that's -- again, the 7 concept there is the ductility has decreased because 8 of the embrittlement and because of that decrease in 9 ductility, the item is more susceptible to cracking 10 under stress with a load?
11 DR. LAHEY: Well, the fracture toughness, 12 how the material can retard crack growth, is 13 decreased. So cracks can grow faster if it's 14 embrittled. So we don't disagree on that. What we 15 disagree on is, do you need a surface crack or not?
16 Is the material weakened or subject to failure if it's 17 embrittled? That's really the crux of it. One thing 18 that --
19 CHAIRMAN MCDADE: Also, and again maybe 20 we'll get into this later, but you used the word 21 failure and there's an inspection failure and then 22 there's a failure of intended use. The failure of 23 intended use involves cracking. But as I understand 24 it, there would be cracking before you had a 25 catastrophic failure, where it would no longer serve NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5115 1 its intended use. I'm just trying to follow in my 2 mind the mechanism along. And bear with me here 3 because I'm plowing new ground for me, I realize that 4 this is old ground for all of you. But I had thought 5 from yesterday that ductility decrease made the item, 6 a bolt for example, more susceptible. I thought that 7 the fracture toughness actually increased with the 8 embrittlement. Do you disagree? Did I --
9 DR. LAHEY: It depends, what do you mean by 10 fracture toughness? If you mean the ability of a 11 crack to propagate, which is the way I think it's 12 being used, then it's not correct. I mean, there is 13 another mechanism in which the strength of the 14 material is increased by irradiation induced 15 embrittlement.
16 CHAIRMAN MCDADE: Okay. And I guess what 17 I'm confused then is the sort of the inter-reaction of 18 these terms and what these terms represent. So if, 19 and again, bear with me here, the strength increases, 20 I'm not really sure what you mean by strength, I had 21 taken strength to be synonymous with fracture 22 toughness. And that the ductility, basically the 23 ability to bend and come back is something entirely 24 different. And although the ductility was decreasing 25 with embrittlement, that there was a strength increase NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5116 1 and there was not a decrease in fracture toughness.
2 Dr. Hiser, is that the Staff's position? Or please 3 educate me here.
4 DR. HISER: This is Allen Hiser with the 5 Staff. What neutron fluence does to stainless steel 6 in particular or the internals is you increase the 7 fluence, the yield strength increases, the ultimate 8 strength increases, the fracture toughness is reduced, 9 and that's normally what we consider neutron 10 embrittlement. It makes the fracture toughness 11 decrease. In addition, the ductility decreases. And 12 that's the bendability that you spoke of. So the 13 fracture toughness does get reduced by neutron 14 fluence.
15 CHAIRMAN MCDADE: And please explain to me 16 very briefly if you can exactly what you mean by 17 fracture toughness --
18 DR. HISER: Fracture toughness --
19 CHAIRMAN MCDADE: -- as opposed to 20 strength.
21 DR. HISER: Well, I guess to separate them 22 maybe in two ways. Strength is in particular relevant 23 if you have an uncracked component. The failure of 24 that component will be directly related to the 25 strength of the material. So if you have an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5117 1 embrittled component or an irradiated component that 2 has no cracks, it will fail at a higher load than one 3 that is unirradiated. And going back to yesterday's 4 discussion, you can always have a higher load and 5 cause that uncracked component to fail.
6 For the fracture toughness, that really 7 relates to the response of a component that has a 8 crack in it. So, for example, if you have two 9 components that are identical, one is irradiated, one 10 is not, they both have the same size of crack, they 11 both are subjected to the same loads, the one that's 12 irradiated will have a lower fracture toughness, the 13 crack will grow more readily in that material than in 14 the unirradiated component. So that is where the 15 fracture toughness comes into place. If you have a 16 crack to start with.
17 CHAIRMAN MCDADE: Okay. And, Dr. Lahey, do 18 you agree with that?
19 DR. LAHEY: Yes, sir. I tried to say that, 20 he said it better than I. And I do want to bring 21 another thing up --
22 ADMIN. JUDGE WARDWELL: Before you go on to 23 that other thing, I want to clarify something else you 24 said to fix this point, if I might.
25 DR. LAHEY: Yes.
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5118 1 ADMIN. JUDGE WARDWELL: And that is, you 2 used the phrase that if it's embrittled and then it's 3 hit with a significant load. Would your same result 4 happen if it was hit with design basis loads? Do you 5 believe those are significant loads to cause the issue 6 that you're dealing with?
7 DR. LAHEY: Yes, many of them are. Many of 8 them are and --
9 ADMIN. JUDGE WARDWELL: That's all I want 10 to know is that it doesn't have to be an excessive 11 load, you believe this will happen under design basis 12 loads. Is that correct?
13 DR. LAHEY: That's correct.
14 ADMIN. JUDGE WARDWELL: Thank you.
15 DR. LAHEY: The difference between what 16 you've heard in the past and what I'm saying is, is 17 that when people have said that they can withstand 18 design basis loads, if you look at the EPRI document 19 and see what they've done, they've applied those in a 20 static way, not in an impulsive way. And I tried to 21 show you in my little cartoon yesterday that you can 22 have significant difference between static and dynamic 23 loads. I want to return to the document where Mr.
24 Lott was an author of. This is the Icone paper that 25 you had asked about. And in that one, the difference NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5119 1 from the one we had talked about in more detail --
2 CHAIRMAN MCDADE: Do you have an exhibit 3 number on that?
4 DR. LAHEY: Do you know the exhibit number?
5 Oh, yes -- NRC 000177.
6 CHAIRMAN MCDADE: Thank you.
7 DR. LAHEY: Okay. In there is a statement 8 which is a little different from what I heard before.
9 This is light water reactor conditions and it says 10 that the increase in tensile strength because of 11 irradiation causes a, in the high cycle fatigue 12 region, low amplitude, an increase in the strength 13 capability, less failure, and a decrease of ductility 14 results in a decrease of fatigue life in the low cycle 15 fatigue region. So this is entirely consistent with 16 a high pressure or the high flux reader reactor data, 17 which was at a higher temperature.
18 They go on to say that the strain 19 amplitude in their test was rather small, 0.6 percent, 20 so they only saw the increase, not the decrease. And 21 we're worried about the decrease, because that's an 22 indication if you have enough amplitude, you can cause 23 the thing to fail early and if you have a larger 24 amplitude because of a shock load, it can have a 25 catastrophic failure. So I wanted to bring that quote NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5120 1 to you. That's on Page 2 under the Effect of Fluence, 2 is the heading.
3 DR. LOTT: And may I, your honor? I just 4 want to say that, that statement about the high strain 5 amplitude data or the low cycle data reflected 6 primarily the limits of what was tested in this 7 program. And we were aware when we wrote the program 8 of exactly the data that Dr. Lahey has suggested here.
9 And just wanted to be cautious about how far we 10 extended the findings of this paper. And I would 11 stick by that. I will also point out, and I think 12 it's again something we may get to under the next 13 Contention, that we don't really expect to see these 14 large strain amplitudes in the reactor internals, at 15 least the reactor internals that have significant 16 irradiation effects.
17 MR. STROSNIDER: Your honor, this is Jack 18 Strosnider for Entergy.
19 ADMIN. JUDGE WARDWELL: I think we're going 20 to move on now, thank you.
21 MR. STROSNIDER: Going to move on? Okay.
22 ADMIN. JUDGE WARDWELL: Yes. Back to my 23 original question, which I really don't want to reread 24 again, but it was all those -- the statement that NRC 25 disagreed with your concerns and pointing out that --
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5121 1 one of the things they pointed out was primary 2 components to be inspected under Entergy's program are 3 those components that are most likely to be affected 4 by synergistic effects if they exist. Do you agree 5 that the primary components will be those most likely 6 to be affected by synergism?
7 DR. LAHEY: Yes. I don't disagree with 8 that.
9 ADMIN. JUDGE WARDWELL: Thank you.
10 DR. LAHEY: Their difference between how I 11 view it and how they view it is, is the cracking 12 required?
13 ADMIN. JUDGE WARDWELL: But if cracking 14 does occur, wouldn't that be a result of whatever 15 mechanisms created that? And so the synergism would 16 be built into that observed crack, would it not?
17 DR. LAHEY: I don't disagree what that 18 either. But remember, what I'm concerned with is 19 either before or after any of the significant surface 20 cracking. If you have an event which loads it 21 impulsively, you can fail the structure. And then 22 once you fail the structure, depending on what it is, 23 it can lead to an uncoolable geometry. That's what 24 I'm concerned with.
25 ADMIN. JUDGE WARDWELL: Okay, thank you.
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5122 1 I think you may have said this, but I'm going to ask 2 it again because maybe you didn't, so I want to make 3 sure we do cover this space. Are you concerned with 4 the interval of the inspections or do you feel that's 5 adequate based on the operating experience to date?
6 DR. LAHEY: You're going back to the ten 7 year --
8 ADMIN. JUDGE WARDWELL: Right.
9 DR. LAHEY: -- interval?
10 ADMIN. JUDGE WARDWELL: Exactly.
11 DR. LAHEY: It seems a little long, but I'm 12 more interested in a baseline inspection, a very 13 thorough baseline inspection as they go into the 14 period of extended operation, because otherwise, an 15 inspection process doesn't make a lot of sense if you 16 don't know where you start. So I think the sooner the 17 better they do that. And then after that, the 18 interval I think would depend on the kind of things 19 that were discussed. What would be the implications 20 of failure and how long do you have to take action 21 before you can have a big problem?
22 ADMIN. JUDGE WARDWELL: Right now, and I 23 believe they're planned for 2016 and 2019, if I was 24 correct in my memory. Do you have any comments in 25 regards to what you would recommend and what's your NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5123 1 justification for recommending anything earlier than 2 -- well, it's not going to happen before 2016 anyhow, 3 but the 2019 one?
4 DR. LAHEY: Well, we had recommended some 5 time ago, but time goes on, so the sooner the better.
6 But if that's the sooner they can do it, so be it.
7 ADMIN. JUDGE WARDWELL: Thank you. I guess 8 you, and as I heard you say, you're more concerned 9 with the extent of that inspection program rather than 10 necessarily any timing between the small years between 11 2016 and 2019?
12 DR. LAHEY: I think that's correct.
13 ADMIN. JUDGE WARDWELL: Thank you. Let's 14 move on now to talk a little bit more about the 15 preventative actions, the corrective actions, and 16 acceptance criteria. Entergy's testimony, 616, Answer 17 203, Page 136 to 137, Entergy has undertaken or will 18 implement several types of preventative actions to 19 manage the effects of aging on reactor vessel 20 internals at Indian Point, including implementing the 21 IPEC water chemistry control program, replacing the 22 split pins at IP2, committing to replace IP2 split 23 pins again in 2016, using the fatigue monitoring 24 program and addressing the action levels Number 8 in 25 the safety evaluation for MRP 227-A, tracking plant NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5124 1 transience and cycles, thereby assuring that fatigue 2 usage from actual plant transience does not exceed 3 ASME code design limits, and implementing neutron flux 4 reduction measures to minimize the neutron fluence on 5 the reactor pressure vessel, which in turn will 6 minimize radiation induced aging effects at high 7 fluence locations within the RVIs. I guess I'd ask 8 the Staff, do you agree that these are considered 9 preventative maintenance activities that are 10 designated by GALL to be part of a consistent AMP?
11 DR. HISER: This is Allen Hiser with the 12 Staff. The water chemistry program clearly is a 13 preventative measure that's implemented as part of 14 their AMP. The replacement of split pins, again, 15 clearly putting in new material would prevent the 16 accumulation of degradation that had occurred with the 17 old pins. Fatigue monitoring is, I'm not sure if I'd 18 call it preventative, but it clearly is an appropriate 19 measure to ensure that the fatigue life is adequately 20 monitored during the operation of the plant. So that 21 would -- I'm not sure if I would call it preventative, 22 but it -- prevent, mitigate, or minimize the effects 23 of aging, I'd say, yes, it's within that umbrella.
24 And finally the flux reduction, again, would help to 25 minimize the aging effects on the vessel internals.
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5125 1 So I would agree with each of these.
2 ADMIN. JUDGE WARDWELL: Okay, thank you.
3 In New York State's Exhibit 496, Attachment 1, which 4 is the AMP, Page 5, the AMP states that the reactor 5 vessel internals program is a condition monitoring 6 program that does not include preventative actions.
7 So I guess I'd ask, and I'll start with Entergy, 8 doesn't that kind of contradict what you've stated in 9 Answer 203 that I quoted from above where you're 10 taking credit for a bunch of preventative actions?
11 MR. COX: Could you repeat that question?
12 ADMIN. JUDGE WARDWELL: Yes. The previous 13 quote I had from your testimony of which I just asked 14 Staff to respond to, which was Answer 203, you listed 15 a number of preventative measurements that you were 16 taking credit for. Your AMP states that, on Page 5, 17 that the reactor vessel internals program is a 18 condition monitoring program that does not include 19 preventative actions. Well, it seems like you just 20 took credit for a bunch of preventative actions and 21 then you're saying in your AMP, it doesn't include 22 them.
23 MR. COX: This is Alan Cox for Entergy. I 24 think it's really a matter of semantics and how you 25 describe things. What was meant by the statement in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5126 1 the reactor vessel internals program was that the 2 water chemistry controls were included in another AMP.
3 I mean, there is a separate AMP that applies not just 4 to reactor vessel internals, but to water chemistry on 5 the primary side as a whole. So that AMP goes beyond 6 just reactor vessel internals. It's treated as a 7 separate program, but it is referenced from the 8 preventative action section here and it says that we 9 do have preventative actions in that AMP that will 10 apply to the reactor vessel internals. It's not a 11 part of the reactor vessel internals AMP in the way it 12 was described because it's a program that covers a 13 whole lot more areas than just reactor vessel 14 internals.
15 ADMIN. JUDGE WARDWELL: Okay, thank you.
16 I guess that explains it as best you can. And I guess 17 I'll turn to Staff. That Criteria 2 of GALL does 18 require preventative actions. So don't there need to 19 be some preventative actions within each individual 20 AMP for each of the different components that are 21 addressed by different AMPs, i.e., doesn't the reactor 22 vessel internals have to have preventative actions 23 associated with it in order to be consistent with 24 GALL?
25 DR. HISER: This is Allen Hiser of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5127 1 Staff. For the vessel internals program, that is 2 correct. Water chemistry is an essential element of 3 the program. Otherwise, the inspection types, 4 frequency, scope, et cetera, would be different, 5 because there's an expectation that there is a program 6 that's helping to minimize the effects of aging.
7 ADMIN. JUDGE WARDWELL: Would the, 8 likewise, the activity in regards to the fuel that 9 reduces the fluence, if you will, but the way it's 10 configured also be considered one of these 11 preventative measures that you're taking account of 12 and taking credit for in your evaluation of whether or 13 not Criteria 2 is met for the RVIs?
14 DR. HISER: This is Allen Hiser of the 15 Staff. I think the flux reduction is not explicit 16 within the GALL AMP, but it is explicit in MRP 227 17 that it was one of the three criteria used for 18 demonstrating that a plant was bounded by the report.
19 And based on that, it is sort of implicit to the GALL 20 AMP that, that is necessary. Again, if flux reduction 21 was not implemented at a plant, than it would be 22 inappropriate for them to use MRP 227 because that is 23 one of the fundamental assumptions because under the 24 aging evaluation, it's in the report, and to 25 demonstrate the adequacy of aging management.
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5128 1 ADMIN. JUDGE WARDWELL: And in your SER, 2 have you specifically addressed each of the criteria, 3 including Criteria 7 for corrective actions and 8 for 4 confirmation process dealing with assuring 5 preventative and corrective actions?
6 MR. POEHLER: This is Jeffrey Poehler of 7 the Staff. Yes, we have specifically addressed each 8 of the ten elements of the GALL Aging Management 9 Program, as modified by the LRISG.
10 ADMIN. JUDGE WARDWELL: Okay, thank you.
11 Entergy's testimony, 616, Answer 146, Page 95, states 12 that once a defect is discovered, its ability to 13 withstand fatigue and combinations of both normal and 14 accidental loads is evaluated by either fracture 15 mechanics analysis or a structural analysis, i.e., an 16 engineering evaluation, using the lower bound fracture 17 toughness, i.e., the evaluation assumes a bounding 18 level of embrittlement of the material.
19 Thus, the program has compensated for any 20 inability to directly determine the level of 21 embrittlement through a conservative assumption 22 employed during evaluation of inspection findings.
23 Thus, reasonable assurance that the effects of aging 24 will be adequately managed is provided without the 25 need for direct observation or measurement of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5129 1 level of embrittlement. I guess my question to 2 Entergy is, how was this bounding level of 3 embrittlement, as you referred to up above, 4 determined?
5 DR. LOTT: Basically, it was determined 6 based on published data on fracture toughness and, 7 effectively, analysis. Which is in -- I'm going to 8 look to my colleagues for the reference numbers, 727?
9 And published by the NRC in the NUREG process.
10 MR. GRIESBACH: 7207?
11 DR. LOTT: 7207, I'm sorry.
12 MR. GRIESBACH: Also MRP 210.
13 DR. LOTT: Yes. Maybe I should let Mr.
14 Griesbach talk.
15 MR. GRIESBACH: This is Tim Griesbach from 16 Entergy. That data and the evaluation method that 17 would be used as an example is given in MRP 210.
18 That's an EPRI MRP program document.
19 ADMIN. JUDGE WARDWELL: And you believe 20 that's an exhibit in this proceeding?
21 DR. LOTT: Yes.
22 MR. GRIESBACH: Yes, it is.
23 ADMIN. JUDGE WARDWELL: Okay. Thank you.
24 DR. LOTT: 646.
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5130 1 hear what the number is shortly from my ace searchers 2 of what that is.
3 MR. KUYLER: Your honor, MRP 210 is Entergy 4 Exhibit 646.
5 ADMIN. JUDGE WARDWELL: Oh, thank you so 6 much. I'd also like to -- in that statement, to 7 refresh your memory from three minutes ago, the 8 program has compensated for any inability to directly 9 determine the level of embrittlement through a 10 conservative assumption. And what is this 11 conservative assumption employed during the evaluation 12 of the inspection findings to which you refer that 13 ensures that the program has compensated for any 14 inability to directly determine the level of 15 embrittlement? And where might that be documented?
16 DR. LOTT: Again, that's documented and 17 it's required based on the procedures in WCAP-170986, 18 which was our internal, we'll be calling methodology 19 and data requirements for determinative engineering 20 evaluations. What we -- again, the bounding value is 21 based on a fracture mechanics analysis where we're 22 looking at this reduced fracture toughness, which is 23 a property of the material. In most cases, it's done 24 on the basis of a linear elastic evaluation, even 25 though we've already testified that it appears that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5131 1 there's some ductility left in these materials, it 2 assumes very low levels of ductility. And it bounds 3 the existing data on fracture toughness of material as 4 a function of fluence, for stainless steel alloys in 5 general.
6 ADMIN. JUDGE WARDWELL: Thank you. Dr.
7 Lahey, why hasn't this approach covered much of some 8 of your uncertainties that you are concerned with in 9 regards to the potential failure modes of what might 10 take place under operational conditions?
11 DR. LAHEY: Does your question -- this is 12 Richard Lahey from New York. Does your question also 13 include the previous discussion with the replacement 14 of split pins and that sort of thing?
15 ADMIN. JUDGE WARDWELL: Sure, if you want 16 to add that --
17 DR. LAHEY: Okay.
18 ADMIN. JUDGE WARDWELL: -- into it, you can 19 talk about that also.
20 DR. LAHEY: Those steps, I believe, are 21 quite prudent. We particularly like the replacement 22 of a degraded component and would highly encourage the 23 replacement of other degraded components, in 24 particular the baffle bolts, which are the most 25 vulnerable. The other part of your question, I guess, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5132 1 relates to the loads and are the loads which are 2 applied to determine the integrity or the time it will 3 take for a component to failure, if they are 4 appropriate loads, if they are impulsive loads, which 5 would have the maximum effect on causing a failure.
6 If they're absolutely correct, then I have no problem 7 with it. So far, I haven't seen that.
8 ADMIN. JUDGE WARDWELL: I think I was more 9 concerned with the bounding level of embrittlement 10 that was selected for this analysis and the 11 conservative assumption employed in evaluation of the 12 inspection findings.
13 DR. LAHEY: Okay. Maybe you can -- one of 14 the exhibits we have, New York State 000495, could we 15 bring that up and we talk about it?
16 ADMIN. JUDGE WARDWELL: What page number 17 would you like?
18 DR. LAHEY: It's Page 3.
19 ADMIN. JUDGE WARDWELL: Okay.
20 DR. LAHEY: Right. So this is a --
21 ADMIN. JUDGE WARDWELL: Well, wait just a 22 second until we get that --
23 DR. LAHEY: Oh, it's not up yet?
24 ADMIN. JUDGE WARDWELL: It's not up in 25 front of me at least.
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5133 1 DR. LAHEY: Oh, it's up here.
2 ADMIN. JUDGE WARDWELL: The ON/OFF button 3 being in the wrong position. We all -- none of us 4 have it, do we?
5 MR. HARRIS: Your honor, this is Brian 6 Harris for the Staff. We don't have it over here 7 either.
8 ADMIN. JUDGE WARDWELL: You don't have it 9 either?
10 MS. SUTTON: Neither does Entergy, your 11 honor.
12 MR. SIPOS: Your honor, for New York, we 13 have it up over here.
14 (Laughter.)
15 ADMIN. JUDGE WARDWELL: You seem pretty 16 smug about that, Mr. Sipos.
17 MR. SIPOS: It's a rare feeling.
18 (Laughter.)
19 ADMIN. JUDGE WARDWELL: You going to rent 20 us out a little sneak views of this or --
21 MR. SIPOS: I could turn the monitor 22 towards your honors direction if you like.
23 ADMIN. JUDGE WARDWELL: For the right 24 price?
25 (Laughter.)
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5134 1 ADMIN. JUDGE WARDWELL: These aren't 2 connected in series are they, like my Christmas 3 lights? Here we go. We've got a winner.
4 (Laughter.)
5 DR. LAHEY: Can you see it? Is it up on 6 your screen now?
7 ADMIN. JUDGE WARDWELL: It is, thank you.
8 DR. LAHEY: Okay.
9 ADMIN. JUDGE WARDWELL: Thank you for 10 waiting.
11 DR. LAHEY: Over there is it okay?
12 ADMIN. JUDGE WARDWELL: Is everyone happy?
13 MR. HARRIS: Yes.
14 MR. SIPOS: Yes, your honor.
15 DR. LAHEY: Okay. So this is a plot that 16 keeps getting passed around. It's in Gary Was's 17 classic book on nuclear metallurgy, it's in a lot of 18 EPRI reports, a lot of U.S. NRC reports. It came from 19 an individual and there's certain assumptions made in 20 it in terms of how you calculate the fluxes, the 21 neutron fluxes. And, in particular, what components 22 you're choosing. But it gives you a pretty good 23 estimate of the type of fluences you're going to see.
24 So on the upper curve is the fluence for greater than 25 one million electron volt neutrons, which are the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5135 1 damaging ones. And on the lower scale is the 2 corresponding displacements per atom, how many times 3 each atom is knocked out of its lattice. So you can 4 see, if you look at the end of PWR life extension, 5 it's expected to be greater than 100 dpa or greater 6 than ten to the 23 --
7 ADMIN. JUDGE WARDWELL: And what are the 8 years for, do you know, that they assume for this life 9 extension?
10 DR. LAHEY: This is the 20 year life 11 extension.
12 ADMIN. JUDGE WARDWELL: Okay. So the 60 13 years total?
14 DR. LAHEY: It's the type we're talking 15 about.
16 ADMIN. JUDGE WARDWELL: Okay, thank you.
17 DR. LAHEY: So you can see there's 18 significant, absolutely significant fluences and 19 damage at that point. And of course, that is where 20 you have to be really concerned because if you look at 21 the data on when bad things happen in terms of 22 embrittlement and all these other issues we've been 23 talking about, you have to get to a displacement per 24 atom of about one or so before bad things start to 25 happen. Then it drops off pretty fast, it really goes NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5136 1 down very fast.
2 So once you start getting out here, you're 3 really in a region where you have to be very careful 4 because you're on what I call the bathtub curve, if 5 you know what I mean by that. You're way out on the 6 other part of the bathtub curve where you're starting 7 to wear things out, you're really beating them up and 8 they're failing. So, hopefully this gives you some 9 insight into what the concerns are, just as a way to 10 benchmark yourself.
11 The NRC uses the criterion for significant 12 embrittlement, as I understand it anyway, you folks 13 can correct me if I'm wrong, but about one times ten 14 to the 21 for fluence, when for stainless steel you 15 start to get significant embrittlement. And other 16 people use like 6.7 times ten to the 20, but it's of 17 that order of magnitude. So you can see, by the end 18 of this thing we're a thousand times greater than the 19 onset of that kind of damage.
20 DR. LOTT: Your honor, may I just -- a 21 moment ago, I had trouble coming up with a reference 22 for you in terms of limiting fracture toughness 23 values. I wanted to point that actually those 24 limiting values, when I think about it, are in MRP 25 227, Section 6, as well as in the other documents we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5137 1 described. I'd also like to perhaps amend a little 2 bit to my statement and say that, one of the other 3 conservatisms in our analysis has been throughout, we 4 tend to apply peak fluence values to entire 5 components. So we don't necessarily -- and when we're 6 doing evaluations, when we've identified components, 7 we have looked at 60 year fluences and looked at the 8 peak location on that and there could be a large 9 gradient in fluence across the component. So I would 10 just suggest that, that's an additionally conservative 11 assumption we've made about determining the fluence on 12 a component.
13 ADMIN. JUDGE WARDWELL: Okay, thank you.
14 Do you have any other comments that you'd like to add 15 in regards to this figure that Dr. Lahey has presented 16 and his comments associated with it?
17 DR. LOTT: I mean, I think I've seen this 18 document many times before as well. It's just a 19 schematic that I think is consistent with actually 20 many of the assumptions we've made here in terms of --
21 I think Dr. Lahey was talking about the threshold 22 values for irradiation embrittlement of seven times 23 ten to the 20, that's exactly the value we used in our 24 evaluations, that's approximately one dpa. So, I 25 think that we're on the same -- we've dealt with and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5138 1 understand these concerns and that our testimony has 2 demonstrated that.
3 ADMIN. JUDGE WARDWELL: Clarify for me, 4 will you, if you are using seven times ten to the 20 5 or about one dpa, as what you assume as a conservative 6 fluence value, how does that relate to what you may 7 actually be experiencing, which is two orders of 8 magnitude higher than that by the end of the life 9 extension?
10 DR. LOTT: Well, first of all, what we're 11 looking at is the threshold for what's the lowest 12 value. I mean, so basically what that determines is 13 whether a component has any irradiation embrittlement 14 concern at all. Above that, all the way up to a 15 million, presumably, it has susceptibility. Our first 16 goal of the threshold values was to determine the 17 lower limit.
18 ADMIN. JUDGE WARDWELL: So it's 19 conservative because it's on the low side, 20 encompassing much more numbers of --
21 DR. LOTT: Right.
22 ADMIN. JUDGE WARDWELL: -- potential of --
23 DR. LOTT: Right, so yes. That's --
24 ADMIN. JUDGE WARDWELL: -- internals.
25 DR. LOTT: -- what the concern is and we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5139 1 identified components with potential irradiation 2 embrittlement. And our basis for that was a 3 relatively low value, the onset of embrittlement, 4 consistent with the curve I showed you earlier this 5 morning on yield stress.
6 ADMIN. JUDGE WARDWELL: Yes. Okay. Thank 7 you.
8 ADMIN. JUDGE KENNEDY: Dr. Lott, this is 9 Judge Kennedy. The baffle, let's see if I'm getting 10 this right, the baffle bolts are, at least components 11 I identify on this chart, at somewhere around ten to 12 the 22, PWR baffle bolt failures?
13 DR. LOTT: Yes. Some --
14 ADMIN. JUDGE KENNEDY: Are there --
15 DR. LOTT: Yes.
16 ADMIN. JUDGE KENNEDY: Are there reactor 17 vessel internal components that experience fluences 18 beyond that value?
19 DR. LOTT: Yes, there are. I mean, it's 20 not like that's a fall off the cliff value. When we 21 have -- and have reported seeing IASCC and baffle 22 bolts in operating plants, I think that, that value is 23 basically trying to demonstrate that it happened 24 roughly at that fluence.
25 ADMIN. JUDGE KENNEDY: Okay. All right, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5140 1 thanks.
2 CHAIRMAN MCDADE: Is this the baffle former 3 bolt or the baffle edge bolt or is this all baffle 4 bolts that they're talking about here?
5 DR. LOTT: Okay. There are basically three 6 different kinds of baffle bolts. There's the baffle 7 -- if you're familiar with the baffle structure, 8 they're plates that surround the core and there's 9 plates behind them, horizontal plates, that hold them 10 in place. The bolt between the baffle plate and the 11 horizontal plate is a baffle bolt. Where the two 12 plates come together, along the edge, along the seam, 13 there may be bolts that go from baffle plate to baffle 14 plate, not baffle plate to the former. Those are the 15 baffle edge bolts. They basically seal up the gap 16 between the two plates to keep water from jetting 17 through there. Not all plants even have baffle edge 18 bolts.
19 And we've done evaluations to determine 20 that in terms of holding the baffle together, in terms 21 of an accident type scenario, we don't take credit for 22 the baffle edge bolts at all. So it's, again, a 23 conservatism in our assumptions when we do these 24 acceptable bolting patterns. So baffle edge bolts.
25 And then there's barrel former bolts, which are from NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5141 1 the round barrel, which is the outside container of 2 the entire internals, that the horizontal plates, the 3 other side of them, attach to the barrel. So barrel 4 former bolts, baffle former bolts, and baffle edge 5 bolts.
6 MR. KUYLER: Your honor, there's a diagram 7 in Entergy's testimony, Exhibit 616, Page 54.
8 CHAIRMAN MCDADE: Okay, thank you. And 9 thank you, Dr. Lott.
10 ADMIN. JUDGE WARDWELL: NRC's testimony, 11 197, Answer 179, Pages 102 to 103, the reactor vessel 12 internals AMP is structured to managing the aging 13 effects such that the intended function of the 14 components will be preserved during the period of 15 extended operation from 40 to 60 years. It 16 accomplishes this task by establishing an Inspection 17 Plan for the relevant components that it's able to 18 identify potential aging effects prior to any loss of 19 function through appropriate schedules and 20 conservative acceptance criteria. And so I'll ask 21 Entergy, does Table 5-5 of the Applicant's Inspection 22 Plan, and that's New York State 496, Attachment 2, 23 contain these acceptance criteria?
24 MR. DOLANSKY: This is Bob Dolansky with 25 Entergy. Yes.
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5142 1 ADMIN. JUDGE WARDWELL: Okay. And does it 2 not state that the examination acceptance criteria for 3 visual examination is the absence of the specific 4 relevant condition?
5 MR. DOLANSKY: Yes.
6 ADMIN. JUDGE WARDWELL: And now the tougher 7 part, what are examples of this relevant condition?
8 And where would one find that?
9 MR. DOLANSKY: For an EVT1, it would be a 10 crack-like indication.
11 ADMIN. JUDGE WARDWELL: So it varies by 12 your inspection technique generally, rather than by 13 component?
14 MR. DOLANSKY: And by the component. In 15 other words, EVT1 is typically looking for cracking, 16 but a VT3 -- this is Bob Dolansky for Entergy. VT3 17 could be looking for either wear or it could be 18 looking for a dimensional change, like void swelling, 19 something like that. So the acceptance criteria would 20 depend on what you're looking for and the method that 21 you're using.
22 ADMIN. JUDGE WARDWELL: And where is that 23 documented anywhere in regards to the various 24 components so that we could turn to that and it would 25 say, yes, it's any indication of cracking or it's any NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5143 1 indication of a dimensional change or --
2 MR. DOLANSKY: Give me one moment, please.
3 ADMIN. JUDGE WARDWELL: If you want to get 4 back to us --
5 MR. DOLANSKY: Okay. If you -- I'll give 6 a cite, one second. New York State 496, Letter 12-7 037.
8 ADMIN. JUDGE WARDWELL: Okay.
9 MR. DOLANSKY: Table 5-5.
10 ADMIN. JUDGE WARDWELL Yes, that's what I 11 was referring to.
12 MR. DOLANSKY: That gives -- you'll see the 13 actual -- under examination acceptance criteria in the 14 table. For instance, for the upper core barrel 15 cylinder girth welds, the specific relevant condition 16 is a detectable crack-like surface indication.
17 ADMIN. JUDGE WARDWELL: Okay. So where 18 ever we see that, that's what it means. As soon as 19 you detect cracking, you're going to take some 20 corrective action.
21 MR. DOLANSKY: Right. So our inspection 22 procedure, that's the people actually doing the 23 inspection use a procedure, their acceptance criteria 24 for a recordable indication would be that.
25 ADMIN. JUDGE WARDWELL: Okay. Thank you.
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5144 1 MR. DOLANSKY: You're welcome.
2 ADMIN. JUDGE WARDWELL: That helps. Just 3 for completeness, Dr. Lahey, any comments in regards 4 to that acceptance criteria and where it's found and 5 the adequacy of it?
6 DR. LAHEY: This is Richard Lahey. I'm 7 sorry, I don't have that document in front of me and 8 I don't recall it.
9 ADMIN. JUDGE WARDWELL: Okay. You have no 10 comment?
11 DR. LAHEY: So I can't really answer right 12 now.
13 ADMIN. JUDGE WARDWELL: If you do later on, 14 if you do want to look at it later on and have some 15 comments --
16 DR. LAHEY: Okay.
17 ADMIN. JUDGE WARDWELL: -- remind me of it 18 and we'll be glad to. I want to make sure you have a 19 chance to anyhow.
20 DR. LAHEY: Thank you.
21 ADMIN. JUDGE WARDWELL: New York State's 22 testimony, 482, Page 79, states that, I believe that 23 the most vulnerable reactor pressure vessel internals 24 need to be carefully identified and repaired or 25 replaced prior to the extended operation since it is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5145 1 beyond the current state of the art to perform 2 realistic and accurate calculations on relocation of 3 failed RVP internals and the resultant potential for 4 core blockages or degraded core cooling. And I guess 5 I'll ask you, Dr. Lahey, that while I understand the 6 basis for your position in regards to this 7 replacement, does not this run counter to the whole 8 idea of managing aging? And doesn't it to a large 9 degree try to circumvent all of the regulations that 10 are geared towards aging management as opposed to 11 prescriptive replacements?
12 DR. LAHEY: That's a very interesting 13 question. In my opinion, aging management is indeed 14 important. But when you get components that look like 15 they're vulnerable and can fail and you're not able to 16 determine with any precision what the effect of that 17 might be, then I think the prudent thing to do is to 18 replace those components. I've spent ten years of my 19 life trying to calculate where things go and it's very 20 hard to do, very difficult to do, there's too many 21 possibilities.
22 But the one thing I know for sure is, once 23 you lose an intact geometry, you've got big problems.
24 So, anything that will preserve that, I think we're 25 way ahead of the game and, to me -- do you know what NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5146 1 a make-buy decision is? You decide is it cheaper to 2 make something or buy something? This is like a make-3 buy decision. Is it cheaper to calculate and inspect 4 and go through litigation and all that or is it 5 cheaper just to replace it and have the problem go 6 away? And, for many of these things, I think it's 7 cheaper and much more prudent to just replace it.
8 ADMIN. JUDGE WARDWELL: The -- I forgot my 9 question now because I was going to change it in 10 regards to your last comment. But with your 11 description of the details which basically are in the 12 AMP and the inspection program that's in there and 13 your applauding of it as not the complete what's 14 needed, but not really a lot of substantive 15 disagreements with the approaches and the extents of 16 what they're doing, given what we've heard in regards 17 to the conservative assumptions, why isn't there a 18 fair degree of reasonable assurance that something, if 19 it does go amiss, would be detected before that 20 intended functionality of the RVIs was lost?
21 DR. LAHEY: This is again Richard Lahey 22 from New York State. When you say conservative 23 assumptions, are you referring to embrittlement alone 24 now or --
25 ADMIN. JUDGE WARDWELL: I'm talking about NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5147 1 all that we've heard about --
2 DR. LAHEY: All of the above?
3 ADMIN. JUDGE WARDWELL: -- to date in 4 regards to our questioning and what's in their 5 testimony in regards to the decisions they make as 6 they prepared their AMP and then how they're 7 implementing their AMP.
8 DR. LAHEY: Right.
9 ADMIN. JUDGE WARDWELL: Does that not give 10 one reasonable assurance that if something does go 11 awry, it's not going to try to stop anything that 12 might go awry, but if it does, isn't there a 13 reasonable assurance that it would be detected prior 14 to the RVI losing all of its intended function, even 15 though it may crack or even do some other things?
16 DR. LAHEY: Not in my opinion. If you have 17 sufficient degraded components that can lead to a 18 destruction of intact geometry, there's a reasonable 19 chance that you can have an unexpected accident.
20 Accidents, by definition, you don't expect them, but 21 they happen, or an earthquake happens just when you 22 don't want it. And then what will that lead to? If 23 it can lead to degraded or materials which fail and 24 relocate, then I'm very concerned about it. Because 25 I just know you can't calculate the consequence of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5148 1 that.
2 So if you can identify materials that are 3 highly irradiated, and we can, and highly fatigued, 4 and we can, and they are things that could lead to 5 destruction of a coolable geometry, then you should 6 take action, the sooner the better. Don't wait for 7 something bad to happen. So that's why I'm very keen 8 on do what you're doing with the -- have Entergy do 9 what they're doing with the split pin, replace it 10 because it's degrading. Replace the things that have 11 a significant effect on the safety of the reactor.
12 There's not that many. I mean, we haven't talked 13 about pressure boundary components yet, but we will, 14 tomorrow I guess. And there's only a few that are 15 really, really crucial. And there's only a few things 16 in core that are really, really crucial. And, 17 happily, these are things that aren't that hard to 18 replace.
19 ADMIN. JUDGE WARDWELL: Thank you, Dr.
20 Lahey.
21 DR. LAHEY: That's why --
22 ADMIN. JUDGE WARDWELL: Thank you.
23 DR. LAHEY: -- I feel that way.
24 ADMIN. JUDGE WARDWELL: Entergy, have you 25 addressed anywhere or evaluated or how did you handle NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5149 1 this potential relocation of failed RVIs and the 2 resultant potential for core blockages and degraded 3 core cooling?
4 MR. DOLANSKY: This is Bob Dolansky with 5 Entergy. When we do an analysis that looks at 6 components, one of the requirements is that we 7 maintain core coolability and core geometry. I mean, 8 that's ultimately what we're trying to do. We want to 9 make sure that, that core -- that's the basis of the 10 whole thing is that the core stays coolable and the 11 geometry stays -- that it maintains core geometry. I 12 mean, that is what we do, that's the exact 13 requirements that we analyze for.
14 DR. LOTT: Yes, this is Randy Lott. And I 15 think, again, we kept coming back to it, but the 16 acceptable baffle bolting pattern analysis is a good 17 example of what we're doing. In that analysis, 18 effectively, you're looking to see that there are 19 enough baffle bolts to keep the baffles from moving, 20 interacting with the fuel, crushing fuel grids, or 21 interfering with the ability to drop control rods.
22 Which is really always our concern, it's maintaining 23 core coolability and being able to insert the control 24 rods, shut down the reaction. That's basically the 25 definition of the safety requirement.
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5150 1 MR. DOLANSKY: Right. And I just want to 2 clarify one thing that Dr. Lahey said. We are 3 replacing split pins. We're not planning to replace 4 baffle bolts. Replacing baffle bolts is not some no 5 nevermind, easy thing to do. They have lock bars 6 welded on, it's an extremely difficult area to get to.
7 Split pins were more designed that they might have to 8 be replaced. So the replacement of the split pins is 9 a little easier, it's more straightforward, something 10 that can be done much more easily.
11 It's easy to sit here and say, just 12 replace the baffle bolts. Actually replacing baffle 13 bolts, although it can be done, is not a simple 14 things. There's a lot of dose involved with that, 15 there's a lot of possibility of loose parts, and 16 there's consequences to replacing things that aren't 17 bad. So, the way we look at is, we will go out and do 18 these inspections on the baffle bolts using a 19 technique that's very good, difficult tooling that was 20 developed just to get us a better UT exam will 21 additionally -- when we do that inspection, as Dr.
22 Lott said, we're going to use an acceptable bolting 23 pattern analysis that ensures that if we find degraded 24 bolts that we can maintain core coolability, core 25 geometry. If we do that inspection and we find that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5151 1 there's very, very few bolts that are degraded, to me, 2 it doesn't make any sense to go out and just wholesale 3 replace baffle bolts. There's too much danger and 4 risk in doing that. If the bolts are good, I don't 5 see any reason to do that.
6 ADMIN. JUDGE WARDWELL: Okay, thank you.
7 CHAIRMAN MCDADE: Okay. And also, just as 8 our discussion moves on for the rest of the week, I 9 mean, we neither have the authority or interest in 10 micromanaging the way that Entergy does its 11 operations. You have to make certain business 12 decisions. Our function is just whether or not the 13 plans that you have put forward provide reasonable 14 assurance that these items will maintain their 15 intended function for the period of extended 16 operation.
17 So, certain issues of whether or not as a 18 matter of policy you replace, that's outside the scope 19 of what we're looking at. Again, the scope of what 20 we're looking at is whether or not the plans that you 21 have put forward provide that reasonable assurance 22 with regard to intended function, period. So, it's 23 not as wide -- it's not just an open-ended discussion.
24 We're trying to focus in on what we have to decide 25 here. Judge Wardwell?
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5152 1 DR. LAHEY: Yes, and your honor, can I say 2 something? What we just heard from Entergy is the 3 most encouraging thing I've heard in the last eight 4 years. I'm very happy to hear it because it sounded 5 like they're going to do something that makes a lot of 6 sense, look at the integrity of it, and if it's not a 7 big issue, then don't do it, if it is a big issue, 8 presumably they will entertain that as a possibility 9 to replace. Up until now, all I heard was, question, 10 do you agree with Dr. Lahey on anything and the answer 11 is, no. Everything, no, no, no.
12 (Laughter.)
13 DR. LAHEY: And so this was encouraging to 14 me.
15 ADMIN. JUDGE WARDWELL: Thank you, Dr.
16 Lahey. NRC's Exhibit 197, Answer 134, Page 82, states 17 that the Staff found Entergy's AMP met the Staff's 18 guidance for corrective actions because detected 19 conditions not satisfied in the examination acceptance 20 criteria will be processed through the plant's 21 corrective action programs. And I guess I'll start 22 with the Staff considering it's your exhibit. What's 23 your understanding of how the plant's corrective 24 action program works and interacts with the AMP for 25 reactor vessel internals?
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5153 1 DR. HISER: This is Allen Hiser of the 2 Staff. The corrective action program would be 3 initiated if there's an inspection finding that 4 exceeds the inspection acceptance criteria, which 5 we've discussed previously. I believe it's Table 5-5 6 of the RVI Inspection Plan from Indian Point. The 7 corrective action program would require that the 8 condition be assessed, I guess in maybe two or three 9 different ways.
10 One is, is the condition that was 11 identified, that condition needs to be resolved, is it 12 acceptable maybe through engineering evaluation, is 13 repair required, is replacement required? So that 14 would be one path that would be followed by the 15 corrective action program. A second would be a 16 consideration of expansion of the inspections 17 consistent with the AMP. So that would follow.
18 Consideration of reinspection interval would be a part 19 of that evaluation as well. So the corrective action 20 program should consider pretty much any aspect that's 21 relevant to the finding itself, be it for that finding 22 or other components.
23 ADMIN. JUDGE WARDWELL: Okay, thank you.
24 DR. HISER: And that is required in 25 Appendix B to 10 CFR Part 50, so it is a regulated NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5154 1 activity, a regulated program of the plant that the 2 NRC inspects periodically.
3 ADMIN. JUDGE WARDWELL: Thank you for your 4 understanding of the plant's corrective action. I'll 5 turn to Entergy and can you now explain how your 6 plant's corrective action program actually works and 7 interacts with the AMP or just state what differences 8 you might have with what Dr. Hiser just presented?
9 MR. AZEVEDO: Yes, this is Nelson Azevedo 10 for Entergy. What was just discussed is correct, 11 that's what we do. So we find an indication then we 12 put in our corrective action program to determine 13 whether we have to repair or replace it or whether 14 it's acceptable. We also do extended condition 15 inspections if it's warranted. So those are the 16 things we look at.
17 MR. DOLANSKY: I just want to -- this is 18 Bob Dolansky with Entergy. Just to add on something 19 I said earlier where I said Westinghouse was 20 developing acceptance criteria for us. The acceptance 21 criteria in the plan are the just evidence of a crack-22 like indication. That would be in the inspection 23 procedure. We're additionally, right now, getting 24 additional acceptance criteria that if there was a 25 crack and it could be a certain size, that's what's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5155 1 being developed. So I misspoke a little bit before, 2 that's not the acceptance criteria in here. That's 3 the --
4 ADMIN. JUDGE WARDWELL: As it exists now.
5 MR. DOLANSKY: -- acceptance criteria after 6 it enters our corrective action program.
7 ADMIN. JUDGE WARDWELL: Okay, thank you.
8 Ready to rock and roll.
9 CHAIRMAN MCDADE: Okay. This may be a good 10 time to break for lunch. Do you have anything before 11 we break?
12 ADMIN. JUDGE KENNEDY: No, I do not.
13 CHAIRMAN MCDADE: Okay. The question is, 14 how long we break for lunch? And I'm not sure if 15 there's any difficulty in people getting lunch within 16 a relatively short period of time. I would propose to 17 come back at 1:30. Is that going to give enough time 18 for people to get lunch and take care of what they 19 need to? NRC?
20 MR. HARRIS: Yes, your honor.
21 CHAIRMAN MCDADE: Entergy?
22 MS. SUTTON: Yes, your honor.
23 CHAIRMAN MCDADE: New York?
24 MR. SIPOS: New York, yes, your honor.
25 CHAIRMAN MCDADE: Riverkeeper?
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5156 1 MS. BRANCATO: Yes, thank you.
2 CHAIRMAN MCDADE: Okay. Do any of the 3 witnesses perceive a problem with restarting at 1:30?
4 Apparently not, we're in recess. We'll come back --
5 DR. LAHEY: No, your honor.
6 CHAIRMAN MCDADE: -- at 1:30. Thank you.
7 (Whereupon, the above-entitled matter went 8 off the record at 12:19 p.m. and resumed at 1:34 p.m.)
9 CHAIRMAN MCDADE: The hearing will come to 10 order. There are a couple of, or a few matters I 11 guess, out there that I just wanted to address before 12 we move forward. I believe there was a question, Mr.
13 Dolansky, to you, as to whether or not you could 14 provide other examples of monitoring of aging effects 15 not observable by inspection. Were you --
16 MR. DOLANSKY: Just the one. Just the one 17 that I gave.
18 CHAIRMAN MCDADE: Okay. And there was a 19 question to Mr. Poehler about an EPRI report and 20 whether or not that was a publically available report.
21 MR. POEHLER: Yes, this is Jeffrey Poehler 22 of the Staff. That report is publically available and 23 it's Exhibit NRC 000208, A through E, there's five 24 parts.
25 CHAIRMAN MCDADE: Okay. And what is the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5157 1 EPRI report number?
2 MR. POEHLER: It's an MRP letter actually, 3 MRP --
4 CHAIRMAN MCDADE: Okay.
5 MR. POEHLER: -- Letter 2014-09.
6 CHAIRMAN MCDADE: Okay. Thank you.
7 MR. POEHLER: You're welcome.
8 CHAIRMAN MCDADE: And, Dr. Lahey, you were 9 asked with regard to questions on Table 5-5 of New 10 York 496. Did you have an opportunity to review that?
11 DR. LAHEY: No, your honor, I haven't been 12 able to get access to it yet.
13 CHAIRMAN MCDADE: Okay. At our next break, 14 we will make sure that you get access to that.
15 DR. LAHEY: Okay. Thank you.
16 CHAIRMAN MCDADE: Mr. Sipos, it looked like 17 you had a question.
18 MR. SIPOS: I had a procedural question, 19 your honor. And it concerns scheduling of New York's 20 second expert, Dr. David Duquette, who is an expert on 21 Contention 38. And I was wondering if the Board could 22 provide any guidance as to when you would like to --
23 or when the Board might start Contention 38? And it's 24 purely a logistical question as to when --
25 CHAIRMAN MCDADE: No, I understand.
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5158 1 MR. SIPOS: -- to tell him to get in the 2 car and --
3 CHAIRMAN MCDADE: And my best estimate at 4 this point in time would be our start time on Thursday 5 morning. That we're still, it's Tuesday afternoon, 6 we're still working on 25, 26 we should get started 7 maybe later today, should run most if not all of 8 tomorrow. And a lot of what had been in 38, is in 38, 9 we have touched in on the testimony on 25 and will 10 touch on some more in the testimony of 26. So, from 11 my standpoint, and I haven't, you just asked the 12 question, I haven't discussed it with my colleagues, 13 I would, for planning purposes, plan on starting on 38 14 on Thursday morning at 8:00.
15 ADMIN. JUDGE WARDWELL: Yes, I guess we 16 just talk out loud. Yes, I would concur, in fact, to 17 the point that even if we got done 26 early tomorrow, 18 it would probably still be later in the day, that we 19 could just not plan on starting 38 until Thursday 20 morning.
21 MR. SIPOS: That's very helpful, thank you.
22 CHAIRMAN MCDADE: Dr. Kennedy?
23 ADMIN. JUDGE KENNEDY: That's fine with me.
24 ADMIN. JUDGE WARDWELL: And do you agree 25 with that, that we just wouldn't --
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5159 1 CHAIRMAN MCDADE: Yes. I mean, I think for 2 planning purposes, we're certainly not going to get 3 done with 26 early tomorrow. So, if we did finish 4 with 26 before 6:00, I don't think anybody would 5 complain if we leave it at 5:50 instead of 6:00. And 6 having Dr. Duquette get here for a relatively short 7 period of time --
8 ADMIN. JUDGE WARDWELL: You want to check 9 with the other parties to make sure they're 10 comfortable with that too? If we waste a couple hours 11 tomorrow afternoon by not starting 38, is that --
12 CHAIRMAN MCDADE: I mean, I don't think 13 it'll be wasting a couple of hours, I think it'll be 14 a few minutes, if anything. And we may still be on 15 26. You all can gauge how well we are keeping a 16 schedule. Ms. Sutton?
17 MS. SUTTON: Yes, your honor. That's fine, 18 we don't have any issues with that.
19 MR. HARRIS: The Staff has no issues with 20 that, your honor.
21 CHAIRMAN MCDADE: Okay.
22 MS. BRANCATO: Riverkeeper has no issues, 23 thanks.
24 MR. SIPOS: Thank you.
25 CHAIRMAN MCDADE: Okay.
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5160 1 ADMIN. JUDGE WARDWELL: Well, I have no 2 more questions on 25.
3 MR. SIPOS: We'll start 38?
4 CHAIRMAN MCDADE: There's no questions on 5 26?
6 MR. SIPOS: All right, let's start 38.
7 CHAIRMAN MCDADE: We're in recess until 8 Thursday morning.
9 (Laughter.)
10 ADMIN. JUDGE WARDWELL: Now, we're going to 11 start looking at some specific materials and 12 components associated with the RVI and starting off 13 with control rods and J-groove welds. Entergy 14 testimony, 616, Answer 98, Page 56, the Inspection 15 Plan NL12-037 Attachment 2 at 62 through 64, provides 16 a complete and corrected list of the RVI subassemblies 17 at Indian Point and breaks those subassemblies down to 18 their constituent components. New York State's 19 testimony, 482, Page 13, states the control rods and 20 the associated components are very important RPV 21 internals and their integrity is an extremely 22 important safety concern. In my opinion, omitting the 23 control rod assemblies and associated fittings from an 24 RPV internals Aging Management Program is a serious 25 and indefensible omission.
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5161 1 Entergy's testimony, 616, Answer 99, Page 2 56, states that control rods are not subject to aging 3 management review for two reasons. First, they 4 perform their intended function with moving parts or 5 a change in configuration. Thus, as the NRC Staff 6 concluded, the control rods are active components not 7 subject to aging management review. Second, control 8 rods are considered consumables and the NRC has 9 excluded from the license renewal review process those 10 components that are subject to replacement based on a 11 qualified life or a specified time period. And I 12 guess I'd ask Dr. Lahey, do you now agree that control 13 rods are not subject to aging management review?
14 DR. LAHEY: I agree that's the rule that 15 has been put in place. I still have the concern that, 16 because I'm looking at everything through the prism of 17 reactor safety, so I have the concern that if you have 18 a significant shock load, you can fracture these 19 highly, highly embrittled structures and they will 20 relocate in some way and you don't know how. And they 21 don't care if they're moving or not, I mean, they're 22 going to be at risk. So I think the reason for the 23 rule is because of what I call silos. Things are 24 being thought of quite separately, rather than in an 25 integrated way.
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5162 1 ADMIN. JUDGE WARDWELL: Thank you.
2 CHAIRMAN MCDADE: But, Dr. Lahey, as I 3 understand the position of Entergy and the NRC Staff, 4 these are consumables. Therefore, they don't need an 5 aging management plan, you are of necessity going to 6 be replacing them at set intervals. So, what would an 7 aging management plan consist of since you're already 8 going to replace them?
9 DR. LAHEY: No, I understand that's why 10 they view it the way they do. What I'm saying though 11 is, let's say a week before they're going to replace 12 them and they're in really bad shape in terms of 13 embrittlement, you have an event which causes them to 14 fail. This causes a big problem. So, I personally 15 believe that when you look at aging management, you 16 ought to look at all the things that affect the safety 17 of the plant, the safety of the plant during the 18 extended operation.
19 CHAIRMAN MCDADE: But the handling of those 20 control rods would be no different during the period 21 of extended operation than during the original 22 license. They'd be subject to the same replacement 23 criteria, would they not? So why does this have to do 24 with the aging management and whether or not there 25 should be a period of extended operation as opposed to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5163 1 just the current operating license of the plant?
2 DR. LAHEY: Well, I'm concerned with what 3 happens during the period of extended operation. And 4 this is a possible event that can happen, it's no 5 different than some of the static components. And 6 just because you can replace them doesn't mean that 7 they couldn't be at risk if you have an event which 8 causes a significant shock load.
9 CHAIRMAN MCDADE: Okay. But my question 10 is, why would that be any different during the period 11 of extended operation than it is during the original 12 licensing period?
13 DR. LAHEY: Oh, I misinterpreted your 14 question. It wouldn't, not for the control rods 15 themselves.
16 CHAIRMAN MCDADE: Okay. Thank you. Dr.
17 Wardwell?
18 ADMIN. JUDGE WARDWELL: Entergy's 19 testimony, 616, Answer 100, Page 57, "the control rod 20 guide tube assemblies, including the guide plates 21 (CRGTs) and the lower flange welds are subject to 22 aging management review and the effects of aging on 23 these components are managed through the RVI AMP."
24 And they're citing again NL12-037 Attachment 2 at 4 to 25 5 and Attachment 1 at 6 to 8. And those two NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5164 1 attachments, if we remember, are the Inspection Plan 2 and the AMP, respectively and that's New York State 3 496. Dr. Lahey, therefore, is incorrect when he 4 asserts that Entergy has claimed that the guide tubes, 5 plates, pins, and welds associated with control rods 6 are not RVIs. Dr. Lahey, do you now agree that the 7 control rod guide tube assemblies, including the guide 8 plates and the lower flange welds are in fact subject 9 to aging management review and part of the AMP for 10 RVIs?
11 DR. LAHEY: Yes, I believe they are.
12 ADMIN. JUDGE WARDWELL: Okay.
13 DR. LAHEY: And they should be.
14 ADMIN. JUDGE WARDWELL: Thanks. New York 15 State's testimony, 482, Page 45, Lines 3 through 9, 16 because of geometric considerations, many pressure 17 water reactors, including IP2 and IP3, cannot meet the 18 U.S. NRC's required minimum coverage for the non-19 destructive testing of so-called J-groove welds. And, 20 thus, the integrity of these important CRD stub tube 21 welds cannot be directly confirmed by inspection.
22 I'll start with you, Dr. Lahey, and what do you mean 23 by the CRD stub tube welds?
24 DR. LAHEY: Well, this is one of the issues 25 we talked about this morning, when we were talking NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5165 1 about the possibility of a leakage forming and boron 2 and water and forming boric acid and that sort of 3 thing. So this is the weld on the inside. And as I 4 understand it, they cannot get a complete inspection 5 of this.
6 ADMIN. JUDGE WARDWELL: Okay. And what 7 does that CRD stand for or --
8 DR. LAHEY: Oh, I'm sorry.
9 ADMIN. JUDGE WARDWELL: No, that's true, I 10 appreciate what you've answered so far, that helped.
11 But I also want to know what is the --
12 DR. LAHEY: Control rod drive.
13 ADMIN. JUDGE WARDWELL: Got you. Got you, 14 okay. Thank you. Entergy's testimony on 616, Answer 15 101, Page 57 and 58, the reactor pressure vessel head 16 penetration nozzle welds, sometimes referred to as J-17 groove welds, are not RVIs or even part of the reactor 18 vessel pressure, but instead part of the reactor 19 pressure vessel head. Aging effects applicable to the 20 J-groove welds on the, and this must mean control rod 21 drive M, head penetrations are managed under the 22 reactor vessel head penetration inspection AMP. So, 23 Entergy, what is the M of the control rod drive stand 24 for?
25 MR. COX: The M -- this is Alan Cox for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5166 1 Entergy. The M stands for mechanism.
2 ADMIN. JUDGE WARDWELL: Okay, mechanism.
3 And, so, back to you, Dr. Lahey, and I think we 4 covered this morning, do you now agree that these are 5 managed under a different Aging Management Program and 6 are not really part of this Contention?
7 DR. LAHEY: Well, they are reactor vessel 8 internals. I mean, they're in the vessel, they're 9 subject to aging. How they want to deal with it is 10 not a great concern to me, as long as it's dealt with.
11 ADMIN. JUDGE WARDWELL: Good, thank you.
12 Let's move on to a couple of materials we're dealing 13 with. The first one we'll deal with is wrought 14 Austenitic stainless steel. Entergy's testimony on 15 Exhibit 616, Answer 105, Page 61, says that the 16 majority of the Indian Point reactor vessel internal 17 components are fabricated from wrought Austenitic 18 stainless steels. Even though an increase in strength 19 and decrease in toughness do occur when exposed to 20 neutron irradiation, these materials retain their 21 resistance to fast fracture within the operating 22 temperature of interest for PWRs. The only exceptions 23 are wrought Austenitic stainless steel materials with 24 high amounts of cold working, and that's greater than 25 20 percent of cold working, but these materials are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5167 1 not present at Indian Point. And I think my first 2 question for Entergy would be that, what do you mean 3 by this cold working and what percentage do you have 4 here at Indian Point?
5 DR. LOTT: Cold working is basically the 6 rolling of the material or the stretching of the 7 material, the working of the material to increase its 8 yield strength. The more you -- for instance, you 9 would take a plate and reduce the area by rolling it, 10 the higher the strength becomes. In general, the 11 reactor internals are not cold worked beyond about 15 12 percent, they're controlled. That's one of our 13 screening criteria that we have been in fact 14 evaluating for.
15 ADMIN. JUDGE WARDWELL: Okay. Thank you.
16 And, Dr. Lahey, do you have any issues with those 17 statements made in their original testimony or what 18 was -- as amplified here?
19 DR. LAHEY: Are you asking me if I agree 20 with the percentage of cold work they have or --
21 ADMIN. JUDGE WARDWELL: Yes. And with 22 that, why they don't see the issue with the Austenitic 23 steel that is exhibited when cold working is greater 24 than 20 percent.
25 DR. LAHEY: But my understanding is, maybe NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5168 1 I misunderstood the statement you read, it was focused 2 on that stainless steel is not subject to stress 3 corrosion or embrittlement if it's not cold worked.
4 Did I misunderstand that?
5 ADMIN. JUDGE WARDWELL: No, this dealt 6 mostly -- well, let me repeat it again. The majority 7 of the reactor vessel components are fabricated from 8 wrought Austenitic stainless steel. Do you agree with 9 that?
10 DR. LAHEY: I agree with that, absolutely.
11 ADMIN. JUDGE WARDWELL: And even though an 12 increase in strength and decrease in toughness do 13 occur when exposed to neutron irradiation --
14 DR. LAHEY: Okay.
15 ADMIN. JUDGE WARDWELL: -- these materials 16 retain their resistance to fast fracture within the 17 operating range of interest for PWRs.
18 DR. LAHEY: I don't agree with that. I 19 mean, certainly, there are components that we've 20 talked about that are highly embrittled and do not 21 satisfy that.
22 ADMIN. JUDGE WARDWELL: Okay. And they say 23 that the only exception are those wrought Austenitic 24 stainless steels with high amounts of cold working, 25 greater than 20 percent. And would you agree that if NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5169 1 you had cold rolling steels greater than 20 percent, 2 they wouldn't necessarily show an increase in strength 3 or decrease in toughness and retain their resistance 4 to fast fractures within the operating temperatures?
5 DR. LAHEY: No, I would not.
6 ADMIN. JUDGE WARDWELL: You would not? So 7 you think they still would retain theirs even if you 8 did have cold working --
9 DR. LAHEY: Yes, I think if they're highly 10 embrittled, they are subject to fracture.
11 ADMIN. JUDGE WARDWELL: Right. So then you 12 would agree that, that is an exception from their 13 abilities to resist fast fracturing?
14 DR. LAHEY: I guess. I mean, I think if 15 you have a material that's stainless steel and it has 16 fluence above a certain level, and we talked about 17 that level this morning, it becomes embrittled and 18 then if subjected to the right kind of load, it can in 19 fact fail. And we can talk about dimple failure, 20 fracture versus --
21 ADMIN. JUDGE WARDWELL: We will in a 22 minute.
23 DR. LAHEY: -- to me that's semantics, it 24 failed.
25 ADMIN. JUDGE WARDWELL: Thank you.
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5170 1 CHAIRMAN MCDADE: A quick question, this is 2 probably going to seem very basic to you all. We've 3 been talking about Austenitic stainless steel, what 4 does that mean? What is Austenitic stainless steel?
5 Dr. Lahey or Dr. Lott or anybody can offer what that 6 definition is as opposed to just stainless steel.
7 MR. GORDON: This is Barry Gordon from 8 Entergy. Austenitic stainless steel is a -- it's 9 named after the fellow who discovered the micro-10 structure. It's a face centered cubic structure.
11 Austenitic stainless steel is not magnetic, a magnet 12 will not stick to it, and it can only be strengthened 13 by cold working. You can't heat treat it like you can 14 some other alloys, like low alloy steel, you can only 15 strengthen it by cold working it to some extent.
16 CHAIRMAN MCDADE: What are the different 17 characteristics of Austenitic stainless steel as 18 opposed to just your typical stainless steel? Are 19 there significant differences?
20 MR. GORDON: Well, there is a ferritic 21 stainless steel and there's a martensitic stainless 22 steel. Martensitic stainless steels can be hardened 23 and that's what -- like you have a stainless steel 24 cutlery at home, those are martensitic stainless 25 steels. And they'll stick, you have a magnetic board NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5171 1 and you can put your knives on it. But Austenitic 2 stainless steel is not a magnetic material.
3 CHAIRMAN MCDADE: Okay. For our purposes 4 right here in the way that it reacts, not how you put 5 it together, not how you work it, but how it reacts 6 over time, is there any specific distinction with it 7 being Austenitic stainless steel?
8 MR. GORDON: It's a very ductile material, 9 it can be embrittled, like Dr. Lahey said, by 10 irradiation, and it can be strengthened by cold 11 working also, and also by irradiation.
12 CHAIRMAN MCDADE: Okay, thank you.
13 ADMIN. JUDGE WARDWELL: As a follow-up --
14 MR. GRIESBACH: Your honor, this is Tim 15 Griesbach --
16 CHAIRMAN MCDADE: I'm sorry, who?
17 ADMIN. JUDGE WARDWELL: A follow-up 18 question on that, which is extremely important to 19 cover. When I buy some stainless steel fittings and 20 stuff, sometimes a magnet will stick, to say, washer 21 fenders and sometimes it won't. Now, it won't be 22 strong, but some of them there's nothing and other 23 times there is and, yet, it's supposed to be a 24 stainless steel. What am I getting?
25 MR. GORDON: It really depends. There's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5172 1 six families of stainless steels and there's also a 2 new family called the super stainless steels that have 3 more alloying elements in it. If you go to like a 4 Best Buy --
5 ADMIN. JUDGE WARDWELL: Yes, I'm going to 6 an Ace Hardware.
7 MR. GORDON: -- that's supposed to be a 8 stainless steel refrigerator and you can bring a 9 magnet if you want -- you really probably want 10 Austenitic stainless steel and that's probably what 11 you want in your sink also. But if you can bring a 12 magnet along, you can see if it's really Austenitic or 13 maybe ferritic stainless steel.
14 ADMIN. JUDGE WARDWELL: But, yet, I've 15 bought these fender washers --
16 MR. GORDON: Yes.
17 ADMIN. JUDGE WARDWELL: -- and some of them 18 it won't stick and other times it'll stick a little 19 bit, it won't stick heavy, but it's not going to be 20 like a --
21 MR. GORDON: Well, some of them, if you 22 cold work Austenitic stainless steel, you get this 23 diffuseness reaction, which you get martensite in the 24 Austenitic stainless steel. Part of it will be 25 transfer -- and you can measure this magnetically.
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5173 1 And this means you're -- it's a way of even measuring 2 the amount of cold working you have by how much 3 martensite you have in the Austenite.
4 ADMIN. JUDGE WARDWELL: So when I bring my 5 magnet to my local Ace Hardware dealer, I can impress 6 him, finally --
7 MR. GORDON: Yes. I mean --
8 ADMIN. JUDGE WARDWELL: -- instead of 9 looking like a fool in the hardware store with my 10 magnet go, oh, well, that one's had some cold working 11 and this is pure --
12 MR. GORDON: Right.
13 ADMIN. JUDGE WARDWELL: -- Austenitic 14 steel.
15 MR. GORDON: That's right. And steel comes 16 --
17 ADMIN. JUDGE WARDWELL: Thank you.
18 MR. GORDON: -- these manufacturers make a 19 stainless steel looking thing, but it's actually made 20 out of carbon steel, but they just put a finish on it.
21 So bring your magnet.
22 ADMIN. JUDGE WARDWELL: I thought we 23 weren't going to get anything out of this hearing and 24 I have been proven wrong.
25 (Laughter.)
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5174 1 ADMIN. JUDGE WARDWELL: Especially when I 2 tell him, gee, I wonder if this is a cast Austenitic 3 stainless steel or a CASS. Which brings us a nice 4 segue into NRC's testimony, 197, Answer 161, Page 92, 5 where the A/LAI, Action Item 7 requires an applicant 6 or a licensee to perform a plant specific analysis of 7 cast RVI components to demonstrate the components will 8 remain capable of performing their intended functions 9 during the period of extended operation. I guess my 10 question for Entergy is that citing your testimony on 11 Answer 175, Page 114, does not the Action Level 7 12 require that this analysis account for the potential 13 loss of fracture toughness of the components due to 14 both thermal embrittlement and irradiation 15 embrittlement?
16 DR. LOTT: Give me a minute to get 17 organized here. Can you give me the citation again?
18 ADMIN. JUDGE WARDWELL: Sure, the citation 19 that it was citing was your testimony, Answer 175, 20 Page 114 in response to that. I'm just asking, does 21 not the A/LAI Number 7, which again you see in MRP 227 22 or in the Aging Management Plan, I may be able to find 23 that, but does not that require, A/LAI 7, does that 24 not require that the analysis account for the 25 potential loss of fracture toughness of the components NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5175 1 due to both thermal embrittlement, or TE as we'll call 2 it, and irradiation embrittlement, IE as we'll call 3 it?
4 MR. DOLANSKY: This is Bob Dolansky for 5 Entergy. The answer's yes.
6 ADMIN. JUDGE WARDWELL: Say again?
7 MR. DOLANSKY: The answer is yes.
8 ADMIN. JUDGE WARDWELL: Okay. It does --
9 MR. DOLANSKY: Yes.
10 ADMIN. JUDGE WARDWELL: -- require that, 11 okay. New York State in their testimony, 576, Page 5, 12 Lines 1 through 17, and through Page 6, Lines 11 13 through 20, that in regards to NUREG/CR-7184, New York 14 State contends that the following observations support 15 previous opinions and testimony. And one of those, 16 the first observation is that for cast materials, 17 synergies may exist between TE, the thermal 18 embrittlement, and the irradiation embrittlement.
19 And, two, embrittled cast materials were observed to 20 experience transgranular brittle cleavage and ductile 21 tearing. And I guess the first question I will have 22 for Entergy is, what does transgranular brittle 23 cleavage and ductile tearing look like and how are 24 they formed and when are they formed and when are they 25 an issue?
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5176 1 DR. LOTT: Okay. We talked earlier about 2 Austenitic stainless steels and ferritic stainless 3 steels, or ferritic steels in general. For instance, 4 a reactor pressure vessel steel is a ferritic steel.
5 It's subject to a brittle fracture at low temperatures 6 that cuts across the grain because it's a very flat 7 surface, it's essentially a cleaving of the crystal 8 structure in the grain, a very brittle failure. That 9 does not tend to happen in the Austenitic stainless 10 steels where the failures remain ductile, they fail by 11 stretching of the material and eventually finding 12 small dimples or ruptures in the material and pulling 13 it apart. So there's a difference in the fracture 14 process between a cleavage failure, which is a very 15 low ductility, lower shelf reactor pressure vessel 16 kind of failure, and the failures we see normally in 17 stainless steels.
18 ADMIN. JUDGE WARDWELL: But say again, what 19 is that transgranular brittle cleavage look like? It 20 looks like a cut surface or a sheared surface or --
21 DR. LOTT: Yes. There's no deformation on 22 the surface, it's very clearly flat and sometimes 23 stepped.
24 ADMIN. JUDGE WARDWELL: So it just looks 25 like a break in the surface?
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5177 1 DR. LOTT: Yes, it's a very -- it's a clean 2 break, I guess is the best way to say it.
3 ADMIN. JUDGE WARDWELL: A crevice in a rock 4 type of thing?
5 DR. LOTT: What's the best way to describe 6 this?
7 ADMIN. JUDGE WARDWELL: Or a crack in a 8 rock?
9 DR. LOTT: Yes. Certain rocks you would 10 fail by cleavage, that's true.
11 ADMIN. JUDGE WARDWELL: And so what is 12 ductile tearing? Is that what the --
13 DR. LOTT: Ductile tearing is the manner in 14 which a crack in a ductile material would advance. So 15 when this begins to fail, you begin to slowly -- it 16 would continue to deform or continue to show some --
17 reconnected and you'd form small voids that would 18 grown into little pockets that would give you these 19 dimpled rupture effects on the surface of the 20 specimen.
21 ADMIN. JUDGE WARDWELL: And so usually 22 you'd say this ductile tearing is associated with 23 dimples at the surface?
24 DR. LOTT: Yes. You can tell by looking at 25 the surface that it's failed in this manner, under a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5178 1 microscope, not necessarily -- well, you might be able 2 to tell by eye.
3 ADMIN. JUDGE WARDWELL: Okay. And this 4 dimpled surface will appear before or after the 5 tearing? I mean, can't you just tell by the split in 6 the tearing or do you see the tear or --
7 DR. LOTT: Well, you can tell, obviously, 8 by the load -- if you're tearing this part and you're 9 measuring the load displacement, you can see that it's 10 continually taking new force to pull it apart.
11 Whereas a brittle fracture, a cleavage fracture, will 12 obviously be real sudden.
13 ADMIN. JUDGE WARDWELL: So a transgranular 14 brittle cleavage would be a definitive break, is that 15 a fair assessment? Where the ductile tearing, you may 16 not see a separation of the material, it just may be 17 a necking down and possibly this dimpling on the 18 surface that you talk about?
19 DR. LOTT: Well, eventually you will form, 20 after it's necked down, you'll begin eventually to 21 form cracks and those cracks will have these dimpled 22 ruptured surface. We're talking about the fracture 23 surface of the specimen, when it separates and you 24 look at it, that's where you'd see the dimples.
25 ADMIN. JUDGE WARDWELL: Okay, thank you.
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5179 1 MR. STROSNIDER: Your honor, this is Jack 2 Strosnider from Entergy. And maybe this will help you 3 visualize it. When you have a cleavage fracture, it's 4 flat because when it's transgranular, it's going along 5 the atoms. Basically it's a shiny, when you look at 6 that surface when it fails, it's shiny because it's 7 very flat. When you have the ductile failure, because 8 of the dimpled, because you have to break little 9 pieces of material, it's got a duller surface and you 10 can see it's not the flat sort of surface that you 11 would see in a cleavage fracture.
12 ADMIN. JUDGE WARDWELL: But both of them 13 are failure surfaces though that you're looking at?
14 MR. STROSNIDER: Oh, yes.
15 ADMIN. JUDGE WARDWELL: Okay.
16 MR. STROSNIDER: It comes apart.
17 ADMIN. JUDGE WARDWELL: Great. Thank you.
18 Now, do you agree with New York's statements in 19 regards to the observations from this NUREG 7184 that 20 the Cast materials synergies may exist between thermal 21 embrittlement and irradiation embrittlement and, two, 22 the embrittled Cast materials were observed to 23 experience these two types of failures?
24 DR. LOTT: I'm not sure that I -- I don't 25 think I do agree with the term synergism. I think NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5180 1 that has a meaning to me that I don't see proven in 2 any of the data. I do agree that irradiation 3 embrittlement and thermal embrittlement may both --
4 either one may happen in a Cast material.
5 ADMIN. JUDGE WARDWELL: Either one or both?
6 DR. LOTT: Well, certainly a material can 7 be subject to both these conditions at the same time.
8 It's just not clear that they interact to me 9 synergistically.
10 ADMIN. JUDGE WARDWELL: Okay. Dr. Lahey, 11 I want to ask you, did the NUREG 7181 use the term 12 synergy between them or how did you -- what did they 13 say in regards to the reference, I guess, to your 14 first statement that for Cast materials synergies may 15 exist between TE and IE?
16 DR. LAHEY: Your honor, this is Richard 17 Lahey, New York. I would actually have to go back and 18 look at it to see if it used that word. This word is 19 used often, in fact it was one of the things that 20 Judge McDade gave me as a homework to read something 21 and it asked about the synergy between these two 22 effects for cast stainless steel. And so it's not 23 something I made up. It may or may not be in NUREG 24 7184, I'd have to check.
25 ADMIN. JUDGE WARDWELL: So that could be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5181 1 your wording in regards to interpreting what they said 2 --
3 DR. LAHEY: Yes. I mean, my --
4 ADMIN. JUDGE WARDWELL: -- in regards to 5 the interaction between the TE and the IE?
6 DR. LAHEY: My personal belief is that you 7 certainly can have both of them and the effect can be 8 larger than each one separate. That's what I mean by 9 synergy.
10 ADMIN. JUDGE WARDWELL: And say that again, 11 what you mean by synergy?
12 DR. LAHEY: Synergy means that they 13 reinforce each other and the combination is more than 14 each effect separate.
15 ADMIN. JUDGE WARDWELL: But wouldn't it be 16 more than just the sum also?
17 DR. LAHEY: Yes.
18 ADMIN. JUDGE WARDWELL: Because that would 19 be the synergy --
20 DR. LAHEY: That's what I meant, I'm sorry.
21 ADMIN. JUDGE WARDWELL: -- you take this 22 TE, you take IE and yes it's more than either one of 23 them --
24 DR. LAHEY: Right.
25 ADMIN. JUDGE WARDWELL: -- but likewise, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5182 1 it's more than the sum of the whole.
2 DR. LAHEY: Exactly.
3 ADMIN. JUDGE WARDWELL: Okay. Thank you.
4 DR. LOTT: Your honor --
5 ADMIN. JUDGE WARDWELL: Yes.
6 DR. LOTT: -- I think in that document, 7 7184, I have some notes on it here. It basically did 8 show that irradiation, relatively low doses 0.08 dpa, 9 can trigger a response in the material that similar to 10 thermal embrittlement, these materials that are 11 normally subject to thermal embrittlement. And what 12 happened effectively was that thermal embrittlement by 13 itself would produce an effect, the irradiation 14 produced a similar effect without the thermal 15 embrittlement, materials that were combined showed 16 about the same. So that's why I'm concerned that they 17 were not necessarily a synergistic effect. Yes, there 18 was an effect of thermal embrittlement, yes, there was 19 an effect of irradiation embrittlement, but to me, the 20 suggestion that it's a synergistic effect would say 21 that it could be larger combined than it is 22 individually.
23 ADMIN. JUDGE WARDWELL: Okay thank you.
24 CHAIRMAN MCDADE: Are you saying that it 25 isn't or you don't know?
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5183 1 DR. LOTT: I just don't -- it doesn't meet 2 my definition for what I would call synergistic.
3 CHAIRMAN MCDADE: Okay. As I understood 4 that, the definition you would use is that it's larger 5 than the sum of the individual parts.
6 DR. LOTT: Right. If I had both, in the 7 specimen if I had both irradiation and thermal 8 embrittlement, the decrease was about the same as it 9 was just due to irradiation and thermal embrittlement 10 alone. I think that's Figure 142 in the document.
11 That's probably 14.2, my handwriting is very bad, I'm 12 sorry.
13 ADMIN. JUDGE WARDWELL: And, Dr. Lahey, I 14 will leave you with this comment, that if you would 15 like to explore 7184 and come up with the wording that 16 they use that would support the synergistic effects of 17 this, meaning that there's some words in there that 18 state that the TE and the IE amplify one another to 19 the point that the effect is greater than the sum of 20 the two, then feel free to present that to us --
21 DR. LAHEY: Yes, sir.
22 ADMIN. JUDGE WARDWELL: -- whenever you 23 have a chance to review that further within the 24 context of why we're here this week.
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5184 1 though, I've read that in a number of Argonne reports 2 from researchers at Argonne.
3 ADMIN. JUDGE WARDWELL: Well, I'm mostly 4 interested in -- I'm only interested in this regard in 5 --
6 DR. LAHEY: Okay.
7 ADMIN. JUDGE WARDWELL: -- regards to this 8 as applied to 7184. Thank you.
9 DR. LAHEY: Thank you.
10 CHAIRMAN MCDADE: Okay. And, Dr. Lott, you 11 referenced Figure 142? Is that what you said?
12 DR. LOTT: We're looking it up right now.
13 Can we get back to perhaps in a --
14 CHAIRMAN MCDADE: Sure.
15 ADMIN. JUDGE WARDWELL: Entergy's 16 testimony, 616 again, Answer 106, Page 64, although 17 stainless steel and nickel alloy RVI materials are 18 also subject to irradiation embrittlement, they do not 19 undergo a ductile to brittle transition or fail by 20 brittle cleavage, even though the neutron exposure 21 levels are much higher than those of the vessel.
22 However, at fluences above the MRPA 175 screening 23 threshold, it is recognized that these Austenitic 24 stainless steels will experience decreases in fracture 25 initiation toughness and in the resistance to ductile NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5185 1 tearing. These effects have been explicitly 2 considered in the MRP 227-A guidelines and in the RVI 3 AMP implementation at Indian Point. And I guess my 4 question for Entergy, I just want to confirm that your 5 use of the word brittle cleavage is the same as this 6 transgranular brittle cleavage that we were talking 7 about earlier?
8 MR. STROSNIDER: Your honor, this is Jack 9 Strosnider for Entergy. That's correct, but --
10 ADMIN. JUDGE WARDWELL: Thank you.
11 MR. STROSNIDER: -- I think the point here 12 that needs to be understood, the point that's being 13 made, we continue to discuss about highly embrittled 14 materials. These Austenitic materials are embrittled 15 and there is a reduction in fracture toughness. The 16 thing you need to understand is when they test that 17 fracture toughness, it still shows ductility, it's not 18 failing by cleavage fracture. They're developing 19 what's called a J-R curve, which requires some 20 ductility in order to fail it. So, I just want to put 21 this highly embrittled in context. There's 22 embrittlement, but there's still ductility in these 23 specimens, they're not failing by cleavage fracture, 24 they don't go through the transition that the ferritic 25 materials go through. It's very important in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5186 1 understanding how these materials can respond to 2 various loads, including the accident loads.
3 ADMIN. JUDGE WARDWELL: And in this Answer 4 106 on Page 64, you say that they do not undergo 5 ductile to brittle transition or failure by, I'm 6 inserting now, the transgranular brittle cleavage, 7 correct?
8 MR. STROSNIDER: Right.
9 ADMIN. JUDGE WARDWELL: So they don't fail 10 by brittle cleavage, but they do still have resistance 11 to ductile tearing, is that correct?
12 MR. STROSNIDER: Jack Strosnider for 13 Entergy. Yes, that is correct.
14 ADMIN. JUDGE WARDWELL: Thank you. So in 15 summary, the aging effects of RVIs do not include 16 transgranular brittle cleavage, but do include ductile 17 tearing. But how about some of the other effects that 18 we've talked about? And I think they include 19 cracking, dimensional changes, wearing, dimpling, 20 stress relaxation, and void swelling cracking. Do 21 they exhibit those types of effects? And we can stay 22 with you, Mr. Jack, I can't see your name so I can't 23 --
24 MR. STROSNIDER: Jack Strosnider for 25 Entergy. Yes, they do. Those are the type of effects NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5187 1 that are identified in the program and managed by the 2 program.
3 ADMIN. JUDGE WARDWELL: Okay, thank you.
4 Entergy's testimony, 722, Answer 8, Page 5, says the 5 existing research also suggests that combined thermal 6 aging and irradiation of representative CASS materials 7 does not appear to lower toughness below what is 8 expected for thermal embrittlement alone. And I guess 9 I'd ask, what's the basis for this statement? Of 10 anyone from Entergy that would like to respond.
11 DR. LOTT: Let me first respond by 12 correcting my previous figure number. It's Figure 98 13 on Page 142. I had them switched in my notes.
14 ADMIN. JUDGE WARDWELL: Okay. Thank you.
15 DR. LOTT: And in fact, that figure I think 16 actually makes the same point that he just said and 17 that I said earlier, which was that the effect of 18 thermal embrittlement and irradiation embrittlement in 19 these high ferrite materials, and we haven't talked 20 about necessarily high ferrite and low ferrite 21 materials, these relatively high ferrite materials do 22 show thermal aging and irradiation caused a similar 23 effect.
24 MR. GRIESBACH: Your honor, this is Tim 25 Griesbach from Entergy. I think one thing that we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5188 1 should point out is that the materials we're talking 2 about at Indian Point in the lower support columns are 3 known to be low in delta ferrite. It's that delta 4 ferrite content that causes the thermal embrittlement 5 that we keep referring to and the materials in these 6 plants are known to not be susceptible to thermal 7 embrittlement based on their low delta ferrite. So, 8 by saying that, they should also not be affected by 9 the synergistic effects of thermal and irradiation 10 embrittlement.
11 ADMIN. JUDGE WARDWELL: Okay. Let me, I 12 guess -- I'm getting swamped by too much information 13 here, I think. And then diversion over to one chart 14 and then I didn't know whether they answered my 15 question or not is the way I interpreted it. You're 16 testimony, Entergy's testimony, 722, Answer 8, Page 5, 17 says the existing research also suggests that the 18 combined thermal aging and irradiation of 19 representative CASS materials does not appear to lower 20 toughness below what is expected for thermal 21 embrittlement alone. And my question to you is, what 22 is the basis for that statement?
23 DR. LOTT: And, again, I think it's 24 illustrated -- I was trying to answer your question 25 with my previous answer. In the data that was shown, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5189 1 it shows exactly what --
2 ADMIN. JUDGE WARDWELL: And I'm sorry, I 3 can't --
4 DR. LOTT: The data that's shown in that 5 figure shows that you start -- and again, for three 6 different materials, all high ferrite materials, where 7 thermal embrittlement would potentially happen, they 8 were either irradiated or thermally treated and the 9 decrease in toughness in the J-integral toughness was 10 basically the same in both cases. In other words, the 11 thermal embrittlement and the irradiation 12 embrittlement and the three of them together all gave 13 very similar results.
14 ADMIN. JUDGE WARDWELL: But the cast 15 materials that we're dealing with aren't high ferric 16 are they?
17 DR. LOTT: The cast materials we're dealing 18 are low ferrite, so we would not expect to see thermal 19 embrittlement, nor would we expect to see any large 20 amount of irradiation embrittlement by the same 21 mechanism.
22 ADMIN. JUDGE WARDWELL: So why does that 23 figure, which deals with high ferric, provide a basis 24 for your statement that the existing research suggests 25 that combined thermal aging and irradiation of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5190 1 representative CASS materials did not appear to lower 2 toughness below what is expected for thermal 3 embrittlement alone?
4 MR. STROSNIDER: This is Jack Strosnider 5 for Entergy and let me see if I can clarify this.
6 ADMIN. JUDGE WARDWELL: Sure.
7 MR. STROSNIDER: I think the first part of 8 the answer that Dr. Lott gave is a generic discussion 9 about the laboratory data --
10 ADMIN. JUDGE WARDWELL: Fine.
11 MR. STROSNIDER: -- which includes high 12 delta ferrite materials. When we talk about the 13 material at Indian Point, in particular the lower 14 columns, they are low delta ferrite not susceptible to 15 thermal aging, and, therefore, this -- what you see in 16 the generic data is really not applicable at Indian 17 Point because of the specific material that's at 18 Indian Point. So a generic part of the response and 19 a plant specific part of the response.
20 ADMIN. JUDGE WARDWELL: And so the plant 21 specific, you've made a statement that, again, I'll 22 read for the third time. The existing research also 23 suggests that combined thermal aging and irradiation 24 of representative CASS materials, which is what you 25 have at Indian Point, correct?
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5191 1 MR. STROSNIDER: Read that again, please?
2 MR. GRIESBACH: This is Tim Griesbach. The 3 --
4 ADMIN. JUDGE WARDWELL: CASS material, you 5 have -- you're dealing with CASS material. For those 6 CASS materials at Indian Point, do not appear to lower 7 toughness below what is expected for thermal 8 embrittlement alone. That was your statement. My 9 question is, where did that come from?
10 MR. DOLANSKY: This is Bob --
11 ADMIN. JUDGE WARDWELL: What is the basis 12 for it? Point me to a graph that demonstrates not 13 what it isn't, I want to see a graph for what it is or 14 some other discussion of why you came up with that 15 statement or how you came up with that statement.
16 MR. DOLANSKY: This is Bob Dolansky with 17 Entergy. I believe the graph that Dr. Lott pointed to 18 contains CF-8 material and that is --
19 ADMIN. JUDGE WARDWELL: Whoa, what's CF-8 20 material? Now we've got another material.
21 (Laughter.)
22 MR. DOLANSKY: -- which that is what we 23 have at Indian Point.
24 ADMIN. JUDGE WARDWELL: That's CASS?
25 MR. DOLANSKY: That's CASS and it's --
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5192 1 ADMIN. JUDGE WARDWELL: Great.
2 MR. DOLANSKY: -- low delta ferrite.
3 ADMIN. JUDGE WARDWELL: Yes, great.
4 MR. DOLANSKY: That is, I believe, and Dr.
5 Lott can correct me if I'm wrong, but that figure that 6 he gave you contains the low delta ferrite CF-8 7 material that's CASS that we have at Indian Point.
8 ADMIN. JUDGE WARDWELL: Okay. And you're 9 referring to Figure 98 on Page 142?
10 MR. DOLANSKY: Yes.
11 ADMIN. JUDGE WARDWELL: Okay. Because by 12 the time we get with putting the record together and 13 look at the transcript, we want to make sure we have 14 it.
15 MR. DOLANSKY: This figure here.
16 ADMIN. JUDGE WARDWELL: And so it's on our 17 screen now?
18 DR. LOTT: Yes.
19 ADMIN. JUDGE WARDWELL: Okay. Right? Is 20 that the one you're referring to?
21 MR. DOLANSKY: Yes.
22 DR. LOTT: Yes.
23 ADMIN. JUDGE WARDWELL: Okay. So how does 24 that support this statement that CASS materials do not 25 appear to lower toughness below what is expected for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5193 1 thermal embrittlement alone?
2 DR. LOTT: Well, I'm not sure what part of 3 your question I'm missing. I mean, if you look at the 4 data here, both irradiation and aging produce a 5 similar decrease in toughness and the material that is 6 both irradiated and aged has a very similar, 7 particularly in the CF-8 material, is a similar 8 toughness. Now, in the CF-8 materials in Indian 9 Point, we don't expect to see these decreases due to 10 thermal aging because the CF-8 material there is low 11 ferrite.
12 MR. COX: This is Alan Cox with Entergy.
13 And let me give you my layman's interpretation of what 14 this drawing shows. It shows --
15 ADMIN. JUDGE WARDWELL: Sure, just get a 16 little closer to your mic though because I can hear 17 you better.
18 MR. COX: The tall bar on the graph is the 19 material at the beginning and then the arrows show --
20 ADMIN. JUDGE WARDWELL: And we're looking 21 at the blue tall bar, which we're going to focus only 22 on the CF-8, is that correct?
23 MR. COX: Sure. I'm just -- the response 24 is similar in all three cases. But if you --
25 ADMIN. JUDGE WARDWELL: Sure.
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5194 1 MR. COX: -- the center one, the blue tall 2 bar is the starting condition. The dotted line that 3 says irradiation shows the decrease in the fracture 4 toughness as you irradiate the material. The dotted 5 line that says aging is the thermal aging. And you 6 can see it lowers -- both of those effects cause a 7 drop in the fracture toughness. The third bar, you 8 see another irradiation on the dotted line that goes 9 from the shorter blue bar to the pink/red colored bar 10 in front, that's the additional effect, that's 11 basically the combined effect of the --
12 ADMIN. JUDGE WARDWELL: Got you.
13 MR. COX: -- thermal aging and then the 14 irradiation. And you see it has a slightly lower 15 value than the irradiation alone, a little bit lower 16 than the thermal alone, but it's certainly not greater 17 than the sum of the two effects.
18 ADMIN. JUDGE WARDWELL: And what is CF-3 19 material?
20 DR. LOTT: It's just a different 21 specification of cast stainless steel.
22 ADMIN. JUDGE WARDWELL: Okay. So all of 23 these are cast stainless steel materials?
24 DR. LOTT: Yes.
25 ADMIN. JUDGE WARDWELL: Dr. Lahey, any NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5195 1 comments on how these graphs apply to what CASS 2 materials appear to do in regards to lowering 3 toughness below what is expected for thermal 4 embrittlement alone when you combine thermal aging 5 with irradiation of those?
6 DR. LAHEY: Just a request for a little bit 7 more information from Entergy. My understanding is 8 the delta ferrite in IP2 is like 14.6 percent, is that 9 correct?
10 ADMIN. JUDGE WARDWELL: Well, why don't I 11 ask the question. You tell me what you're interested 12 in and -- I want to make sure you're not --
13 DR. LAHEY: Well, the screening criteria is 14 15 percent and I'm trying to understand if that's what 15 they mean by low delta ferrite, the 14.6.
16 ADMIN. JUDGE WARDWELL: What do you mean by 17 low delta ferrite?
18 MR. AZEVEDO: Yes, your honor. This is 19 Nelson Azevedo for Entergy. It is true that both 20 Units 2 and 3 have low delta ferrite as measured by 21 the screening criteria 15 percent. So they're both 22 below 15 percent.
23 ADMIN. JUDGE WARDWELL: And do you have any 24 idea how far below?
25 MR. AZEVEDO: Yes. Unit 2 is 14 and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5196 1 change, Unit 3 is a little bit lower, like 11, 12.
2 ADMIN. JUDGE WARDWELL: Okay. Thank you.
3 Dr. Lahey, any other comments or --
4 DR. LAHEY: Yes. My comment is I wish my 5 colleague Dr. Duquette was here because this is his 6 field, he's a world class metallurgist. And he would 7 have a lot of comment on this and all I can give you 8 is secondhand information because this is not my 9 field.
10 ADMIN. JUDGE WARDWELL: But he's not even 11 a witness for this Contention, so even if he was here, 12 he would be a spectator.
13 DR. LAHEY: Okay. Well, I can tell you 14 what he would say if you want to hear that.
15 ADMIN. JUDGE WARDWELL: No. No, I'd like 16 to hear --
17 CHAIRMAN MCDADE: Just tell us what you 18 would say.
19 ADMIN. JUDGE WARDWELL: -- what your 20 professional interpretation might be of anything, but 21 only from what you're comfortable saying in regards to 22 your professional background.
23 DR. LAHEY: Well, concerning the data that 24 we have on the screen, the data speaks for itself. If 25 this is good data, that's what it is.
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5197 1 ADMIN. JUDGE WARDWELL: Okay. Thank you.
2 CHAIRMAN MCDADE: Okay. It may speak for 3 itself, but it doesn't speak loud enough for me to 4 understand it. Okay. We start off --
5 ADMIN. JUDGE WARDWELL: You're never going 6 to understand this chart.
7 (Laughter.)
8 CHAIRMAN MCDADE: Okay. The low delta 9 ferrite, start from the premise that it's below 15 10 percent, it's 14 or 11 percent. How does that inform 11 your conclusion here?
12 DR. LAHEY: Well, if you want me to give 13 you some secondhand information, then I can do that.
14 But I can't give you any firsthand information other 15 than looking at the graph and what it shows to me is 16 there's obviously no synergism shown in this data.
17 CHAIRMAN MCDADE: Okay. There's no 18 synergism shown in this data, but from your 19 experience, do you believe when you have the low delta 20 ferrite that you would expect synergism?
21 DR. LAHEY: Not necessarily. I think it 22 depends on how the duplex structure is arranged. If 23 it's arranged in a certain way, it can have a 24 different effect than if it's arranged in a different 25 way. And unfortunately the only way you know that is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5198 1 with a destructive examination. So, I think you do 2 have to rely on data that shows, here -- when I am 3 talking about synergism, there's a number of other 4 experiments, not necessarily for these type of 5 percentages of delta ferrite, which do show synergism.
6 But it may be, apparently is, for the materials used 7 in Indian Point, this is the result you get.
8 CHAIRMAN MCDADE: So you're saying that 9 based on data that you have observed, if the 10 percentage was not 11 percent, but was 20 percent or 11 40 percent, it would have a different effect on your 12 conclusion with regard to synergism?
13 DR. LAHEY: That's my understanding of the 14 reading that I've done, yes.
15 CHAIRMAN MCDADE: Okay, thank you.
16 ADMIN. JUDGE KENNEDY: Dr. Lahey, this is 17 Judge Kennedy. If I looked at the material CF-3, look 18 at the data there, does that show more synergism than 19 CF-8 does?
20 DR. LAHEY: Well, if I understand the way 21 they're doing it, it would show less. The one on the 22 right, the little pink or red or whatever color that 23 is on the right is lower than the one on the left. So 24 it's a little odd that the thermal aging plus 25 irradiation would give you lower, but it depends on NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5199 1 the level of irradiation, I suppose.
2 ADMIN. JUDGE KENNEDY: All right, thank 3 you.
4 ADMIN. JUDGE WARDWELL: Yes. I guess that 5 raises a question I hadn't really come up with. What 6 is this J-integral that we're plotting on the Y axis?
7 DR. LOTT: It's a measure of the ductile 8 fracture toughness.
9 ADMIN. JUDGE WARDWELL: So that's the 10 measure of the toughness?
11 DR. LOTT: The toughness being the 12 resistance to crack initiation. And, again, I'll use 13 that word, let me qualify. In a J-integral test, you 14 would start with a crack specimen and you would 15 basically be measuring how much it would take to 16 reinitiate and grow that crack.
17 ADMIN. JUDGE WARDWELL: To reinitiate and 18 what that crack?
19 DR. LOTT: Grow it.
20 ADMIN. JUDGE WARDWELL: Grow that crack.
21 So what do low values of J mean? That it's not 22 susceptible to cracking or susceptible to crack or --
23 DR. LOTT: It's not the susceptibility to 24 cracking, it's susceptibility -- the impact of 25 cracking on the ability to maintain a load.
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5200 1 ADMIN. JUDGE WARDWELL: On the ability --
2 so what are --
3 MR. STROSNIDER: This is Jack --
4 ADMIN. JUDGE WARDWELL: -- the low values 5 mean?
6 MR. STROSNIDER: This is Jack Strosnider 7 for Entergy, let me see if -- first of all, in terms 8 of the J-integral and what it is, it can be related to 9 the amount of energy that's required to allow the 10 crack to tear through the material, the preexisting 11 crack. It can be related to the energy for that. So 12 the lower the J value, the less energy it takes. But 13 this J-integral approach was developed specifically 14 for ductile materials. If the material is not 15 ductile, if it's going to fail in the cleavage mode 16 that we talked about earlier, there's a different 17 measure that's used for that. So, in this case, it's 18 related to energy, the lower that value, the lower the 19 energy. But in every case, it's showing some 20 ductility.
21 ADMIN. JUDGE WARDWELL: And so, while the 22 CF-8 shows that we reach the same level of the J-23 integral with irradiation as we do with aging and 24 irradiation, the CF-3 material shows that we end up 25 with a lower value with the combination of the aging NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5201 1 and the irradiation --
2 MR. STROSNIDER: Right.
3 ADMIN. JUDGE WARDWELL: -- than just 4 irradiation, which means less energy is needed to grow 5 the crack. Am I interpreting that correctly?
6 DR. LOTT: I supposed you could interpret 7 it that way. I must admit, I think it's within the 8 scatter in the data, within the accuracy of the data.
9 To make that conclusion would be difficult in my mind.
10 CHAIRMAN MCDADE: Okay, Dr. Lott, excuse 11 me. Could you either move yourself closer to the 12 microphone or the microphone closer to you?
13 DR. LOTT: Okay, sorry.
14 CHAIRMAN MCDADE: Okay, thank you.
15 ADMIN. JUDGE WARDWELL: So you think that 16 the data is plus or minus almost 100 percent because 17 it seems to be about half of what it is before and if 18 that's the noise, then that's accuracy of what we're 19 dealing with here in this graph? Can you see how I 20 reached that conclusion?
21 DR. LOTT: Yes, I see how you reached that 22 conclusion. It's hard for me to answer that question.
23 ADMIN. JUDGE WARDWELL: Fine, thank you.
24 DR. LOTT: Again, I think it's important to 25 point out that basically the materials that we're NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5202 1 dealing with, we don't expect to see large amounts of 2 thermal embrittlement at all. Now, these materials 3 are higher in ferrite and we would expect to see that 4 happen. I think all materials measured at least at 5 levels greater than 20 percent.
6 ADMIN. JUDGE WARDWELL: Boy, you had me 7 right to the very end. These materials, what are you 8 referring to as these materials?
9 DR. LOTT: The materials in this graph, the 10 CF-3, the CF-8, and the CF-8M. The particular 11 materials that were tested.
12 ADMIN. JUDGE WARDWELL: But all of these --
13 now we're back to where we started I think.
14 DR. LOTT: I'm sorry.
15 ADMIN. JUDGE WARDWELL: What is the ferric 16 content of these materials on this graph?
17 DR. LOTT: The ferrite content in all three 18 cases I believe is greater than 20 percent, measured.
19 MR. DOLANSKY: Your honor, maybe I could 20 help a little bit. This is Bob Dolansky with Entergy.
21 CF-8 can have a range of delta ferrite. Indian Point 22 3 has CF-8 materials. We actually went and did the 23 research, pulled our CMTRs, our delta ferrite was 24 below 15 percent at both IPEC units. That doesn't 25 mean that all CF-8 is below 15 percent. Does that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5203 1 help?
2 ADMIN. JUDGE WARDWELL: And so, if I looked 3 at the rest of this --
4 MR. COX: This is Alan Cox --
5 ADMIN. JUDGE WARDWELL: -- Exhibit 488B, is 6 that what it is? And that is an 83 page report, it 7 would tell me what the ferric content of this is?
8 DR. LOTT: I believe so.
9 MR. COX: Judge Wardwell, this is Alan Cox 10 for Entergy. The paragraph immediately above this 11 graph says that all the samples tested were high delta 12 ferrite samples. And the report may say somewhere 13 what that is in terms of a number, but it does say 14 it's high delta ferrite.
15 DR. HISER: Dr. Wardwell, the information 16 you're looking for is in New York State 488A, Page 5, 17 Table 1, provides chemical composition of the 18 materials. And for the CF-3, it's measured 24 19 percent, CF-8 measured 23 percent, and the 8M is 28 20 percent.
21 ADMIN. JUDGE WARDWELL: So would -- now I'm 22 going to go back to Entergy. Thank you for that, Dr.
23 Hiser, from the Staff. Back to Entergy, you would 24 call these samples that were used to generate this 25 figure in 488 high ferric content samples?
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5204 1 DR. LOTT: And perhaps I could offer 2 further explanation. In the evaluation not in the 3 reactor internals, but in other reactor components, 4 particularly piping components, the guidelines are 5 effectively to consider thermal embrittlement in 6 materials containing more than 20 percent ferrite.
7 So, in our minds I guess, that is the threshold. I'm 8 not sure there's an absolute threshold, but certainly 9 there's no requirement in analyzing piping materials 10 to consider the loss of toughness due to thermal aging 11 for materials that are less than 20 percent ferrite.
12 ADMIN. JUDGE WARDWELL: So this graph is 13 not representative of the CASS materials we have in 14 the reactor vessel internals at Indian Point?
15 DR. LOTT: No, because we would not in that 16 case expect to see thermal embrittlement, or certainly 17 not this level of thermal embrittlement.
18 ADMIN. JUDGE WARDWELL: So, which gets me 19 back to the original question I had.
20 DR. LOTT: Well, I guess I could only 21 suggest that the only data had to offer one way or the 22 other on the original question, which was is there a 23 synergistic effect, was this data. I don't see, 24 again, in this data -- what it suggests to me quite 25 frankly is that the properties of the duplex steel, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5205 1 this material that has 80 percent Austenite and 20 2 percent ferrite or whatever, is affected by the 3 embrittlement of the ferrite phase. There are two 4 ways that you could end up embrittling the ferrite 5 phase. One is by thermal aging, the other is by 6 irradiation. This would indicate that it doesn't 7 really matter what you do to embrittle that phase, 8 once you have it embrittled, you have a similar effect 9 on the material itself. But that's an interpretation 10 --
11 CHAIRMAN MCDADE: Okay. Following --
12 DR. LOTT: -- the best I can give you.
13 CHAIRMAN MCDADE: Following up, Dr. Lott, 14 using this chart. The delta ferrite for the CF-8, 23 15 percent based on the testimony that Dr. Hiser just 16 gave, when you look at the difference between the 17 effect of irradiation and aging plus irradiation, you 18 have very little difference, minimal, drawing your 19 conclusion that there is no synergistic effect or 20 minimal synergistic effect. In the Indian Point 21 situation, we have, at IP3, the delta ferrite is 22 approximately 11 percent. So this chart suggests to 23 you there is no synergistic effect, but the fact that 24 there's a lower delta ferrite, meaning it would be 25 even less susceptible to the heat aging, would NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5206 1 indicate to you that there would be even a lesser 2 impact on the metal used at Indian Point 2 and 3 at 14 3 and 11 percent?
4 DR. LOTT: Yes.
5 CHAIRMAN MCDADE: Am I --
6 DR. LOTT: Yes. I --
7 CHAIRMAN MCDADE: I just want to -- I 8 repeat this just to make sure that I'm hearing what 9 you're saying and there's not something --
10 DR. LOTT: I believe you've well 11 interpreted what I'm trying to say.
12 CHAIRMAN MCDADE: Okay, thank you.
13 ADMIN. JUDGE WARDWELL: But then I'll go 14 back to my original question based on your original 15 statement in your testimony, which this is where I'm 16 trying to get to. This chart does not apply to Indian 17 Point materials, correct?
18 DR. LOTT: No, we have no measurements on 19 the Indian Point materials.
20 ADMIN. JUDGE WARDWELL: The statement that 21 was made on your testimony, 722, Answer 8, Page 5, 22 you've made the statement, the existing research also 23 suggests that combined thermal aging and irradiation 24 of representative CASS materials, and I assume you 25 mean representative of what's there at Indian Point, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5207 1 does not appear to lower toughness below what is 2 expected for thermal embrittlement alone. And so my 3 question is, what is the basis for that statement?
4 And if you're going to use this figure, I'd like to 5 know why you can use this figure to extrapolate for 6 something that isn't representative of what's at 7 Indian Point.
8 MR. COX: Let me take a shot at that and 9 then Dr. Lott can chime in. But I think the way that 10 I would look at that graph is that if you have little 11 thermal aging, I mean, you can imaging the aging line 12 on the graph would show much less of a decrease than 13 what you show there. So it seems very reasonable to 14 say if you have a lot of irradiation embrittlement, a 15 lot of thermal embrittlement, and you see very little 16 difference in the combined effect than you do from the 17 irradiation effect, if you reduce the thermal aging 18 embrittlement, like you would expect to see from 19 material with a low delta ferrite, you would have no 20 reason to expect a larger difference when you combine 21 irradiation with the thermal.
22 So, I mean, I think this is not the exact 23 same material, but knowing what we know about the 24 behavior of the material under irradiation and thermal 25 embrittlement, we would expect the same result. And NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5208 1 that is very little change when you combine the two 2 effects since the thermal embrittlement would be even 3 less of an effect, there would be very little change 4 from the combined effect and what you see from the 5 irradiation effect alone.
6 DR. LOTT: Can I request -- I offered this 7 slide because I thought it was going to quickly 8 address your issue, obviously it did not. If we could 9 have some time to look at the answer to your question 10 and come back to you with it, because I think it's 11 going to require us to go through more than one 12 document in order to put the answer together.
13 ADMIN. JUDGE WARDWELL: That's fine.
14 DR. LAHEY: Your honor --
15 ADMIN. JUDGE WARDWELL: Yes.
16 DR. LAHEY: -- could I say, if I understand 17 this graph correctly now, the CF-3 shows the synergism 18 that you are asking about, whereas the CF-8 does not.
19 And so I think that's what they're talking about, but 20 for a lower percentage. But over here you can clearly 21 see the synergism in terms of the integral or the J, 22 it puts the onset of the crack occurs.
23 ADMIN. JUDGE WARDWELL: In your opinion?
24 DR. LAHEY: Well, if I believe the heights, 25 you go down with irradiation, you go down with thermal NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5209 1 aging, and if you add on the irradiation, you're 2 significantly below than just each one.
3 CHAIRMAN MCDADE: But what you're saying, 4 and again, I'm repeating this to make sure I hear what 5 you're saying and understand it, is that the CF-3 6 shows the synergy even though the CF-8 does not on 7 this graph.
8 DR. LAHEY: It appears so.
9 CHAIRMAN MCDADE: Okay. And would you 10 explain the rationale for, theoretically, why the CF-3 11 would show apparently some synergistic effect whereas 12 the CF-8 did not? Dr. Lott?
13 DR. LOTT: Well, I think there's two issues 14 here. One, as I indicated before, there's a fair 15 amount of scatter in any measurement of toughness and 16 in CASS materials in particular. So I'm not sure 17 about the -- I'm not trying to cut these numbers that 18 fine. And, again, I'm not sure that -- and I don't 19 want to get into a semantic argument about the meaning 20 of synergism, but again, these effects are not larger 21 than the combined individual effects. The effects 22 seem to be similar for a large portion of the 23 behavior.
24 ADMIN. JUDGE WARDWELL: But to also be sure 25 that I'm clear here that my question had nothing NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5210 1 necessarily to do -- my question had nothing to do 2 with synergism. My question had to do with where is 3 the basis for your statement that you made in your 4 testimony? And if it's a professional estimate, 5 that's fine. If it's a wild guess, that's fine. If 6 it's based on something else, I would like to hear 7 about it. Thank you.
8 Entergy's testimony, 722, same Answer on 9 the same Page, because it goes on to state, we may 10 have to pull this back too, that given the ongoing 11 research in this area, the Electric Power Research 12 Institute (EPRI) Materials Reliability Program (MRP) 13 developed conservative screening criteria to identify 14 components that are potentially susceptible to the 15 effects of such mechanisms. And let me ask you, are 16 these such mechanisms the thermal embrittlement and 17 the irradiation embrittlement and the combination of 18 the two? Is that what is meant by the such 19 mechanisms?
20 DR. LOTT: Yes, I believe that's true.
21 ADMIN. JUDGE WARDWELL: And what is that 22 conservative screening criteria that EPRI developed?
23 Was that the 15 percent screening criteria or is it 24 some other criteria?
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5211 1 has proposed I believe is that 20 percent value with 2 a fluence level of about one dpa.
3 ADMIN. JUDGE WARDWELL: You tailed off at 4 the end and --
5 DR. LOTT: I'm sorry, my voice is tender.
6 No, I do not believe it's the 15 percent. I believe 7 the EPRI proposal is 20 percent. I might have to 8 check to see if that number is not in our file 9 testimony. I'll have to look and see.
10 ADMIN. JUDGE WARDWELL: What screening 11 criteria are we talking about that EPRI is proposing?
12 A screening for what?
13 DR. LOTT: Basically a screening for 14 thermal and irradiation embrittlement that's used to 15 identify components that would be -- again, we'd have 16 to look at the effect of -- when we go to the -- first 17 of all, when we identified in our screening process 18 originally for MRP 227, we identified all of the cast 19 materials as potentially susceptible to thermal 20 embrittlement because we did not have any ferrite 21 contents for them. And so we couldn't say if they 22 were even above or below 20 percent.
23 Since that time, we've done a large amount 24 of working with customers such as Entergy and we've 25 identified that no materials that are greater than 20 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5212 1 percent in the Westinghouse internals components that 2 are cast stainless steel. So, that would be -- but 3 that's not in our -- it's a conservatism, because 4 we've never undid the questions about thermal 5 embrittlement. And, in fact, that's what the EPRI 6 concerns are about is the, what are the threshold 7 values for identifying thermal and irradiation 8 embrittlement? Should there be a different fluence 9 level for determining irradiation embrittlement 10 susceptibility in cast stainless steels?
11 ADMIN. JUDGE WARDWELL: And do you need to 12 time to also --
13 DR. LOTT: Let me locate that --
14 ADMIN. JUDGE WARDWELL: -- determine what 15 that EPRI screening criteria is?
16 DR. LOTT: Yes, right. Yes, let me --
17 ADMIN. JUDGE WARDWELL: Can I ask this 18 question of, I don't know who answered it before, who 19 stated it before, but about a half hour ago when we 20 started on these two simple questions that I thought 21 we were going to zoom by --
22 DR. LOTT: Yes.
23 ADMIN. JUDGE WARDWELL: -- this 15 percent 24 of ferric content was brought up as a screening 25 criteria. Is that a different screening criteria or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5213 1 did I hear that wrong or where did that come from?
2 MR. GRIESBACH: This is Tim Griesbach from 3 Entergy. Molybdenum also plays a big part in this 4 thermal embrittlement of cast materials. So there is 5 --
6 ADMIN. JUDGE WARDWELL: How are you 7 simplifying this discussion?
8 (Laughter.)
9 MR. GRIESBACH: There are two different 10 screening criteria depending on the molybdenum 11 content. If it's less than one half percent or 12 greater than one half percent. If it's greater than 13 one half percent --
14 ADMIN. JUDGE WARDWELL: Of molybdenum?
15 MR. GRIESBACH: -- of molybdenum and high 16 or low delta ferrite, there are different criteria.
17 And then there's a separate criteria based on fluence 18 also. So we will produce that for you and hopefully 19 that will simplify things, especially as it applies to 20 the Indian Point materials.
21 ADMIN. JUDGE WARDWELL: Can I get one of my 22 questions out of the way beforehand, and that is, what 23 was this 15 percent we brought and should I just 24 forget about that and you'll trump it with more 25 specific --
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5214 1 MR. AZEVEDO: Yes, your honor --
2 ADMIN. JUDGE WARDWELL: I see Mr. Azevedo 3 warming up to the old mic.
4 MR. AZEVEDO: This is Nelson Azevedo for 5 Entergy. Yes, so let me see if I can explain real 6 shortly. The 20 percent came from EPRI, that's EPRI 7 developed. The 15 percent came from the NRC.
8 ADMIN. JUDGE WARDWELL: And the 20 percent 9 is ferric content?
10 MR. AZEVEDO: Ferrite content, right. So 11 EPRI proposed 20 percent, the NRC used 15 percent.
12 ADMIN. JUDGE WARDWELL: Dr. Hiser, do you 13 agree that you used 15 percent and EPRI used 20 14 percent for a screening criteria?
15 DR. HISER: Yes, that's correct.
16 ADMIN. JUDGE WARDWELL: Okay.
17 DR. HISER: Fifteen percent is for 18 irradiated cast stainless steel.
19 ADMIN. JUDGE WARDWELL: Okay. Great. And 20 so you will still get back to us with the molybdenum 21 and anything else you want to add to that, for 22 whatever you want to? Great. Moving on to other 23 welds, New York State testimony, 576, Page 4, Line 8 24 through 23, and moving through to Page 5, Lines 1 25 through 17, states that in regard to NUREG/CR--7185, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5215 1 New York State contends that, one, cast and Austenitic 2 stainless steel welds have a duplex structure and may 3 experience thermal embrittlement, which may increase 4 the hardness and tensile strength of a material, but 5 decrease ductility, fracture toughness, and impact 6 strength of cast materials and Austenitic stainless 7 steel welds.
8 Two, IE is a concern for CASS components 9 for fluences greater than two times ten to the 20, and 10 that's N per square centimeters, so I believe that's 11 neutrons per square centimeter, allegedly equivalent 12 to ten displacements per atom, or dpa, and irradiation 13 makes cast materials and Austenitic stainless steel 14 welds more susceptible to irradiation assisted stress 15 corrosion cracking, or the IASCC. And, three, that 16 IASCC increases the crack growth rate of cracks 17 induced by stress corrosion cracking, but there is 18 allegedly virtually no data above the ten dpa, 19 although some reactor vessel internal components may 20 experience several hundred dpa.
21 And, fourth, that there is possibly 22 synergy between TE and IE, although the report 23 stresses the need for more information to develop 24 reliable failure curves. And, five, that TE could 25 make the welds associated with the pressurizer spray NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5216 1 nozzles vulnerable to seismic and thermal pressure 2 shock loads. Let me start with Dr. Lahey who made 3 this statement. To your knowledge, is a pressurizer 4 spray nozzle an RVI component covered by the RVI AMP 5 within the list of those components?
6 DR. LAHEY: Not --
7 ADMIN. JUDGE WARDWELL: Or you don't know 8 for sure?
9 DR. LAHEY: It's not a reactor vessel 10 internal at all. It's an external pressure boundary.
11 ADMIN. JUDGE WARDWELL: Okay.
12 DR. LAHEY: And I never called it an RVI.
13 ADMIN. JUDGE WARDWELL: Okay.
14 DR. LAHEY: Also, there was a typo in my 15 thing, it's not ten dpa. If you look at that chart we 16 showed before, it's more like 0.3 and that was pointed 17 out by Entergy --
18 ADMIN. JUDGE WARDWELL: Okay.
19 DR. LAHEY: -- and rightly so. So, other 20 than that, the new information to me that I didn't 21 have when I expressed this concern was the composition 22 of the weld rods that they actually used. They used 23 308 weld rods for stainless steel 304 and 309 weld 24 rods for stainless steel 316. And this does give a 25 duplex structure, but the net effect is more like less NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5217 1 than ten percent. So it's likely not subject to 2 thermal embrittlement. And what I was made aware of 3 based on the feedback from Entergy or NRC, I don't 4 recall which one, was that Dr. Chopra had a very high 5 delta ferrite content in his welds. And that made the 6 difference.
7 ADMIN. JUDGE WARDWELL: So is your net 8 conclusion that you don't believe this is an issue 9 now?
10 DR. LAHEY: I don't believe the thermal 11 embrittlement of those particular welds is an issue, 12 if in fact that is how they did their welds.
13 ADMIN. JUDGE WARDWELL: Okay. And I'll 14 turn to Entergy, as just represented by Dr. Lahey, is 15 that how the welds were done? Or do you have some 16 other nuances to discuss in regards to how the welds 17 were performed?
18 MR. AZEVEDO: This is Nelson Azevedo for 19 Entergy. No, your honor, what's been said is correct.
20 ADMIN. JUDGE WARDWELL: Okay. Thank you.
21 So there was a list that I have here, without reading 22 through it I guess -- if you don't believe that welds 23 are now an issue, you're comfortable with the welds 24 that they were performed as represented by Entergy?
25 DR. LAHEY: Well, the welds we were just NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5218 1 talking about were the welds associated with the 2 pressurizer spray nozzle. There may be concern with 3 the welds inside of the reactor vessel internal --
4 ADMIN. JUDGE WARDWELL: Okay.
5 DR. LAHEY: -- component welds, like the 6 core barrel.
7 ADMIN. JUDGE WARDWELL: Okay, good. So 8 let's continue then. I guess I'll go with Staff now.
9 Are the pressurized spray nozzles to safe end welds 10 considered part of RVIs?
11 MR. POEHLER: This is Jeffrey Poehler for 12 the Staff. No, they are not considered part of RVI.
13 ADMIN. JUDGE WARDWELL: And are they 14 handled under some other Aging Management Program?
15 MR. POEHLER: Yes.
16 ADMIN. JUDGE WARDWELL: Do you agree that 17 there is a decrease in ductility, fracture toughness, 18 and impact strength of cast materials and Austenitic 19 stainless steel weld? And, if so, does this drive any 20 changes need to thermal embrittlement screening 21 criteria or other aging management procedures?
22 MR. POEHLER: Can you clarify -- under what 23 conditions are you referring to? Combined irradiation 24 and thermal exposure or one or the other?
25 ADMIN. JUDGE WARDWELL: One or the other.
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5219 1 MR. POEHLER: In cast Austenitic stainless 2 steel, yes, we agree there is a decrease in fracture 3 toughness that can occur to either irradiation or due 4 to thermal aging.
5 ADMIN. JUDGE WARDWELL: Maybe I should back 6 up a bit because I'm a little bit confused. What 7 welds are left that are part of the RVIs? Are there 8 a whole host of them or are there only a few isolated 9 ones that part of RVIs or --
10 MR. POEHLER: Are we discussing welds or 11 castings?
12 ADMIN. JUDGE WARDWELL: Well, I believe it 13 was welding of cast material.
14 MR. POEHLER: Yes, there are some welds in 15 cast materials, such as the lower support columns.
16 ADMIN. JUDGE WARDWELL: Such as the what?
17 MR. POEHLER: The lower core support 18 columns.
19 ADMIN. JUDGE WARDWELL: And are those -- do 20 you handle those as separate or do you handle them 21 within the component itself, the support column 22 itself?
23 MR. POEHLER: We handle those as part of 24 the component itself.
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5220 1 welds, though -- well, let me ask Dr. Lahey. In 2 regards to welds as cast material, given the 3 representation of Entergy on how those welds were 4 performed, is it your understand that, that is also 5 how the reactor vessel internals were welded, for any 6 welds that were needed for reactor vessel internals?
7 DR. LAHEY: Your honor, what we were 8 talking about before was welds outside of the pressure 9 vessel using wrought Austenitic stainless steel and 10 the type of weld rods that they used to perform that.
11 Which was one of my concerns until I understood what 12 they actually did. Inside the reactor pressure vessel 13 for some of these other components, it's a different 14 story.
15 ADMIN. JUDGE WARDWELL: And is that story 16 the same story you believe applies to the welds as 17 well as the -- that you also have for the internals 18 themselves?
19 DR. LAHEY: Well, depending on the delta 20 ferrite content, as long as it's above a certain 21 level, you heard 15 percent, then you could have this 22 thermal embrittlement as well as irradiation 23 embrittlement. And there's certainly welds that can 24 be subjected to both of those effects.
25 ADMIN. JUDGE WARDWELL: But if it's below NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5221 1 the -- if it's screened out below the 15 percent 2 ferric content, then you believe that thermal 3 embrittlement is not an issue, it's just the 4 irradiation embrittlement?
5 DR. LAHEY: Generally, yes, but there are 6 some exceptions if you have a linkage within the weld 7 material itself of the delta ferrite. But generally 8 you have to have more of it to have this effect.
9 ADMIN. JUDGE WARDWELL: Let me turn to 10 Staff, how would anyone handle leakage out of a weld 11 and the impacts of that? Is there a need to evaluate 12 potential leakage out of a weld?
13 MR. POEHLER: This is Jeffrey Poehler from 14 the Staff. Are you referring to pressure boundary 15 welds or --
16 ADMIN. JUDGE WARDWELL: I'm referring to 17 welds of the RVIs that we're dealing with under the 18 RVI AMP, whatever they are.
19 MR. POEHLER: Since the welds in the RVI do 20 not serve a pressure boundary function, then leakage 21 is not a failure criteria for those welds.
22 ADMIN. JUDGE WARDWELL: And in your review 23 of the AMP and the components that are in the RVI AMP, 24 have any of the components been welded such that the 25 weld and the component itself have a higher ferric NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5222 1 content than 15 percent? Or were they all screened 2 out?
3 MR. POEHLER: We didn't look at the ferrite 4 content of the welds. So weld filler metals used for 5 Austenitic stainless steel welds, they tend to have 6 lower ferrite content in general than cast stainless 7 steels. So thermal aging is not generally an issue, 8 has not generally been considered an issue for those 9 welds. So in the screening criteria used to develop 10 the AMP at Indian Point and MRP 227-A, they didn't 11 screen-in thermal embrittlement for the welds.
12 ADMIN. JUDGE WARDWELL: How many RVIs are 13 made out of -- are composed of cast materials?
14 MR. POEHLER: Basically, there's a handful 15 of components, the lower core support columns and then 16 at Indian Point it's only the upper portion of those 17 columns, which they call the column cap, is cast 18 stainless steel. You also have the lower core support 19 forging at the very bottom of the core barrel. There 20 may be a couple other ones, but those are the main 21 ones.
22 ADMIN. JUDGE WARDWELL: Let me turn to 23 Entergy and see if they want to clarify. Well, let's 24 start off with, how many of the RVIs are, to your 25 knowledge, cast materials versus the wrought NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5223 1 Austenitic?
2 DR. LOTT: I believe that we identified 3 that in our filed testimony. I'm looking at Page 62, 4 I think it's Question 105, and it identifies the upper 5 instrumentation conduit supports --
6 ADMIN. JUDGE WARDWELL: Sorry, you're going 7 to have speak into the mic and a little slower.
8 DR. LOTT: Okay. It identifies the upper 9 instrumentation conduits and supports, upper support 10 column assemblies, the lower support casting. That 11 should be lower support column assemblies. Oh, no, 12 I'm sorry, that's right. The upper instrumentation 13 conduits, upper support column assemblies, and the 14 lower support casting.
15 ADMIN. JUDGE WARDWELL: And how many of 16 those were screened out due to ferric content? Or 17 moly content or whatever?
18 DR. LOTT: In the original screening 19 process, none of them were screened out because 20 everything was screened in based on lack of knowledge 21 at that time of the ferric content.
22 ADMIN. JUDGE WARDWELL: Okay. I'll wait, 23 I guess, until I get to those individual components.
24 Because I thought somewhere, it was my impression that 25 only those upper caps of the lower supports were cast NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5224 1 materials that were part of the aging management plan, 2 but I don't know that. I'll have to --
3 DR. LOTT: But they are cast materials part 4 of the aging management plan, but they were not 5 necessarily -- well, first of all, most of them are 6 not irradiated at all. So this question of 7 irradiation embrittlement doesn't come in. And I 8 think there was also -- some of them were screened out 9 based on lack of structural requirement. In other 10 words, there were just no --
11 ADMIN. JUDGE WARDWELL: Well, I'll be able 12 to get to that quote once I get to it. I just can't 13 find it and I don't want to spend time looking for it.
14 I'll come upon it as I work my way through. But I 15 want to get back to welds, I think. How many -- how 16 do you handle the welds in your AMP, for both the 17 wrought and the cast materials? Are they part of the 18 individual component or are they considered 19 separately?
20 DR. LOTT: There are some key welds that 21 are considered separately. And those welds were, for 22 instance, the core barrel welds, we knew we were 23 concerned about them, they're identified separate.
24 There are other components that contain welds and in 25 the screening process, one of the things we identified NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5225 1 was which components contained welds and which did 2 not. And then the primary concern with welds was 3 large structural welds where concern for stress 4 corrosion cracking, not necessarily embrittlement.
5 ADMIN. JUDGE WARDWELL: And, so, how many 6 of these special welds were designated as individual 7 components when you split this out?
8 DR. LOTT: I'd have to go back and count, 9 I --
10 ADMIN. JUDGE WARDWELL: Okay.
11 DR. LOTT: -- couldn't answer that.
12 MR. DOLANSKY: This is Bob Dolansky from 13 Entergy. The Table 1 on Page 87 of our testimony 14 lists the primary components and the expansion 15 components. The primary components lists if it's a 16 weld or not. Does that answer your question?
17 ADMIN. JUDGE WARDWELL: To a certain 18 degree. I was hoping you had an approximate number of 19 those that were there that related to split out welds, 20 but that's fine. I don't need that right now.
21 MR. DOLANSKY: Okay.
22 ADMIN. JUDGE WARDWELL: I would like that 23 number to get a feeling for how many there are and 24 what's being -- so I can then refer to that table to 25 see what's being handled for them.
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5226 1 MR. DOLANSKY: I could go through this 2 table if you wanted and --
3 ADMIN. JUDGE WARDWELL: Yes, at some point 4 and then just come up with a number and that will be 5 sufficient so I make sure I see the same number that 6 you do when I -- if it does come up in our decision 7 writing, I'm able to refer to that and handle any 8 discussion associated with it.
9 MR. DOLANSKY: Just to clarify, what I'll 10 do the research on is for each primary component, if 11 it's a weld, I'll tell you how many there are.
12 ADMIN. JUDGE WARDWELL: Okay.
13 MR. DOLANSKY: That's basically what you're 14 looking for --
15 ADMIN. JUDGE WARDWELL: That would be 16 great.
17 MR. DOLANSKY: -- to understand?
18 ADMIN. JUDGE WARDWELL: Thank you.
19 MR. DOLANSKY: Okay.
20 ADMIN. JUDGE WARDWELL: Yes, let's do it 21 now.
22 CHAIRMAN MCDADE: Okay. It might be an 23 appropriate time to take a short ten minute break. A 24 couple of things. First of all, Mr. Sipos, we had a 25 discussion with Dr. Lahey about the Table 5-5 and he NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5227 1 hadn't had an opportunity to review it. Do you have 2 handy New York 496, which is the document that Table 3 5-5 is in? If not, I've got it right here and I can 4 provide it to you for Dr. Lahey to look at.
5 MR. SIPOS: Your honor, I believe we have 6 it with us. Thank you.
7 CHAIRMAN MCDADE: Okay. If you don't, then 8 at the next break I can provide it to you and --
9 MR. SIPOS: Thank you.
10 CHAIRMAN MCDADE: -- you can give that to 11 Dr. Lahey. Another thing that I actually should have 12 mentioned earlier, in Exhibit 616, which Entergy 13 provided, at the beginning of it, there's a table of 14 abbreviations, which is extremely helpful. Whoever 15 prepared it, I really want to thank them. I just 16 wanted to mention it, as the court reporter is going 17 through this, if you have not reviewed it, Exhibit 18 616, the table of abbreviations I think is going to be 19 very helpful to you in make sure that you -- well, 20 okay. Then that isn't necessary, apparently he found 21 it without my assistance. Do we have anything else we 22 need to take up before we take a short break?
23 ADMIN. JUDGE WARDWELL: But I think ISR was 24 missing from that table though, but I echo how much I 25 did use that for both other testimony, I would go back NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5228 1 to the Entergy abbreviations or the NRC ones, whoever 2 had them --
3 CHAIRMAN MCDADE: Okay. So it's ten 4 minutes after 3:00, we'll break until 20 minutes after 5 3:00. Thank you.
6 (Whereupon, the above-entitled matter went 7 off the record at 3:09 p.m. and resumed at 3:23 p.m.)
8 CHAIRMAN MCDADE: Let's get started. The 9 hearing will come to order.
10 ADMIN. JUDGE WARDWELL: Okay. Moving on to 11 baffle former bolts, the NRC testimony, 197, Answer 12 80, Page 59, says, an example of an augmented 13 inspection is the baseline volumetric examination 14 using ultrasound testing, UT, of the baffle former 15 bolts between 25 and 35 effective full power years, 16 which you use the acronym with the letters E-F-P-Y, 17 there's probably some fancy way to say that, that I 18 don't know, within a subset examination on a ten year 19 interval, as specified in Table 4.3 of MRP 227-A.
20 CHAIRMAN MCDADE: Wait, whose testimony is 21 this?
22 ADMIN. JUDGE WARDWELL: I'm sorry?
23 CHAIRMAN MCDADE: Which exhibit are you 24 reading from?
25 ADMIN. JUDGE WARDWELL: This is NRC's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5229 1 Exhibit 197, A 80, Page 59.
2 CHAIRMAN MCDADE: NRC's?
3 ADMIN. JUDGE WARDWELL: Yes. And I guess 4 I'd ask Entergy, how long -- have you initiated any UT 5 testing of the baffle former bolts at Indian Point?
6 MR. DOLANSKY: This is Bob Dolansky for 7 Entergy. No.
8 ADMIN. JUDGE WARDWELL: And has the 9 industry itself and do you have data on that?
10 MR. DOLANSKY: The industry has performed 11 inspections on baffle former bolts. They have -- when 12 you say, do we have data on that, what kind of data 13 are you looking for?
14 ADMIN. JUDGE WARDWELL: I'm just curious on 15 what the experience is with any of the failure rates 16 associated with these in other Westinghouse reactors.
17 MR. DOLANSKY: I would characterize it as 18 most people who have inspected have found some 19 degraded bolts, but not enough that they were required 20 to replace any bolts.
21 ADMIN. JUDGE WARDWELL: Okay, thank you.
22 DR. LOTT: May I suggest, there is some 23 data on that in Entergy Exhibit 650, which summarizes 24 the experience to date with various reactor internal 25 exams, including the baffle bolt exams.
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5230 1 ADMIN. JUDGE WARDWELL: Okay, thank you.
2 NRC's testimony on 197, Question 222, Page 124, 3 states, what does this operating experience tell us 4 about the probability of cracks existing in the PWR 5 RVIs? And the Answer says, this result summarized in 6 the presentation indicate no cracking has been found 7 with the exception of some cracking of bolts, about 8 1.5 percent of Westinghouse baffle former bolts. And 9 there I might ask Staff, what might be the cite for 10 this 1.5 percent cracking rate for the baffle former 11 bolts that you state in your testimony?
12 DR. HISER: This is Allen Hiser of the 13 Staff. I believe it's NRC 207.
14 ADMIN. JUDGE WARDWELL: Okay. And does 15 that also give the total number of baffle bolts that 16 they looked at in regards to this, that generated the 17 1.5 percent figure?
18 DR. HISER: If you'll indulge me, I can 19 pull it up and review it real quickly.
20 ADMIN. JUDGE WARDWELL: Okay. And what are 21 you pulling up now?
22 DR. HISER: From Exhibit 207.
23 ADMIN. JUDGE WARDWELL: Okay. As you're 24 pulling that up, let me ask this question also, is it 25 your understanding or do you know if this industry NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5231 1 experience with the inspection of baffle former bolts 2 is also documented in Appendix A of MRP 227, which is 3 NRC 114C?
4 DR. HISER: It is document in the MRP 227, 5 but that was as of several years ago. The Exhibit 207 6 represents data through the fall of 2014.
7 ADMIN. JUDGE WARDWELL: Okay.
8 DR. HISER: So it's probably the most 9 recent information. And this indicates that 8,887 10 baffle bolts have been UT inspected.
11 ADMIN. JUDGE WARDWELL: And that's 12 generated the 1.5 percent in regards to the cracking 13 rate?
14 DR. HISER: Yes, that's correct.
15 ADMIN. JUDGE WARDWELL: And do you know 16 what was the criteria used to say that a bolt had been 17 cracked? Is there any parameters that was given?
18 Does it have to be separated, loose? Is there just a 19 crack detected? Or a visual crack would apply to 20 that?
21 DR. HISER: In this case, I would expect it 22 at least was any indication of a crack from the UT 23 exam.
24 ADMIN. JUDGE WARDWELL: Dr. Lahey, it seems 25 like this experience has shown that very few cracked NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5232 1 or failed baffle former bolts have been detected 2 during these examinations and in most cases no cracked 3 or failed bolts were detected at all. Do you have any 4 other experience that dictates that they are more of 5 an issue --
6 DR. LAHEY: Well, there have been --
7 ADMIN. JUDGE WARDWELL: -- of these than 8 there seem to be?
9 DR. LAHEY: There have been failures 10 reported overseas as well. So, it's not been massive 11 failures, but there have definitely been failures.
12 While we're talking about baffle bolts, I was asked to 13 look at Table 5-5 and if you want me to -- which 14 includes the acceptance criteria for such things as 15 baffle bolts. Do you want me to report on that now or 16 not?
17 ADMIN. JUDGE WARDWELL: That was one of 18 your homework assignments wasn't it?
19 DR. LAHEY: During the break, I thought.
20 ADMIN. JUDGE WARDWELL: Yes. Well, we 21 allow homework to be done during a break --
22 DR. LAHEY: Right.
23 ADMIN. JUDGE WARDWELL: -- we don't 24 restrict that.
25 DR. LAHEY: Shall I report on that?
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5233 1 ADMIN. JUDGE WARDWELL: Sure. This is 2 probably as good an opportunity.
3 DR. LAHEY: Okay. And it has to do -- I 4 looked at the various items in the Table 5-5 and the 5 one that I have some concern about is the baffle 6 bolts. And, in particular, when it talks about 7 additional examination acceptance criteria, it's not 8 yet specified. It says, the examination acceptance 9 criteria for the UT of the bolts shall be established 10 as part of the examination technical justification.
11 So it's still sort of open-ended and since it's such 12 an important component and has safety significance.
13 That's a little unsettling.
14 ADMIN. JUDGE WARDWELL: And that's the 15 examination criteria, is that correct?
16 DR. LAHEY: Yes.
17 ADMIN. JUDGE WARDWELL: And may I turn --
18 DR. LAHEY: Table 5-5.
19 ADMIN. JUDGE WARDWELL: Yes. Turn to 20 Entergy in regards to that examination criteria. What 21 does that mean in regards to the statement made in 22 Table 5-5 for the baffle former bolts? For Entergy, 23 anyone at Entergy who wants to answer.
24 MR. DOLANSKY: The examination acceptance 25 criteria for baffle former bolts would be any NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5234 1 indication of cracking.
2 ADMIN. JUDGE WARDWELL: And what is the --
3 is there a difference between examination criteria and 4 inspection criteria or acceptance -- I guess we have 5 acceptance criteria and examination criteria. Are 6 they different or are those two sayings for the same 7 thing?
8 MR. AZEVEDO: This is Nelson Azevedo for 9 Entergy. The examination criteria is the criteria for 10 the inspectors. So anything that exceeds the 11 examination criteria, they must report it and then we 12 enter into our corrective action process. The 13 acceptance criteria or the engineering acceptance 14 criteria, if you will, it's how many bolts, for 15 example, can we afford to lose without impacting the 16 ability of the structure to perform its intended 17 safety function?
18 ADMIN. JUDGE WARDWELL: Great. Thank you 19 for that clarification. And did you have anything 20 more, Dr. Lahey, that you wanted to talk about in 21 regards to Table 5-5?
22 DR. LAHEY: No. With regards to that 23 table, that was our specific concern, the number of 24 bolts, because that is significant.
25 ADMIN. JUDGE WARDWELL: And then just to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5235 1 close the other item, to make sure -- you don't have 2 any hard numbers in regards to percent of these baffle 3 bolts that failed overseas or anywhere else that would 4 contradict the 1.5 percent that was observed by 5 Westinghouse?
6 DR. LAHEY: I've never tabulated it, no.
7 ADMIN. JUDGE WARDWELL: Okay. Thank you.
8 NRC's testimony, 147, Page 47, in order to maintain 9 the intended function, only about 20 to 30 percent of 10 the baffle former bolts need to remain intact. And I 11 guess I'll ask Entergy -- well, it's an NRC statement, 12 so I'll ask NRC. What's the basis for this 20 to 30 13 percent figure? And if it's just that it came from 14 the Applicant, so say, or is it something that you are 15 familiar with in regards to the generation of these 16 particular values?
17 MR. POEHLER: Right. This is Jeffrey 18 Poehler from the Staff. It did not come from the 19 Applicant. It is based on some -- well, there's a 20 couple of Topical Reports, WCAP reports, where they 21 provide a methodology for performing these types of 22 minimum pattern analyses for the baffle former bolts.
23 Several plants have actually used that methodology and 24 the results came out in the 20 to 30 percent range of 25 intact bolts needed.
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5236 1 ADMIN. JUDGE WARDWELL: What do you mean 2 by, they used that and came out with a 20 to 30 3 percent?
4 MR. POEHLER: They used a preapproved --
5 the NRC had approved the methodology for doing these 6 analyses and it was submitted as a Topical Report --
7 ADMIN. JUDGE WARDWELL: And this is --
8 MR. POEHLER: -- to the NRC.
9 ADMIN. JUDGE WARDWELL: This is an 10 engineering analysis, it has nothing to do with 11 inspection or anything or number of bolts, it's just 12 -- number of bolts that have cracked or anything, this 13 has strictly an analysis of how many are needed to 14 maintain the intended function of the baffle, is that 15 correct?
16 MR. POEHLER: Correct. And how many and in 17 what positions, what type of patterns --
18 ADMIN. JUDGE WARDWELL: Okay.
19 MR. POEHLER: -- that would be needed to 20 maintain the design function of the baffle former 21 assembly.
22 ADMIN. JUDGE WARDWELL: And that analysis 23 resulted in the 20 to 30 percent needed?
24 MR. POEHLER: When that methodology was 25 applied for specific plants.
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5237 1 ADMIN. JUDGE WARDWELL: Has that -- was 2 Indian Point one of those specific plants?
3 MR. POEHLER: I don't -- to my knowledge, 4 they weren't. If they've done that type of analysis, 5 I'm not aware that their results have been submitted.
6 ADMIN. JUDGE WARDWELL: Okay. Well, let's 7 turn to Entergy. Has such an analysis been done at 8 Entergy or do you just use that 20 to 30 percent as an 9 accepted figure based on the analysis that has been 10 done by others?
11 MR. DOLANSKY: This is Bob Dolansky from 12 Entergy. We're having a plant specific acceptable 13 bolting pattern analysis performed for us right now 14 that will determine plant specific. So the 15 methodology that -- the Topical Report is a general 16 methodology. We then took that general methodology 17 and put in our specific accident loads, LOCA loads, 18 all those things, and did the analysis. Well, it's 19 ongoing. It'll be ready before we perform the 20 inspections in the spring.
21 ADMIN. JUDGE WARDWELL: Do the --
22 MR. STROSNIDER: This Jack Strosnider for 23 Entergy. If I just could, this question has come up 24 a number of time about Topical Reports. I just want 25 to make sure people understand how that process works, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5238 1 if I could take just a second? When the industry sees 2 a generic issue, such as this issue, and they know 3 that a lot of plants are going to have to deal with 4 it, they'll go contract the vendor, like Westinghouse 5 or somebody, to develop a methodology to do some pilot 6 plants to do a case study. And then they'll submit 7 that to the NRC.
8 The NRC reviews that and, if they approve 9 it with whatever approvals they make, then a specific 10 plant, a utility, can come in and reference that 11 report when they do their plant specific analysis.
12 And the important part of that is that the methodology 13 has been approved by NRC. So as was just indicated, 14 that includes all the loading and how you do the 15 calculations and how you demonstrate functionality.
16 So what's happening is, at the plant specific level 17 then is just to do the plant specific evaluation.
18 When the NRC issues their safety evaluation, they will 19 include in it any specific action items that have to 20 be addressed on a plant specific basis. So I hope 21 that's of help to you because there's a number of 22 questions that have come up on various Topical Reports 23 and that's how that whole process works.
24 ADMIN. JUDGE WARDWELL: Thank you. And, 25 Dr. Hiser, has NRC approved this methodology for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5239 1 determining the number of bolts needed and the 2 patterns for the baffle former bolts?
3 DR. HISER: Yes, we have. And the exhibits 4 are ENT 654 and 655.
5 ADMIN. JUDGE WARDWELL: Thank you. And 6 previously, if Indian Point is consistent with what 7 was done before, they should arrive at -- if it's 8 consistent with what our understanding is now, it 9 would be 20 to 30 percent of the bolts are needed, is 10 that correct?
11 DR. HISER: I would expect that to be the 12 case.
13 ADMIN. JUDGE WARDWELL: I think I'll go 14 back to Entergy though to ask this question. Do you 15 know if that 20 to 30 percent -- does not the intended 16 function of a component, like the baffles and the 17 former bolts that are attached to it, include some 18 type of, in my field I'd call it a safety factor, 19 other people will consider it an error margin or 20 whatever else? Isn't there some sort of safety factor 21 or error margin applied when you estimate how many of 22 the bolts are needed?
23 MR. DOLANSKY: Yes. If I could just take 24 one minute and explain? We're getting a site specific 25 acceptable bolting pattern analysis performed. That's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5240 1 going to tell us some very small number of bolts that 2 are allowed to be degraded without requiring 3 additional analysis. So let's just say one -- there's 4 832 baffle former bolts. Let's just say one, so we 5 could find one and we would be okay. But if we find 6 more than one, then we will have to do a thing called 7 a real time analysis and that basically is, they take 8 the actual place that we found the degraded bolts and 9 they put them into the analysis.
10 A big part of -- it's taken roughly a year 11 to do this analysis. They develop a computer program, 12 they'll then take the specific bolts that we actually 13 found degraded, put those into the computer program, 14 run it with our specific loads and site specific 15 requirements, and come out with, is that actual what 16 we found acceptable or not?
17 ADMIN. JUDGE WARDWELL: And by acceptable, 18 does that have an error margin or a safety margin 19 around it?
20 MR. DOLANSKY: Yes, it does.
21 ADMIN. JUDGE WARDWELL: And my question is, 22 do you know if the 20 to 30 percent figure of the 23 acceptable bolts that have been done in the past, that 24 number generated in the past, does that include that 25 safety factor or could it possibly be that, that is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5241 1 what needed with a safety factor of only one?
2 MR. DOLANSKY: I can't answer that.
3 ADMIN. JUDGE WARDWELL: Okay. Staff, would 4 you be able to answer that question?
5 DR. HISER: This is Allen Hiser of the 6 Staff. In the generic methodology, Topical Report, 7 there would be safety factors included in that.
8 ADMIN. JUDGE WARDWELL: So you believe that 9 --
10 DR. HISER: It includes safety factors.
11 ADMIN. JUDGE WARDWELL: -- the 20 to 30 12 percent includes some sort of safety factor such that 13 it could be even less and the thing would still hold 14 together?
15 DR. HISER: If you use a safety factor of 16 one, yes, I expect you could go to fewer bolts than 17 the 20 to 30 percent.
18 ADMIN. JUDGE WARDWELL: Right. Or 19 conversely, if the safety factor was two that they 20 actually used, you could really go down to 10 to 15 21 percent?
22 DR. HISER: Well, safety factor is not on 23 the number of bolts --
24 ADMIN. JUDGE WARDWELL: Just in general 25 terms, just to put a handle on --
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5242 1 DR. HISER: Yes, that's correct. It would 2 be a lower number.
3 ADMIN. JUDGE WARDWELL: The point I'm 4 trying to get is I want to make sure that, that 20 to 5 30 percent doesn't pertain to a safety factor of one.
6 And, likewise, when you do your analysis, I'm asking 7 Entergy, will a number you have be the number that is 8 absolutely needed in order to maintain safety with no 9 extra margin or will it include some margin associated 10 with it?
11 MR. DOLANSKY: This is Bob Dolansky with 12 Entergy. First, when we're doing all this, we're 13 really concerned about reactor safety, maintaining 14 core coolability, maintaining the ability to insert 15 the control rods. That's factored into the analysis 16 and there are safety margins on that safety analysis.
17 MR. STROSNIDER: This is Jack Strosnider --
18 ADMIN. JUDGE WARDWELL: So when you -- no, 19 I'd like to stay with my thoughts. When you come up 20 with an acceptable pattern, that pattern will have 21 some margin built into it, such that in actuality if 22 you knew truth, you could end up with lesser number of 23 bolts, but because we don't know truth, you're going 24 to have some margin in there that allows for the 25 uncertainty associated with not knowing truth in that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5243 1 pattern. Is that a fair assessment?
2 MR. DOLANSKY: Yes. And the reason why I 3 spoke before about the real time analysis is we don't 4 want to know, when we don't know what we actually 5 have, come up with some bolting pattern analysis. We 6 want to wait until we find out what we actually have 7 out there and then run it through the program and make 8 sure that it's robust and is perfectly acceptable. So 9 that's why we actually run the real time analysis.
10 MR. AZEVEDO: Your honor, this is Nelson 11 Azevedo. If I may add, we use the NRC approved 12 methodologies, so the same safety margins are going to 13 be used.
14 MR. STROSNIDER: And this is Jack 15 Strosnider. Just to expand on that, typically what 16 you're going to see in these Topical Reports when they 17 do this type of analysis, is they're going to work to 18 maintain the safety margins that were in the original 19 licensing basis. So in this case, if you're talking 20 about the structural criteria, it's going to be the 21 ASME code factors of safety during normal and accident 22 conditions. And when you look at the other aspect of 23 this, we get into some of the core cooling and 24 accident evaluations and they would need to maintain 25 the margins that are in that licensing basis in terms NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5244 1 of margin to core damage and that sort of thing. So 2 they work to maintain the current licensing basis and 3 the margins that were in those.
4 ADMIN. JUDGE WARDWELL: And so it's your 5 professional opinion that the 20 to 30 percent that 6 has historically been generated by Westinghouse at 7 other plants includes those safety factors associated 8 with the ASME code?
9 MR. STROSNIDER: Yes, it would have those 10 margins in it.
11 ADMIN. JUDGE WARDWELL: Thank you. Your 12 testimony, Entergy, 616, Answer 152, Page 99, says, to 13 prepare for these inspections, as explained in the 14 Supplemented SER Number 2, that the UT examination 15 acceptance criteria for the baffle former bolts will 16 be developed as part of the technical justification 17 for the inspections. Page 100, Answer 154, that the 18 examination acceptance criteria for the individual 19 baffle former bolts will be no defect that could be 20 detectable via UT inspections. And in parentheses, 21 you say that detectable versus UT inspections are a 22 defect exceeding 30 percent of the bolt cross-23 sectional area.
24 The TJ will merely demonstrate that the UT 25 inspections at Indian Point will be capable of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5245 1 detecting such cracking. And I guess my first 2 question to you is, from where did this 30 percent of 3 the cross-sectional area come from in regards to -- I 4 gather that is what is needed before you're able to 5 detect it with the UT inspections.
6 DR. LOTT: I believe the last thing you 7 said is basically true, that the UT inspections, a 8 standard UT inspection, very reliably detect 30 9 percent. I believe that the requirement is stated in 10 MRP 228, if I'm not mistaken, that, that minimum 30 11 percent is there based on the judgement of the 12 inspectors.
13 ADMIN. JUDGE WARDWELL: But that's a result 14 of what you can get out of it. That's what you need 15 before you're -- that's the sensitivity of your UT 16 device --
17 DR. LOTT: Right.
18 ADMIN. JUDGE WARDWELL: -- if you will. Is 19 that correct?
20 DR. LOTT: Yes.
21 ADMIN. JUDGE WARDWELL: And, so, that means 22 that anything under that won't be detected, is that 23 correct?
24 DR. LOTT: Well, maybe not necessarily 25 won't, but might not be.
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5246 1 ADMIN. JUDGE WARDWELL: The odds are it 2 won't. I mean, there will be some that needs more 3 than 30 percent that won't be detected also, using the 4 same phraseology?
5 DR. LOTT: Yes. It's a very high 6 probability of detection, I believe, at 30 percent.
7 ADMIN. JUDGE WARDWELL: And is this 8 acceptance criteria -- has the 30 percent figure shown 9 up anywhere in the Inspection Plan or anything else?
10 Or is that just part of the testimony that you've 11 created here for this hearing?
12 MR. DOLANSKY: This is Bob Dolansky with 13 Entergy. I believe that 30 percent will be in the 14 technical justification --
15 ADMIN. JUDGE WARDWELL: Okay.
16 MR. DOLANSKY: -- that will be used by --
17 that has to be documented and reviewed by our NDE 18 Level III at the site before they perform the 19 inspection.
20 ADMIN. JUDGE WARDWELL: This is all stuff 21 you're going to get prepared and done prior to 22 starting your inspection, I think you talked about 23 earlier today.
24 MR. DOLANSKY: Correct.
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5247 1 have any comments on what you heard in regards to the 2 baffle former bolts --
3 DR. LAHEY: Yes.
4 ADMIN. JUDGE WARDWELL: -- inspection and 5 --
6 DR. LAHEY: Yes, your honor. This was a 7 very interesting discussion for me. I have not seen 8 that report or the results, I'd be very interested to 9 see it. In particular, I heard nothing about the type 10 of loads that were applied to assure the integrity of 11 these patterns with reduced number of bolts and would 12 be very interested in hearing about what type of 13 impulsive loads were used, if they were used. And 14 what type of codes might have been used to generate 15 these.
16 ADMIN. JUDGE WARDWELL: Okay, thank you.
17 DR. LOTT: Some of that information, I will 18 suggest, would be available in ENT 0655, which is 19 basically the same document that Dr. Hiser just 20 referred to in terms of the methodology for 21 determining acceptable bolt patterns. It discusses 22 the types of loads that are included, particular the 23 multi flex code and how they're transferred to the --
24 ADMIN. JUDGE WARDWELL: Have those loads 25 been documented in the Topical Reports that the NRC NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5248 1 has approved?
2 DR. LOTT: Yes, in great detail.
3 CHAIRMAN MCDADE: Okay. Dr. Lott, what 4 exhibit did you just reference?
5 DR. LOTT: Entergy 655.
6 CHAIRMAN MCDADE: Thank you.
7 ADMIN. JUDGE WARDWELL: When did you start 8 this endeavor in regards to planning for the 9 inspections of the baffles and going through this site 10 specific process of determining your pattern for the 11 baffle bolts?
12 MR. DOLANSKY: Just the baffle former bolts 13 --
14 ADMIN. JUDGE WARDWELL: Yes.
15 MR. DOLANSKY: -- you're asking about?
16 ADMIN. JUDGE WARDWELL: Yes.
17 MR. DOLANSKY: This is Bob Dolansky for 18 Entergy. The contract for the inspection of the 19 baffle former bolts was issued, I'm going to say in 20 spring of 2015. The contract to develop the 21 acceptable bolting pattern analysis, I think, was 22 January. It takes a little over a year to develop 23 that acceptable bolting pattern analysis. So I'm 24 pretty sure it was January and the final report will 25 be ready for us in February before our inspections in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5249 1 March.
2 ADMIN. JUDGE WARDWELL: For Dr. Hiser of 3 the NRC, when was the Topical Report approved by you 4 that --
5 DR. HISER: I believe it was --
6 ADMIN. JUDGE WARDWELL: -- highlighted 7 these inspection procedures or the whole methodology, 8 is what I guess the best word is?
9 DR. HISER: This is Allen Hiser of the 10 Staff. Is the question the methodology used to 11 demonstrate the minimum bolt pattern?
12 ADMIN. JUDGE WARDWELL: Yes.
13 DR. HISER: That was I believe 1999.
14 ADMIN. JUDGE WARDWELL: So my question to 15 Entergy will be, why haven't you started earlier in 16 this process then in order to -- if the methodology 17 has been outlined in regards to coming up with this 18 pattern, why wasn't it done earlier considering your 19 License Renewal Application was submitted in 2007?
20 MR. DOLANSKY: We didn't feel that we 21 needed the acceptable bolting pattern analysis -- Bob 22 Dolansky with Entergy -- until we were going to 23 perform the inspection. Basically, until you perform 24 the inspection and determine if you have any degraded 25 bolts, the acceptable bolting pattern analysis doesn't NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5250 1 do anything for you. You don't -- there's no value to 2 it until you --
3 ADMIN. JUDGE WARDWELL: I guess that gets 4 back to the question I asked earlier, why didn't you 5 start to do the inspection on the bolts earlier?
6 Which I think we've already talked about, so, okay.
7 MR. AZEVEDO: Your honor --
8 ADMIN. JUDGE WARDWELL: Mr. Azevedo, it 9 looked like you were -- you weren't, you were just 10 stretching or something?
11 MR. AZEVEDO: No, I was going to say what 12 Mr. Dolansky just said.
13 ADMIN. JUDGE WARDWELL: Okay. That's a cop 14 out. No, I believe you.
15 MR. KUYLER: Your honor, if I may? I 16 believe a moment ago Dr. Lott may have misspoke. The 17 Exhibit that I think he might have been referring to 18 is Entergy 654, that provides the accident loads.
19 DR. LOTT: Yes, 655 is the non-proprietary 20 version.
21 ADMIN. JUDGE WARDWELL: Okay.
22 DR. LOTT: Either way it works.
23 CHAIRMAN MCDADE: Dr. Lott had mentioned 24 655 and counsel is suggesting it was actually Entergy 25 654, is that --
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5251 1 MR. KUYLER: Yes, your honor.
2 CHAIRMAN MCDADE: Okay, thank you.
3 ADMIN. JUDGE WARDWELL: Moving on to the 4 clevis bolt --
5 CHAIRMAN MCDADE: Actually, could I clear 6 something up before --
7 ADMIN. JUDGE WARDWELL: Sure.
8 CHAIRMAN MCDADE: -- you move on to --
9 ADMIN. JUDGE WARDWELL: Yes.
10 CHAIRMAN MCDADE: -- a new topic, just 11 really quickly. Mr. Dolansky, before the break, you 12 were asked with regard -- questioned on Table 1, Page 13 87, regarding welds. And that particular document, 14 it's Entergy Exhibit 616, it identifies four different 15 kinds of welds. The upper core baffle flange, the 16 barrel cylinder girth welds, the lower core barrel to 17 lower support casting welds, and the core barrel 18 outlet nozzle welds. But it doesn't have any 19 indication as to for each of these categories, how 20 many welds we're talking about. Are we talking about 21 a single weld each time or are we talking about scores 22 or hundreds or thousands?
23 MR. DOLANSKY: This is Bob Dolansky with 24 Entergy. Looking at Table 87, the upper core barrel 25 flange weld is a single weld. The upper and lower NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5252 1 core barrel cylinder girth welds are two individual 2 welds, an upper and a lower core barrel cylinder girth 3 weld. The lower core barrel to support casting weld 4 is a single weld. The only ones that are multiple 5 welds are the lower flange welds on the control rod 6 guide tubes, which are at the top there, the second 7 box. That -- give me one second and if we could bring 8 up on the screen, let's see what, number seven -- NRC 9 114A through C and we're actually going to be looking 10 at 114B, Bravo, I believe. And we want Page 4-60.
11 It'll help clarify.
12 CHAIRMAN MCDADE: 4-60?
13 MR. DOLANSKY: 4-60, yes, your honor. Yes.
14 Stop right there, that's fine. So, this is a 15 depiction of the control rod guide tube assembly. If 16 you look on the bottom, so what we call the lower 17 flange welds in the table on Page 87, if you look on 18 the bottom of that, it shows two arrows pointing to 19 lower flange welds, the bottom one, let's say, there's 20 discrete, very small ribs and each of those ribs has 21 welds. So these components have multiple welds on 22 each component. But that's the only one that really 23 has multiple welds on each component.
24 CHAIRMAN MCDADE: Okay, thank you. And I 25 wondered, since we're at a break right here, pose a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5253 1 question to -- I'm sorry, Mr. Cox, did you have a --
2 MR. COX: Yes, I was just going to clarify 3 that the next page of that exhibit shows the other 4 welds that you had asked about. Mr. Dolansky 5 indicated there were one or two welds in those cases.
6 I just wanted to point out, those welds are around the 7 entire circumference of the core barrel, so it's -- I 8 mean, some of these guys could probably tell you how 9 many inches of weld that is, but it's several hundred 10 probably.
11 MR. DOLANSKY: Right. They're large -- I'm 12 sorry, I didn't mean to imply that they were small 13 welds. They're very large, long welds around a big 14 robust component.
15 CHAIRMAN MCDADE: Okay.
16 ADMIN. JUDGE WARDWELL: Does one guy do 17 them? Or do they have multiple guys?
18 MR. DOLANSKY: Well, if they're EVT1, they 19 have to -- because there's criteria on scanning speed 20 and so forth, it's actually done by special tooling.
21 But there are people watching screens for each of 22 them.
23 ADMIN. JUDGE WARDWELL: But the guy who 24 first built it, was it one guy doing one weld and he 25 can take ownership of that weld and say, there's my NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5254 1 weld?
2 MR. DOLANSKY: Probably. Back then, 3 probably true.
4 ADMIN. JUDGE WARDWELL: Okay.
5 CHAIRMAN MCDADE: Okay. And the other 6 thing I wanted to do is just pose a question, not be 7 answered right now, but to move on with Judge 8 Wardwell's question, but to raise it for Dr. Lott and 9 Dr. Lahey who had testified earlier, addressing New 10 York Exhibit 488, which is a NUREG/CR-7184, it's from 11 December of 2014. And, basically, my question is 12 this, we've had testimony that for these low ferrite 13 stainless steel material, they won't show a meaningful 14 combined effect from thermal aging and irradiation.
15 And in that NUREG it seems to suggest, 16 well, it states, the radiation and the reduction of 17 fracture toughness was more significant in the unaged 18 than in the thermally aged specimens. And it goes on 19 to have a further discussion of that. And I'd like 20 to, perhaps at the beginning of tomorrow, come back 21 and discuss whether or not this supports the 22 hypothesis of the -- Dr. Lahey's hypothesis or whether 23 or not there's an explanation explaining why this 24 doesn't support the synergistic effect of the thermal 25 aging and the neutron embrittlement. But I don't want NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5255 1 to take that up right now, I just wanted to address 2 that particular document to you so you would have a 3 chance to look at it this evening and perhaps you 4 could take it up briefly tomorrow. Dr. Lahey?
5 DR. LAHEY: Yes, this is Richard Lahey, New 6 York. You had asked me to look at this report, NUREG-7 7184, and to look at the word synergistic and I did 8 look at this report and it's very interesting, the 9 later version of it was redlined out. It was 10 synergistic, but it was redlined out and combined 11 effect was put it its place. So at one point it said 12 synergistic and I guess the final issue would have 13 said combined effect.
14 CHAIRMAN MCDADE: Okay. But you would view 15 those two words as synonyms, but I don't want to take 16 it up right now and I want to let --
17 DR. LAHEY: Right.
18 CHAIRMAN MCDADE: -- Judge Wardwell move 19 on. But what I'm looking for is to whether or not, 20 not just a particular word, but whether or not this 21 NUREG and the data presented there, what impact, if 22 any, does it have on the hypothesis of the synergistic 23 effects? And, again, I don't want to take it up right 24 now, we want to move on. Take a look at it, have Dr.
25 Lott and Dr. Hiser take a look at it as well overnight NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5256 1 and we can discuss it very briefly in the morning.
2 DR. HISER: Judge McDade, quick question on 3 that. Allen Hiser of the Staff. What page number of 4 7185 were you quoting from?
5 CHAIRMAN MCDADE: Well, I was quoting from 6 the abstract --
7 DR. HISER: Okay.
8 CHAIRMAN MCDADE: -- which is I think 9 especially like a little I or a little double I --
10 DR. HISER: Okay.
11 CHAIRMAN MCDADE: -- I was just quoting 12 from the abstract, but it then goes on.
13 MR. SIPOS: And your honor, for New York, 14 I think --
15 CHAIRMAN MCDADE: I didn't say that, I said 16 I was quoting from the abstract.
17 MR. SIPOS: Your honor, John Sipos for the 18 record. I think you were referring to 7184. I just 19 want to make sure that's clear for the record, rather 20 than 7185, which I think Dr. Hiser just mentioned.
21 CHAIRMAN MCDADE: Wait, now --
22 ADMIN. JUDGE WARDWELL: You initially said 23 7184.
24 MR. SIPOS: And that is Exhibit New York 25 488 --
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5257 1 CHAIRMAN MCDADE: Hold on one second.
2 MR. SIPOS: -- I believe.
3 ADMIN. JUDGE WARDWELL: While you're doing 4 that, I want to correct something else that you put 5 words in Dr. Lahey's mouth and he said right and I 6 want to verify he meant to say right to what you said.
7 Judge McDade said that you believe that synergism and 8 combined are the same. That isn't what I heard you 9 say earlier today. I heard you say today synergism 10 was more than the combined.
11 DR. LAHEY: My understanding of what he 12 said and why I said right, was he indicated that they 13 were the same as far as I was concerned. What I 14 thought was synergism and what they called combined 15 was the same.
16 ADMIN. JUDGE WARDWELL: But it's not the 17 sum of the two components in your opinion is 18 synergism. Synergism is more than that, is it not?
19 DR. LAHEY: I think it can be, yes. But it 20 may also be the same.
21 ADMIN. JUDGE WARDWELL: Be what?
22 DR. LAHEY: Well, no, I think it can be 23 either one.
24 ADMIN. JUDGE WARDWELL: You think synergism 25 can be just the sum of the two?
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5258 1 DR. LAHEY: Can have two things --
2 synergism can mean two things going on at the same 3 time or --
4 ADMIN. JUDGE WARDWELL: And it's equal to 5 only the sum of the two individual contributions?
6 DR. LAHEY: It can be, I believe, and it 7 can be --
8 ADMIN. JUDGE WARDWELL: Greater?
9 DR. LAHEY: -- more.
10 ADMIN. JUDGE WARDWELL: Okay. So you 11 believe it's both the sum and/or?
12 DR. LAHEY: I believe so. I don't know 13 what the author or --
14 ADMIN. JUDGE WARDWELL: No, what I'm asking 15 you is what's your --
16 DR. LAHEY: Yes.
17 ADMIN. JUDGE WARDWELL: -- when you say the 18 word synergism --
19 DR. LAHEY: That's my view.
20 ADMIN. JUDGE WARDWELL: -- do you mean it 21 has to be more than the sum or can it be the sum or 22 more?
23 DR. LAHEY: Yes.
24 ADMIN. JUDGE WARDWELL: No, those are 25 choices.
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5259 1 (Laughter.)
2 DR. LAHEY: It can be either or.
3 ADMIN. JUDGE WARDWELL: Thank you. Moving 4 on to see if we can confuse some other stuff or I can 5 confuse some other stuff. I want to go back to the --
6 CHAIRMAN MCDADE: Well, let me clarify. I 7 was referring to New York Exhibit 488, NUREG/CR-7184.
8 ADMIN. JUDGE WARDWELL: Right.
9 CHAIRMAN MCDADE: If I said something 10 different, I just misspoke and I apologize.
11 ADMIN. JUDGE WARDWELL: Well, you said it 12 right, Dr. Hiser said 85 and we wanted to make sure --
13 DR. HISER: I apologize.
14 MR. COX: Could I add a clarification on 15 the report numbers?
16 (Laughter.)
17 CHAIRMAN MCDADE: Pardon?
18 MR. COX: I just -- this is Alan Cox with 19 Entergy. I just wanted to point out there's two 20 versions of the NUREG/CR-7184.
21 CHAIRMAN MCDADE: I was reading from the 22 December 2014. Is that the latest?
23 MR. COX: There is I believe a later 24 version. It is New York State, you said 574?
25 CHAIRMAN MCDADE: No, I said 488.
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5260 1 MR. COX: Okay. There's a 574, it's also, 2 it's a New York State 000574, it's NUREG/CR-7184. And 3 I think Dr. Lahey alluded to two versions of this 4 report, so those may be the two versions. One of 5 those may have the word synergy and the other may not.
6 CHAIRMAN MCDADE: Okay. Thank you, Mr.
7 Cox.
8 ADMIN. JUDGE WARDWELL: Now we get to move 9 to what we've already discussed, but I want to clarify 10 again in regards to the clevis bolts that New York 11 State testimony, 482, Page 56 to 57, and 56 it's Lines 12 20 to 23 and 57 it moves on to Lines 1 through 9. It 13 says that failures of the clevis insert bolts 14 apparently caused by PWSCC were detected at a 15 Westinghouse designed reactor in 2010. Out of the 48 16 bolts in this reactor, 29 were partially or completely 17 fractured, but only seven of those damaged bolts were 18 visually detected as having failed.
19 Despite this high rate of failure, about 20 60 percent of the total bolts were damaged, and a low 21 rate of visual detection, only 24 percent of the 22 damaged bolts were detected, the Applicant proposes to 23 managing the aging degradation of clevis insert bolts 24 with visual VT3 inspections rather than preemptive 25 replacement. And I guess I'll reaffirm again that, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5261 1 Entergy, do you agree with these numbers in regards to 2 what was previously reported from Westinghouse?
3 DR. LOTT: I agree with the scale of the 4 numbers, I haven't checked the actual --
5 ADMIN. JUDGE WARDWELL: There's no reason 6 not to believe those numbers?
7 DR. LOTT: No reason not to believe those 8 numbers.
9 ADMIN. JUDGE WARDWELL: And how does that 10 low percentage, again, support the use of VT3 as an 11 inspection, when in fact almost three-quarters of 12 damaged ones will go undetected?
13 DR. LOTT: It was never our intention to 14 inspect the bolts, it was our intention to verify the 15 stability and location of the clevis itself. As we 16 talked before, the clevis can function perfectly well 17 without the bolts if it's in place. So, again, our 18 inspection is basically -- and we made recommendations 19 in the Tech Bulletin that we issued on this subject to 20 basically inspect the clevis for its seating into the 21 lug on the vessel wall and to make sure that it had 22 not moved, that there was not undue wear on the 23 surfaces that would indicate that it had moved or was 24 free to move. But we did not anticipate that a visual 25 inspection would necessarily detect cracking of a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5262 1 bolt.
2 ADMIN. JUDGE WARDWELL: But did you not 3 also go on and testify that you're considering or 4 evaluating whether or not to replace any of the bolts 5 that may have been damaged?
6 DR. LOTT: Well, first of all, at that 7 particular plant a number of the bolts were replaced.
8 They were replaced, again, because of our concern 9 about the commercial aspects, the potential that those 10 clevises might become dislodged, which would make it 11 difficult for them to, as we said, reinsert the 12 barrel, restart the plant. So we certainly were 13 advising them, and I think it explains this in the 14 Technical Bulletin, that it might be wise to replace 15 the bolts, but not because of a safety concern.
16 ADMIN. JUDGE WARDWELL: Dr. Lahey, in your 17 testimony, 482, Page 58, Lines 5 through 9, you state 18 that rather than taking proactive steps to replace the 19 degraded clevis bolts prior to failure, the Applicant 20 proposes to wait for failures to occur before taking 21 steps to address the problem, an approach that is 22 totally unacceptable in my opinion. Hearing what you 23 heard now, do you still believe that your demand for 24 wholesale replacement of these bolts are a reasonable 25 expenditure of effort?
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5263 1 DR. LAHEY: Your honor, this is Richard 2 Lahey from New York. When we had this discussion 3 earlier today, it was my understanding that to replace 4 a few of the bolts is straightforward, if you have a 5 lot of the bolts, it requires a massive effort in 6 terms of realignment. And, so, I believe based on 7 that and the safety significance of it, it would 8 really depend on the degree of failure of the bolts.
9 What would make sense in any event, you don't want 10 loose parts rattling around.
11 ADMIN. JUDGE WARDWELL: In regards to that 12 last statement, Entergy, do these failures of the 13 bolts end up with loose parts? Or are they kind of 14 contained with the --
15 DR. LOTT: It's contained within the 16 system. It's difficult for the bolt head -- first of 17 all, there's locking bars. If there was wear through 18 or failure of the locking bars, still the bolt head 19 could not escape and the threaded part of the bolt 20 can't get past the head.
21 ADMIN. JUDGE WARDWELL: Thank you. Just 22 reading through here to -- Staff, is there any place 23 within your SER or SE that you evaluated the clevis 24 insert bolts and what is an appropriate action level, 25 if any, for these documented --
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5264 1 MR. POEHLER: This is Jeffrey Poehler for 2 the Staff. Yes, we do address that in our SER in the 3 discussion of operating experience, which is located 4 right after the ten element discussion. And we --
5 yes, we reviewed that quite extensively and we agreed 6 with continuing to do VT3 examination as an acceptable 7 means of managing the potential for bolt failures.
8 ADMIN. JUDGE WARDWELL: Okay, thank you.
9 New York State's testimony, 482, Page 57, Line 20 10 through 23, and then moving over to 58, Line 1, says, 11 the Applicant's analysis of the effects of clevis bolt 12 failures assumes that all of the components will be 13 functioning according to their design specification 14 and does not consider the fact that other components 15 may also be undergoing degradation from various 16 interacting mechanisms.
17 Entergy's testimony, Exhibit 616, Answer 18 166, Page 107, says, Entergy is not required to assume 19 without evidence that other components that are within 20 the scope of the reactor vessel internals AMP or any 21 other AMP are also degraded when it evaluates the 22 functionality of the clevis insert bolts. And I guess 23 considering we have the two competing statements, I'll 24 go to Staff and ask them. What do you look for and 25 what do you consider in regards to the potential for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5265 1 other components outside of the AMP, the RVI AMP, that 2 might not be functioning along with these clevis bolts 3 and how are they assessed? Or is there no requirement 4 to do that and no need for safety to do that?
5 MR. POEHLER: This is Jeffrey Poehler from 6 the Staff. We would not require them to assume that, 7 for instance, the part that interfaces with the clevis 8 insert is called the radial key and that's attached to 9 the bottom of the core barrel, and we would not 10 require them to assume that, that had failed or expect 11 them to do that when doing their functionality 12 evaluation of the clevis or the lower radial support 13 system. And in addition to that, they're both 14 redundant components, both the clevis inserts and the 15 radial keys, I believe there's six. So the likelihood 16 of a significant number of those failing at once is 17 low. And the radial keys were not even a -- they were 18 a no additional measure component, which means there 19 were no inspection requirements in MRP 227-A.
20 ADMIN. JUDGE WARDWELL: Are there any 21 components that do have inspection requirements that 22 are associated with one another where you have looked 23 at the potential failure of both in the reactor vessel 24 internals?
25 DR. HISER: In general --
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5266 1 ADMIN. JUDGE WARDWELL: Not necessarily 2 with the clevis system.
3 DR. HISER: This is Allen Hiser with the 4 Staff. We do not require postulation of failures of 5 other components when assessing a finding of a 6 degraded component. That's not a part of the 7 regulatory process.
8 ADMIN. JUDGE WARDWELL: And do you know any 9 technical basis for supporting that position?
10 DR. HISER: I think technical basis is just 11 Commission position that it's not required as a part 12 of the license renewal review and that annotation of 13 license renewal.
14 ADMIN. JUDGE WARDWELL: Thank you. Let's 15 move on to the lower support columns. NRC's 16 testimony, 197, Answer 27 on Pages 35 to 36, the fuel 17 assemblies are supported inside the lower internals 18 assembly on top of the lower core plate, that's LCP, 19 and the function of the LCP are to position and 20 support the core and provide a metered control of 21 reactor coolant flow into each fuel assembly. The 22 support columns transmit vertical fuel assembly loads 23 from the LCP to the much thicker lower support casing.
24 The function of the lower support casing is to provide 25 support for the core. My question to Staff, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5267 1 considering it is your testimony, are these support 2 columns you're referring to the lower support columns?
3 Are they one and the same?
4 MR. POEHLER: This is Jeffrey Poehler of 5 the Staff. That's correct.
6 ADMIN. JUDGE WARDWELL: And for these 7 columns, what is the primary mechanism for developing 8 flaws from aging? Is it driven by normal operating 9 conditions?
10 MR. POEHLER: It would be normal operating 11 conditions.
12 ADMIN. JUDGE WARDWELL: And would other 13 operating conditions, such like seismic or LOCA 14 events, likely be the primary contributor to service 15 induced flaws and, if not, why not?
16 MR. POEHLER: No, those other events, like 17 seismic events would not be a significant contributor 18 because those events occur very infrequently. So the 19 --
20 ADMIN. JUDGE WARDWELL: Your testimony on 21 197, Answer 171, Page 92, says that the Action Level 22 7 requires an applicant or a licensee to perform a 23 plant specific analysis of the cast Austenitic 24 stainless steel RVI components to demonstrate that 25 components will remain capable of performing their NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5268 1 intended functions through the end of the plant life.
2 Your Answer on 163, Page 93, goes on to state that 3 Entergy identified the only components requiring a 4 response to Action Level 7 for IP2 and 2 are the lower 5 support columns.
6 Only the upper portion of the lower 7 support columns, known as the column cap, is made from 8 CASS. The lower support columns are an expansion 9 component of MRP 227-A and the associated linked 10 primary component is a control rod guide tube lower 11 flange welds. So, let me make sure I understand this 12 correctly. Is this not saying to me that the only 13 CASS materials in the population of reactor vessel 14 internals are the upper portion of the lower support 15 columns, i.e., what's called the column cap?
16 MR. POEHLER: It's not the only CASS 17 component in the internals, or the only type of CASS 18 component, I should say, at Indian Point. It is the 19 one that is of most concern to the Staff because of 20 the high irradiation levels that parts of it can 21 experience.
22 DR. HISER: Dr. Wardwell, this is Allen 23 Hiser of the Staff. Action Item 7 specifically 24 identifies the lower support column bodies as within 25 the scope of that Action Item. And it is due to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5269 1 potential for thermal embrittlement and irradiation 2 embrittlement of those. And those are the only ones 3 that were identified that would be within the scope of 4 irradiation embrittlement. So that's why it's limited 5 only to the lower core support columns.
6 ADMIN. JUDGE WARDWELL: But does the Action 7 Level 7 state that only that component is required or 8 is it more general that it requires an applicant or a 9 licensee to perform a plant specific analysis of CASS 10 components?
11 DR. HISER: Those are specified within the 12 Action Item, the lower core support columns for 13 Westinghouse plants. The intent of that Action Item 14 was to identify CASS that's within the high fluence 15 field that could lead to irradiation embrittlement.
16 Within the context of what was evaluated in MRP 227, 17 that was the only generic CASS component. Referring 18 back to Action Item 2, where plants are to evaluate 19 differences between the plant specific configuration 20 and MRP 227, if CASS was atypically used in a place at 21 Indian Point different from MRP 227, then that would 22 have been identified in Action Item 2 and also should 23 show up in Action Item 7.
24 ADMIN. JUDGE WARDWELL: And so what's the 25 criteria for being under 2 again? For an applicant to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5270 1 highlight whether a CASS composed material, an 2 internal composed of CASS, that's the way I want to 3 say it, I guess, would or would not require the 4 additional attention brought on by Action Level 7?
5 DR. HISER: It says that applicants should 6 review the information in Tables 4-12 and 4-2 in MRP 7 189 Rev 1, Tables 4-4 and 4-5 in MRP 191, and identify 8 whether these tables contain all the RVI components 9 that are within the scope of license renewal for their 10 facilities. I think as a part of that, because 11 material is critical, that, that would be identified 12 as well as a difference between the plant specific 13 configuration and the MRP 227 assumptions.
14 ADMIN. JUDGE WARDWELL: I guess I'm still 15 a little confused. Are the upper portion of the lower 16 support columns the only RVIs under license renewal 17 that are made of CASS? And, if not, how did the 18 others get screened out and what was the criteria for 19 screening them out, for not being highlighted under 20 Action Level 7?
21 DR. HISER: This is Allen Hiser of the 22 Staff. For Indian Point, our understanding is those 23 are the only CASS components that are subject to 24 irradiation embrittlement above the threshold level 25 used in the development of MRP 227. So they would be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5271 1 the only CASS components that would be subject to the 2 potential susceptibility of thermal embrittlement and 3 irradiation embrittlement.
4 ADMIN. JUDGE WARDWELL: And what is that 5 screening criteria that would separate out those that 6 were and were not susceptible to thermal embrittlement 7 and irradiation embrittlement?
8 MR. POEHLER: This is Jeffrey Poehler of 9 the Staff. The MRP screening criteria was one dpa.
10 ADMIN. JUDGE WARDWELL: Okay. And 11 Westinghouse, is everything that was stated consistent 12 with your approach and is, as a net result, the upper 13 portion of the lower column supports, that is made of 14 CASS material, the only RVI that does not meet the 15 screening criteria of one dpa of fluence and, 16 therefore, is part of and falls under Action Level 7?
17 DR. LOTT: For the most part, I believe 18 that's true. I'm recalling --
19 ADMIN. JUDGE WARDWELL: Well, then the 20 other parts I'm interested in.
21 DR. LOTT: I know and I'm trying to get 22 there. I'm not -- really I'm trying to get there.
23 And there's one component that I need to check on and 24 that might be the upper support column, which I 25 believe in my previous testimony indicated that it was NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5272 1 potentially CASS. I don't know if it is or not.
2 ADMIN. JUDGE WARDWELL: Let's put that on 3 your homework list for tomorrow morning then to 4 determine whether there are any RVIs --
5 DR. LOTT: Yes. I would point out that 6 there are some other CASS components in the system and 7 that part of the evaluation procedure that led to this 8 screening process in MRP 191 was an evaluation of the 9 impact of those components or of those degradation 10 mechanisms. Of course, they would have been 11 identified for thermal embrittlement anyway.
12 ADMIN. JUDGE WARDWELL: Do you agree that 13 the one dpa is the threshold screening criteria or is 14 it not?
15 DR. LOTT: Yes.
16 ADMIN. JUDGE WARDWELL: It is?
17 DR. LOTT: Yes, it is. But --
18 ADMIN. JUDGE WARDWELL: So we're back to --
19 DR. LOTT: -- what I would say, what the 20 classification process that identified components to 21 be in MRP 227 or not also had this evaluation of the 22 structural significance of the component. So it's 23 possible -- and as I understand the documentation in 24 Action Item 7, it basically said that the NRC had 25 reviewed those decisions in MRP 191 and said that if NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5273 1 things were appropriately dealt with there, there was 2 not a need for functionality analysis under Action 3 Item 7. So it grandfathered in things that were 4 identified within MRP 191 as potentially CASS and 5 evaluated and screened out, are not required to go 6 through this process. And that I'm believe is true of 7 the upper support columns.
8 ADMIN. JUDGE WARDWELL: Well, that's -- I'm 9 interested in whether or not all of those that haven't 10 either been grandfathered out by 191 or haven't been 11 screened out due to the criteria of the one dpa are 12 now present in or not highlighted by 7 now, the 13 reasons why they aren't? Well, I do not want to read 14 that question in the transcript at all.
15 (Laughter.)
16 DR. LOTT: I think I can -- I don't think 17 that, that description you -- there are any components 18 that meet that description that you just said.
19 ADMIN. JUDGE WARDWELL: I just want to have 20 you verify that you have captured all of those that 21 need to be captured and that the lower support column 22 is the only one that needs to be captured by 7.
23 DR. LOTT: Okay. Action Item 7, that I 24 believe is true.
25 MR. STROSNIDER: This is Jack Strosnider NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5274 1 for Entergy. I just want to say, one other parameter 2 that may come into consideration here is the 3 resistance of CASS material to any type of formation 4 of any type of corrosion cracks. And we haven't 5 talked about that, but this material, although it can 6 permeate and we talked about that earlier, depending 7 on the ferrite content, et cetera, but the good thing 8 about this material is that it's very hard to crack.
9 And I'm not sure that there's any operating experience 10 in which we actually have seen cracking in these CASS 11 materials. To my knowledge, we haven't, unless 12 there's something fairly recently. So, that may also 13 factor into how this is treated and we'll have to look 14 at that.
15 ADMIN. JUDGE WARDWELL: Thank you. Back to 16 Entergy again. The upper portion of the lower support 17 column is made of CASS, what's the lower portion made 18 of?
19 DR. LOTT: The wrought stainless steel.
20 ADMIN. JUDGE WARDWELL: The raw?
21 DR. LOTT: Wrought stainless steel. I'm 22 sorry, I'll get closer to the microphone again.
23 ADMIN. JUDGE WARDWELL: I thought I heard 24 you say raw. Okay, wrought stainless steel. Thank 25 you. Dr. Lahey, do you have any knowledge of any NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5275 1 potential RVIs that you believe should fall under 2 A/LAI 7 in regards to the Aging Management Program?
3 DR. LAHEY: Your honor, I have reviewed 4 EPRI 191 and I'll have to look back on it, it's in the 5 other room. But my recollection is there were core 6 plates, upper and lower, that were castings, but I'll 7 have to verify that.
8 ADMIN. JUDGE WARDWELL: Okay. So, Entergy, 9 you might note that also that those plates are of 10 interest and comment on that if you would.
11 DR. LOTT: Okay.
12 ADMIN. JUDGE WARDWELL: Thank you, Dr.
13 Lahey. Answer of NRC's 197, 163 on Page 94, states 14 that the lower support columns for IP2 and 3 are made 15 from type CF-8 stainless steel, which is a low 16 molybdenum contact grade of cast stainless steel. Low 17 molybdenum cast grades are less susceptible to thermal 18 embrittlement than the high moly cast grades, such as 19 types CF-8M. Entergy determined the lower support 20 columns at IP2 and 3 were not susceptible to thermal 21 embrittlement. And, again, do these statement refer 22 only to the cap as being not susceptible to thermal 23 embrittlement because the lower wrought iron has 24 already been screened out, it's not part of A/LAI 7?
25 DR. LOTT: Yes. The lower cast of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5276 1 Austenitic stainless steel, wrought stainless steel, 2 is not subject to thermal embrittlement. It would 3 only be the upper cast portion.
4 ADMIN. JUDGE WARDWELL: And so whenever we 5 talk about -- whenever you reference lower support 6 under this section of the CASS portion, you mean just 7 the upper part? Let me rephrase that question. When 8 you are referring to the cap of the lower support 9 system, that's what you're referring to is that upper 10 part that's made out of CASS, is that correct?
11 DR. LOTT: Yes.
12 ADMIN. JUDGE WARDWELL: And is that the top 13 half of the column? Is it at the top little piece of 14 the column? Or is it a top quarter of the column?
15 DR. LOTT: It's more like quarter to a 16 third of the column. I don't think it --
17 ADMIN. JUDGE WARDWELL: But it's not that 18 little, I saw a diagram of the column and there's a 19 little rectangular piece on the top, that's not the 20 cap, it's more than that?
21 DR. LOTT: No, it's more than that.
22 ADMIN. JUDGE WARDWELL: It's a --
23 DR. LOTT: Yes.
24 ADMIN. JUDGE WARDWELL: -- major portion --
25 DR. LOTT: It's --
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5277 1 ADMIN. JUDGE WARDWELL: -- a portion of 2 that --
3 DR. LOTT: Right.
4 ADMIN. JUDGE WARDWELL: -- entire column?
5 MR. DOLANSKY: It has some length to it.
6 ADMIN. JUDGE WARDWELL: Yes, okay. Just so 7 we have a feeling of it. And by screening it out as 8 being unsusceptible to thermal embrittlement, that by 9 necessity means that the caps cannot be susceptible to 10 the combined effects of irradiation embrittlement and 11 thermal embrittlement, it's just the irradiation 12 embrittlement, is that correct? Entergy?
13 MR. DOLANSKY: Yes, that's correct. This 14 is Bob Dolansky. Yes, that's correct.
15 ADMIN. JUDGE WARDWELL: And, Staff, as far 16 as you're concerned, you agree with that conclusion?
17 MR. POEHLER: This is Jeffrey Poehler with 18 the Staff. Yes, we agree with that conclusion.
19 ADMIN. JUDGE WARDWELL: Dr. Lahey, do you 20 agree that just the upper portion of this column cap 21 is --
22 DR. LAHEY: Yes, sir, that's my 23 understanding as well.
24 ADMIN. JUDGE WARDWELL: Thank you. The A 25 163, Page 95 goes on to say that screening criteria NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5278 1 would allow the low molybdenum cast with delta ferrite 2 content of 20 percent or less to be screened out, that 3 is not susceptible to thermal embrittlement. Since 4 portions of the lower support column will have neutron 5 fluences at the end of life greater than one times ten 6 to the 17 neutrons per square centimeter, the Staff 7 did not accept Entergy's screening and the Staff 8 updated the criteria of low molybdenum statically cast 9 CASS with a ferric content of less than 15 percent can 10 be screened out for thermal embrittlement and any 11 synergistic effects of TE and IE and as low molybdenum 12 statically cast CASS is only susceptible to 13 irradiation embrittlement at fluences greater than one 14 times ten to the 21 neutrons per square centimeter or 15 1.5 displacements per atom.
16 Staff, could you try to condense that 17 whole confusing thing I just read, or a bit confusing 18 to me, in regards to how these lower columns were 19 specifically, and specifically the lower portion of 20 this that's called the cap -- the upper portion of the 21 lower column, were screened out and assessed by you?
22 DR. HISER: Okay. This is Allen Hiser of 23 the Staff. The top of that paragraph talks about the 24 20 percent or less screening for thermal 25 embrittlement.
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5279 1 ADMIN. JUDGE WARDWELL: And that's for 2 delta ferrite content, correct?
3 DR. HISER: Right. That's correct. And 4 that was proposed for really CASS that was not subject 5 to neutron embrittlement. So it was subject to low 6 neutron fluence. With the possibility of higher 7 neutron fluence levels combined with the potential for 8 thermal embrittlement, we went back and looked at the 9 screening, the 20 percent, and concluded that we 10 should reduce that level by five percent to 15 11 percent. And we thought that, that was a reasonable 12 screen to preclude the potential synergism of thermal 13 embrittlement and irradiation embrittlement.
14 When Indian Point then came in with their 15 measured ferrite values, or their ferrite values for 16 their lower core support column caps, and it was below 17 15 percent, we thought it was reasonable to screen 18 that out then from potential synergistic effects of 19 thermal embrittlement and irradiation embrittlement.
20 So what that leaves then for those components is only 21 a concern with the irradiation embrittlement on the 22 fracture toughness for those materials.
23 ADMIN. JUDGE WARDWELL: And the fluence 24 screening and any molybdenum screening didn't play a 25 part in this?
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5280 1 DR. HISER: The molybdenum screening did 2 not because this was a low molybdenum material. The 3 fluence level did from the perspective that the 4 fluence exceeded one dpa and, therefore, the 15 5 percent delta ferrite screening level was implemented 6 instead of the 20 percent that would apply for lower 7 fluence CASS material.
8 ADMIN. JUDGE WARDWELL: Thank you. Dr.
9 Lahey, do you agree that this was an appropriate 10 screening process and that the lower column caps have 11 been successfully screened and are being assessed in 12 the way they should be for irradiation embrittlement 13 alone?
14 DR. LAHEY: Yes, your honor. Given the 15 criteria that's been established, it sounds like it 16 has been properly done.
17 ADMIN. JUDGE WARDWELL: Thank you, sir.
18 New York State's testimony on 482, Page 18, Lines 16 19 through 22, discusses a recent report prepared by 20 Argonne National Laboratory for the U.S. NRC and 21 acknowledges with respect to CASS materials that are 22 used for the lower support column caps that combined 23 effect of thermal aging and irradiation embrittlement 24 could reduce the fracture resistance even further to 25 a level neither of these degrading mechanisms can NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5281 1 impart alone.
2 And I guess I would just ask, and we've 3 already verified that it's been screened correctly, so 4 really what my question is here, I want to make sure 5 what Argonne National Laboratory work and report that 6 you were referring to in this statement. Do you 7 recall, Dr. Lahey? Is it NUREG/CR-7184 or is there 8 some other one? Because you go on, on Page 20, to 9 talk about NUREG-7184 and then even the Chopra report, 10 the degradation of light water reactor core internal 11 materials due to neutron irradiation, which is NUREG-12 7027. And I'm just trying to sort out all these 13 reports.
14 DR. LAHEY: Your honor, I'd have to go back 15 and look to know for sure.
16 ADMIN. JUDGE WARDWELL: Okay. Add that on 17 to your list.
18 DR. LAHEY: Yes.
19 ADMIN. JUDGE WARDWELL: And do try to get 20 to bed by at least 3:00 a.m. if you could. Because 21 we're starting here tomorrow at 4:00 aren't we?
22 (Laughter.)
23 CHAIRMAN MCDADE: Thirty.
24 ADMIN. JUDGE WARDWELL: 4:30, I'm sorry.
25 (Laughter.)
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5282 1 ADMIN. JUDGE WARDWELL: Yes, thank you.
2 And specifically to get back to that --
3 CHAIRMAN MCDADE: I digress for a second, 4 but Judge Wardwell has stayed on Central European Time 5 forever. So he insists that we do start early and 6 would prefer that we start at 4:00 Eastern.
7 ADMIN. JUDGE WARDWELL: No, because I 8 wouldn't be ready. I need that time for being ready.
9 (Laughter.)
10 ADMIN. JUDGE WARDWELL: It's your testimony 11 in 482 on Page 18, Line 16 through 22, and then it's 12 your testimony Page 20 on Lines 6 through 10 where 13 these various -- on 16 through 22, you talk about the 14 Argonne National Laboratory report and then you go on, 15 on Page 20 to cite both 7184 and 7027.
16 DR. LAHEY: Okay.
17 ADMIN. JUDGE WARDWELL: NRC's testimony on 18 197, Answer 164, Pages 96 to 97, states that the Staff 19 found that Entergy adequately addressed Action Level 20 7 based on the following. One, Entergy evaluated the 21 CASS components of the RVIs and that is limited to the 22 lower support column caps. Entergy screened the 23 column caps for TE and IE using plant specific 24 materials data and determined that the column caps are 25 not susceptible to TE. And Staff confirmed the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5283 1 results of the screening using its own screening 2 criteria.
3 Three, Entergy provided information on 4 fabrication, non-destructive NDE demonstrating that 5 preexisting flaws are unlikely to exist in the column 6 caps. Four, Entergy provided information on expected 7 stresses and neutron fluences for the column caps that 8 demonstrated the service induced cracking due to 9 irradiation assisted stress corrosion cracking is 10 unlikely. And, five, Entergy modified its reactor 11 vessel internals program to include a link to a lead 12 component that is an appropriate predictor of IASCC 13 and IE for column caps with an appropriate schedule 14 for performing the expansion inspection if necessary.
15 Therefore, the Staff found that the 16 formation provided by Entergy provides reasonable 17 assurances that the functionality in the column caps 18 will be maintained during this period of extended 19 operation. Entergy goes on in their testimony and 20 states in Exhibit 616 under Answer 197, Page 131, that 21 in regards to the inspection criteria, the lower core 22 barrel cylindrical girth weld reveals that it is more 23 likely to experience this irradiation assisted stress 24 corrosion cracking than the lower support column caps.
25 So it is appropriate to link the two with the weld as NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5284 1 a primary component and the LSCC as the expansion 2 component.
3 And I guess I'll start with Staff on this 4 and say that, in your approval of this as summarized 5 in that, that I just read, you mentioned about linking 6 to a component that is the primary component and I 7 thought earlier, which I think I read, you stated that 8 the control rod guide tubes lower flange welds were 9 the primary link to the lower support column caps.
10 And now Entergy is claiming that it's the lower core 11 barrel cylinder girth weld is the one. Which is it?
12 MR. POEHLER: At this time, it's both 13 because they added the girth weld as a plant specific 14 primary link. So in MRP 227-A, the generic primary 15 link for the lower core support columns is the control 16 rod guide tube lower flange. When the Staff was 17 reviewing the Action Item 7 information submitted by 18 Entergy, we were satisfied with what they had given 19 us, but we still had a lingering concern that for 20 irradiation embrittlement, because irradiation 21 embrittlement still applies even though thermal 22 embrittlement was screened out, we still had a 23 lingering concern that irradiation embrittlement 24 wasn't appropriately predicted by the standard MRP 25 primary link.
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5285 1 Because that control rod guide tube lower 2 flange weld is a relatively low fluence component, 3 didn't even screen-in for irradiation embrittlement or 4 irradiation assisted stress corrosion cracking, which 5 are two mechanisms that do apply to the lower support 6 columns. So we asked another RAI to Entergy regarding 7 an appropriate -- can you propose an appropriate 8 primary link for these mechanisms? And that's when 9 they proposed the lower core barrel girth weld. And 10 we reviewed that information that they submitted and 11 concluded that it was a more appropriate primary link 12 for those mechanisms of irradiation embrittlement and 13 irradiation assisted stress corrosion cracking.
14 ADMIN. JUDGE WARDWELL: Okay. Thank you.
15 Dr. Lahey, do you have any reason to not accept those 16 as appropriate primary links to these expansion 17 components?
18 DR. LAHEY: The use of the girth weld as a 19 proxy for the core cap, it seems an odd choice 20 actually because of the materials used and how they 21 would respond to thermal and irradiation effects. And 22 I know they needed to use something because they 23 weren't able to do an inspection of the core caps, but 24 it just seems it's a very odd choice. And I must say 25 when I was reading the ACRS testimony on this, they NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5286 1 had exactly the same opinion. So I know what they 2 did, it just seems odd.
3 ADMIN. JUDGE WARDWELL: And do you feel 4 it's also odd to use the control rod guide tube lower 5 flange welds as the other primary?
6 DR. LAHEY: It's not a casting. Yes.
7 ADMIN. JUDGE WARDWELL: So, Entergy, why 8 have you proposed the lower core barrel cylinder girth 9 weld as the primary?
10 DR. LOTT: Well, as Mr. Poehler explained, 11 it was basically a request coming from the NRC to 12 consider an alternative to the lower control rod 13 guide, I can't say it either myself, the CRGT lower 14 flange welds. And we looked at similarities in 15 materials and we understood that -- we looked at the 16 lower support column caps as a low ferrite material, 17 not susceptible to thermal embrittlement. And we 18 looked at the core barrel weld material and realized 19 it also would have a duplex structure, a low ferrite 20 content, and then some very similar characteristics.
21 Both components had similar fluences and in an 22 abundance of caution here, I think even though we 23 probably could have demonstrated that it would screen-24 out for irradiation assisted stress corrosion 25 cracking, I think we just wanted to make sure that we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5287 1 were looking for the most likely locations for cracks.
2 ADMIN. JUDGE WARDWELL: Okay.
3 DR. LOTT: And let me just one further --
4 I think the tipping point that made the core barrel 5 the primary location for this particular situation was 6 that based on the potential for residual stresses in 7 the core barrel weld, we expect the stresses, 8 particularly on the surface, to be much higher in the 9 core barrel weld, where as we've just explained, we 10 looked through the normal operating stresses in the 11 lower support columns and found they're very limited.
12 So we would expect to see cracking in these kinds of 13 materials at these fluences first in the core barrel 14 and later in the lower support column, if ever.
15 ADMIN. JUDGE WARDWELL: Okay. What about 16 the control rod guide tubes lower flange welds, will 17 they also show up any cracking prior to the lower 18 support column caps?
19 DR. LOTT: Well, again, they have some 20 potential to have some residual stresses and some 21 other concerns that might put them ahead on the list.
22 But because they weren't irradiated, we felt that this 23 addition that Entergy has of using both as a primary 24 component made the most sense.
25 ADMIN. JUDGE WARDWELL: Now, then for Staff NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5288 1 --
2 CHAIRMAN MCDADE: If I could interrupt for 3 a second. Dr. Lott, when we come back to you, if you 4 could just state your name, otherwise --
5 DR. LAHEY: Okay.
6 CHAIRMAN MCDADE: -- Mr. Cox is going to 7 get --
8 DR. LAHEY: I'm sorry.
9 CHAIRMAN MCDADE: -- blamed in the 10 transcript for things that you say.
11 DR. LAHEY: Yes, I'm sorry.
12 CHAIRMAN MCDADE: Okay.
13 ADMIN. JUDGE WARDWELL: Take credit. Dr.
14 Hiser, this discussion raises a question in my mind, 15 why shouldn't these lower support column caps be 16 elevated to a primary component? And what is the 17 criteria that puts something as a primary component as 18 opposed to an expansion component, if I've used the 19 right two terminologies?
20 DR. HISER: This is Allen Hiser of the 21 Staff. One of the reasons is that the column caps are 22 a highly redundant structure. And, therefore, it 23 would require multiple failures for the functionality 24 of the lower core support system to be challenged. In 25 this case, as Dr. Lott mentioned, the residual NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5289 1 stresses in the core barrel weld create a situation 2 where we expect that cracking would initiate much 3 sooner than in the column caps. Therefore, we believe 4 it's a reasonable linkage with the column caps being 5 an expansion component predicated on inspection 6 findings from the core barrel girth welds.
7 ADMIN. JUDGE WARDWELL: And why still hold 8 with the control rod guide tubes lower flange welds?
9 Why should that still be a primary for this? It 10 sounds like you don't feel comfortable that it will 11 supply any advanced notice of what might happen or 12 substitute notice for what might happen at the lower 13 support column caps.
14 DR. HISER: This is Allen Hiser with the 15 Staff. The CRGT welds, I believe, are still a primary 16 link for the column caps. Again, because of weld 17 residual stresses, we would expect the likelihood of 18 stress corrosion cracking to be higher there. So 19 they, again, if you were to look at a hierarchy of 20 where you expect things to occur first, the CRGT welds 21 and the lower core barrel girth weld would be higher 22 than the column caps.
23 ADMIN. JUDGE WARDWELL: Is it possible to 24 inspect the lower support column caps?
25 DR. HISER: From discussions we've had with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5290 1 Westinghouse, I believe it is possible, yes.
2 ADMIN. JUDGE WARDWELL: Okay.
3 DR. HISER: Our understanding is that it 4 would be an extremely difficult inspection to do, but 5 it is certainly possible and would be necessary 6 pending the results from the primary inspections, if 7 expansion was required then the plant would be 8 required to do those expansion inspections.
9 ADMIN. JUDGE WARDWELL: Dr. Lahey, after 10 hearing this additional explanation, do you have any 11 other comments in regards to the expansion monitoring 12 plan for the lower support column caps?
13 DR. LAHEY: No. I understand what they 14 said and I understand the rationale. Just personally, 15 I think if you're interested in core cap, it's better 16 to do the inspection on the core cap, even if it's a 17 more difficult inspection.
18 ADMIN. JUDGE WARDWELL: Okay, thank you.
19 New York State's testimony, 482, Page 59, Lines 1 20 through 3, states that the Applicant, based on a lack 21 of documented fractures of core support columns, 22 assumed that only a limited number of columns could 23 actually contain flaws of significant size. And I 24 guess I'll ask Entergy, did you assume only a limited 25 number of lower support column caps contained flaws or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5291 1 was this the result of any analysis or conclusion or 2 was it just an assumption?
3 DR. LOTT: I'm not sure exactly where we 4 think we -- my name is Randy Lott, I'm sorry, thank 5 you. I'm not sure exactly --
6 ADMIN. JUDGE WARDWELL: Alan doesn't want 7 to take credit for what you're saying or Mr. Cox 8 doesn't want to take credit for what you're saying.
9 DR. LOTT: I'm not sure exactly where we're 10 saying we made this assumption. We certainly did do 11 analyses as Dr. Hiser has suggested to look at 12 effectively the redundancy in the system, in the 13 system of calculations that are very similar to what's 14 done for the baffle former bolts to look at minimum 15 requirements, and found that there was a large degree 16 of redundancy in the system. That's all reported in 17 -- it's a proprietary report, but it's PWROG14-048, 18 which is ENT 667.
19 ADMIN. JUDGE WARDWELL: Did you have any 20 other knowledge of any potential flaws or lack thereof 21 in the lower support column caps?
22 DR. LOTT: We have no expectation of flaws.
23 MR. DOLANSKY: This is Bob Dolansky with 24 Entergy. I just want to add, we also -- for the lower 25 support column caps, when they were installed they had NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5292 1 detailed NDE radiography performed on them, which 2 would give very, very high confidence that there were 3 no flaws in those when they were installed.
4 ADMIN. JUDGE WARDWELL: And what is that 5 testing composed o?
6 MR. DOLANSKY: Basically similar to an x-7 ray on a person. It's an x-ray of the lower support 8 column cap, was performed on all column caps, and 9 determined that there was no flaws in -- it's the, I 10 don't want to say the best, but it's an extremely good 11 NDE technique used to verify that there's no flaws.
12 ADMIN. JUDGE WARDWELL: Thank you.
13 CHAIRMAN MCDADE: If I could jump in here 14 quick between questions and get something clarified.
15 A while ago, a few minutes, we were talking about, and 16 I believe that you said, Dr. Lahey, that you thought 17 it was an odd choice of using the core barrel girth 18 weld as a proxy for another item. Do you recall that 19 testimony?
20 DR. LAHEY: Yes.
21 CHAIRMAN MCDADE: Okay. And --
22 DR. LAHEY: For the column caps.
23 CHAIRMAN MCDADE: -- the control guide tube 24 flange, was it or --
25 DR. LAHEY: Well, I believe we were talking NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5293 1 about that and the cast column cap that was --
2 CHAIRMAN MCDADE: Okay. But what I'm 3 trying to get at is, what I thought I heard you say is 4 that you thought that it was an odd proxy because of 5 the differences in the materials that they were made 6 out of. And then I thought I heard Dr. Lott testify 7 that part of the reason that they were chosen is 8 because of the similarity of the materials. So, let 9 me go to Dr. Lott, would you state for the record what 10 components we're talking about and what they're made 11 of?
12 DR. LOTT: Okay. This is Randy Lott for 13 Entergy. We're talking about the large structural 14 weld basically around the middle, I'll call it the 15 belt line, of the core barrel and these column caps, 16 which obviously we've been discussing all along, which 17 are CF-8 material, but with a relatively low ferrite 18 content of about 15 percent. They're castings, which 19 means that effectively they're melt, cast, and cooled 20 in the system, which gives us this duplex grain 21 structure. In fact, we need some duplex structure in 22 order to get a successful casting, so that's -- in a 23 weld, you get a very similar micro-structure or 24 material structure because, again, it's a material 25 that's welded and solidified.
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5294 1 And you get very similar behavior in terms 2 of duplex construction of those material, albeit 3 sometimes a slightly lower ferrite content. So, in 4 the sense that they're both these duplex Austenitic 5 ferritic structures, there's some real similarities in 6 the contents. Particularly, it seemed to us, because 7 there was particularly low ferrite content at Indian 8 Point, it made sense to couple up these two materials.
9 Then, as I said, the controlling factor for us was the 10 large potential residual stresses, which would drive 11 much more cracking in the core barrel, much sooner, 12 than it would in the lower support columns.
13 CHAIRMAN MCDADE: Okay. Thank you, Dr.
14 Lott. And, Dr. Lahey, what were the dissimilarities 15 that you thought made it an odd proxy?
16 DR. LAHEY: This is Richard Lahey from New 17 York State. The dissimilarities are the metals 18 themselves. The weld has typically significantly 19 lower delta ferrite than would be in the casting. And 20 the stresses on the structure are significantly 21 different. As I understood it, one of the things they 22 liked about the two were the potential for corrosion 23 cracking. But it's a stretch, I just think it's quite 24 a stretch to look at one and say, it's going to happen 25 here before it happens there. So they're looking for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5295 1 the canary in the coal mine to decide what they might 2 want to do on the core cap. And I, frankly, I think 3 it's better to look at the coal mine, go down and 4 check the core cap.
5 CHAIRMAN MCDADE: Okay. Thank you, Dr.
6 Lahey.
7 ADMIN. JUDGE WARDWELL: Do you have any 8 reason to believe that the cracking would not occur 9 sooner in the lower core barrel cylinder girth weld 10 prior to anything happening with the lower support 11 column caps?
12 DR. LAHEY: Your honor, I think the 13 stresses on the things are -- there's no reason for 14 them to be similar. You have one loaded supporting 15 structure, you have the other one loaded in this way.
16 It's just -- to me it's a very odd choice. I heard 17 what they said, I do understand why they decided to do 18 it, but primarily I think it's driven by 19 accessibility.
20 ADMIN. JUDGE WARDWELL: Thank you.
21 DR. HISER: Judge Wardwell, could we 22 supplement that a little bit?
23 ADMIN. JUDGE WARDWELL: Yes, you may.
24 DR. HISER: I think you've heard a little 25 bit -- this is Allen Hiser of the Staff. You've heard NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5296 1 a little bit about the reasons that the Applicant 2 thought that was acceptable. The Staff basis is sort 3 of similar, but there may be at least two additional 4 factors. As Dr. Lahey mentioned, the column caps will 5 be in compression. The core barrel girth weld, the 6 CRGT welds will not. So it's much more likely that 7 they will exhibit stress corrosion cracking because of 8 the weld residual stresses first and the higher 9 membrane stresses from the operational loads.
10 In addition, the column caps are cast 11 Austenitic stainless steel, which has generally been 12 found to have a very, very low likelihood of 13 initiating stress corrosion cracks. So, therefore, 14 the material in the column caps is much better from a 15 likelihood of stress corrosion cracking. The other 16 materials have much higher stresses and so they will 17 be much more likely to have cracking prior to the 18 column caps. So that provides a more complete 19 picture, I think, of the basis for the Staff finding 20 that to be acceptable.
21 ADMIN. JUDGE WARDWELL: Thank you, Dr.
22 Hiser. New York State's testimony, 482, Page 61, 23 Lines 11 through 17, states that the reactor vessel 24 internal components made from non-cast stainless steel 25 will also experience the combined effects of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5297 1 irradiation induced embrittlement, corrosion, and 2 other aging mechanisms. The Applicant has failed to 3 evaluate the mechanisms that occur for many of the 4 other important and vulnerable RVI components, such as 5 the core baffles, the baffle bolts, and the formers.
6 Entergy's testimony, Exhibit 616, Answer 7 202, Page 135 to 136, states that these components are 8 not made of cast materials, so they are not 9 susceptible to thermal embrittlement. MRP 227-A and 10 the reactor vessel internals AMP identify irradiation 11 assisted stress corrosion cracking which, as the name 12 implies, is actually the result of multiple underlying 13 mechanisms itself as the aging mechanism of concern 14 for these components, again referring to the core 15 baffles, baffle bolts, and formers. And that these 16 components are all designated for primary inspections 17 under the RVI AMP and Inspection Plan. And I guess 18 I'll start with Entergy just quickly to confirm that 19 when you were referring to these components, you were 20 referring to the core baffles, baffle bolts, and 21 formers that you claim are designated as primary 22 inspections under the RVI AMP and Inspection Plan. Is 23 that correct?
24 DR. LOTT: Yes, I believe that's true.
25 ADMIN. JUDGE WARDWELL: Say --
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5298 1 DR. LOTT: Yes, I believe that's true.
2 This is Randy Lott.
3 ADMIN. JUDGE WARDWELL: Thank you. And, 4 Dr. Lahey, do you now agree that those core baffles 5 and the baffle bolts and the formers are handled by 6 the RVI AMP?
7 DR. LAHEY: It's clear that they are being 8 inspected using the techniques in MRP 227-A, if that's 9 what you mean.
10 ADMIN. JUDGE WARDWELL: Yes.
11 DR. LAHEY: Yes.
12 ADMIN. JUDGE WARDWELL: Okay. Do you have 13 any other comments on that? And they're being 14 inspected as the primary components, is that your 15 understanding also?
16 DR. LAHEY: Yes, that's my understanding.
17 ADMIN. JUDGE WARDWELL: Okay, thank you.
18 That's all I have.
19 ADMIN. JUDGE KENNEDY: This is Judge 20 Kennedy. Just going back to the baffle former bolts 21 and maybe this most recent discussion answered the 22 question, but earlier in today's testimony, you had a 23 colorful expression for the baffle former bolts being 24 subject to shock loads and you referred to unzipping 25 the rest of the bolting.
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5299 1 DR. LAHEY: Yes.
2 ADMIN. JUDGE KENNEDY: Is this still a 3 concern of yours given what we've gone through this 4 afternoon and all the other testimony on the baffle 5 former bolts?
6 DR. LAHEY: Yes, your honor, it is.
7 ADMIN. JUDGE KENNEDY: Within your 8 testimony, do you have any support for the assertion 9 that these bolts, when subject to a shock loading, 10 would fail catastrophically?
11 DR. LAHEY: If they are significantly 12 weakened, we're talking as we get out in time and 13 they've been irradiated significantly, they've been 14 subjected to fatigue, they've been subjected to 15 irradiation assisted stress corrosion cracking, and if 16 they're significantly weakened in that way and they're 17 subjected to a strong shock load, yes, I have a 18 serious concern about it.
19 ADMIN. JUDGE KENNEDY: Is this combination 20 of aging mechanisms all applicable to these bolts?
21 DR. LAHEY: I'm sorry, could you --
22 ADMIN. JUDGE KENNEDY: Is this combination 23 of aging mechanisms all applicable to these --
24 DR. LAHEY: I believe they are, yes.
25 ADMIN. JUDGE KENNEDY: Entergy, do you feel NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5300 1 that this combination of aging mechanisms would be 2 applicable to the baffle former bolts?
3 DR. LOTT: Yes. I mean, to the baffle 4 former bolts -- can you name the mechanisms again for 5 me -- I'm sorry, I'm -- they're subject to irradiation 6 embrittlement, they're subject to irradiation assisted 7 stress corrosion cracking, they're subject to 8 irradiation induced stress relaxation, they're 9 potentially subject to void swelling. So, yes, 10 they're -- and we model all of those components.
11 ADMIN. JUDGE KENNEDY: I guess the other 12 one I heard was metal fatigue. I don't know --
13 DR. LOTT: Metal fatigue is, yes, certainly 14 a possibility.
15 ADMIN. JUDGE KENNEDY: So is this now back 16 to the question of how much synergism there is between 17 all these aging mechanisms, if there is any?
18 DR. LOTT: Well, I mean, I will point out 19 that our experience with baffle -- we have experience 20 with failure rates in baffle former bolts, a fair 21 amount. We have a fair amount of data now on baffle 22 former bolts. So, we have -- I don't see a lot of 23 surprises coming forward in this testimony and because 24 we're monitoring, as we said, the effects and if we're 25 looking for the effects on cracking, whatever combined NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5301 1 effects are there, are there in that data base or in 2 that operating experience.
3 ADMIN. JUDGE KENNEDY: Does this data 4 include bolts subjected to higher fluences, towards 5 the end of design life? Maybe not extended life, but 6 design life at least?
7 DR. LOTT: Yes. I mean, some of these 8 examinations have been performed as part of similar 9 plant life license renewal applications, PEOs. So, 10 yes, they've been towards the end of plant life.
11 ADMIN. JUDGE KENNEDY: So it does include 12 some bolting that has been examined during the period 13 of extended operation for other plants?
14 DR. LOTT: At or near.
15 ADMIN. JUDGE KENNEDY: Okay. At or near.
16 I don't want to put words in your mouth.
17 MR. STROSNIDER: This is Jack Strosnider 18 for Entergy. I'd like to come back to the first part 19 of your question on loads for just a minute if I 20 could.
21 ADMIN. JUDGE KENNEDY: Okay, sure.
22 MR. STROSNIDER: We talked earlier about 23 the WCAP report that NRC reviewed and approved with 24 regard to the methodology for establishing the bolting 25 patterns, et cetera. That WCAP report evaluated the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5302 1 design basis dynamic loads associated with design 2 basis accidents, such as the loss of coolant, et 3 cetera. So to the extent that Dr. Lahey's concern 4 about dynamic loads, that those loads are within the 5 original design basis, they have been addressed. If 6 there's something over and above what was in the 7 original design basis, I'm not -- it's not clear, to 8 me at least, what they are and also I would suggest 9 that they're outside the space of license renewal.
10 But it's just not clear what loads he's talking about 11 that would be over and above what's in the licensing 12 basis. And if they're within the licensing basis, 13 they were evaluated.
14 ADMIN. JUDGE KENNEDY: And would those 15 bolts that were subjected to the design basis loads 16 include the effects of these various aging mechanisms?
17 Consideration of the effects of these various aging 18 mechanisms? Which seems to be the other part of Dr.
19 Lahey's concern.
20 MR. GRIESBACH: Your honor, this is Tim 21 Griesbach for Entergy. If you're asking whether those 22 multiple aging mechanisms have been considered as part 23 of the Aging Management Program for the baffle former 24 bolts, as Dr. Lott said, the answer's definitely yes.
25 They've also been identified as primary components for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5303 1 inspections and, to the best of my knowledge, there's 2 been at least 19 PWR units that have performed those 3 inspections already for MRP 227 and very few failed 4 bolts have been found throughout those inspections.
5 ADMIN. JUDGE KENNEDY: Is the foundation of 6 the strength of this bolting subject to these aging 7 mechanisms the lack of cracking? I mean, if it hasn't 8 cracked, it still maintains its design fatigue life or 9 design life?
10 MR. GRIESBACH: This is Tim Griesbach again 11 for Entergy. The answer is yes. Those materials 12 undergo strengthening, in fact, due to the irradiation 13 effects. So they are stronger than they would have 14 been initially and still maintain the margins that 15 were there from the beginning for that reason, if they 16 are uncracked.
17 ADMIN. JUDGE KENNEDY: And we're back to 18 the beginning. All right. Thank you very much.
19 DR. LAHEY: Your honor, I think I've been 20 asked the question many times, are my loads different 21 than your loads? They're the same loads. The only 22 difference is how impulsive these loads are. So it 23 has to do with how they're calculated. I haven't seen 24 the document which gives the details on that. If they 25 have a shock capturing routine, like adaptive grid, or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5304 1 they're done by method of characteristics or some way 2 where they don't smear out things, instead of hitting 3 it with a hammer, you're hitting it with a powder 4 puff, then I'm happy that everything's surviving. But 5 I haven't seen that. And I've done a lot of work in 6 the past on RELAP and TRAC and RETRAN, so if you're 7 using those kinds of codes, it's more like a powder 8 puff than it is a hammer. So that's the difference.
9 ADMIN. JUDGE KENNEDY: Would Entergy like 10 to respond to that? I mean, again, this is a question 11 of how you analyze the dynamic loads.
12 DR. LOTT: Well, I guess --
13 CHAIRMAN MCDADE: Dr. Lott?
14 DR. LOTT: I'm sorry. This is Randy Lott.
15 That information, I think, is dealt with in the 16 methodology documents that we talked about earlier, 17 and I'm sorry I don't have the number right in front 18 of me, but I think we discussed earlier the approved 19 methodology for analysis of the bolts, the computer 20 programs and the process are in that document. We 21 believe it's more than adequate and consistent with 22 the licensing basis.
23 MR. STROSNIDER: This is Jack Strosnider 24 for Entergy. I'd like to add to that, that the codes 25 that are used for doing these analyses are codes that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5305 1 have been reviewed and approved by NRC as acceptable 2 for analyzing the design basis loads. So without 3 going into all the details of how those calculations 4 are done, which would require a code expert, you 5 should at least recognize that the NRC has done that 6 type of review and concluded that these codes are 7 acceptable for the application.
8 ADMIN. JUDGE KENNEDY: Thank you. Dr.
9 Lott, I believe earlier Entergy testified that the 10 Topical Report is an exhibit in this proceeding. Is 11 that true? The bolting methodology?
12 DR. LOTT: The approved -- yes.
13 ADMIN. JUDGE KENNEDY: Dr. Lahey, did you 14 get a chance to review this exhibit as part of your --
15 DR. LAHEY: You're talking about the WCAP 16 report that they --
17 ADMIN. JUDGE KENNEDY: Yes, sir.
18 DR. LAHEY: -- talked about? I haven't 19 seen it yet, I was planning to look at it this evening 20 if I can get my hands on it. I mean, one thing that 21 --
22 ADMIN. JUDGE KENNEDY: Did I miss one of 23 your homework assignments? Is this --
24 DR. LAHEY: Yes, I know, it's just adding 25 up. One thing you folks should know though is, I was NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5306 1 very involved with the development of RELAP and TRAC.
2 So I'm very familiar with that methodology, what it 3 does and what it doesn't do. And let me say one last 4 time, hopefully the last time, after we assured 5 ourselves that these significant shock loads did not 6 distort the core geometry, we then focused on, do you 7 have the right mass in the right place, is the heat 8 transfer right, could you cool the core adequately?
9 That's what these codes are intended for. It's true, 10 they calculate pressure versus time and temperature 11 versus time, but they're in no way sharp shock load 12 codes. They're not intended for that purpose.
13 ADMIN. JUDGE KENNEDY: All right, thank 14 you. Appreciate it. I have nothing further.
15 CHAIRMAN MCDADE: Okay. Very quickly, this 16 may have been covered, but it's sort of a question in 17 my mind and I just want to clarify it. You've done 18 the UT examination on baffle former bolts for the past 19 20 years. Yet, in the testimony, it talks about the 20 UT examination acceptance criteria for the baffle 21 former bolts will be developed. And the question is, 22 what are you using now?
23 MR. DOLANSKY: This is Bob Dolansky with 24 Entergy. The plants that have performed UT of baffle 25 bolts would have that technical justification. Indian NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5307 1 Point has not performed UT of baffle bolts yet. So 2 what we're getting is a plant specific technical 3 justification for the UT of the baffle bolts, that's 4 why we don't have it. So when we say that 8,000 bolts 5 have been inspected, the technical justifications for 6 those, I can't speak specifically for each plant, but 7 I would assume each plant has one. I know that when 8 we go to do our baffle bolts inspection, we will have 9 a plant specific technical justification. And that's 10 what's being developed for us now.
11 CHAIRMAN MCDADE: Okay. One of the 12 concerns that we have is the differentiation between 13 the development and implementation of a plan. That 14 the development of the plan being something to be done 15 that we can look at, the implementation is going to go 16 on and is going to be monitored by the NRC during the 17 period of extended operation if the license is 18 granted. But is there assurance that these acceptance 19 criteria that are to be developed will be adequate to 20 ensure the continued operation?
21 MR. DOLANSKY: I believe the acceptance 22 criteria is contained now in the AMP. This is Bob 23 Dolansky, I'm sorry. So if we go to our --
24 CHAIRMAN MCDADE: I mean, here's the thing, 25 and correct me if I'm wrong, any time you use the term NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5308 1 will be developed, it sort of is a red flag for us.
2 And as I understand, the term that you're using is 3 that you basically have a procedure in existence now, 4 but the plan is that you will be using perhaps more up 5 to date technology down the road and that the 6 acceptance criteria will recognize that increased 7 data, increased knowledge at the time. But it's not 8 that there are no acceptance criteria now, it's that 9 just simply they're going to be updated based on newer 10 technology, additional information. Am I incorrect in 11 that?
12 MR. DOLANSKY: I want -- at the first part 13 of your question, you said that we have a procedure 14 now to perform baffle bolt inspections. Did I 15 understand you correctly, is that what --
16 CHAIRMAN MCDADE: Let me ask the question.
17 In the AMP as it exists right now, are there 18 acceptance criteria for the UT examination of the 19 baffle former bolts?
20 MR. DOLANSKY: No. Right now, the 21 examination acceptance criteria for UT of the baffle 22 former bolts shall be established as part of the exam 23 technical justification.
24 CHAIRMAN MCDADE: Okay. So to Dr. Hiser, 25 how does the NRC determine now that these acceptance NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5309 1 criteria to be developed will be adequate to ensure 2 the proper functioning of the baffle former bolts over 3 the period of extended operation?
4 DR. HISER: This is Allen Hiser of the 5 Staff. Our understanding is that for individual 6 baffle bolt examinations, any indication of cracking 7 from the UT exam indicates that, that bolt is no 8 longer functional. So it represents a failure. I 9 believe the analysis that Mr. Dolansky is discussing 10 is post-inspection, when they have identified each 11 bolt, whether it is acceptable or unacceptable, and 12 they're looking more at the configuration whether it 13 is acceptable. So the acceptance criteria for the 14 final configuration post-inspection is, I believe, 15 what they are still working on. But for each 16 individual bolt, it's crack/no crack, 17 unacceptable/acceptable, is our understanding.
18 MR. DOLANSKY: That's correct. Any 19 indication of cracking will be considered a degraded 20 bolt and will be put into our corrective action 21 program.
22 DR. HISER: So I think -- this is Allen 23 Hiser again. I think there's just been a confusion 24 between acceptance criteria on a piece inspection 25 basis versus the assembly basis after one has NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5310 1 inspected all of the bolts. And I think that the 2 terminology has gotten a little bit mixed up. So the 3 inspection acceptance criteria is any crack is a 4 failed bolt. The assembly acceptability criteria is 5 different. That is to be determined. From the NRC 6 perspective, that's a corrective action that is within 7 the purview of the applicant. If they require 8 approval of any of that, then they would come to the 9 NRC for that.
10 CHAIRMAN MCDADE: Okay. But the way you've 11 just described it, that corrective action may or may 12 not be adequate to ensure the continued viability and 13 the continued operation, the utility of that 14 particular part. So how does the NRC assure itself 15 now, and New York and Riverkeeper assure themselves 16 now, that something that still has to be developed 17 will be adequate? Do you understand what my concern 18 is?
19 DR. HISER: I think so. The corrective 20 actions, there are three potential paths that they 21 could take. If they were to find degraded bolts, they 22 could replace the bolts. Repair really is not an 23 option in this case, but they could replace the bolts.
24 If they wanted to continue to operate with degraded 25 bolts, they would need to submit an analysis to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5311 1 NRC that justified their configuration that they want 2 to continue operating with. From that perspective, we 3 still have approval of any operation with the degraded 4 condition. So I think from that perspective, I 5 believe that there are sufficient controls on this.
6 MR. STROSNIDER: This is Jack Strosnider 7 for Entergy. The one other thing I think we should 8 add here is what we talked about earlier. The 9 methodology that's being used to develop the plant 10 specific bolting pattern, that acceptable bolting 11 pattern, that methodology was reviewed in the WCAP 12 report that was submitted to NRC, it's an extensive 13 review with all the Requests for Additional 14 Information, the back and forth, the safety evaluation 15 written. So they're using what the NRC has looked at 16 and concluded that, that's an acceptable methodology.
17 You do it on a plant specific basis to make sure that 18 if anything plant specific comes up that you've 19 accounted for it. But you should be able to move 20 forward with that methodology and come up with an 21 acceptable technical evaluation for the acceptable 22 bolting pattern.
23 MR. COX: This is Alan Cox --
24 MR. STROSNIDER: That's the whole purpose 25 of, as I explained earlier, the Topical Report process NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5312 1 and reviewing these methodologies up front. The NRC 2 does not want to look at one of these -- do that same 3 review over and over and over again for every plant.
4 So they look at the generic methodology, if the plant 5 can apply it within the boundary conditions that, that 6 methodology is good for, then they can go do the plant 7 specific evaluation and you should have confidence in 8 it.
9 MR. COX: This is Alan Cox with Entergy.
10 Let me add one more thing. I believe the Standard 11 Review Plan provides for, when it talks about 12 acceptance criteria, it says the acceptance criteria 13 doesn't necessarily have to be a formal value. It can 14 be a description of the method that will be used to 15 establish that value. And I would say that it's not 16 practical to do all that in advance. You're talking 17 about finding a -- you're going to find a certain 18 pattern of bolts where you have a failed bolt, there's 19 832 bolts, you can't predict or it would be 20 impractical to try to do an analysis for every 21 combination of which bolt might be failed where.
22 So you have a method established, you take 23 the results of the inspection, you look at the bolts 24 that are failed, and you apply that data using the 25 method that's established and that's how you determine NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5313 1 your acceptance criteria. It would be impractical to 2 evaluate all of those combinations ahead of time that 3 you could have with the inspection results.
4 MR. DOLANSKY: This is Bob Dolansky. I 5 just want to add one other thing. The inspection of 6 baffle bolts, the fact that it's been going on for 7 years and the fact that other plants have done it 8 means that all this stuff has been looked at before.
9 So, I mean, it's not like a first of its kind. We're 10 not the first ones doing it, other plants within the 11 industry have performed these inspections, they've had 12 NRC reviews of both during the inspection and after 13 the outage. So not only has the Topical Report been 14 reviewed by the Staff, but the actual implementation 15 has been reviewed by the Staff at other plants.
16 CHAIRMAN MCDADE: And the preface to my 17 question assumed that and recognized that these 18 inspections have been going on for more than 20 years 19 and that there has been a method established. And 20 what I'm trying to just make sure that I want to 21 cement in my mind is what is currently established and 22 what, if anything, by way of methodology still needs 23 to be developed? As Mr. Cox pointed out, I think it's 24 relatively obvious that you can't say what you're 25 going to do as the result of an inspection until you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5314 1 get the results of that inspection and then you have 2 to assess the meaning of that.
3 But in developing a plan that provides 4 assurances that you're going to have continued safe 5 operation, you look, is the methodology used adequate 6 to develop an appropriate response? And all I'm 7 trying to do is cement for the record of, what is the 8 current status of development? Again, the language 9 you used, will be developed, which sort of, well, more 10 than sort of, which suggests that there is still a 11 sort of, we'll figure this out in the future, we hope.
12 And from what Dr. Hiser has said, it isn't that and I 13 just wanted to get as clear as possible what we 14 currently have and what it is that still needs to be 15 done so we can determine whether or not there exists 16 now reasonable assurance of continued safe operation.
17 Can you elaborate on that?
18 MR. AZEVEDO: Yes, your honor. This is 19 Nelson Azevedo for Entergy. Maybe it's already clear, 20 but just to make sure it is clear. The methodology is 21 established, has been approved by the NRC, that's what 22 we're using. What is going on right now is taking 23 that methodology in developing a model specifically 24 for Indian Point. But the methodology itself, how it 25 is done, is established, it's been approved by the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5315 1 NRC, and it's what Indian Point is using.
2 CHAIRMAN MCDADE: And that's part of the 3 AMP for these baffle former bolts? Where do we look 4 to see where that methodology is and the fact that 5 it's been approved by the NRC?
6 MR. AZEVEDO: It's referenced in MRP 227, 7 that's what Mr. Cox was telling you.
8 MR. COX: This is Alan Cox. Chapter 6 of 9 MRP 227 has a discussion on evaluation of bolts and 10 pins. And basically it says that you don't have to do 11 the individual evaluation, you have to look at the 12 effect on the assembly. And then Section 6.4 gives 13 you the guidance on doing assembly level evaluations.
14 CHAIRMAN MCDADE: Okay. And, Dr. Hiser, if 15 that were to change, is that something that the 16 Applicant would be able to do and inform the NRC? Is 17 it something that would be subject to the 50.59 18 procedure? Or is it something that would require a 19 license amendment?
20 DR. HISER: This is Allen Hiser of the 21 Staff. My guess is it would be subject to 50.59.
22 Given the safety implications and -- my expectation 23 would be that it would not pass 50.59 and would 24 require a submittal to the NRC.
25 CHAIRMAN MCDADE: Okay. And once the plant NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5316 1 specific details are worked out, what is the vehicle 2 for NRC review of that plant specific? Do they submit 3 it to you, is it subject to any kind of a formal 4 review and approval? Or is it simply a situation 5 where you would have the ability to review and 6 comment?
7 DR. HISER: This is Allen Hiser of NRC. My 8 expectation is that, that would require at a minimum 9 a submittal to the NRC. And could involve a license 10 amendment. But I'm not certain of that.
11 CHAIRMAN MCDADE: Okay. Two separate 12 things. One, we were talking about the methodology 13 set out in Chapter 6 of MRP 227. And we were talking 14 about what would be necessary if that methodology were 15 changed. My question right now is not the methodology 16 having been changed, but rather, as Entergy explained, 17 they are taking that methodology and are currently 18 working to make a plant specific program. And my 19 question is, once they have completed the details of 20 that plant specific, how is that subject to review by 21 the NRC?
22 DR. HISER: This is Allen Hiser. My prior 23 statement really still holds. Whether it's an 24 approved generic methodology or a plant specific 25 methodology would require the same approval, if NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5317 1 approval is necessary. The only difference in that 2 case is for the plant specific methodology, we would 3 have to review the entire methodology instead of 4 relying on the generic approval from the generic 5 methodology that previously was approved.
6 CHAIRMAN MCDADE: And if you viewed it to 7 be unacceptable, would this be going at that point 8 through the 50.59 procedure or is this something less 9 than the 50.59 procedure?
10 DR. HISER: I'm sorry, if we found what to 11 be unacceptable?
12 CHAIRMAN MCDADE: Their plant specifics.
13 If you identified problems with the plant specifics, 14 how does that work through?
15 DR. HISER: If it is a license amendment 16 request, then they would need to modify their approach 17 to become acceptable. In the absence of approval of 18 the license amendment, they could replace the bolts 19 that they found to be degraded.
20 CHAIRMAN MCDADE: Okay. But I -- at least 21 I had not anticipated at the level of a -- if it's a 22 license amendment, then it's a situation where New 23 York gets a notice of an opportunity for a hearing and 24 has the opportunity to challenge the adequacy. What 25 I'm trying to get at is once they take the methodology NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5318 1 that is in existence, that the NRC has reviewed, has 2 approved, is satisfied with, and apply it specifically 3 to the plant, what does the NRC do then and what is 4 the nature of review and the public aspect of that, if 5 any?
6 DR. HISER: This is Allen Hiser again. If 7 we can just discuss for a moment here.
8 CHAIRMAN MCDADE: Well, we're basically at 9 the end of the day. Do you want to answer that at the 10 beginning --
11 DR. HISER: Sure.
12 CHAIRMAN MCDADE: -- of tomorrow rather 13 than -- I mean, to me this is something that's 14 important and I don't want an answer off the top of 15 your head that I think this is the way it should be.
16 So, why don't we leave that and take that up first 17 thing in the morning?
18 DR. HISER: Okay. That's acceptable.
19 CHAIRMAN MCDADE: Okay. Do you have 20 anything further for --
21 ADMIN. JUDGE WARDWELL: I do not.
22 CHAIRMAN MCDADE: Okay. I would propose 23 that we would start tomorrow at 8:30. Before we 24 break, NRC is there any administrative matters or 25 other matters that you want to take up before we break NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5319 1 for this evening?
2 MR. HARRIS: No, your honor.
3 CHAIRMAN MCDADE: Entergy?
4 MR. KUYLER: Yes, your honor. A few 5 minutes ago, we were discussing the Westinghouse 6 proprietary reports regarding the baffle former bolts 7 and the minimum bolting pattern analysis. And I 8 believe one of the things that Dr. Lahey said was that 9 he had not reviewed those reports. And we just wanted 10 to note that those have been in the record for several 11 months and they've been disclosed several years ago.
12 And, so, for us to look at that for the first time 13 afresh tomorrow would be challenging, your honor.
14 CHAIRMAN MCDADE: Well, I mean, these are 15 exhibits, are they not?
16 MR. KUYLER: Yes, they are.
17 CHAIRMAN MCDADE: So, I mean, if there's 18 something that's already admitted as an exhibit that 19 will help clarify Dr. Lahey's testimony, we will allow 20 him to refer to that and will of course give Entergy 21 the opportunity, if their experts need some additional 22 time to respond. Quite frankly, so far I've been 23 amazed at the capacity of all of our witnesses to know 24 a rather voluminous record and to be able to testify 25 with regard to the contents of the literally tens of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5320 1 thousands of pages of documents that we have. So I 2 think you may have underestimated your experts, but in 3 any event, we would certainly give them time and the 4 opportunity to refresh themselves with the documents, 5 with the reports in order to comment on Dr. Lahey's 6 testimony.
7 MR. KUYLER: Yes, your honor. But I would 8 just note that we have not seen any specific 9 criticisms of those documents from Dr. Lahey in the 10 past. So for him to introduce those on the very last 11 day of the hearing for the first time, we would object 12 to that, your honor.
13 CHAIRMAN MCDADE: We don't know whether 14 he's going to criticize them. He may look at them and 15 say, this is great. This allays my fears, if I had 16 read this before, if I had studied it harder, I 17 wouldn't be here. We don't know, we'll find out 18 tomorrow. Mr. Sipos, anything before we break for 19 tomorrow?
20 MR. SIPOS: Nothing further from New York 21 at this time.
22 CHAIRMAN MCDADE: Ms. Brancato?
23 MS. BRANCATO: No, your honor.
24 CHAIRMAN MCDADE: Okay. Does 8:30 tomorrow 25 pose a problem for anybody?
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5321 1 MS. SUTTON: No, your honor.
2 CHAIRMAN MCDADE: Okay, apparently not.
3 We'll see you all at 8:30 in the morning. Thank you.
4 (Whereupon, the above-entitled matter went 5 off the record at 5:36 p.m.)
6 7
8 9
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Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
Entergy Nuclear Operations, Inc.
Indian Point Nuclear Generating Station Docket Number: 50-247-LR and 50-286-LR ASLBP Number: 07-858-03-LR-BD01 Location: Tarrytown, New York Date: Tuesday, November 17, 2015 Work Order No.: NRC-2016 Pages 5002-5321 NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.
Washington, D.C. 20005 (202) 234-4433
5002 1 UNITED STATES OF AMERICA 2 U.S. NUCLEAR REGULATORY COMMISSION 3 + + + + +
4 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 5 + + + + +
6 ________________________________
7 In the Matter of: : Docket No.
8 ENTERGY NUCLEAR OPERATIONS, INC. : 50-247-LR 9 (Indian Point Nuclear Generating : 50-286-LR 10 Station, Units 2 and 3) : ASLBP No.
11 ________________________________ : 07-858-03-LR-BD01 12 Tuesday, November 17, 2015 13 14 Doubletree Tarrytown 15 Westchester Ballroom 16 455 South Broadway 17 Tarrytown, New York 18 19 20 21 BEFORE:
22 LAWRENCE G. MCDADE, Chairman 23 MICHAEL F. KENNEDY, Administrative Judge 24 RICHARD E. WARDWELL, Administrative Judge 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5003 1 APPEARANCES:
2 On Behalf of the U.S. Nuclear Regulatory 3 Commission:
4 DAVID E. ROTH, ESQ.
5 SHERWIN E. TURK, ESQ.
6 BRIAN HARRIS, ESQ.
7 U.S. Nuclear Regulatory Commission 8 Office of General Counsel 9 Mail Stop 15 D21 10 Washington, D.C. 20555 11 david.roth@nrc.gov 12 sherwin.turk@nrc.gov 13 brian.harris@nrc.gov 14 301-415-2749 (Roth) 15 301-415-1533 (Turk) 16 301-415-1392 (Harris) 17 18 On Behalf of Entergy Nuclear Operations, Inc.:
19 KATHRYN M. SUTTON, ESQ.
20 PAUL M. BESSETTE, ESQ.
21 RAPHAEL "RAY" KUYLER, ESQ.
22 Morgan, Lewis & Bockius, LLP 23 1111 Pennsylvania Avenue, NW 24 Washington, DC 20004 25 202-739-5738 (Sutton)
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5004 1 202-739-5796 (Bessette) 2 202-739-5146 (Kuyler) 3 ksutton@morganlewis.com 4 pbessette@morganlewis.com 5 rkuyler@morganlewis.com 6
7 On Behalf of the State of New York:
8 JOHN J. SIPOS, ESQ.
9 LISA S. KWONG, ESQ.
10 MIHIR A. DESAI, ESQ.
11 of: New York State 12 Office of the Attorney General 13 Environmental Protection Bureau 14 The Capitol 15 Albany, NY 12224 16 brian.lusignan@ag.ny.gov 17 18 On Behalf of Riverkeeper Inc.:
19 DEBORAH BRANCATO, ESQ.
20 Riverkeeper, Inc.
21 20 Secor Road 22 Ossining, New York 10562 23 800-21-RIVER 24 info@riverkeeper.org 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5005 1 On Behalf of Westchester County:
2 CHRISTOPHER INZERO, ESQ.
3 Assistant County Attorney 4 Westchester County Government 5 148 Martine Avenue 6 Room 600 7 White Plains, New York 10601 8 914-995-2000 9
10 On Behalf of Westinghouse Electric Company:
11 RICHARD J. COLDREN, ESQ.
12 Westinghouse Electric Company 13 1000 Westinghouse Drive 14 Cranberry Township, Pennsylvania 16066 15 412-374-6645 16 17 18 19 20 21 22 23 24 25 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5006 1 P R O C E E D I N G S 2 8:31 a.m.
3 CHAIRMAN MCDADE: This hearing will come to 4 order. We will continue with discussion on Contention 5 Number 25. Before we get started however, there was 6 one administrative matter that I forgot from yesterday 7 that I'm embarrassed that I forgot. Looking at the 8 Statement of Position of Entergy, this is Exhibit 615, 9 Page 33, third paragraph down. First word of the 10 paragraph is stricken from the record.
11 MR. KUYLER: Your honor, could you say 12 again the page?
13 CHAIRMAN MCDADE: Page 33, third paragraph.
14 MR. KUYLER: Yes, your honor.
15 CHAIRMAN MCDADE: We are striking from the 16 record as grossly inappropriate the first word of that 17 paragraph. Are you ready to proceed?
18 ADMIN. JUDGE WARDWELL: What about the 19 homework assignments? Should we start with Dr. Lahey 20 first and then we'll assume Entergy has something in 21 regards to the data? If not, we can give you more 22 time if you need it. But let's start with Dr. Lahey.
23 DR. LAHEY: So, he wanted these.
24 ADMIN. JUDGE KENNEDY: Well, we don't what 25 these are, Dr. Lahey.
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5007 1 CHAIRMAN MCDADE: Yes.
2 DR. LAHEY: Oh, they're -- you asked me to 3 --
4 CHAIRMAN MCDADE: Dr. Lahey, excuse me. If 5 you're going to talk, you've got to be sitting with 6 the microphone. Otherwise, it's going to get lost for 7 the record.
8 DR. LAHEY: Okay.
9 CHAIRMAN MCDADE: And we can get someone --
10 DR. LAHEY: Richard Lahey. Your honor, you 11 asked me to give you the references that I had cited 12 yesterday. So I have copies of them for you.
13 CHAIRMAN MCDADE: Well, could you just tell 14 us what the cites are?
15 DR. LAHEY: What the references are?
16 CHAIRMAN MCDADE: Yes.
17 DR. LAHEY: Okay. There's three of them.
18 One of them is Kanaski, the other one is Arai, and the 19 other one is Korth, et al. And these had to do with 20 low cycle fatigue versus high cycle fatigue and the 21 effect of this on failure for testing components that 22 were irradiated.
23 ADMIN. JUDGE WARDWELL: And were these 24 previous exhibits that you had submitted?
25 DR. LAHEY: Yes. They were in my NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5008 1 testimony, but Judge McDade asked me if I would -- I 2 thought it was my homework assignment if I would make 3 him a copy of them, so I did.
4 MR. SIPOS: And Judge, this is John Sipos, 5 we have the exhibit numbers if you would like.
6 ADMIN. JUDGE WARDWELL: And that's all we 7 needed.
8 MR. SIPOS: Would that be helpful?
9 CHAIRMAN MCDADE: Yes.
10 ADMIN. JUDGE WARDWELL: That's what I think 11 we were trying to imply is we just needed the cites 12 for those, we didn't need the cites.
13 MR. SIPOS: So the first one, New York 14 State 564 is Arai, A-R-A-I. Another one is 15 Riverkeeper 152 and that's Korth, K-O-R-T-H. And the 16 third is NRC 177 and the first author is Kanasaki, K-17 A-N-A-S-A-K-I.
18 CHAIRMAN MCDADE: Okay. Thank you, Mr.
19 Sipos.
20 ADMIN. JUDGE WARDWELL: And thank you, Dr.
21 Lahey, for digging out those specific ones. We only 22 really needed the cites though. Sorry for the 23 misunderstanding, we appreciate your effort.
24 DR. LAHEY: Okay.
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5009 1 anything you'd like to offer in regards to the data 2 that demonstrates the effect of, I believe it was 3 looking at the effects of embrittlement on fatigue in 4 some fashion or fracture toughness.
5 DR. LOTT: Yes. My understanding was there 6 were two questions that came up yesterday related to 7 data. That being one of them, the irradiation effects 8 on fatigue. And there was another question about high 9 fluence properties in general and the survivability of 10 materials at high fluences. As to the question of the 11 effects of radiation on fatigue, I must admit, I've 12 been reliant on NRC NUREG/CR-6909, which in Section 13 1.32 has a discussion of irradiation embrittlement and 14 fatigue.
15 Let me say that I have a difficult time 16 with the word synergism when we talk about these 17 particular relationships. I understand that all 18 irradiated materials are embrittled and the question 19 to me is whether or not irradiation also has effect on 20 the strain life of the material or on the fatigue 21 resistance of the material. I don't see that as a 22 synergism, that's just a question of irradiation 23 embrittlement and irradiation fatigue life.
24 I can give you a summary of what I think 25 it says in NRC 6909. It does reference the papers by NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5010 1 Korth and Harper. I'll point out that, that data in 2 that paper is basically generated for fast meter 3 conditions, it's not at a temperature that we would 4 generally think would be applicable for PWR 5 interactions. The paper by Arai was actually 6 reporting on irradiation embrittlement work that was 7 done at Westinghouse. I'm not sure that there's any 8 fatigue data in that paper at all, I don't recall any.
9 And the paper by Kanasaki, I'll point out 10 I'm a co-author on that paper, so we're certainly 11 aware of it. It is the one paper that, and we can put 12 the data -- do you want to discuss the data or just 13 note that the data is there? That is the paper that 14 does show that at low strain amplitudes, below I 15 believe it's 0.6, the data that was tested of PWR 16 relevant conditions all showed an increase in fatigue 17 life with irradiation. We're not to discussing the 18 CUFs in Contention 26 yet, but when we do, I think 19 we'll find that there's, A, a limited number of 20 materials of the internals which have CUF values 21 calculated are irradiated. And those materials, I do 22 not expect to have a large strain amplitude. So I 23 think the data in the Kanasaki paper is directly 24 relevant to the conditions we're talking about.
25 ADMIN. JUDGE WARDWELL: Okay. When you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5011 1 first started off referencing a NUREG was it?
2 DR. LOTT: It's CR-6909 and I'm sorry, I'll 3 find you the --
4 ADMIN. JUDGE WARDWELL: And is there an 5 exhibit number. If you just --
6 DR. LOTT: Yes, there is.
7 ADMIN. JUDGE WARDWELL: -- have an exhibit 8 number, that would suffice because then we can find it 9 through the --
10 MR. STEVENS: Your honor, Jerry Stevens of 11 the Staff. That's New York State 490 Alpha.
12 ADMIN. JUDGE WARDWELL: Okay.
13 DR. LOTT: Yes.
14 ADMIN. JUDGE WARDWELL: Thank you.
15 DR. LOTT: Of which Mr. Stevens is an 16 author.
17 ADMIN. JUDGE WARDWELL: Okay. Is there 18 anything else you'd like to offer in regards to 19 sources of data backing your claims from yesterday?
20 DR. LOTT: Well, there is, in addition, the 21 issue of high fluence properties. There was some 22 question in our mind whether the question was high 23 fluence data on fracture toughness or high fluence 24 data on the yield stress of materials. My observation 25 again would be that, when we come to the very highly NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5012 1 irradiated components in the plant, particularly the 2 baffle former bolts, and you look at strategies for 3 dealing with the baffle former bolts, anything that we 4 observe with a crack, we assume has failed. So we 5 never actually use a fracture toughness value for very 6 highly irradiated materials because we have this 7 failure assumption in our acceptance criteria.
8 So if you want to know about those 9 materials at very high fluences, I would recommend to 10 you and I'm going to find it here, it's Exhibit 11 Entergy 000646, MRP 210, has in there in Figure 1-4 a 12 plot of yield strength versus neutron fluence that 13 goes out to fluences data certainly greater than 90 14 dpa, actually one data point closer to 120 some dpa.
15 Which would be well beyond what we would expect the 16 fluence on the vast majority of the reactor internals 17 ever to see.
18 ADMIN. JUDGE WARDWELL: Thank you. And 19 again, is there a -- did you reference an exhibit 20 number in your --
21 DR. LOTT: Yes, I believe I did. ENT 646.
22 ADMIN. JUDGE WARDWELL: Oh, ENT, okay.
23 Thank you.
24 MS. BRANCATO: Your honors, this is Debra 25 Brancato from Riverkeeper. Dr. Hopenfeld has NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5013 1 indicated to me he has some helpful input to this 2 discussion as well if you care to entertain some 3 testimony from him at this time.
4 CHAIRMAN MCDADE: Well, probably not at 5 this time, but before we leave the topic.
6 MS. BRANCATO: Okay.
7 ADMIN. JUDGE WARDWELL: Okay. So let's 8 continue on with some questioning and dealing with 9 Contention 25 --
10 CHAIRMAN MCDADE: But before you go on, Dr.
11 Hopenfeld is, as I said yesterday, sometimes things 12 move and we hear an awful lot of things. Take a card, 13 write down the comment to remind yourself, and you 14 will be testifying later. Thank you.
15 ADMIN. JUDGE WARDWELL: And I'll refer 16 again to Entergy's Exhibit 616, their testimony, 17 Answer 174, Pages 113 to 114. During its technical 18 review of MRP 227, the NRC Staff specifically 19 requested additional information on how the program 20 accounts for synergistic effects. And I guess I'll 21 ask again for Entergy, how does your AMP look at and 22 include actual analyses which have addressed the 23 change in fatigue strength as a function of varying 24 degrees of embrittlement of the specimen that occurs 25 with time?
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5014 1 DR. LOTT: Well, again, to me there are two 2 questions in mind there. One is when we talk about 3 embrittlement, if we're talking about the loss of 4 fracture toughness, we calculate or look at lower 5 bound fracture toughness curves based on the dose of 6 the material. So we're using what we believe to be 7 bounding fracture toughness curves and then analyze 8 the crack that has grown. So there's certainly an 9 interaction there between the tolerance of the 10 material and the embrittlement and the fatigue crack 11 growth, they're tied into the same calculation.
12 ADMIN. JUDGE WARDWELL: But that's a 13 proposal of what you would do if a crack was observed, 14 right? And until a crack is observed and as part of 15 the development of your AMP, you haven't done any 16 analysis in regards to that, is that correct?
17 DR. LOTT: Well, we've certainly done 18 analysis to show that we do not expect to see cracks 19 occur in these materials whatsoever. Again, the CUF 20 would be a good example of that. We've looked at 21 extensive analysis of IASCC, particularly in the 22 highly rated components we generated a fairly complex 23 model of the reactor internals to look at the aging of 24 those reactor internals and predict where IASCC might 25 occur. The only place that we actually predicted it NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5015 1 would occur was in the baffle former bolts, which we 2 discussed. There's a fairly detailed process for 3 that.
4 In our processing and recommendations for 5 inspections, we certainly looked at susceptibility to 6 multiple degradation mechanisms as part of that and we 7 based our inspection program on determination of the 8 effect of those mechanisms. So as far as we're 9 concerned, we're obviously inspecting for irradiation 10 fatigue, IASCC, or SCC, all three of which are 11 cracking mechanisms and the inspection program doesn't 12 care which one of those caused that, it's just simply 13 looking for that. So I think in our prioritization of 14 inspections, we certainly have looked at that. We've 15 looked at that in the design of our program in 16 general.
17 ADMIN. JUDGE WARDWELL: I'm just trying to 18 think of whether I'll wait until 26 to discuss this 19 further or not, but I'll bring it up again now, I 20 guess, because we did talk about it yesterday and I 21 want to fix again, I guess with you, Dr. Lott. I'm 22 trying to get a grasp on the handle between the peak 23 strength and fatigue durability. As I heard our 24 discussion yesterday, I can understand why someone may 25 say embrittlement won't come into effect really until NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5016 1 there's a crack due to excessive peak strength.
2 You've got to get past the peak strength and then 3 embrittlement comes into play to a certain degree.
4 But fatigue durability doesn't rely on loads that 5 exceed the peak strength. It's a repetitive loading 6 that causes the types of cracks, is that correct?
7 DR. LOTT: Yes.
8 ADMIN. JUDGE WARDWELL: And so, again, I'm 9 back to the question of where is anyone looked at 10 whether or not that fatigue durability is influenced 11 by embrittlement and to what degree is it?
12 DR. LOTT: Well, again, this is, at least 13 from my perspective, I don't necessarily -- to me, 14 embrittlement and fatigue life, well, again, 15 embrittlement is not a property in and of itself. The 16 properties are yield stress, ultimate stress, 17 ductility measurements, such as total elongation, 18 fracture toughness. And I would see, again, as 19 another issue, would be the question of what's the 20 impact on the S-N curve, the number of --
21 ADMIN. JUDGE WARDWELL: Well, can't we use 22 fracture toughness as an indication of embrittlement?
23 DR. LOTT: It's an indication that the 24 material has been irradiated and irradiated materials 25 have a decrease in that. But it does not necessarily NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5017 1 indicate how the material will survive under fatigue 2 conditions. So basically, we have a test for that, 3 just like we have a test to determine yield stress.
4 That is the fatigue life test to generation of the S-N 5 curve. We actually take a specimen, strain it at some 6 strain amplitude, we put it on repeated cycles, and 7 count the number of the cycles it takes to fail the 8 specimen or to have a load drop in the specimen 9 actually. So those fatigue life curves, again, are 10 behind basically the calculation of the CUF factors 11 based on a design curve, which is bounding to all of 12 the measurements of fatigue life.
13 ADMIN. JUDGE WARDWELL: Let me ask you 14 this, if I did a fatigue test, whatever the fatigue 15 test might be, if I try to fatigue it, as a specimen 16 that's not irradiated whatsoever, and I repeated that 17 test under different degrees of radiation exposure, 18 what would you expect the results of the fatigue test 19 to do? Remain the same, improve, or degrade?
20 DR. LOTT: That's exactly what we did in 21 the Kanasaki paper that, again, I'm a co-author on.
22 ADMIN. JUDGE WARDWELL: Okay.
23 DR. LOTT: We looked at that. We looked at 24 that at some fairly high fluences. So, if you'll --
25 the fluences are in two different papers reported in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5018 1 different units. If you'll take my word for it, in 2 the Kanasaki paper the data is about 20 dpa. At 20 3 dpa, if you look at the yield stress or the fracture 4 toughness curves, you'll see that there's a sharp 5 decrease in the fracture toughness of the material, 6 there's a large increase in the yield stress and, yet, 7 the fatigue life in those specimens that we tested for 8 Kanasaki, the fatigue life of that material got 9 longer. So the fatigue life improved at the same time 10 that the fluences were such that the yield stress 11 would increase and the fracture toughness decreased.
12 ADMIN. JUDGE WARDWELL: Thank you. Dr.
13 Lahey, do you have any other types of cites or 14 evidence in your testimony that differs from what they 15 have just expounded upon and/or what would you believe 16 would be the change in that fatigue property as the 17 materials are irradiated?
18 DR. LAHEY: Right. This is Richard Lahey 19 for New York State. I don't disagree with anything he 20 said except he didn't say everything. If you look at 21 the Korth paper, which admittedly was for high 22 pressure or higher temperature conditions for the 23 reactor, what it says is for high cycle fatigue where 24 you have larger amplitudes, you can have a reduction 25 by a factor of one half of the cycles to failure for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5019 1 irradiated materials, irradiated up to 1.1 times ten 2 to the 22 neutrons per centimeter square.
3 It is inconclusive. The reason people say 4 it's inconclusive is there's other data that shows if 5 you have high cycle fatigue or have low amplitude, 6 then it strengthens and you're in the region where 7 things get better. Here, things get significantly 8 worse. So all the real experts, the people who do 9 research on this, say the same thing. We don't really 10 have any good data for light water reactor conditions 11 and it's sorely needed. And this is why in the 12 sustainability program they're doing those kind of 13 tests. So I don't disagree with that at all.
14 What I am concerned with is there's 15 evidence, admittedly it's not perfect, but there's 16 evidence it can have a significant degrading effect.
17 And then the question is, what do you do in the 18 meantime until you can definitively tell and quantify 19 the effect? Do you just ignore it and keep looking 20 until you get a crack? Or do you put on some sort of 21 factor to account -- put a cushion in to account --
22 ADMIN. JUDGE WARDWELL: In your cite that 23 you just gave us, would that temperature have a big 24 difference and is that temperature representative of 25 what might be experienced in a PWR at IPEC?
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5020 1 DR. LAHEY: It's somewhat higher than the 2 temperature that you would have in a light water 3 reactor for sure. Therefore, it's not a perfect data 4 set.
5 ADMIN. JUDGE WARDWELL: And would that 6 likely have an influence on the data results, the 7 extra temperature?
8 DR. LAHEY: It's hard to say.
9 ADMIN. JUDGE KENNEDY: Dr. Lahey, this is 10 Judge Kennedy. I'm curious now that I've heard from 11 Dr. Lott, we've got two conflicting views of the data, 12 one for slightly different conditions. It appears 13 that Dr. Lott's paper addressed the right conditions 14 and addressed different levels of irradiation. How 15 would you challenge his paper? You've offered up the 16 higher temperature data, but what would you say to the 17 data that he's presented? I mean, I recognize it's 18 one of the three references you provided.
19 DR. LAHEY: Right. Well, my understanding 20 -- I mean, he's the author, so --
21 ADMIN. JUDGE KENNEDY: But I think you're 22 --
23 DR. LAHEY: -- he has a little advantage 24 there, but --
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5021 1 offering his same paper for a different conclusion.
2 DR. LAHEY: Yes. I mean, it is one of the 3 few papers that addressed fatigue and irradiation.
4 And as I understood the focus of his paper was more 5 into what happens once you get a crack and how does it 6 propagate and the initiation of the crack, rather than 7 the cycles to failure.
8 ADMIN. JUDGE KENNEDY: But isn't that what 9 he just said? That the number of cycles to failure 10 increases with the irradiation of the sample material?
11 DR. LAHEY: Well, that's what --
12 ADMIN. JUDGE KENNEDY: That the fatigue 13 life went up?
14 DR. LAHEY: It depends entirely on the 15 amplitude. It depends on the amplitude of the fatigue 16 cycle.
17 ADMIN. JUDGE KENNEDY: Are you suggesting 18 that the amplitude in his test data or in his 19 calculations using the test data are not the right 20 characterization of the amplitudes that would be 21 present at Indian Point Units 2 or 3?
22 DR. LAHEY: I mean, I don't -- I mean, he 23 should say what the purpose of his test was. But as 24 I understand it, it wasn't to specifically address 25 that.
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5022 1 ADMIN. JUDGE KENNEDY: And maybe this is a 2 more general question about the term high cycle/low 3 amplitude and low cycle/high amplitude has been pushed 4 around here. I'm not sure sitting up here, are both 5 applicable to the operating conditions at Indian 6 Point? Is one grouping, I don't know if it was -- is 7 high cycle/low amplitude more applicable to Indian 8 Point? Or is low cycle/high amplitude more 9 applicable? Or are they both? And is there a hole in 10 the data as Dr. Lahey is suggesting?
11 DR. LAHEY: Well, as I under -- this is 12 Richard Lahey again. As I understand high cycle, it's 13 things like flow induced vibration, turbulence 14 induced, flow induced vibration. This is not a great 15 concern. It is in the steam generators and what that 16 might do to fretting and things like that, but it's 17 not in the primary side. It's more low cycle and many 18 of the kind of transients they have would in my 19 opinion give a larger amplitude. So it's more like 20 low cycle, larger amplitude fatigue.
21 CHAIRMAN MCDADE: Okay. Dr. Lahey, let me 22 interrupt for a second here. The document that you 23 referenced was the Korth paper, K-O-R-T-H.
24 DR. LAHEY: Right.
25 CHAIRMAN MCDADE: And that was Riverkeeper NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5023 1 152. That paper was originally presented back in June 2 of 1974. Is that data still valid? I mean, isn't 3 there something more recent that you can address us to 4 as far -- you're talking about the absence of data 5 here and that particular study is more than 40 years 6 old.
7 DR. LAHEY: Well, that's correct. But 8 nobody that we could find in a literature search has 9 really systematically done that and as a consequence 10 that's why they took it on in the Light Water Reactor 11 Sustainability Program. I agree there's data needed.
12 CHAIRMAN MCDADE: Okay. So that particular 13 study, although it's more than 40 years old, it's your 14 position that since then there has been no significant 15 work that has generated more informative data?
16 DR. LAHEY: I haven't been able to find it 17 if there has been. And I don't know anybody else that 18 has. All the -- this has been discussed by the NRC 19 because they had input, like I'm giving you, from some 20 of their experts at Argonne and they took all that 21 input in, looked at this data, the other data, and 22 then decided it's inconclusive and so we'll do our 23 inspection program, but not do anything separate in 24 terms of putting a penalty factor in. But that's 25 their job, in my view, as regulators. That's their NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5024 1 opinion.
2 ADMIN. JUDGE KENNEDY: I guess -- this is 3 Judge Kennedy. I'm still curious about low cycle/high 4 amplitude and high cycle/low amplitude. Is this worth 5 discussing in more detail? Is it relevant to the 6 metal fatigue for reactor vessel internals?
7 DR. LOTT: Well, I think probably what's 8 relevant to this particular discussion is the limits 9 that are in the Kanasaki paper in general, which is 10 this 0.6 percent strain amplitude. We believe those 11 conditions were chosen such that they would be 12 relevant to reactor internals, that was the intent of 13 the testing in the first place. And, again, as I 14 indicated, did a quick check of the CUF values that 15 are reported for the reactor internals and those that 16 are in components that also see irradiation, which is 17 only a fraction of the total. And having surveyed 18 those, I believe that we'll find that all of the 19 strain amplitudes are within the limits that are in 20 the Kanasaki paper. If you look at the number of 21 cycles to failure in Figure 8 of that paper, you'll 22 see that in some cases they are as few as 1,000 23 cycles. So it's not like this is thousands of cycles 24 per year kind of numbers that we're talking about.
25 ADMIN. JUDGE KENNEDY: All right. Thank NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5025 1 you.
2 ADMIN. JUDGE WARDWELL: Moving on from 3 that, Entergy's testimony, Exhibit 616, Answer 144, 4 Pages 93 to 94, present eight age relating degradation 5 mechanisms. And they include stress corrosion 6 cracking, irradiation assisted stress corrosion 7 cracking, wear, fatigue, thermal aging embrittlement, 8 irradiation embrittlement, void swelling and 9 irradiation growth, thermal and irradiation enhanced 10 stress relaxation, or irradiation enhanced creep.
11 They then go on to say that for each of 12 these eight mechanism, MRP 227 identifies the 13 resulting aging effect, which will then be managed 14 through inspections under MRP 227-A guidelines.
15 Notably, in most cases, the key effects are cracking, 16 dimensional changes, or wear, but in all cases, as 17 explained below, the inspections specified in MRP 227-18 A are designed to detect potential aging effects 19 applicable to each reactor vessel internal component 20 regardless of the underlying mechanism.
21 And I guess I'd address this to Entergy, 22 whoever would like to answer, if you say that in most 23 cases the key effects are cracking, dimensional 24 changes, or wear, but given that the effects are 25 managed, not the mechanism, what are the key effects NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5026 1 in the other minority cases and are they detectable by 2 your AMP? I'm addressing your throw away statement in 3 this testimony that in most cases the key effects are 4 cracking, dimensional changes, et cetera. What I'm 5 asking is, what about in those other cases that aren't 6 included in that, what are those key effects because 7 it is true you are claiming to be monitoring for 8 effects and not mechanisms?
9 DR. LOTT: I may have to take a minute to 10 think about this one. I suspect that, that is, as you 11 suggested, a throw away sentence that we were probably 12 just overly cautious. I don't --
13 MR. DOLANSKY: Dr. Lott?
14 DR. LOTT: Yes.
15 MR. DOLANSKY: This is Bob Dolansky with 16 Entergy. Perhaps one example would be the internals 17 hold down spring.
18 DR. LOTT: Okay.
19 MR. DOLANSKY: We're actually taking 20 measurements of the hold down spring, we're not 21 looking for cracking or wear. We're going to take 22 dimensional measurements of the hold down spring and 23 use that to determine whether that is acceptable. Is 24 that --
25 DR. LOTT: Yes.
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5027 1 MR. DOLANSKY: Would that be one example?
2 DR. LOTT: That would be a good idea.
3 ADMIN. JUDGE WARDWELL: Okay. So in that 4 particular case then for that spring, you're doing 5 some actual measurement rather than just observing for 6 a crack?
7 MR. DOLANSKY: Correct.
8 ADMIN. JUDGE WARDWELL: Do you have any 9 other examples of this?
10 MR. DOLANSKY: Give me one --
11 ADMIN. JUDGE WARDWELL: And let's just 12 assign that as a homework assignment so you don't 13 break out in beads of sweat panicking --
14 DR. LOTT: Right.
15 ADMIN. JUDGE WARDWELL: -- trying to find 16 everything you possibly can at this moment, I know the 17 feeling all too well being on that side of the table 18 often. Let's just -- when you feel comfortable, get 19 back with a response if there's any other examples you 20 can give in regards to those. Because I think it is 21 somewhat important because, again, you are monitoring 22 for the effects and I want to make sure we don't have 23 some giant hole of something else that's out there 24 that we would like to be able to track.
25 DR. LOTT: May I say that we did go through NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5028 1 this FMECA process which looked at each component and 2 asked ourselves what are the appropriate failure modes 3 and consequences? So we certainly did look at that 4 and asked ourselves the questions in the process.
5 ADMIN. JUDGE WARDWELL: So it sounds like 6 your reference to most cases, you were referring to in 7 most of the internals, we are looking for some type of 8 strain and deformation or a crack or something in that 9 neighborhood.
10 DR. LOTT: Yes. Well, we --
11 ADMIN. JUDGE WARDWELL: And in some of your 12 -- and then other cases, you are doing something else 13 besides looking for some sort of strain.
14 DR. LOTT: Yes. We tried to identify in 15 each case what the effect was that we thought we were 16 looking for in the prescribed inspection. So if it 17 was a wear inspection, we would be looking for loss of 18 material or evidence of wear on the surfaces. If it 19 was something that was subject to a cracking 20 mechanism, we would say, we're looking for cracks.
21 And this is where the question of irradiation 22 embrittlement would come up as well.
23 ADMIN. JUDGE WARDWELL: Yes. And I kind of 24 view wear as a strain, it's not really a strain, but 25 it's a loss of the dimension if nothing else.
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5029 1 DR. LOTT: Loss of --
2 ADMIN. JUDGE WARDWELL: A change in a 3 dimension.
4 DR. LOTT: Yes. Opening of a gap, 5 displacement of a component one with respect to the 6 other in some small way. We tried to find the places 7 where that was most evidence where we could see it.
8 ADMIN. JUDGE WARDWELL: Is it not true 9 though that your AMP is based on the fact of 10 monitoring for effects and not trying to deal with the 11 mechanism that caused those effects?
12 DR. LOTT: Yes. I think that was the 13 instruction of the Aging Management Program in 14 general.
15 ADMIN. JUDGE WARDWELL: And, NRC, would you 16 agree that, that's the motive behind the inspection 17 program for the reactor vessel internals AMP?
18 DR. HISER: This is Allen Hiser with the 19 NRC. Yes, that's correct. It's monitoring, 20 inspecting for aging effects. Mechanisms create 21 effects and they're important to understand what aging 22 effects you need to manage, but the mechanisms 23 themselves are not managed.
24 MR. STROSNIDER: This is Jack Strosnider 25 for Entergy. If I could just add there, if you want NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5030 1 a citation on that, you can go to the Statements of 2 Consideration that were issued when the License 3 Renewal Part 54 was issued in 1995. And it explicitly 4 discusses this notion of managing effects rather than 5 mechanisms. And I think part of the reason for that 6 is to recognize that the effect doesn't care what 7 synergisms, if there are any, that are happening. If 8 you see a crack, you see a crack and whatever 9 contributed to it, contributed to it and then you need 10 to take the right corrective actions. But I just 11 wanted to get you that citation in case you want to 12 look at that.
13 ADMIN. JUDGE WARDWELL: How does one 14 observe embrittlement?
15 MR. STROSNIDER: This is Jack Strosnider of 16 Entergy. You asked -- is that question directed at 17 me?
18 ADMIN. JUDGE WARDWELL: Yes.
19 MR. STROSNIDER: Okay. So the discussion 20 in the AMP is that embrittlement is not directly 21 observed, but it is managed through the detection of 22 cracks. And, as we've been discussing, if you find a 23 crack, you then need to assess it considering the 24 material properties that would be associated with 25 whatever level of embrittlement that component has NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5031 1 experienced. So it's an indirect approach. There is 2 not an embrittlement meter, if you will, that you can 3 go in and look. You have to do this indirectly and 4 that's what's laid out in the program.
5 ADMIN. JUDGE WARDWELL: Thank you.
6 DR. LOTT: May I add? This is Randy Lott.
7 In MRP 227, when there were components that were 8 subject to irradiation or thermal embrittlement, we 9 tried to note them in that way. There was effects 10 listed and then there would be the conditional note 11 that says, this effect should be -- aging management 12 for irradiation embrittlement, thermal embrittlement 13 would be required in this component. So we would mark 14 those places where this would be a concern.
15 ADMIN. JUDGE WARDWELL: Okay. Thank you.
16 Dr. Lahey --
17 CHAIRMAN MCDADE: Okay. If I could, just 18 -- and, Dr. Lahey, as I understand your position is 19 that given the fact that you can't directly observe or 20 monitor embrittlement, that there is a significant 21 change, even in the absence of cracks, that could set 22 up the reactor vessel internal for failure in the 23 event of a shock load.
24 DR. LAHEY: The reactor vessel internals, 25 is that what you're talking about? Yes, I agree with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5032 1 what they said that, that's what they're monitoring.
2 They're not looking at incipient failure, they're 3 looking at failures that have occurred. And I'm 4 concerned with the mechanism of microcracks, there are 5 plenty of cracks, but they're microcracks, and how 6 this weakens the material in the event that you have 7 an impulsive load on this material.
8 CHAIRMAN MCDADE: Okay. And it's your view 9 that there's no specific inspection technique 10 currently available that would be able to identify the 11 effects of embrittlement prior to failure and that, in 12 your view, the only way to adequately manage the 13 effect of aging is to have a reasonable replacement or 14 repair system. Am I correct in summarizing --
15 DR. LAHEY: Yes. I think you are correct.
16 I mean, the problem is, if we had all the data that we 17 really need to have, we wouldn't be having this 18 discussion. We would know what to do, the NRC would 19 be requiring it, Entergy would be complying with it, 20 and everything would be fine. But we don't. We have 21 some fragmentary data, which indicates concerns, and 22 so how do you deal with that? To me, the easiest way 23 to deal with it, for things like bolts, they're 24 relatively easy to replace, would be just replace 25 them. Get rid of the problem rather than try to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5033 1 calculate how many bolts can be failed and still keep 2 running and that sort of thing. I think that's a very 3 dangerous game, you're playing with fire when you do 4 that.
5 CHAIRMAN MCDADE: Okay. And let me address 6 this in turn to Dr. Hiser and then to Dr. Lott, it's 7 your view, Dr. Hiser, that in the absence of cracking, 8 there is no reason to believe that you are on the 9 verge of failure, even in the event of a shock load, 10 with these reactor vessel internals and that, 11 therefore, the inspections that are currently 12 available are adequate to monitor the aging of these 13 reactor vessel internals? Is that correct?
14 DR. HISER: This is Allen Hiser. Yes, that 15 is correct.
16 CHAIRMAN MCDADE: Okay. And, Dr. Lott, do 17 you agree with what Dr. Hiser just affirmed?
18 DR. LOTT: Yes, I do.
19 CHAIRMAN MCDADE: Okay. So that's 20 basically the difference of opinion here between the 21 NRC Staff, Entergy, and the position of New York's 22 witness, Dr. Lahey, as you see it Dr. Hiser?
23 DR. HISER: If I could just -- this is 24 Allen Hiser. If I could just elaborate a little bit 25 because --
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5034 1 CHAIRMAN MCDADE: Please.
2 DR. HISER: -- some of the discussion 3 yesterday related to ductility of the materials at 4 high fluence and a couple of the exhibits, New York 5 487 shows data, they're up to about 70 dpa that have 6 measurable fracture toughness, which is indicative of 7 ductility, NRC 209 has data that are up to about 12 8 dpa, that again show reasonable fracture toughness.
9 So there is still ductility in the material at these 10 fluence areas of interest. And, again, there also is, 11 my understanding, there was no exhibits provided to, 12 no data that we've seen that would indicate that 13 fatigue weakens a material or a component in the 14 absence of cracks.
15 ADMIN. JUDGE WARDWELL: And what types of 16 fluences do we expect after 60 years of operation?
17 DR. HISER: My understanding is for the 18 internals on the upwards of 75 dpa for the maximum for 19 the baffle bolts. If you go beyond that area, go 20 beyond the baffle assembly, they're much lower. So 21 the baffle bolts --
22 ADMIN. JUDGE WARDWELL: But the ones you 23 just quoted were well below that, weren't they? You 24 say it was over --
25 DR. HISER: The data in New York State 487 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5035 1 is upwards of 60, 70 dpa.
2 ADMIN. JUDGE WARDWELL: Okay. Thank you.
3 CHAIRMAN MCDADE: Okay. And, Dr. Hiser, 4 we're talking perhaps about different things. We've 5 been talking about cracks. Dr. Lahey mentioned 6 microcracks, cracks that are there, but are not 7 observable given current inspection techniques. Do 8 you agree that there's probably microcracks in many of 9 these reactor vessel internals, such as the baffle 10 former bolts?
11 DR. HISER: This is Allen Hiser. As the 12 CUF gets much closer to one, I think the likelihood 13 increases that you could have microcracks. But I 14 think the impact of those microcracks on the fracture 15 response of the component is negligible. I think 16 that's been demonstrated through many tests.
17 CHAIRMAN MCDADE: So the degradation in 18 fracture toughness would be minimal in your view?
19 DR. HISER: Well, I think the effect of the 20 degradation of fracture toughness would not be 21 significant. There may be reductions in fracture 22 toughness, but the impact of that in the presence of 23 even microcracks is not significant.
24 CHAIRMAN MCDADE: In your view, would not 25 be of consequence?
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5036 1 DR. HISER: Yes, that's correct.
2 CHAIRMAN MCDADE: Dr. Lott, do you agree 3 with that?
4 DR. LOTT: Yes, and I believe that the data 5 that Dr. Hiser just cited basically shows you that 6 these components can survive with actual cracks, 7 macrocracks, measurable cracks, not just the 8 microcracks that are suggested here. Certainly if we 9 can withstand cracking and we can demonstrate 10 stability of the component with a crack in it, concern 11 about microcracks does not, to my view, seem to be 12 important.
13 CHAIRMAN MCDADE: Okay. And this is the 14 data in New York State 487 and NRC 209, correct, Dr.
15 Hiser?
16 DR. HISER: Yes, that's correct.
17 CHAIRMAN MCDADE: Okay. Dr. Lahey?
18 DR. LAHEY: All right. I want to try to 19 clear up something we talked about yesterday and here.
20 Dr. Kennedy brought up the question, have I invented 21 new loads, do I need new loads to show that these 22 things can fail or not? And my answer was, no, the 23 existing type of accidents and seismic events are 24 sufficient. What's happened, if you go back in 25 history, is there was a point in time when we were NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5037 1 real concerned about design basis LOCA event and the 2 decompression wave that went in and what that would do 3 to the structures. Because in a particular instant, 4 you can have atmospheric, essentially atmospheric 5 pressure on one side and several thousand PSI on the 6 other and that would give you a very large impulsive 7 load.
8 So codes were written to address that, 9 method of characteristic kind of codes, like the WHAM 10 code that Stan Fabic wrote in Westinghouse, and 11 detailed analysis was done that showed for ductile 12 structures, they can withstand it. And we confirmed 13 that in the Loft experiment, which was an experiment 14 which we ran a simulated loss of coolant accident.
15 And so that concern was mitigated and as a 16 consequence, now people do the analyses using codes 17 such as RELAP and TRAC, those kind of codes, and they 18 really smear out this type of shock. They don't give 19 you the kind of shock loads that you would get if you 20 tracked the wave, the rarefaction waves, throughout 21 the vessel.
22 So I'm concerned with the real shock 23 loads. It's time, in my view, to go back and take a 24 look at this again with degraded materials. These are 25 weakened materials, you can't have lots of microcracks NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5038 1 in there and not have it weakened. It depends on how 2 much weaker it is, that's just due to fatigue. And 3 then irradiation makes it weaker yet, makes it more 4 brittle, more subject to failure. So that's the real 5 difference. That's why when I say shock loads, I may 6 mean something quite different than what they're 7 talking about with shock loads because they're talking 8 about the normal safety analyses using these big 9 system codes, which really are intended to look at the 10 inventory of the liquid and the coolability and that 11 sort of stuff. They're not very good at giving 12 instantaneous loads, either the thermal or pressure 13 loads.
14 CHAIRMAN MCDADE: Okay. But according to 15 Dr. Hiser and Dr. Lot, even in the situation where you 16 have macrocracks, observable cracks as opposed --
17 there still would not be sufficient degradation in 18 order to create a real risk, a significant risk of 19 failure. And they rely on, I believe it was, what was 20 it, New York 487, which was an Argonne Lab study from 21 2010. Somewhat more recent data than the one that you 22 had cited. Does that not alleviate your fears that 23 this material would continue to be sufficiently 24 robust?
25 DR. LAHEY: That's not what the Argonne NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5039 1 experts are saying. When they give input, they're 2 saying very similar things to what I'm saying. They 3 have the same kind of concerns about the lack of data 4 and the effect of it and what it may imply. And 5 they're, of course, hoping to get funding to run more 6 experiments. And I agree, more experiments are 7 needed.
8 CHAIRMAN MCDADE: Okay. So you don't 9 disagree with the study, it's just what you take away 10 from that Argonne Laboratory study is different from 11 what Dr. Hiser and Dr. Lott take away from it.
12 DR. LAHEY: Apparently. I mean, I don't 13 agree with what he said. I think a degraded structure 14 is inherently weak and more subject to failure.
15 CHAIRMAN MCDADE: Okay. But what we're 16 dealing with here is basically a professional 17 disagreement. We're looking at the same data and 18 you're interpreting it, it creates more concern in 19 your mind as to the potential for failure than what 20 was expressed by Dr. Hiser and Dr. Lott. Is that 21 correct?
22 DR. LAHEY: I suppose.
23 CHAIRMAN MCDADE: Well, I'm not trying to 24 -- this is new material for me and I'm trying to make 25 sure that I understand the relative positions here and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5040 1 what the basis of the disagreement is. It seems to me 2 that it is an interpretation of the data, the 3 significance of the data that's available and the data 4 that isn't available. And that you seem to be very 5 concerned with an absence of data that leaves 6 questions in your mind. Am I correct?
7 DR. LAHEY: Well, not entirely. I'll tell 8 you why I feel the way I do is I'm on a science 9 council for a program called CASL, which is based in 10 Oak Ridge. It's a very large program, which is funded 11 by DoE, to develop computational capability for 12 nuclear reactors. It involves many national labs and 13 many universities. One of the members of the board 14 was also an executive or the person in charge of the 15 Light Water Reactor Sustainability Program, so I've 16 had ample opportunity to talk to the people who are 17 working in that program. And I know from the comments 18 that I have there's people who would like to think, 19 that's just for operating out beyond where we're 20 talking about. We're talking about 60 years, they're 21 going on to 80 years.
22 But in nobody's mind is there a sharp 23 demarcation. Those are the same concerns, they have 24 the same concerns that I do, and they're working on 25 it. And we're now looking at how to take the code NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5041 1 that we've developed under CASL, which is a three 2 dimensional neutronics, thermohydraulics, fuel, water 3 chemistry, crud deposition, everything, code, and 4 apply it to these kind of issues for relicensing of 5 nuclear reactors. So there are certainly things you 6 can do, like the water chemistry, what is the water 7 chemistry during transience and how does that affect 8 FN?
9 So these are things we'll do in the next 10 -- I think we're going to get to in the next issue.
11 But it's not -- I haven't just made this stuff up. I 12 mean, these are valid concerns by people who are 13 working in it, some of them are even working under NRC 14 funding. So, I think it's an honest professional 15 disagreement and nobody has the perfect data set right 16 now to say, here's the answer. But there's 17 significant concerns about these type of things.
18 CHAIRMAN MCDADE: Okay. Thank you, Dr.
19 Lahey.
20 ADMIN. JUDGE KENNEDY: Dr. Lahey, since you 21 responded to my design basis question -- this is Judge 22 Kennedy. I'm just curious, it seems like the scenario 23 that you painted in terms of analysis methodologies 24 would be just as applicable to non-irradiated material 25 shock loading analysis as irradiated material shock NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5042 1 loading analysis. Are you -- am I correct in that?
2 Do you have the same concern with non-irradiated 3 material analysis, shock loading, as you do with the 4 irradiated material shock loading analysis?
5 DR. LAHEY: I didn't explain it very well.
6 Originally, that was the concern and so they were more 7 ductile materials. These were newer reactors and so 8 the concern was, will they maintain a coolable 9 geometry during these type of events? And so, I think 10 that question has been settled and now the issue is, 11 given an intact geometry, can you cool them with 12 emergency core cooling engineered safety systems? The 13 new thing is, now you have degraded structures, which 14 you never had to deal with before, highly degraded 15 both from fatigue and irradiation, and it's time to 16 now relook at that. Because they can be, not only 17 deformed, they can be failed and relocated and then 18 the concern is core coolability.
19 ADMIN. JUDGE KENNEDY: I guess I thought I 20 understood you to say that there was a problem with 21 the analysis techniques that are being utilized to 22 analyze the shock loadings, that they were deficient.
23 Is that not the case? Is that not what you were 24 trying to say?
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5043 1 and used previously were focused on tracking 2 depressurization fronts and their effect in terms of 3 loading. They're rather expensive to run method of 4 characteristic type codes, but they give you good 5 answers. Since then, people have concluded that, at 6 least for ductile geometry, the core will stay intact, 7 so given this geometry, the geometry is assumed, now 8 we can use codes that do not do that, they're more 9 control volume kind of codes, like RELAP and TRAC. So 10 when you do the analyses for plants like Indian Point, 11 you don't look at the deformation of the core, you 12 have a certain geometry. What you look at is where's 13 the liquid, how's the cooling, what's the peak clad 14 temperature, that sort of thing, to see if you're in 15 the compliance with the safety regulations.
16 ADMIN. JUDGE KENNEDY: All right. Thank 17 you.
18 ADMIN. JUDGE WARDWELL: Getting back to the 19 inspections where we were, do you have any criticisms, 20 Dr. Lahey, of their approach of trying to monitor for 21 effects rather than mechanisms? Strictly this issue, 22 not in regards to your overall issue, but as an expert 23 witness before us, do you agree that monitoring for 24 effects is a good approach rather than trying to 25 monitor for any mechanism?
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5044 1 DR. LAHEY: The answer to that, your honor 2 -- this is Richard Lahey for New York State. As I 3 tried to say yesterday, I think the program that they 4 have in place, which is inspection based, is a pretty 5 good program. So I'm all for it. I just don't think 6 it addresses all the real concerns.
7 ADMIN. JUDGE WARDWELL: You've said that, 8 thank you. In your testimony, 482, Page 17, Lines 6 9 through 17, you talk about the rather complex and 10 interacting metal degradation mechanisms associated 11 with fatigue, irradiation, and corrosion interact is 12 still an area of active research. And you point to a 13 DoE, USNR, in conjunction with various other national 14 laboratories that have recently embarked on a program 15 to understand and resolve issues related to these 16 interacting and synergistic effects. And you 17 reference NUREG/CR-7153, which is entitled Expanded 18 Materials Degradation Assessment, or EMDA --
19 DR. LAHEY: Right.
20 ADMIN. JUDGE WARDWELL: -- Aging of Core 21 Internals. And in that particular NUREG, do you know 22 the age of the study of how far out they were looking?
23 Was it within the license renewal period or was it 24 beyond that, to your knowledge?
25 DR. LAHEY: That report is part of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5045 1 Light Water Reactor Sustainability Program. So 2 they're looking beyond the design life of 40 years, 3 they're going out farther. They're actually charged 4 with seeing if you can get out to 80 years. But the 5 phenomena has no sharp demarcation at any point in 6 time. So they're looking at it all the way out.
7 ADMIN. JUDGE WARDWELL: But any conclusions 8 they reach, if they're making them in regards to how 9 it is out to 80, that would be certainly different 10 than if it was out to 60, wouldn't it?
11 DR. LAHEY: Yes, but if they find phenomena 12 that is of concern at 50, they will communicate that 13 and it will be relevant to what we're talking about 14 here.
15 ADMIN. JUDGE WARDWELL: And did they have 16 any conclusions from that report that related to out 17 to 50 or 60 years that demonstrate or support your 18 positions?
19 DR. LAHEY: Dr. Busby is one of the authors 20 of that report and one of the things that I recall 21 from that report was he was greatly concerned about 22 irradiated assisted stress corrosion cracking and what 23 the impact of that may be.
24 ADMIN. JUDGE WARDWELL: Okay, thank you.
25 You go on, in fact on that same page and extending NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5046 1 over to Page 18 with lines 17 through 22 and then 1 2 through 2 on the top, that the federal government has 3 also embarked on a fairly large research program. And 4 this is what you've termed -- or not you -- yes, you 5 call this Light Water Reactor Sustainability Program, 6 which includes research into whether the different 7 materials and light water reactor components can 8 continue to perform their intended functions during 9 the period of operations. This report that you cite, 10 which is Exhibit New York State 485, was in August of 11 2014. And, again, what were the periods of years that 12 they were looking at under that particular document?
13 DR. LAHEY: From now until operating out to 14 80 years.
15 ADMIN. JUDGE WARDWELL: Okay, thank you.
16 DR. LAHEY: I mean, it's being done at 17 various national labs and they put out monthly 18 newsletters on where they stand on various things.
19 ADMIN. JUDGE WARDWELL: Okay. Thank you.
20 NRC's testimony, 197, Answer 124, Page 75, states 21 that, "Entergy has also implemented a low leakage core 22 design for IP2 and IP3 prior to 30 calendar years of 23 operation, which reduces the potential for irradiation 24 driven aging mechanisms, such as the IASCC, the IE 25 void swelling, and ISR." I guess I'll go to Staff NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5047 1 considering this was your exhibit. I knew what all 2 those other acronyms were for, but what about the ISR?
3 MR. POEHLER: This is Jeffrey Poehler of 4 the Staff. So ISR is irradiation assisted stress 5 relaxation or irradiation stress relaxation.
6 ADMIN. JUDGE WARDWELL: Irradiation stress 7 relaxation, is that what the ISR is for?
8 MR. POEHLER: Correct.
9 ADMIN. JUDGE WARDWELL: Okay, thank you.
10 How is this low leakage core design achieved?
11 MR. POEHLER: This is Jeffrey Poehler of 12 the Staff. So, basically the core design is such that 13 you have fuel assemblies that have higher, I guess, 14 levels of burn up or depletion or placed around the 15 outside of the periphery of the core so that the newer 16 fuel assemblies are concentrated more towards the 17 middle of the core.
18 ADMIN. JUDGE WARDWELL: And what does this 19 do for you supposedly?
20 MR. POEHLER: It reduces the, I guess, the 21 leakage of -- it reduces the fluence levels.
22 ADMIN. JUDGE WARDWELL: That's what I was 23 wondering. Okay, thank you.
24 CHAIRMAN MCDADE: I'm sorry. It releases 25 the -- I just didn't hear.
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5048 1 ADMIN. JUDGE WARDWELL: It reduces the 2 fluence --
3 CHAIRMAN MCDADE: Fluence.
4 ADMIN. JUDGE WARDWELL -- or the flux 5 actually, so over time the fluence will be less.
6 ADMIN. JUDGE WARDWELL: Dr. Lahey, do you 7 have any comments on the low leakage core design?
8 DR. LAHEY: No. No, I understood what they 9 were doing and why. I mean, it's certainly helpful to 10 the concern we talked about yesterday with the core 11 plates and the pressure vessel to try to reduce the 12 fluence.
13 ADMIN. JUDGE WARDWELL: Okay. Thank you.
14 Entergy's Exhibit 616, testimony, Answer 182 on Page 15 120, failure of a component without a pre-existing 16 crack is governed by the mechanical properties of the 17 material, the yield strength, the ultimate strength in 18 particular. Irradiation increases the yield and 19 ultimate strengths, and we talked about this. I guess 20 my question for Entergy, does this increase in 21 strength with irradiation occur without any bound? I 22 mean, will it continue on forever, the longer you 23 irradiate it, will it continue to gain strength --
24 DR. LOTT: Okay. This is Randy --
25 ADMIN. JUDGE WARDWELL: -- keep on going NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5049 1 unlimited?
2 DR. LOTT: This is Randy Lott for Entergy.
3 I think it might be clearer if you examine the exhibit 4 I offered this morning about yield stress in material 5 versus fluence. In general, most of the changes that 6 happen to the mechanical properties of the material 7 happen within the first five to ten dpa. So they 8 actually happen early in life and saturate such that 9 there are much less changes. That's true both of the 10 increase in yield stress and the ultimate stress, the 11 decrease in ductility. Most of the action happens, I 12 would say, at less than ten dpa. Which, for your 13 highly irradiated -- ten dpa may be the end of life 14 fluence for some components, for others it's as much 15 as 60, for others it's one. It just depends on where 16 the component is.
17 ADMIN. JUDGE WARDWELL: And are you 18 referring to all components or just the reactor vessel 19 internals?
20 DR. LOTT: The internals. The internals 21 are the only ones that are going to see enough neutron 22 exposure to exceed one dpa. We talk about milli-dpa 23 when get out to the reactor pressure vessel.
24 ADMIN. JUDGE WARDWELL: Okay. Thank you.
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5050 1 are you claiming those still aren't caused by 2 irradiation or where would those microcracks be 3 occurring from?
4 DR. LOTT: The microcracks, as I understand 5 the argument, are one of the precursors to crack 6 formation in the fatigue specimens. So, as where it 7 was offered into evidence, and I think this is also 8 explained in NRC NUREG-6909, at least in the draft 9 version, they're sort of stages in the process of 10 forming an observable crack and one of those early 11 stages is the microcracking stage. So, again, yes, 12 it's a form of -- I mean, obviously, if there is a 13 limiting number of cycles, then something must be 14 changing over the course of time. It's been a 15 struggle to identify those things in these materials, 16 but this microcracking and some of these other small 17 micro-structural changes are just the evidence of 18 accumulating fatigue in material.
19 ADMIN. JUDGE WARDWELL: Okay, thank you.
20 Does irradiation decrease the resistance to the crack 21 propagation?
22 DR. LOTT: Okay. Let me ask, I want to 23 clarify here. When we talk about crack propagation, 24 are we talking about crack propagation due to loading 25 or crack propagation due to corrosion cracking or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5051 1 crack propagation due to fatigue? There are multiple 2 kinds of ways that, that term might be used.
3 ADMIN. JUDGE WARDWELL: Answer for each 4 situation.
5 DR. LOTT: Okay. So, I've created my own 6 question now, didn't I?
7 (Laughter.)
8 DR. LOTT: In terms of mechanical 9 properties, we talked and I think Dr. Hiser talked 10 earlier about the fracture mechanics mechanisms that 11 we look at or fracture toughness values that we use.
12 And in general, those toughness values are measured in 13 terms of J-resistance curves. That would tell you 14 effectively how much work it takes to advance a crack 15 in the material mechanically. And the slope of that 16 J-resistance curve gives you an idea of what the, 17 again, the resistance the material is to crack 18 advance. And in general, as the irradiation goes up, 19 that number goes down. So there is some decrease in 20 the resistance. The good news, however, is if you're 21 even measuring resistance, you're into ductile 22 failure, not into the brittle failures we had before.
23 Fatigue, I think we talked about the 24 effect of radiation on fatigue earlier and it's our 25 contention that where we have relevant data, that data NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5052 1 says that the resistance to fatigue initiation is 2 decreased. Fatigue propagation, I think, follows 3 similar kinds of rules, but it is a different 4 mechanism. I think we'd have to go back -- I'd have 5 to do a little more research if you wanted to know 6 about that.
7 ADMIN. JUDGE WARDWELL: But by different 8 rules, I mean, does irradiation decrease that 9 resistance to crack propagation? Does a crack 10 propagate faster when it's been irradiated under 11 fatigue cracking?
12 DR. LOTT: Not necessarily. I think we'll 13 get back to the same kind of data that we discussed in 14 terms of the initiation. And I'd have to go back and 15 review that data for you in detail. And, again, we 16 talked about -- so there's fatigue stress corrosion 17 cracking, again, that effectively is irradiation 18 assisted stress corrosion cracking, that's the concern 19 we have is that there will be crack formation and 20 growth due to irradiation.
21 ADMIN. JUDGE WARDWELL: Okay. Thank you.
22 Dr. Lahey, do you believe that irradiation decreases 23 the resistance to crack propagation once a crack forms 24 from whatever mechanism?
25 DR. LAHEY: Yes, definitely the crack will NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5053 1 propagate faster.
2 ADMIN. JUDGE WARDWELL: Thank you.
3 Entergy's Exhibit 616, Answer 114, Page 71, says RVIs 4 have no pressure retaining function. A pressurized 5 thermal shock, PTS, transient, therefore, does not 6 subject the RVI components to the sustained membrane 7 stresses characteristics of the effects of a PTS event 8 on a reactor pressure vessel itself. I guess my 9 question to Entergy is, why wouldn't the internals 10 still feel the pressure wave from a PTS transient if 11 one occurred?
12 DR. LOTT: Well, and again, I'm not sure if 13 someone else from Entergy panel wants to -- I'll step 14 forward first, I guess, and they can help me out as I 15 go along. Effectively, the pressurized thermal shock 16 is a repressurization, it's not necessarily -- it's a 17 long time in developing. It's not, I don't believe, 18 the same kind of process we're talking about here.
19 And one of the key elements of it is that you've 20 cooled down the vessel at a time when the internal 21 pressure on the vessel, the membrane stresses, remain 22 high because you've not seen the depressurization of 23 the system. So, the pressure itself in the vessel 24 creates the baseline high loads that the thermal shock 25 challenges. I'm not sure I've -- perhaps somebody can NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5054 1 help me out with my words.
2 MR. AZEVEDO: Yes, this is Nelson Azevedo 3 from Entergy. What Dr. Lott said was correct, the 4 repressurization during a PTS event is not an 5 instantaneous event like a pipe break, we have a 6 fraction wave traveling through the system. This is 7 a repressurization that takes some time to 8 repressurize the system.
9 ADMIN. JUDGE WARDWELL: But even so, 10 wouldn't the vessel internals feel however small 11 gradual change there is regardless? And I'm trying to 12 understand your statement that, as I assume you're 13 trying to imply on this Answer 114, is that we don't 14 have to worry about PTS because it's not a pressure 15 retaining -- these aren't pressure retaining 16 components, but they're in and amongst the pressurized 17 area and so why wouldn't it still feel that change, 18 whatever shock that does occur?
19 MR. AZEVEDO: This is Nelson Azevedo for 20 Entergy. And, yes, that's correct, but the way I 21 visualize at least, for a pressure boundary component, 22 like the reactor vessel, you have 2,200 pounds on the 23 inside and essentially zero on the outside. So you 24 have that whole delta P across the component. When 25 you're talking about a reactor vessel internal, like NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5055 1 a column, it's pressurized the same amount all the way 2 around, so there's no pressure differential. That's 3 why they're differentiated between pressure boundary 4 components and vessel internals.
5 ADMIN. JUDGE WARDWELL: Thank you, that's 6 helpful. And, Dr. Lahey, do you agree with those 7 statements? Any disagreement?
8 DR. LAHEY: I was somewhat surprised when 9 I got comments on my concerns about thermal shock, 10 significant thermal shocks, because I was not worried 11 about pressurized thermal shocks, that's a pressure 12 vessel phenomena. I was worried about cold water, for 13 example, coming in and some of the internal structures 14 would then be suddenly changed in temperature, it 15 would hit, it would shock the surface, it would try to 16 contract, it would crack, it could fail. That was my 17 concern.
18 ADMIN. JUDGE WARDWELL: Sure. And I just 19 want to verify that you don't have any disagreement 20 with the pressurized thermal shock specifically that 21 we just discussed now.
22 DR. LAHEY: No, I have no disagreement with 23 it.
24 ADMIN. JUDGE WARDWELL: Okay, great.
25 DR. LAHEY: I wasn't concerned with it.
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5056 1 ADMIN. JUDGE WARDWELL: Just want to make 2 sure they're not trying to spoof me over here with 3 some voodoo.
4 DR. LAHEY: No.
5 ADMIN. JUDGE WARDWELL: That's why I'm 6 going to you to see if you can agree with those small 7 points. That's why I go back and forth. But I 8 understand your comments about the other thermal 9 shock.
10 DR. LAHEY: Okay.
11 ADMIN. JUDGE WARDWELL: That was a 12 different issue. Thank you. Entergy's Exhibit 616, 13 Question and Answer 185, Page 122. The Question 185 14 says, do accident loads need to be considered as a 15 contributor to the effects of aging on reactor vessel 16 internals? And the Answer in 185 says, no, aging is 17 a gradual, long-term degradation of a component 18 resulting from sustained environmental conditions, 19 that is applied loads and residual stresses. And then 20 goes on to talk about the ASME codes and Staff 21 Guidance.
22 While I can understand why the loads 23 aren't a contributor to aging, I just want to make 24 sure that it's clear that I haven't confused this by 25 saying that still those loads, those design basis NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5057 1 loads, are considered in your AMP in regards to any 2 evaluation that you might be doing to look at any 3 changes in strength or whatever else occurs during the 4 aging process. Is that correct? And I'll address 5 this to Entergy, I guess, because it was your 6 testimony.
7 DR. LOTT: This is Randy Lott for Entergy.
8 Yes, what we were trying to say, I think, is exactly 9 what you have said that it does not contribute to the 10 condition that you would observe at any time of the 11 components, but if we're looking at the ability of the 12 components to withstand an accident mode, that's a 13 different question.
14 ADMIN. JUDGE WARDWELL: You're still --
15 DR. LOTT: It's not defined as the aging.
16 ADMIN. JUDGE WARDWELL: You still will 17 consider those loads --
18 DR. LOTT: Right.
19 ADMIN. JUDGE WARDWELL: -- design basis 20 loads. And aren't the design basis loads -- do not 21 the design basis loads also include LOCAs?
22 DR. LOTT: Yes.
23 ADMIN. JUDGE WARDWELL: Thank you.
24 Entergy's testimony, 616, Answer 185, Page 122, the 25 ASME Code Section 3 compares accident loads such as NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5058 1 large break LOCAs and large main steam line breaks to 2 the stress allowables to ensure that the affected 3 components remain capable of performing their intended 4 safety function during and after the event. And so I 5 want to just confirm again with you that doesn't the 6 degraded age strength of the PEO determine these 7 stress allowables in your analysis?
8 DR. LOTT: I -- go ahead.
9 MR. AZEVEDO: Yes, this is Nelson Azevedo 10 for Entergy. The stress allowables are directly 11 obtained from the ASME code. So based on what 12 material the component is made out of, the ASME code 13 specifies what the allowables are, we don't get to 14 choose those.
15 DR. LOTT: Yes, and I think that what 16 you're talking about is basically the design section 17 of the code, right? So it's --
18 MR. AZEVEDO: Right. So this would be 19 discussed in ASME Section 3, the original analysis, 20 and the analysis has been revised since then.
21 ADMIN. JUDGE WARDWELL: So isn't it taking 22 advantage of the full virginal strength of this 23 material if it's part of your -- or, I guess not, if 24 it was an allowable.
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5059 1 for Entergy. You're right. The original Section 3 2 design process uses ASME minimum yield strength and 3 code allowables in evaluating the margins that are 4 intended to be there. What we've described 5 previously, those tensile strengths and yield 6 strengths may actually increase due to the irradiation 7 and aging process. That's not taken into account, you 8 still use the code required properties of the virgin 9 material when you do those analyses. They're still 10 valid, in fact.
11 ADMIN. JUDGE WARDWELL: But if you observed 12 a crack and now you're trying to evaluate where you go 13 from there, does this statement not apply to those 14 types of analyses that you may or may not perform as 15 part of your corrective measure for the observation of 16 that crack?
17 MR. AZEVEDO: Yes, this is Nelson Azevedo 18 for Entergy. The design process, the ASME Section 3 19 specifically that we're talking about right now, does 20 not allow cracks. So cracks are not allowed during 21 the design phase. If you find cracks, then you 22 evaluate them either under ASME Section 11, which is 23 the operating version of the code, or another NRC 24 approved methodology. But the original design that 25 we're talking about, stress calculations do not allow NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5060 1 cracks.
2 ADMIN. JUDGE WARDWELL: Thank you, that's 3 helpful. Let's move now on to talk a little bit more 4 about these inspections. Yes, Dr. Lahey, would you 5 like to -- I hadn't gotten back to you after telling 6 you why I was getting back to you all the time.
7 DR. LAHEY: This is Richard Lahey again.
8 I think that was a great question because it really 9 captures a fundamental difference in our view, or my 10 view and the view that they have expressed. I'm 11 concerned with having degraded properties and the 12 ability if you have high enough strain to fail those 13 properties, those components in core. And I don't 14 believe it's adequate just to do a safety analysis 15 using ductile materials. I don't believe it's 16 adequate at all. That's the fundamental difference in 17 our view.
18 ADMIN. JUDGE WARDWELL: Okay. Thank you.
19 CHAIRMAN MCDADE: Excuse me. Judge 20 Wardwell, before we move on, I've got a question, 21 perhaps going back a little bit, that I need some 22 clarification on. Maybe a little bit off point, but 23 Dr. Lahey had raised an issue at Pages 45 and 46 of 24 his testimony, about issues at Davis Besse. And in 25 the response, in Answer Number 102, by Entergy, you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5061 1 indicated that -- dismissed it by saying, it does not 2 appear that conditions similar to those which led to 3 the degradation at Davis Besse are present at Indian 4 Point. But there isn't any explanation as to why the 5 conditions are different or how the conditions are 6 different. This is your Answer 102 on Page 59 of your 7 testimony. Anybody from Entergy can address what the 8 differences are between the conditions at Davis Besse 9 and Indian Point and why you believe the concern of 10 Dr. Lahey is unwarranted?
11 MR. AZEVEDO: Yes, this is Nelson Azevedo 12 for Entergy. The events at Davis Besse, I'm assuming 13 you're talking about the corrosion that the reactor 14 vessel had --
15 CHAIRMAN MCDADE: Yes.
16 MR. AZEVEDO: -- at Davis Besse.
17 CHAIRMAN MCDADE: And I realize that's 18 covered by a different AMP, but --
19 MR. AZEVEDO: Yes, so the events at Davis 20 Besse occurred because the reactor vessel had 21 penetrations, which are LI600, cracked and they were 22 undetected for a period of time. And that caused 23 leakage onto the reactor vessel head, which eventually 24 corroded the base metal itself. Indian Point does not 25 have this issue and one of the reasons is because we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5062 1 do inspections every outage. We inspect those 2 penetrations every outage and we have not found any 3 cracks to date. So we know there is no leakage going 4 on similar to what happened at Davis Besse.
5 MR. DOLANSKY: This is Bob Dolansky with 6 Entergy. Additionally, we not only inspect the 7 penetrations for cracking, but we also do what's 8 called a bare metal visual inspection on the top of 9 the head, on the outside surface of the head, where we 10 go around every penetration and look visually and make 11 sure that there's no evidence of any corrosion. So we 12 actually look for where cracking could start inside 13 the head and then we also verify by a bare metal 14 visual inspection that there is in fact no corrosion 15 going on like there was at Davis Besse.
16 CHAIRMAN MCDADE: Okay. But, Dr. Lahey, am 17 I correct, what your concern was, was the mechanism 18 that caused the cracking as opposed to the failure of 19 inspecting?
20 DR. LAHEY: Well, I'm concerned about what 21 happens internally and what happens if it's a through-22 crack. So it's heartening to hear that they do those 23 types of inspections. I think that's very 24 responsible. But there are some welds in the inside 25 that they can't do full inspection of and if they NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5063 1 weaken for various degradation mechanisms, like stress 2 corrosion cracking or whatever happens, you can have 3 a concern, particularly for some of the stub tubes 4 associated with the control rod drives. So this is 5 one of the concerns that we had.
6 CHAIRMAN MCDADE: Okay, but are we 7 concerned about the mechanism that caused the 8 cracking? As I understand it, at Davis Besse, there 9 was an issue that many of these locations were either 10 inaccessible to inspection or very difficult to 11 inspect and that they didn't inspect and that's why 12 the problem was able to reach the level that it had.
13 Here, Entergy is indicating that they do have an 14 inspection program that identifies these potential 15 problems and are able to ensure that there has not 16 been cracking. What I'm concerned with is the 17 mechanism that would have caused the cracking in the 18 first place, not the ability to identify it, but the 19 mechanism that would cause the cracking. And would 20 there be any difference in the situation leading to 21 that mechanism between Davis Besse and Indian Point?
22 DR. LAHEY: Well, one of -- this is Dr.
23 Lahey again. One of the things in the conclusions in 24 the document that you asked about before, which was 25 authored by Dr. Busby, was that irradiated assisted NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5064 1 stress corrosion cracking was still one of the 2 greatest concerns that we have. And in fact, EPRI had 3 put that out as a statement about a year before as 4 well. So it is an issue, I mean, that's the mechanism 5 as I understand it that's of most concern and since 6 you can't do full inspection inside, it remains a 7 concern.
8 CHAIRMAN MCDADE: Okay. When --
9 MR. STROSNIDER: This is Jack Strosnider 10 for Entergy, if I could just comment on this. First 11 of all, there could be differences in the 12 susceptibility to cracking based on who manufactured 13 the vessel head and the specific configuration. I 14 can't speak to that without a lot of details, but 15 there could be differences. But I think the important 16 thing to recognize is that GALL says that the 17 potential for cracking does exist and that's why these 18 inspections are done.
19 So, if you look at the overall framework, 20 the GALL report doesn't say you're not going to see 21 that kind of cracking, it says, it's a potential and 22 you need to go look for it and they're doing two 23 different types of inspections. And, by the way, at 24 Davis Besse, they had several outages of opportunity 25 to identify the problem that was occurring there, but NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5065 1 the inspections weren't being done that needed to be 2 done.
3 CHAIRMAN MCDADE: And all I'm trying to 4 find out is just to make sure I understand, when you 5 use the term, does not appear to have conditions 6 similar, are we talking about only the fact that you 7 have an effective inspection program, and specifically 8 your reactor vessel head penetration inspection AMP, 9 that it is effective versus the one at Davis Besse?
10 Or is there something different about the reactor 11 itself where when you use the term conditions similar 12 which would result in cracking as opposed to just your 13 ability to identify it?
14 MR. AZEVEDO: Yes, sir. It's Nelson 15 Azevedo for Entergy. There are differences between 16 the way the reactor vessel heads were fabricated for 17 Davis Besse versus Indian Point. Davis Besse's 18 reactor vessel head was made by B&W. Our heads were 19 made by Combustion Engineering. So the process they 20 used to install the penetrations and the way they 21 strength the penetrations, the way the welding was 22 done, was different. Also, the materials were 23 different. They're both LI600, but the Davis Besse 24 materials were B&W tubular products, ours are 25 Huntington alloy materials.
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5066 1 And if you look at the OE, at the 2 operating history for the leaks, essentially all the 3 leaks occurred in heads made by B&W. If you look at 4 Oconee, Davis Besse, most of the other leaks were by 5 B&W heads. Again, ours is different. So they're both 6 LI600, so they're both material susceptible, but I 7 feel that our material is much less susceptible than 8 the Davis Besse heads. Also, Oconee --
9 CHAIRMAN MCDADE: Excuse me. Why and is 10 there anything here relevant that carries over to the 11 reactor vessel internals, which we're focusing on?
12 MR. AZEVEDO: Yes, that's what it was going 13 to say. As far as the irradiation on the upper head, 14 there's no -- the fluence is very, very low, less than 15 one times ten to 17. And, no, I don't think there's 16 anything that carries over to the reactor vessel 17 internals.
18 CHAIRMAN MCDADE: Okay. Dr. Lahey?
19 DR. LAHEY: Yes, I misspoke, I should have 20 said primary water stress corrosion cracking. He's 21 absolutely correct.
22 CHAIRMAN MCDADE: Okay, thank you. Dr.
23 Wardwell?
24 ADMIN. JUDGE WARDWELL: Entergy's 25 testimony, 616, Answer 139, Page 86, states at Table NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5067 1 5-2 of Attachment 2 to NL12-037, and that Attachment 2 2037 is the Inspection Plan, so this is Table 2 of the 3 Inspection Plan, specifies a required timing of the 4 first inspections and subsequent intervals for the 5 primary components in the reactor vessel internals 6 AMP. For most components, the first planned 7 inspection at Indian Point are scheduled for two 8 refueling outages from the beginning of the PEO, i.e.,
9 the Spring of 2016 for IP2 and the Spring of 2019 for 10 IP3.
11 You go on to state in Answer 142 of Page 12 92 that the NRC Staff, in a safety evaluation for MRP 13 227-A, acknowledged the justification for the timing 14 of the initial PEO and subsequent inspections and 15 found the inspection intervals acceptable, and 16 referencing the SE that Staff put out for MRP 227, 17 which I believe is Entergy's Exhibit 230. And I guess 18 I'll start with Entergy. What's your technical basis 19 for justifying not performing the first inspections 20 during the first refueling outage?
21 DR. LOTT: This is Randy Lott for Entergy.
22 As I think someone said earlier in the day, we did a 23 lot of evaluations for these components and we did not 24 identify any component where degradation, shall I say, 25 fell off the edge of the table. That it was a gradual NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5068 1 process, it was a process that seemed to be 2 appropriately managed. We wanted to see that it was 3 integrated into -- that there were baseline conditions 4 set up within the program. And so we thought within 5 the first two years of PEO was appropriate.
6 It also gave a chance to coordinate some 7 of these examinations with the ASME Section 11 8 examinations, gave some flexibility to that schedule, 9 which I think is important to the implementation of 10 these exams. It's proven to be -- and we've already 11 done a number of baseline exams and I think we've 12 shown that our number of findings have been extremely 13 low and it seems an appropriate response.
14 ADMIN. JUDGE WARDWELL: But I guess I don't 15 understand the timing need. I mean, the application 16 was submitted in 2007, so they knew this was coming 17 up. Why did they need more time to get ready for the 18 first inspection besides the first refueling outage 19 after the PEO started?
20 MR. DOLANSKY: This is Bob Dolansky with 21 Entergy. To perform the MRP 227-A inspections, we'll 22 remove the component called the core barrel. We also 23 do that when we do the ten year ISI inspections. So 24 we wanted to do both of those together. And the ten 25 year ISI inspection, which inspects the reactor vessel NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5069 1 itself, the welds that make up the reactor vessel, 2 that's required on a ten year frequency for ASME 3 Section 11. So the reason we're doing it during the 4 second outage of the PEO is to allow us to do both of 5 those together so we only have to remove the core 6 barrel one time.
7 ADMIN. JUDGE WARDWELL: NRC Staff, would 8 you like to comment on why you are convinced through 9 their justification that the second refueling outage 10 was adequate?
11 DR. HISER: This is Allen Hiser. The Staff 12 -- I think some of the bases relate to the 13 expectations of degradation and the desire to allow 14 for a higher likelihood that degradation would be 15 detectable. Some of the mechanisms, the analyses by 16 Westinghouse, it led to MRP 227, the degradation that 17 they -- the levels that they were using relate to 60 18 year fluences and things like that. At 40 years, the 19 fluences aren't to that level. So I think we just 20 thought that it was reasonable to delay the baseline 21 inspections into the PEO.
22 As Dr. Lott mentioned, plants have done 23 inspections so far, so far there have been very 24 limited, if any, indications of degradation identified 25 at all. So I think that reinforces -- if the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5070 1 inspections that had been done at other plants 2 indicated problems, then I think the NRC would have 3 likely pushed to accelerate the inspections at Indian 4 Point and similar plants.
5 ADMIN. JUDGE WARDWELL: Back to Entergy.
6 I believe your Table 5-2 and 5-3 of your Inspection 7 Plan shows inspection intervals of ten years for many 8 components. Were there any other shorter intervals 9 incorporated into your Inspection Plans beside the ten 10 year cycle that you can remember?
11 MR. DOLANSKY: No.
12 ADMIN. JUDGE WARDWELL: And what's the 13 justification for what seems like a pretty long 14 interval between inspections considering the 15 importance of which you're placing on these 16 inspections?
17 DR. LOTT: This is Randy Lott for Entergy.
18 Again, there were two issues. One was coordination of 19 schedules because we felt that, that was actually 20 important, it's a fairly difficult operation to remove 21 the core barrel and it's not something people want to 22 do lightly. And I have to point out that it was our 23 feeling that, particularly as an industry, a single 24 inspection gives us information about a wide variety, 25 for instance, the 800 baffle former bolts in the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5071 1 plant, there's factors of three and four in the 2 fluences, so we see lead components, we see a whole 3 range of fluences in a single inspection. As we 4 gather data, that gives us a much broader data base.
5 And I'll say that we looked at in 6 particular one of the components that had us most 7 concerned were the baffle former bolts themselves.
8 That, as Dr. Hiser has kind of alluded to here, that 9 drove a lot of our inspection schedule decisions. As 10 well, we found because of the assumptions in our 11 analysis and the way that the analysis was put 12 together that the rate of degradation of the baffle 13 former bolts was actually slower in the last part of 14 our irradiations and that it took at least, I think it 15 was 25 effective full power years of operation before 16 we had a reasonable number of predicted failures such 17 that we thought there would be actually something to 18 see. We have experience on baffle former bolt 19 inspections from the 1990s when things were done at 20 lower fluences and very little is seen in any of those 21 plants. So, again, we just felt that it gave us the 22 best opportunity to collect data and that we would get 23 a robust data base from the industry.
24 MR. STROSNIDER: This is Jack Strosnider 25 for Entergy. I'd also like to add to this and what NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5072 1 was mentioned a minute ago that I think it's helpful 2 if you look at this from a fleet perspective. There's 3 a lot of plants doing these inspections at different 4 times and they have similar designs, similar 5 environments. So if something shows up in the 6 operating experience, in accordance with the Program 7 Element 8, Operating Experience portion, that's going 8 to be reviewed and evaluated and if it required a 9 change in the inspection frequency, then they could do 10 that. So there's a lot more data than is coming just 11 from the inspections at this point.
12 CHAIRMAN MCDADE: Okay. If I can clarify 13 in my own mind here, so the difference between the 14 initiation and propagation of a crack, do you have 15 sufficient data to determine, for example you inspect 16 today, tomorrow a crack is initiated, it's another ten 17 years before you inspect again. At the rate of 18 propagation of that crack, do you have data that would 19 indicate how long it would take after initiation for 20 a crack in the bolt to become problematic?
21 DR. LOTT: Well, again, our presumption in 22 the bolts was that any bolt that had a crack would 23 fail. So we basically did not do a calculation of 24 crack propagation in the bolts. We basically said, if 25 it's cracked, it's gone. We do have --
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5073 1 CHAIRMAN MCDADE: Well, what I'm getting 2 at, Dr. Lott, is you do the inspection today, you 3 don't see any cracks in a baffle former bolt. You now 4 don't inspect it for another ten years. You don't 5 know when during that period of time a crack may 6 initiate. If it initiates nine years nine months 7 after the inspection, there's no reason to believe 8 there will be significant propagation by the time you 9 inspect the next time.
10 My question is, is there any way of 11 knowing, for example, if the crack initiates a day, a 12 week, a month after the inspection of whether it is 13 subject to failure within that ten year period before 14 it's inspected again? Is there data that would lead 15 you to believe that the propagation would be at a rate 16 slow enough that it would not be problematic prior to 17 the time of the next inspection? Do you understand 18 where my --
19 ADMIN. JUDGE WARDWELL: And if it helps 20 you, because this was my question I was going to ask, 21 don't relate it to the bolts, relate it to anything.
22 What's to say that a crack that is just ready to be 23 initiated when you inspected it, but hadn't occurred 24 yet, or at least wasn't large enough to be visible, 25 and the next day, it became visible, you don't inspect NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5074 1 it now for ten more years, how do we know there won't 2 be some catastrophic type of result associated with 3 that crack over that ten year interval? What gives 4 you confidence that, that interval is sufficient 5 enough to still be within the range of it being able 6 to maintain its intended function?
7 DR. LOTT: I guess in response to that, I 8 would point out that in preparation for these 9 examinations, we're working with Entergy to develop 10 inspection acceptance criteria. Those inspection 11 acceptance criteria have built into them, again, how 12 large a crack would be allowable and that includes an 13 allowance for crack growth. So that would be starting 14 with a crack, that means there is an allowable size.
15 I would suggest to you that if a crack were to 16 initiate now, it would be less than that allowable 17 size because there's ten years of growth in the 18 acceptance criteria. So as long as --
19 ADMIN. JUDGE WARDWELL: But what is the 20 allowable size in your acceptance criteria? I thought 21 it as soon as a crack appears, you have to take some 22 -- that is --
23 DR. LOTT: Well, we need to --
24 ADMIN. JUDGE WARDWELL: -- that isn't an 25 acceptance criteria.
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5075 1 DR. LOTT: Well, that would trigger an 2 engineering evaluation. What I'm talking about is the 3 engineering evaluation, the basis for the engineering 4 evaluation.
5 ADMIN. JUDGE WARDWELL: Yes, but the ten 6 years is passed, you're not doing that evaluation, 7 you're not doing anything on it, but the crack is 8 already there and off and running. And what's to lead 9 you to believe that it will maintain its intended 10 function for ten years as this crack propagates?
11 DR. LOTT: Well, again, within the 12 evaluation, we started with a finite crack length and 13 allowed it to grow ten years and showed at the end of 14 ten years it would still be acceptable. That's part 15 of the engineering evaluation.
16 ADMIN. JUDGE WARDWELL: And you know that 17 for all the RVIs that they will still be able to 18 maintain their function as a crack initiates, assuming 19 that the criteria for evaluating it occurs the day 20 after you inspect it, when you didn't see anything, 21 and you do nothing during that period, do you have 22 enough data to comfort yourself that it won't reach 23 some -- it will still be able to maintain its intended 24 function ten years later?
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5076 1 Dolansky with Entergy. I don't think I can say right 2 now that I've reviewed and looked at that for every 3 component. I mean, I know that -- what Dr. Lott is 4 saying, Westinghouse is doing work for Entergy right 5 now coming up with acceptance criteria for the 6 inspection. I don't know for every single component 7 what that acceptance criteria is. I do remember for 8 some of the components where there is an acceptance 9 criteria, where there's an acceptable crack length, 10 that it says crack below this length are acceptable.
11 That means that they have done the calculation out to 12 ten more years before we would inspect it again and 13 determined that, that's acceptable. But I can't say 14 right now if -- I'd have to go back and -- I don't 15 think we have -- we don't have the final acceptance 16 criteria at this time. I'd have to go ask 17 Westinghouse if they've --
18 ADMIN. JUDGE WARDWELL: You've got 19 acceptance criteria in your AMP, don't you?
20 MR. DOLANSKY: The methodology and -- yes.
21 Acceptance criteria, but not the detailed plant 22 specific where we look at the plant specific loads, 23 the plant specific licensing basis, where they 24 actually did the engineering evaluation. You'd have 25 to ask --
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5077 1 ADMIN. JUDGE WARDWELL: Well, we've got 2 more questions, we'll get into the acceptance criteria 3 in more detail a little bit later. But I guess I'd 4 like to turn to Staff in regards to, what did you 5 review that comforts you that you could accept this 6 program knowing that there's a potential for a crack 7 to occur the day after an inspection was finished and 8 it wouldn't be looked at again for ten more years that 9 it would still maintain its intended function?
10 DR. HISER: This is Allen Hiser with the 11 NRC. I was not involved in the specific review of MRP 12 227 and their program is based on 227. In general 13 though -- and there's nothing in the SCR that really 14 describes the basis for acceptability, the ten year 15 reinspection interval. In general, I think there's an 16 expectation that given the knowledge that we have of 17 crack growth rates and things like that, that a flaw 18 that could initiate the day after the inspection as 19 you mentioned, that it would not have sufficient time 20 to propagate to where it could cause a failure of that 21 piece, maybe not even the assembly, but of that 22 individual piece, before the next inspection.
23 ADMIN. JUDGE WARDWELL: And there's data to 24 support that or is that just based on, because that's 25 what we've doing all along?
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5078 1 DR. HISER: I don't believe it was, that's 2 what we've been doing all along. I believe there was 3 more consideration to it. Fundamentally though, it 4 came, my guess is that it came down to engineering 5 judgment. There is nothing in the SCR that really 6 provides a roadmap to that, but I think given our 7 knowledge of crack growth rates, evaluations that have 8 been done for similar components, for example in BWR 9 internals, that, that was found to be a reasonable 10 reinspection interval to provide reasonable assurance 11 of the integrity of the RVI.
12 ADMIN. JUDGE WARDWELL: Some of these 13 inspections have been going on as part of your current 14 licensing basis, haven't these, for some of these 15 internals?
16 DR. HISER: That's correct.
17 ADMIN. JUDGE WARDWELL: And what's the 18 frequency for those?
19 DR. HISER: Those would be every ten years.
20 ADMIN. JUDGE WARDWELL: That's what I 21 thought. With the ten year interval, wouldn't it mean 22 that these inspections basically are going to occur 23 just once over the PEO?
24 DR. HISER: Well, they would occur twice.
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5079 1 53. So they would occur twice, every ten years during 2 the PEO.
3 ADMIN. JUDGE WARDWELL: Well, you've got a 4 20 year PEO period. That's going to be the end of the 5 --
6 DR. HISER: Right. So --
7 ADMIN. JUDGE WARDWELL: -- 53 is going to 8 be the end, you're going to shut down then. You're 9 going to inspect them as you shut down and tear it 10 down?
11 DR. HISER: Year 43 would be the first 12 inspection under this program. Year 53, 13 approximately, would be the second inspection. So it 14 would be two times during the PEO.
15 MR. DOLANSKY: This is Bob Dolansky with 16 Entergy. Just to clarify, the first inspection would 17 be within two outages of the beginning of the PEO.
18 ADMIN. JUDGE WARDWELL: Right.
19 MR. DOLANSKY: That's the first. And then 20 the second one --
21 ADMIN. JUDGE WARDWELL: Oh, that's what 22 you're -- I'm calling the first -- that's what I call 23 a baseline inspection. The first subsequent interval 24 inspection will occur --
25 MR. DOLANSKY: Ten years.
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5080 1 ADMIN. JUDGE WARDWELL: -- 53 and then, so 2 you have only have one interval inspection is what I'm 3 saying beyond this baseline inspection.
4 MR. DOLANSKY: Okay. We count the baseline 5 as the first. We say that we're doing the MRP 227-A 6 exam in Spring of 2016 at Indian Point 2.
7 ADMIN. JUDGE WARDWELL: That's your 8 baseline, correct?
9 MR. DOLANSKY: Yes, but we --
10 ADMIN. JUDGE WARDWELL: Right.
11 MR. DOLANSKY: We treat that as our first 12 227-A exam.
13 ADMIN. JUDGE WARDWELL: Fine. Semantics 14 and that's -- I understand the difference between our 15 discussion. Okay.
16 CHAIRMAN MCDADE: Okay. And I understand 17 the, I think, genesis for the ten year period of the 18 difficulties and what is involved in the inspection.
19 And trying to become sanguine that the ten year 20 inspection is adequate. And I was using the term 21 crack propagation, Dr. Hiser, you used the term crack 22 growth rates. Where do you look to determine the 23 anticipated crack growth rates for these kinds of 24 materials? To satisfy you that a ten year period is 25 adequate?
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5081 1 MR. POEHLER: This is Jeffrey Poehler of 2 the Staff. So MRP 227-A has guidance on crack growth 3 rates to be used for engineering evaluations and 4 that's in Chapter 6 of that report. Also, the 5 industry plans to use WCAP-17096, which is under 6 review by the Staff and we mentioned that yesterday.
7 But that provides methodologies for engineering 8 evaluations when degradation is found. And that 9 report gives guidance on which crack growth rates to 10 use. But they are referencing industry accepted crack 11 growth rates that have been developed for -- well, I 12 probably shouldn't say because it's proprietary at 13 this point.
14 CHAIRMAN MCDADE: Okay. But basically, as 15 I understand what you're saying, is that you look 16 right now to MRP 227, the crack growth rates that are 17 projected there. That your review suggests those are 18 valid and you're willing to rely on those crack growth 19 rates --
20 MR. POEHLER: Right.
21 CHAIRMAN MCDADE: -- and your review of the 22 validity of those crack growth rates in determining 23 that a ten year period is sufficient to ensure that 24 these reactor vessel internals will maintain their 25 intended function.
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5082 1 MR. POEHLER: The Staff believes or it's 2 our opinion that the crack growth rates that they're 3 recommending would be conservative for pressurized 4 water reactors, given the --
5 CHAIRMAN MCDADE: Would be or are?
6 MR. POEHLER: Are.
7 CHAIRMAN MCDADE: Okay. Okay, Dr. Lahey, 8 do you have some input here with regard to crack 9 growth rates and MRP 227 and the validity of the ten 10 year period?
11 DR. LAHEY: Yes. Your honor, we have the 12 same concern as you expressed in your questions. Two 13 inputs that I would give to this discussion is there's 14 been publications which indicate the baffle former 15 bolts, when you do ultrasound inspections, you're 16 unable to detect up to or below 30 percent through-17 crack. So that means you could already start out with 18 a significant weakened bolt at that point. And I want 19 to remind you that on top of all of this, this is an 20 inspection based program and I agree with the concerns 21 that you have raised, but on top of this, at any time 22 during this event, if you have some of these highly 23 degraded structures subject to shock loads, you can 24 fail them. That's the concern that I have. Because, 25 as you know, I'm focused, the bottom line is on the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5083 1 safety of the plant. And the safety of the plant 2 means to be me core coolability.
3 CHAIRMAN MCDADE: It means core --
4 DR. LAHEY: Coolability.
5 CHAIRMAN MCDADE: -- coolability?
6 Coolability, you're saying?
7 DR. LAHEY: Yes. Once you lose an intact 8 geometry, all bets are off as to core coolability.
9 And by far the most vulnerable reactor vessel 10 internals are these baffle bolts.
11 CHAIRMAN MCDADE: Which of course Entergy 12 and the Staff represent can have a 50 percent failure 13 rate without impacting the integrity.
14 DR. LAHEY: Well, my understanding of how 15 that conclusion was drawn was really based on the kind 16 of loads you get during steady state operation and the 17 redundancy to hold them in place. But during accident 18 loads, if you have significant loads, you can unzip 19 the rest of your bolts. That's what I'm worried 20 about.
21 CHAIRMAN MCDADE: Okay.
22 MR. STROSNIDER: This is Jack Strosnider 23 for Entergy. I just should comment that those bolting 24 patterns that were analyzed in the WCAP, and I don't 25 remember the exact number right now, that they're NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5084 1 based on accident loads.
2 DR. HISER: This is Allen Hiser with the 3 Staff.
4 ADMIN. JUDGE WARDWELL: What was the --
5 DR. HISER: I can fill in. Entergy 654 and 6 655 are exhibits that are, at least in one case, is an 7 NRC approved report that looks at the bolting patterns 8 under accident loads.
9 ADMIN. JUDGE WARDWELL: Okay. Well, we'll 10 talk about baffle former bolts in more detail this 11 afternoon too. But before that, I do want to ask the 12 Staff in regards to what's their understanding of the 13 current acceptance criteria that's in the AMP as it 14 stands now?
15 DR. HISER: This is Allen Hiser with the 16 Staff. The acceptance criteria are provided in the 17 RVI Inspection Plan for Indian Point. I think there 18 maybe was some confusion with some of the discussion 19 earlier. The acceptance criteria that are under 20 development by Westinghouse would be an engineering 21 analysis option under corrective actions. So it's --
22 maybe if we say inspection acceptance criteria are in 23 the Inspection Plan and they are definitive.
24 ADMIN. JUDGE WARDWELL: And what do they 25 state there?
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5085 1 DR. HISER: I think in general for 2 cracking, it's no cracking. Any identified cracking 3 is subject to corrective actions.
4 ADMIN. JUDGE WARDWELL: Thank you.
5 CHAIRMAN MCDADE: Okay. Let me -- Dr.
6 Hiser, mentioning that any identifiable cracks, we've 7 talked quite a bit over this morning and yesterday 8 about microcracks. The existence of microcracks that 9 could be present even during an inspection, but not 10 identified. Does the possibility of those microcracks 11 affect your consideration of this, of the frequency of 12 inspection?
13 DR. HISER: No, I don't believe so.
14 Because a microcrack would be subsumed under the 15 inspection -- or depending on the size of the 16 microcrack. I mean, microcracks, if they're below the 17 inspectability limit of the NDE method would not --
18 clearly you would not be able to detect those. And I 19 don't believe that microcracks would have a 20 significant impact on the integrity of the RVI 21 components, sort of as a starting point. So with the 22 analyses --
23 CHAIRMAN MCDADE: But isn't the microcrack 24 basically the crack initiation and then you have the 25 propagation. So the question is, being able to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5086 1 identify where you are on that spectrum at the time of 2 the inspection. The clearer picture you have, the 3 better way to follow it. Now, you made reference to 4 the crack growth rates in MRP 227, does that take into 5 consideration that at the time of the inspection there 6 may well be significant, and perhaps maybe not use the 7 word significant, but microcracks that are approaching 8 visibility, but not yet visible?
9 DR. HISER: Yes. This is Allen Hiser of 10 the Staff. When the report was reviewed for 11 acceptability, my expectation is that the 12 consideration was along the lines of what kinds of 13 flaws could be missed by the inspection given the 14 knowledge that we have of crack growth rates. Was it 15 likely that there would be a challenge to 16 functionality at the end of the ten year reinspection 17 interval? Based on that analysis, be it -- and my 18 guess is this may have been an engineering judgment 19 based that there was not thought to be a significant 20 concern. There was reasonable assurance that the RVI 21 would still maintain their functionality.
22 CHAIRMAN MCDADE: Okay. Thank you, Dr.
23 Hiser.
24 ADMIN. JUDGE WARDWELL: Entergy's 25 testimony, 616, Answer 209, Page 140. The results of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5087 1 the IPEC inspections will be available to the NRC 2 Staff for onsite inspection. Are these results also 3 publically available?
4 MR. DOLANSKY: This is Bob Dolansky for 5 Entergy. These inspection results are put into a 6 report and supplied to EPRI for the whole industry and 7 EPRI rolls that up. I don't know if that's publically 8 available. I don't think it's publically available, 9 but it's available to the industry for sure.
10 ADMIN. JUDGE WARDWELL: So publically it's 11 not available?
12 MR. DOLANSKY: I don't believe so.
13 ADMIN. JUDGE WARDWELL: Okay, thank you.
14 MR. POEHLER: And, your honor, I would just 15 like to add to that. And that report is also 16 submitted to the NRC Staff for our information. So, 17 that's in the MRP 227 implementing process that they 18 will do that. So we will get a chance to review what 19 the operating experience has been and if there's been 20 any trends in failures, then we'll know about it.
21 ADMIN. JUDGE WARDWELL: And is that 22 available to the public or is it considered 23 proprietary information?
24 MR. POEHLER: I don't know the answer to 25 that offhand, I can check.
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5088 1 ADMIN. JUDGE WARDWELL: Do that, would you 2 please? And I think now would be a good time to take 3 a break.
4 CHAIRMAN MCDADE: Okay. Would anybody need 5 more than ten minutes? Okay, it's now -- why don't we 6 come back at 10:40.
7 (Whereupon, the above-entitled matter went 8 off the record at 10:26 a.m. and resumed at 10:44 9 a.m.)
10 CHAIRMAN MCDADE: We're back on the record.
11 ADMIN. JUDGE WARDWELL: Let's turn to some 12 inspection actions now. NRC's Exhibit 197, testimony, 13 Answer 80, Page 51, says inspection techniques include 14 ultrasonic UT testing, EVT1 enhanced visual 15 examinations, and VT3 visual examinations. And so, 16 considering it's NRC's testimony I'm referring to, 17 I'll let them answer. Could you explain each type of 18 test and its applicability for the various reactor 19 vessel internals?
20 MR. POEHLER: Okay. This is Jeffrey 21 Poehler of the Staff. So, just to clarify the 22 question, you want us to explain the applicability of 23 each type of inspection technique?
24 ADMIN. JUDGE WARDWELL: I want you to 25 explain what's a UT test, what's an EVT1 test, and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5089 1 what's the VT3 test.
2 MRP: So UT test is an ultrasonic 3 examination. And it's basically using sound to detect 4 cracks in a material. Do you want a detailed 5 explanation or --
6 ADMIN. JUDGE WARDWELL: How do they show 7 up? How does a crack show up? So, you put an --
8 MR. POEHLER: So you have a transducer that 9 generates ultrasonic sound. It has to be in contact 10 with the material and it puts sound into the material 11 and you get echos back from the material basically 12 that are detected by either the same transducer or a 13 separate transducer. And those are processed 14 electronically such that you get a signal. So if you 15 have a discontinuity, like let's say in a bolt, you 16 have a partially cracked bolt, that's going to reflect 17 the sound back and be detected and it'll be processed 18 by the electronics such that you can determine the 19 location of that discontinuity.
20 And with certain ultrasonic techniques, 21 you can create images, you can image the 22 discontinuity. So ultrasonic is used throughout the 23 nuclear industry for piping weld exams to detect 24 cracking and also in vessels, in large structural 25 welds. So it's a very established technique for welds NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5090 1 and it's also for bolting, such as baffle former 2 bolts, that's been used -- it was first used back in 3 the 1990s when baffle former bolt cracking was first 4 detected.
5 ADMIN. JUDGE WARDWELL: So that's a type of 6 test that's used for the baffle former bolts?
7 MR. POEHLER: It's a pretty well 8 established technique for bolts.
9 ADMIN. JUDGE WARDWELL: So is it used for 10 the clevis bolts that we'll talk about later also?
11 MR. POEHLER: It is not specified for the 12 clevis bolts at this time. It could be, but it's not 13 --
14 ADMIN. JUDGE WARDWELL: Okay.
15 MR. POEHLER: -- has been determined not to 16 be necessary.
17 ADMIN. JUDGE WARDWELL: So is the baffle 18 former bolts the only one it's used for in regards to 19 reactor vessel internals?
20 MR. POEHLER: The only component in a 21 Westinghouse design reactor internals that it's 22 currently used for is the baffle former bolts.
23 ADMIN. JUDGE WARDWELL: Okay.
24 MR. POEHLER: You also asked about visual 25 examination.
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5091 1 ADMIN. JUDGE WARDWELL: How about the 2 enhanced visual examination, EVT1, what is that 3 composed of?
4 MR. POEHLER: Well, so enhanced --
5 ADMIN. JUDGE WARDWELL: Well, let me ask 6 one other question about the UT. So is this a sensor 7 that you put in at the time you do the inspection or 8 are these permanently mounted so you can turn it on?
9 Or how are the mechanics of this, the logistics of 10 this achieved?
11 MR. POEHLER: Right, it's a sensor that has 12 to be put in. It's put in for the time of the 13 inspection and there's special tooling to access the 14 bolting. And so it's not permanent.
15 ADMIN. JUDGE WARDWELL: Okay, thank you.
16 CHAIRMAN MCDADE: What percentage of the 17 baffle former bolts are accessible to the UT 18 inspection?
19 MR. POEHLER: For the UT inspection, 20 essentially 100 percent of them are accessible. There 21 are minimum coverage requirements in MRP 227-A, so you 22 can credit 100 percent -- you can take credit for an 23 examination if you exam 75 percent of a population in 24 general. And that would include both the accessible 25 and inaccessible members of the population. But, to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5092 1 my knowledge, I don't think there are any major 2 obstructions that would prevent you from accessing all 3 the baffle former bolts.
4 CHAIRMAN MCDADE: To Entergy, are 100 5 percent of the baffle former bolts accessible to the 6 ultrasound inspection?
7 MR. DOLANSKY: Yes. This is Bob Dolansky 8 with Entergy. We believe 100 percent will be 9 accessible. We won't know for sure until we actually 10 get out there and do the exam, but based on all of our 11 drawing reviews and so forth, we expect to get 100 12 percent of all the bolts and we expect, there's 832 of 13 them, we expect all 832 to be accessible.
14 CHAIRMAN MCDADE: So unless somebody 15 changed the design since you last looked at it?
16 MR. DOLANSKY: Well, there could -- in 17 theory, there could be -- let me give a little bit of 18 background. The technique to perform this inspection 19 is a submarine that goes down underwater, it's done 20 underwater. It basically docks up against the baffle 21 plate and there's a special head that goes in and 22 inspects the bolt. So although all the bolts should 23 be accessible, we're not 100 percent sure we can reach 24 every one with this tooling technique. So that's why 25 there's a possibility that we wouldn't get every one.
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5093 1 But other Westinghouse plants have got 100 percent, we 2 expect to get 100 percent, but there could be some 3 interference that would limit us, but only a very 4 small number.
5 CHAIRMAN MCDADE: Okay. Thank you.
6 ADMIN. JUDGE WARDWELL: Enhanced visual 7 examinations?
8 MR. POEHLER: Yes. This is Jeffery Poehler 9 of the Staff. So enhanced visual examinations are 10 specified for examining welds for cracking. Or 11 anytime you're looking for cracking specifically in 12 MRP 227-A. For example, the core valve girth welds in 13 a Westinghouse design, reactor internals, you use 14 enhanced visual testing. And that's a visual 15 examination that has a fairly stringent detection or 16 resolution requirement. So you have to be able to --
17 the way that they test this in situ is that they have 18 to be able to identify a character, like a letter, 19 letter C or A or O, and they have to identify that 20 letter and the letter has a 0.04 inch size on the 21 card. So it's a pretty small letter. So that's how 22 they determine the resolution is adequate in the 23 environment that you're going to do the testing.
24 ADMIN. JUDGE WARDWELL: And what's this 25 card?
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5094 1 MR. POEHLER: It's like a plaque --
2 ADMIN. JUDGE WARDWELL: A plaque, a little 3 plaque?
4 MR. POEHLER: Yes, it's a --
5 ADMIN. JUDGE WARDWELL: Okay.
6 MR. POEHLER: It's something that -- yes.
7 ADMIN. JUDGE WARDWELL: Is it on each 8 internal or is it just a test plaque for you to start 9 off to verify that you've got enhanced --
10 MR. POEHLER: It's just a test plaque. You 11 could -- analogous to calibration standard.
12 ADMIN. JUDGE WARDWELL: Calibration coupon 13 --
14 MR. POEHLER: Right.
15 ADMIN. JUDGE WARDWELL: -- shall we say?
16 MR. POEHLER: Right. And so then you have 17 to -- so that's basically how that's qualified. And 18 you would do that before you start the examination.
19 ADMIN. JUDGE WARDWELL: This is a camera 20 that you put --
21 MR. POEHLER: Yes, remote --
22 ADMIN. JUDGE WARDWELL: -- in there, it's 23 not someone's eyeball that you're calibrating.
24 MR. POEHLER: I think they used to -- the 25 submarine type delivery system, or I think for the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5095 1 core valve girth weld, they have special tooling to go 2 up under the thermal shield to deliver the camera. So 3 those are used for when you're specifically looking 4 for cracks such as stress corrosion cracks, IASCC, 5 irradiation assisted stress corrosion cracks, in welds 6 or other components. Other than bolting, it would not 7 be used for bolting because of the location. Where 8 the cracks are would be at the head, the shank area in 9 the bolt, which you're not going to be able to look at 10 side-on. So that's why ultrasonic is used for those 11 bolts. So you also asked about --
12 ADMIN. JUDGE WARDWELL: VT3, visual 13 examination.
14 MR. POEHLER: So VT3 is another visual 15 examination technique. The main difference between 16 VT3 and EVT1 is that it is -- VT3 has a slightly lower 17 resolution requirement and it's used for general 18 mechanical and structural conditions. So you're 19 looking for gross failure, like broken components, 20 broken bolts, and other distortion in structures such 21 as it's used for the baffle former assembly to look 22 for effects of void swelling. But the VT3 visual is 23 only specified in MRP 227-A when it's for either a 24 redundant population of components or components that 25 have been changed to be highly flaw tolerant, such NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5096 1 that they could tolerate cracks. So that's VT3.
2 ADMIN. JUDGE WARDWELL: And VT3 implies 3 there's a VT1 and a VT2 and where did they go and are 4 they ever used at Indian Point?
5 MR. POEHLER: Yes. VT1 is not used for, 6 and Entergy may correct me, but I don't believe it's 7 used for any of the Westinghouse RVI exams. But in 8 some other designs it's used for looking for gaps.
9 But VT1 is also a more -- would have a higher 10 resolution requirement than a VT3. But it's an 11 examination that's called out in ASME code, so it's 12 defined in there. But it's not used very much for the 13 internals. And VT2 is another type of visual 14 examination where you specifically look for leakage of 15 pressure boundary components. So that's not 16 applicable and they don't use that for reactor 17 internals.
18 ADMIN. JUDGE WARDWELL: In New York State's 19 testimony on 482 at 62, lines 3 through 8, Dr. Lahey 20 criticizes the use of VT3 in visual inspections as 21 inadequate for use in inspections for cracking, 22 stating that there are significant shortcomings of 23 this technique to detect material cracking, 24 degradation, or wear prior to failure, as illustrated 25 by the visual detection of only seven out of 29 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5097 1 fractured clevis insert bolts at the Westinghouse PWR 2 in 2010. Why I don't I turn to -- well, I'll stay 3 with you. So, why do we have any confidence in VT3 4 for any components and what components beside the 5 clevis bolts is this technique used and what's it's 6 track record for the other components?
7 MR. POEHLER: Well, for example, it's used 8 for the baffle former assembly, for the general 9 examination for the assembly for distortion due to 10 void swelling. So that wouldn't be something you 11 would use -- it's the most appropriate for looking for 12 that type of aging deformation.
13 CHAIRMAN MCDADE: Okay. Well, let me jump 14 in here a second and to Entergy, in your Exhibit 616 15 at Page 87, your Table 1, you list the various items 16 to be inspected and how you're going to be inspecting 17 them. Is there any difference in the, for lack of a 18 better phrase, degree of difficulty in the EVT1 and 19 the VT3? I mean, it appears that the EVT1 gives you 20 greater resolution, more information. Is there any 21 reason why you use the VT3 for certain items, like 22 baffle edge bolts, as opposed to the EVT1?
23 MR. DOLANSKY: This is Bob Dolansky with 24 Entergy. An EVT1 is looking for cracking. A VT3 is 25 not looking for cracking. So, for example, the baffle NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5098 1 edge bolts, if the cracking is down in the bolt, a 2 visual inspection is not going to see it. So if you 3 do an EVT1 on the baffle edge bolts, it's not going to 4 give you any more information than a VT3 would and it 5 is in fact a more difficult exam to do because you 6 have -- the character requirements are more stringent 7 and --
8 CHAIRMAN MCDADE: I'm sorry, I just didn't 9 hear. The what requirements?
10 MR. DOLANSKY: The character card 11 requirements are more stringent and you have to 12 inspect at a certain speed. You typically use video 13 enhancement to do that inspection. So, for the baffle 14 edge bolts, since you can't see the area where they 15 would be cracking anyway, an EVT1 doesn't buy you 16 anything. What a VT3 tells you about baffle edge 17 bolts is that if you look at all the edge bolts, if 18 you saw that there was -- the two plates where they go 19 together are shifted or moved, a VT3 is very good for 20 that. You're not --
21 ADMIN. JUDGE WARDWELL: But I thought you 22 used the UT for baffle bolts.
23 MR. DOLANSKY: For baffle former bolts, a 24 UT is used.
25 ADMIN. JUDGE WARDWELL: Yes.
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5099 1 MR. DOLANSKY: But a baffle edge bolt --
2 ADMIN. JUDGE WARDWELL: Oh, okay, sorry.
3 MR. DOLANSKY: -- it's not.
4 ADMIN. JUDGE WARDWELL: Sorry.
5 MR. DOLANSKY: A VT3 is used for the baffle 6 edge bolts. Did that answer your questions, your 7 honor --
8 ADMIN. JUDGE WARDWELL: Yes.
9 MR. DOLANSKY: -- about the differences?
10 MR. AZEVEDO: Your honor, this is Nelson 11 Azevedo. Maybe I can add a little bit to that. A VT1 12 is done up close. So, for example, if you're looking 13 for deformation of a component, see if a component is 14 bent, if you're up real close to a component, you may 15 not be able to see if the component is bent. So you 16 actually, if you step back, you can be further from 17 the component, in a lot of cases, you actually get a 18 better assessment of what the component is 19 experiencing rather than be within a couple inches of 20 the component. So that's another reason why sometimes 21 VT3 are used versus VT1.
22 CHAIRMAN MCDADE: Sort of why I take my 23 glasses off to thread a needle?
24 MR. AZEVEDO: If that's the example you 25 want to use.
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5100 1 CHAIRMAN MCDADE: Do you thread needles?
2 (Laughter.)
3 MR. AZEVEDO: Occasionally.
4 CHAIRMAN MCDADE: Sorry. Staff, do you 5 thread needles?
6 (Laughter.)
7 ADMIN. JUDGE WARDWELL: Back to the clevis 8 insert bolts, and I'll direct this to, I guess I'll 9 stay with Entergy, it doesn't seem like the seven out 10 of 29 fractured clevis insert bolts that were detected 11 by this process is a very high percentage of success.
12 Well, let me rephrase that, you look confused. New 13 York State claimed that only seven out of 29 fractured 14 clevis insert bolts were detected in 2010 at a 15 Westinghouse PWR. Are you aware of that and, if so, 16 why is VT3 a successful inspection technique for seven 17 out of 29?
18 MR. DOLANSKY: This is Bob Dolansky with 19 Entergy. We definitely were aware of that. There was 20 a Technical Bulletin that was issued as a result of 21 the clevis insert bolting. I would say that the VT3 22 is what initially found it at the other plant. I 23 think it was sufficient, they did find it, they didn't 24 find every one, but it showed them that there was 25 something going on and then they did additional exams NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5101 1 to determine that. To me, with the clevis insert 2 bolts, and Dr. Lott can step in and help me with this, 3 but the clevis insert bolts don't have any safety 4 function. So there's no -- the fact that they cracked 5 and that they detected that with VT3 shows that, as 6 part of the Section 11 program, shows that, that is in 7 fact working in my estimation. Do you want to add 8 anything about that?
9 DR. LOTT: I don't think we ever believed 10 or contended that the clevis insert bolts, the VT3 11 inspection would identify failed clevis insert bolts.
12 What we were worried about, as Mr. Dolansky said, was 13 that the location and the securing of the clevis in 14 the log on the reactor pressure vessel -- and the 15 Technical Bulletin that he referred to, which is 16 Westinghouse Technical Bulletin 14-5, is ENT Exhibit 17 656, just for the record.
18 And we looked very closely at the clevis 19 insert bolts in that case and made additional 20 recommendations about inspections that would help us 21 determine what we thought was the key feature, which 22 was the clevis being still in place. Once the reactor 23 is loaded, the lower core plate is in, the clevis is 24 effectively locked in place and the safety function of 25 the clevis, as long as it's there when you start up, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5102 1 there's no place for it to go when you're basically 2 running the reactor. We can go through this in more 3 detail if you want to, but we'd probably have to pull 4 up some of the diagrams.
5 ADMIN. JUDGE WARDWELL: Yes. I think I at 6 least have the picture of that. How did you determine 7 that eventually there was 29 -- or how did 8 Westinghouse determine that there were 29 fractured 9 bolts, not just the seven that were detected by the 10 VT3?
11 DR. LOTT: Well, Westinghouse's involvement 12 in this program, we had advised the utility that, when 13 they observed the first, or you told me the number 14 seven, broken bolts, advised them that they should 15 think about replacing the bolts, not because of a 16 safety concern, but our concern was that if that were 17 to become dislodged in the refueling cycle, it would 18 be extremely expensive and difficult to put it back in 19 place, it's a precision fit part. So therefore --
20 ADMIN. JUDGE WARDWELL: This is the clevis 21 itself, the insert itself?
22 DR. LOTT: The clevis itself. The clevis 23 insert is specifically machined to match the keys in 24 the core barrel.
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5103 1 it correctly, these bolts merely hold that in place 2 until that superstructure or whatever you call it, the 3 --
4 DR. LOTT: The core barrel.
5 ADMIN. JUDGE WARDWELL: -- the core barrel 6 coming in and --
7 DR. LOTT: Yes, once the core barrel is 8 engaged --
9 ADMIN. JUDGE WARDWELL: -- resting on that, 10 which --
11 DR. LOTT: -- in the key, it's effectively 12 restrained by multiple factors that we can go into if 13 you want to. And so, therefore, we determined it 14 wasn't a safety concern, but that it might be in the 15 best interest of the utility to replace those bolts 16 because if it were to come loose, it would be 17 incredibly difficult to replace.
18 ADMIN. JUDGE WARDWELL: Are either the --
19 any of these bolts, whether it's the clevis or the 20 baffle edge or the baffle former bolts, if they fail, 21 do they become dislodged and have a potential to 22 impact the geometry, if you will, and the coolability 23 of the core itself? Or a function of any of the other 24 vessel internals?
25 DR. LOTT: Well, one of the things and one NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5104 1 of the advantages of the VT3 examination is that it 2 allows you to determine whether the lock bars, which 3 hold the heads of all the bolts in place, lock bars 4 that are locked across the heads, are in place. And 5 as long as those lock bars are in place, there's no 6 way for the bolt to escape. In particular, the clevis 7 insert bolts are, again, geometrically constrained.
8 Once they're in place, there's no way for that head to 9 escape from this very small tight dimensions of the 10 object.
11 ADMIN. JUDGE WARDWELL: Well, couldn't the 12 bolt itself crack and fall away from the head and go 13 down below?
14 DR. LOTT: Well, it's threaded in and it 15 can't -- there's nowhere for it to go except for out 16 past the head.
17 ADMIN. JUDGE WARDWELL: Okay, thank you.
18 While we're on it, Dr. Lahey, would you agree with the 19 statements that were made in regards to the clevis 20 bolt inserts and do you have comments about the seven 21 out of 29?
22 DR. LAHEY: Well, our concern was it was 23 not a very effective non-destructive testing technique 24 and the accuracy of it wasn't very effective. I'm not 25 very concerned about the safety aspects of the clevis NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5105 1 bolts, but the technique itself had some deficiencies.
2 ADMIN. JUDGE WARDWELL: Thank you.
3 CHAIRMAN MCDADE: Okay. If I could, just 4 to clarify. We've talked about these clevis bolts and 5 not having a safety consideration. If you could 6 elaborate for me, as I understand from that, Dr. Lott, 7 you referred to the Westinghouse Technical Bulletin, 8 which was Entergy Exhibit 656, and it talks about 9 maintenance risk due to the complexity and cost of 10 repair and the required level of contingency planning.
11 That inability to remove the core barrel or need to 12 replace the insert to reestablish customized design 13 gaps. Could you explain to me, what really --
14 elaborate a bit on the implications of a failure of 15 these clevis bolts?
16 DR. LOTT: Sure. The clevis inserts 17 themselves are actually, they're positioned -- they 18 locate these keys and they're relatively high 19 tolerance components, such that if it -- when the 20 reactor is actually built, they're precision machined 21 to match. And then they're shrunk fit into the lugs 22 in the component, so they're frozen, put in -- so 23 they're a tight fit. That's one of the reason they 24 don't come out is because they're shrunk fit in place.
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5106 1 in the plant and relocate it into the exact location 2 where it was in the first place if the bolts were to 3 come loose and you wanted to even put the same one 4 back in, much less replacement, is a very difficult 5 job. So if we were to lose it, it would be hard to 6 replace. Is that --
7 MR. DOLANSKY: Dr. Lott, can I just add 8 something?
9 CHAIRMAN MCDADE: Hard to or impossible and 10 then --
11 DR. LOTT: I don't think I would say 12 impossible to replace, but, again, this was a 13 maintenance concern, not a safety concern. So it was 14 really a matter of what did it make sense for the 15 utility to do at that point?
16 MR. DOLANSKY: This is Bob Dolansky with 17 Entergy. Just to try to clarify a little bit, when 18 you refuel the reactor vessel, basically you do 19 everything remotely, so you're not unbolting things.
20 Basically, everything slides together and slides 21 apart. So when you're removing the core barrel, the 22 clevis insert basically is like a key and there's a 23 key on the core barrel that just slides in there.
24 Once it's in there, there's no safety function, it has 25 no safety function.
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5107 1 But if I as the owner am planning to 2 refuel the reactor vessel and remove the core barrel 3 and that clevis insert moves, then I cannot get that 4 back in. That's very expensive for me to then try to 5 get that fixed, just so I can get it back in, it has 6 no safety function, but just to be able to put it back 7 together. So that's why it's a commercial issue and 8 not a safety issue. And that's why what Dr. Lott is 9 saying is, if three out of six bolts or four out of 10 six bolts failed, but two are still holding it, I can 11 get those four that are failed out and put new ones in 12 without that thing shifting. Once it shifts, to try 13 to get it lined up exactly again, underwater, would be 14 very, very difficult. It could be done, absolutely 15 could be done and would be done, but it just would be 16 much more difficult. So for us, it makes much more 17 sense to try to get those things replaced before it 18 causes us a commercial issue. Does that help?
19 CHAIRMAN MCDADE: Yes. So basically what 20 you're saying is, it's not an inability to remove the 21 core barrel, it just makes it a much -- the degree of 22 difficulty increases significantly?
23 MR. DOLANSKY: Getting it out or putting it 24 back together. Putting it back together, if it's 25 cocked, it wouldn't slide, it's a very close machine NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5108 1 tolerance. The core barrel may not slide back into 2 the key if it were totally -- if it had moved.
3 CHAIRMAN MCDADE: Okay. Thank you.
4 ADMIN. JUDGE WARDWELL: Entergy's 5 testimony, 616, Answer 188, Page 124, states that the 6 basis for the adequacy of these inspection techniques 7 is described in the companion document to MRP 227-A, 8 which is MRP 228. This document describes the 9 standards to be met by each specific examination 10 method. The standards in MRP 228 reflect the latest 11 information and regulatory documents, such as 12 NUREG/CR-6943, which addresses visual examinations, 13 including remote visual examinations, and describes 14 the characteristics of flaws to be detected in nuclear 15 reactor components. In particular, such critical 16 characteristics as the crack opening displacement.
17 And I guess I just want to verify, is 228 an exhibit 18 in this proceeding, Entergy, as far as you know?
19 DR. LOTT: I believe it is.
20 MR. DOLANSKY: I believe it is.
21 ADMIN. JUDGE WARDWELL: Okay, thank you.
22 MR. KUYLER: Your honor? MRP 228 is 23 Exhibit Entergy 645.
24 ADMIN. JUDGE WARDWELL: Thank you. I guess 25 I'll ask Dr. Lahey, have you provided any evidence NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5109 1 challenging these details on the basis for the 2 development of Entergy's Inspection Plan or that 3 Entergy's Inspection Plan does not meet all the 4 requirements of MRP 227?
5 DR. LAHEY: Are you talking about other 6 than the inability to detect the clevis bolts?
7 ADMIN. JUDGE WARDWELL: Right. Just --
8 DR. LAHEY: I would have to look back, I 9 don't know off hand.
10 ADMIN. JUDGE WARDWELL: Okay. Thank you.
11 CHAIRMAN MCDADE: Excuse me one second, and 12 perhaps I just got confused. Did you say, MRP 227?
13 ADMIN. JUDGE WARDWELL: Yes.
14 CHAIRMAN MCDADE: Okay. Because 645 is 15 EPRI 228.
16 ADMIN. JUDGE WARDWELL: Right. That was my 17 first question.
18 CHAIRMAN MCDADE: Thank you. I'm just --
19 thank you.
20 ADMIN. JUDGE WARDWELL: Okay. We switched 21 MRPs on you. NRC 197, testimony, Answer 183, Page 22 105, states that we disagree with Mr. Lahey's concern 23 regarding synergistic effects because, one, primary 24 components are to be inspected under Entergy's program 25 are those components which are most likely to be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5110 1 effected by synergistic effects, if they exist. Two, 2 that Entergy's AMP inspects for and will detect 3 cracking whether a single aging mechanism or multiple 4 or synergistic aging mechanisms contribute to 5 cracking.
6 Three, embrittlement alone will not cause 7 failure without the presence of a crack and the 8 inspections performed by Entergy's AMP are sufficient 9 to detect cracking. Four, Dr. Lahey has not 10 identified any tests or operating experience which 11 demonstrates that synergistic effects are significant 12 for PWR RVIs and existing laboratory test data on 13 synergistic effects is inconclusive. And, five, the 14 industry reactor vessels internal program in which 15 Entergy is participating is a living program which 16 shares operating experience among all PWR licensees 17 and, thus, any occurrence of an unexplained or 18 accelerated degradation due to synergistic effects 19 will be identified and adjustments to the industry 20 guidance and the Entergy AMP will be made to ensure 21 continued integrity of the RVI across the fleet.
22 I think with this statement, I'd like to 23 start off fixing one other point again. And this is 24 this statement that embrittlement will not occur 25 without a crack. And I'll address this to Entergy.
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5111 1 Does this statement mean that embrittlement won't have 2 any affect until cracking is caused by some other 3 mechanism or is this a statement that says that 4 embrittlement won't alone cause failure without a 5 crack? Meaning that embrittlement can cause a crack, 6 but it won't result in failure unless there is a crack 7 demonstrated from the embrittlement?
8 DR. LOTT: This is Randy Lott for Entergy.
9 The statement, I believe, is saying that if a material 10 is embrittled, that fact alone will not result in the 11 failure of the component, it needs to be combined with 12 a crack and a load that would challenge the stability 13 of the component.
14 ADMIN. JUDGE WARDWELL: With embrittlement, 15 can embrittlement eventually cause a crack in and of 16 itself?
17 DR. LOTT: No. It's not identified as one 18 of the crack causing mechanisms in the component.
19 ADMIN. JUDGE WARDWELL: What is the effect 20 of embrittlement then?
21 DR. LOTT: What is the effect? The change 22 in the -- for me, the discussion of embrittlement is 23 a discussion of the change in properties that happen 24 when the material is irradiated. And we've listed 25 them several times before, it's the increase in yield NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5112 1 stress, the decrease in ductility, the decrease in 2 fracture toughness, those combined effects.
3 ADMIN. JUDGE WARDWELL: Is the resulting 4 effect?
5 DR. LOTT: Those are -- yes. If I had a 6 test specimen, those are measurable, objective things 7 I could measure.
8 ADMIN. JUDGE WARDWELL: So regardless --
9 DR. LOTT: I can't do that test non-10 destructively.
11 ADMIN. JUDGE WARDWELL: So regardless of 12 how long or how much fluence there is at a location, 13 a piece of metal there will not exhibit a crack due to 14 just embrittlement causes? Let's say you have a test 15 coupon and --
16 DR. LOTT: If I had a test coupon and I 17 were to just -- we certainly irradiated lots of 18 specimens to the fluences that we've talked about 19 lots. We've irradiated specimens to the fluences that 20 we've talked about here, the 60, 70 dpa, and they do 21 not just spontaneously crack.
22 ADMIN. JUDGE WARDWELL: Okay.
23 DR. HISER: And, your honor, this is Allen 24 Hiser of the Staff. I just want to make sure, the 25 Staff did not imply that embrittlement can cause NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5113 1 cracking. The statement that you read just says 2 embrittlement cannot cause failure unless there is a 3 crack.
4 ADMIN. JUDGE WARDWELL: Correct.
5 DR. HISER: So the Staff did not imply 6 embrittlement causes cracks.
7 ADMIN. JUDGE WARDWELL: Right.
8 DR. HISER: Okay. Just to make sure.
9 ADMIN. JUDGE WARDWELL: Yes. I didn't mean 10 to imply that you implied that.
11 DR. HISER: Okay.
12 CHAIRMAN MCDADE: Okay. And --
13 DR. HISER: Maybe I inferred that from you.
14 CHAIRMAN MCDADE: And either Dr. Hiser or 15 Dr. Lahey, let me -- and, again, I want to make sure 16 I understand this before we go off to try to make a 17 decision here. Embrittlement is a mechanism, an 18 effect of embrittlement is a decrease in ductility.
19 And it's Dr. Lahey's position that, that decrease in 20 ductility makes the item more susceptible to cracking.
21 Dr. Lahey, is that correct?
22 DR. LAHEY: The effect of embrittlement is 23 -- I mean, fundamentally, the difference is no crack 24 no problem. I mean, there are people who believe no 25 crack no problem, to me that was the quotable quote NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5114 1 from yesterday. I don't believe that. I believe with 2 or without a crack, if you have an embrittled 3 structure and you hit it with a significant impulsive 4 load, you can fail the structure. And that's the 5 concern.
6 CHAIRMAN MCDADE: But that's -- again, the 7 concept there is the ductility has decreased because 8 of the embrittlement and because of that decrease in 9 ductility, the item is more susceptible to cracking 10 under stress with a load?
11 DR. LAHEY: Well, the fracture toughness, 12 how the material can retard crack growth, is 13 decreased. So cracks can grow faster if it's 14 embrittled. So we don't disagree on that. What we 15 disagree on is, do you need a surface crack or not?
16 Is the material weakened or subject to failure if it's 17 embrittled? That's really the crux of it. One thing 18 that --
19 CHAIRMAN MCDADE: Also, and again maybe 20 we'll get into this later, but you used the word 21 failure and there's an inspection failure and then 22 there's a failure of intended use. The failure of 23 intended use involves cracking. But as I understand 24 it, there would be cracking before you had a 25 catastrophic failure, where it would no longer serve NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5115 1 its intended use. I'm just trying to follow in my 2 mind the mechanism along. And bear with me here 3 because I'm plowing new ground for me, I realize that 4 this is old ground for all of you. But I had thought 5 from yesterday that ductility decrease made the item, 6 a bolt for example, more susceptible. I thought that 7 the fracture toughness actually increased with the 8 embrittlement. Do you disagree? Did I --
9 DR. LAHEY: It depends, what do you mean by 10 fracture toughness? If you mean the ability of a 11 crack to propagate, which is the way I think it's 12 being used, then it's not correct. I mean, there is 13 another mechanism in which the strength of the 14 material is increased by irradiation induced 15 embrittlement.
16 CHAIRMAN MCDADE: Okay. And I guess what 17 I'm confused then is the sort of the inter-reaction of 18 these terms and what these terms represent. So if, 19 and again, bear with me here, the strength increases, 20 I'm not really sure what you mean by strength, I had 21 taken strength to be synonymous with fracture 22 toughness. And that the ductility, basically the 23 ability to bend and come back is something entirely 24 different. And although the ductility was decreasing 25 with embrittlement, that there was a strength increase NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5116 1 and there was not a decrease in fracture toughness.
2 Dr. Hiser, is that the Staff's position? Or please 3 educate me here.
4 DR. HISER: This is Allen Hiser with the 5 Staff. What neutron fluence does to stainless steel 6 in particular or the internals is you increase the 7 fluence, the yield strength increases, the ultimate 8 strength increases, the fracture toughness is reduced, 9 and that's normally what we consider neutron 10 embrittlement. It makes the fracture toughness 11 decrease. In addition, the ductility decreases. And 12 that's the bendability that you spoke of. So the 13 fracture toughness does get reduced by neutron 14 fluence.
15 CHAIRMAN MCDADE: And please explain to me 16 very briefly if you can exactly what you mean by 17 fracture toughness --
18 DR. HISER: Fracture toughness --
19 CHAIRMAN MCDADE: -- as opposed to 20 strength.
21 DR. HISER: Well, I guess to separate them 22 maybe in two ways. Strength is in particular relevant 23 if you have an uncracked component. The failure of 24 that component will be directly related to the 25 strength of the material. So if you have an NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5117 1 embrittled component or an irradiated component that 2 has no cracks, it will fail at a higher load than one 3 that is unirradiated. And going back to yesterday's 4 discussion, you can always have a higher load and 5 cause that uncracked component to fail.
6 For the fracture toughness, that really 7 relates to the response of a component that has a 8 crack in it. So, for example, if you have two 9 components that are identical, one is irradiated, one 10 is not, they both have the same size of crack, they 11 both are subjected to the same loads, the one that's 12 irradiated will have a lower fracture toughness, the 13 crack will grow more readily in that material than in 14 the unirradiated component. So that is where the 15 fracture toughness comes into place. If you have a 16 crack to start with.
17 CHAIRMAN MCDADE: Okay. And, Dr. Lahey, do 18 you agree with that?
19 DR. LAHEY: Yes, sir. I tried to say that, 20 he said it better than I. And I do want to bring 21 another thing up --
22 ADMIN. JUDGE WARDWELL: Before you go on to 23 that other thing, I want to clarify something else you 24 said to fix this point, if I might.
25 DR. LAHEY: Yes.
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5118 1 ADMIN. JUDGE WARDWELL: And that is, you 2 used the phrase that if it's embrittled and then it's 3 hit with a significant load. Would your same result 4 happen if it was hit with design basis loads? Do you 5 believe those are significant loads to cause the issue 6 that you're dealing with?
7 DR. LAHEY: Yes, many of them are. Many of 8 them are and --
9 ADMIN. JUDGE WARDWELL: That's all I want 10 to know is that it doesn't have to be an excessive 11 load, you believe this will happen under design basis 12 loads. Is that correct?
13 DR. LAHEY: That's correct.
14 ADMIN. JUDGE WARDWELL: Thank you.
15 DR. LAHEY: The difference between what 16 you've heard in the past and what I'm saying is, is 17 that when people have said that they can withstand 18 design basis loads, if you look at the EPRI document 19 and see what they've done, they've applied those in a 20 static way, not in an impulsive way. And I tried to 21 show you in my little cartoon yesterday that you can 22 have significant difference between static and dynamic 23 loads. I want to return to the document where Mr.
24 Lott was an author of. This is the Icone paper that 25 you had asked about. And in that one, the difference NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5119 1 from the one we had talked about in more detail --
2 CHAIRMAN MCDADE: Do you have an exhibit 3 number on that?
4 DR. LAHEY: Do you know the exhibit number?
5 Oh, yes -- NRC 000177.
6 CHAIRMAN MCDADE: Thank you.
7 DR. LAHEY: Okay. In there is a statement 8 which is a little different from what I heard before.
9 This is light water reactor conditions and it says 10 that the increase in tensile strength because of 11 irradiation causes a, in the high cycle fatigue 12 region, low amplitude, an increase in the strength 13 capability, less failure, and a decrease of ductility 14 results in a decrease of fatigue life in the low cycle 15 fatigue region. So this is entirely consistent with 16 a high pressure or the high flux reader reactor data, 17 which was at a higher temperature.
18 They go on to say that the strain 19 amplitude in their test was rather small, 0.6 percent, 20 so they only saw the increase, not the decrease. And 21 we're worried about the decrease, because that's an 22 indication if you have enough amplitude, you can cause 23 the thing to fail early and if you have a larger 24 amplitude because of a shock load, it can have a 25 catastrophic failure. So I wanted to bring that quote NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5120 1 to you. That's on Page 2 under the Effect of Fluence, 2 is the heading.
3 DR. LOTT: And may I, your honor? I just 4 want to say that, that statement about the high strain 5 amplitude data or the low cycle data reflected 6 primarily the limits of what was tested in this 7 program. And we were aware when we wrote the program 8 of exactly the data that Dr. Lahey has suggested here.
9 And just wanted to be cautious about how far we 10 extended the findings of this paper. And I would 11 stick by that. I will also point out, and I think 12 it's again something we may get to under the next 13 Contention, that we don't really expect to see these 14 large strain amplitudes in the reactor internals, at 15 least the reactor internals that have significant 16 irradiation effects.
17 MR. STROSNIDER: Your honor, this is Jack 18 Strosnider for Entergy.
19 ADMIN. JUDGE WARDWELL: I think we're going 20 to move on now, thank you.
21 MR. STROSNIDER: Going to move on? Okay.
22 ADMIN. JUDGE WARDWELL: Yes. Back to my 23 original question, which I really don't want to reread 24 again, but it was all those -- the statement that NRC 25 disagreed with your concerns and pointing out that --
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5121 1 one of the things they pointed out was primary 2 components to be inspected under Entergy's program are 3 those components that are most likely to be affected 4 by synergistic effects if they exist. Do you agree 5 that the primary components will be those most likely 6 to be affected by synergism?
7 DR. LAHEY: Yes. I don't disagree with 8 that.
9 ADMIN. JUDGE WARDWELL: Thank you.
10 DR. LAHEY: Their difference between how I 11 view it and how they view it is, is the cracking 12 required?
13 ADMIN. JUDGE WARDWELL: But if cracking 14 does occur, wouldn't that be a result of whatever 15 mechanisms created that? And so the synergism would 16 be built into that observed crack, would it not?
17 DR. LAHEY: I don't disagree what that 18 either. But remember, what I'm concerned with is 19 either before or after any of the significant surface 20 cracking. If you have an event which loads it 21 impulsively, you can fail the structure. And then 22 once you fail the structure, depending on what it is, 23 it can lead to an uncoolable geometry. That's what 24 I'm concerned with.
25 ADMIN. JUDGE WARDWELL: Okay, thank you.
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5122 1 I think you may have said this, but I'm going to ask 2 it again because maybe you didn't, so I want to make 3 sure we do cover this space. Are you concerned with 4 the interval of the inspections or do you feel that's 5 adequate based on the operating experience to date?
6 DR. LAHEY: You're going back to the ten 7 year --
8 ADMIN. JUDGE WARDWELL: Right.
9 DR. LAHEY: -- interval?
10 ADMIN. JUDGE WARDWELL: Exactly.
11 DR. LAHEY: It seems a little long, but I'm 12 more interested in a baseline inspection, a very 13 thorough baseline inspection as they go into the 14 period of extended operation, because otherwise, an 15 inspection process doesn't make a lot of sense if you 16 don't know where you start. So I think the sooner the 17 better they do that. And then after that, the 18 interval I think would depend on the kind of things 19 that were discussed. What would be the implications 20 of failure and how long do you have to take action 21 before you can have a big problem?
22 ADMIN. JUDGE WARDWELL: Right now, and I 23 believe they're planned for 2016 and 2019, if I was 24 correct in my memory. Do you have any comments in 25 regards to what you would recommend and what's your NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5123 1 justification for recommending anything earlier than 2 -- well, it's not going to happen before 2016 anyhow, 3 but the 2019 one?
4 DR. LAHEY: Well, we had recommended some 5 time ago, but time goes on, so the sooner the better.
6 But if that's the sooner they can do it, so be it.
7 ADMIN. JUDGE WARDWELL: Thank you. I guess 8 you, and as I heard you say, you're more concerned 9 with the extent of that inspection program rather than 10 necessarily any timing between the small years between 11 2016 and 2019?
12 DR. LAHEY: I think that's correct.
13 ADMIN. JUDGE WARDWELL: Thank you. Let's 14 move on now to talk a little bit more about the 15 preventative actions, the corrective actions, and 16 acceptance criteria. Entergy's testimony, 616, Answer 17 203, Page 136 to 137, Entergy has undertaken or will 18 implement several types of preventative actions to 19 manage the effects of aging on reactor vessel 20 internals at Indian Point, including implementing the 21 IPEC water chemistry control program, replacing the 22 split pins at IP2, committing to replace IP2 split 23 pins again in 2016, using the fatigue monitoring 24 program and addressing the action levels Number 8 in 25 the safety evaluation for MRP 227-A, tracking plant NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5124 1 transience and cycles, thereby assuring that fatigue 2 usage from actual plant transience does not exceed 3 ASME code design limits, and implementing neutron flux 4 reduction measures to minimize the neutron fluence on 5 the reactor pressure vessel, which in turn will 6 minimize radiation induced aging effects at high 7 fluence locations within the RVIs. I guess I'd ask 8 the Staff, do you agree that these are considered 9 preventative maintenance activities that are 10 designated by GALL to be part of a consistent AMP?
11 DR. HISER: This is Allen Hiser with the 12 Staff. The water chemistry program clearly is a 13 preventative measure that's implemented as part of 14 their AMP. The replacement of split pins, again, 15 clearly putting in new material would prevent the 16 accumulation of degradation that had occurred with the 17 old pins. Fatigue monitoring is, I'm not sure if I'd 18 call it preventative, but it clearly is an appropriate 19 measure to ensure that the fatigue life is adequately 20 monitored during the operation of the plant. So that 21 would -- I'm not sure if I would call it preventative, 22 but it -- prevent, mitigate, or minimize the effects 23 of aging, I'd say, yes, it's within that umbrella.
24 And finally the flux reduction, again, would help to 25 minimize the aging effects on the vessel internals.
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5125 1 So I would agree with each of these.
2 ADMIN. JUDGE WARDWELL: Okay, thank you.
3 In New York State's Exhibit 496, Attachment 1, which 4 is the AMP, Page 5, the AMP states that the reactor 5 vessel internals program is a condition monitoring 6 program that does not include preventative actions.
7 So I guess I'd ask, and I'll start with Entergy, 8 doesn't that kind of contradict what you've stated in 9 Answer 203 that I quoted from above where you're 10 taking credit for a bunch of preventative actions?
11 MR. COX: Could you repeat that question?
12 ADMIN. JUDGE WARDWELL: Yes. The previous 13 quote I had from your testimony of which I just asked 14 Staff to respond to, which was Answer 203, you listed 15 a number of preventative measurements that you were 16 taking credit for. Your AMP states that, on Page 5, 17 that the reactor vessel internals program is a 18 condition monitoring program that does not include 19 preventative actions. Well, it seems like you just 20 took credit for a bunch of preventative actions and 21 then you're saying in your AMP, it doesn't include 22 them.
23 MR. COX: This is Alan Cox for Entergy. I 24 think it's really a matter of semantics and how you 25 describe things. What was meant by the statement in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5126 1 the reactor vessel internals program was that the 2 water chemistry controls were included in another AMP.
3 I mean, there is a separate AMP that applies not just 4 to reactor vessel internals, but to water chemistry on 5 the primary side as a whole. So that AMP goes beyond 6 just reactor vessel internals. It's treated as a 7 separate program, but it is referenced from the 8 preventative action section here and it says that we 9 do have preventative actions in that AMP that will 10 apply to the reactor vessel internals. It's not a 11 part of the reactor vessel internals AMP in the way it 12 was described because it's a program that covers a 13 whole lot more areas than just reactor vessel 14 internals.
15 ADMIN. JUDGE WARDWELL: Okay, thank you.
16 I guess that explains it as best you can. And I guess 17 I'll turn to Staff. That Criteria 2 of GALL does 18 require preventative actions. So don't there need to 19 be some preventative actions within each individual 20 AMP for each of the different components that are 21 addressed by different AMPs, i.e., doesn't the reactor 22 vessel internals have to have preventative actions 23 associated with it in order to be consistent with 24 GALL?
25 DR. HISER: This is Allen Hiser of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5127 1 Staff. For the vessel internals program, that is 2 correct. Water chemistry is an essential element of 3 the program. Otherwise, the inspection types, 4 frequency, scope, et cetera, would be different, 5 because there's an expectation that there is a program 6 that's helping to minimize the effects of aging.
7 ADMIN. JUDGE WARDWELL: Would the, 8 likewise, the activity in regards to the fuel that 9 reduces the fluence, if you will, but the way it's 10 configured also be considered one of these 11 preventative measures that you're taking account of 12 and taking credit for in your evaluation of whether or 13 not Criteria 2 is met for the RVIs?
14 DR. HISER: This is Allen Hiser of the 15 Staff. I think the flux reduction is not explicit 16 within the GALL AMP, but it is explicit in MRP 227 17 that it was one of the three criteria used for 18 demonstrating that a plant was bounded by the report.
19 And based on that, it is sort of implicit to the GALL 20 AMP that, that is necessary. Again, if flux reduction 21 was not implemented at a plant, than it would be 22 inappropriate for them to use MRP 227 because that is 23 one of the fundamental assumptions because under the 24 aging evaluation, it's in the report, and to 25 demonstrate the adequacy of aging management.
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5128 1 ADMIN. JUDGE WARDWELL: And in your SER, 2 have you specifically addressed each of the criteria, 3 including Criteria 7 for corrective actions and 8 for 4 confirmation process dealing with assuring 5 preventative and corrective actions?
6 MR. POEHLER: This is Jeffrey Poehler of 7 the Staff. Yes, we have specifically addressed each 8 of the ten elements of the GALL Aging Management 9 Program, as modified by the LRISG.
10 ADMIN. JUDGE WARDWELL: Okay, thank you.
11 Entergy's testimony, 616, Answer 146, Page 95, states 12 that once a defect is discovered, its ability to 13 withstand fatigue and combinations of both normal and 14 accidental loads is evaluated by either fracture 15 mechanics analysis or a structural analysis, i.e., an 16 engineering evaluation, using the lower bound fracture 17 toughness, i.e., the evaluation assumes a bounding 18 level of embrittlement of the material.
19 Thus, the program has compensated for any 20 inability to directly determine the level of 21 embrittlement through a conservative assumption 22 employed during evaluation of inspection findings.
23 Thus, reasonable assurance that the effects of aging 24 will be adequately managed is provided without the 25 need for direct observation or measurement of the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5129 1 level of embrittlement. I guess my question to 2 Entergy is, how was this bounding level of 3 embrittlement, as you referred to up above, 4 determined?
5 DR. LOTT: Basically, it was determined 6 based on published data on fracture toughness and, 7 effectively, analysis. Which is in -- I'm going to 8 look to my colleagues for the reference numbers, 727?
9 And published by the NRC in the NUREG process.
10 MR. GRIESBACH: 7207?
11 DR. LOTT: 7207, I'm sorry.
12 MR. GRIESBACH: Also MRP 210.
13 DR. LOTT: Yes. Maybe I should let Mr.
14 Griesbach talk.
15 MR. GRIESBACH: This is Tim Griesbach from 16 Entergy. That data and the evaluation method that 17 would be used as an example is given in MRP 210.
18 That's an EPRI MRP program document.
19 ADMIN. JUDGE WARDWELL: And you believe 20 that's an exhibit in this proceeding?
21 DR. LOTT: Yes.
22 MR. GRIESBACH: Yes, it is.
23 ADMIN. JUDGE WARDWELL: Okay. Thank you.
24 DR. LOTT: 646.
25 ADMIN. JUDGE WARDWELL: And I assume I'll NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5130 1 hear what the number is shortly from my ace searchers 2 of what that is.
3 MR. KUYLER: Your honor, MRP 210 is Entergy 4 Exhibit 646.
5 ADMIN. JUDGE WARDWELL: Oh, thank you so 6 much. I'd also like to -- in that statement, to 7 refresh your memory from three minutes ago, the 8 program has compensated for any inability to directly 9 determine the level of embrittlement through a 10 conservative assumption. And what is this 11 conservative assumption employed during the evaluation 12 of the inspection findings to which you refer that 13 ensures that the program has compensated for any 14 inability to directly determine the level of 15 embrittlement? And where might that be documented?
16 DR. LOTT: Again, that's documented and 17 it's required based on the procedures in WCAP-170986, 18 which was our internal, we'll be calling methodology 19 and data requirements for determinative engineering 20 evaluations. What we -- again, the bounding value is 21 based on a fracture mechanics analysis where we're 22 looking at this reduced fracture toughness, which is 23 a property of the material. In most cases, it's done 24 on the basis of a linear elastic evaluation, even 25 though we've already testified that it appears that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5131 1 there's some ductility left in these materials, it 2 assumes very low levels of ductility. And it bounds 3 the existing data on fracture toughness of material as 4 a function of fluence, for stainless steel alloys in 5 general.
6 ADMIN. JUDGE WARDWELL: Thank you. Dr.
7 Lahey, why hasn't this approach covered much of some 8 of your uncertainties that you are concerned with in 9 regards to the potential failure modes of what might 10 take place under operational conditions?
11 DR. LAHEY: Does your question -- this is 12 Richard Lahey from New York. Does your question also 13 include the previous discussion with the replacement 14 of split pins and that sort of thing?
15 ADMIN. JUDGE WARDWELL: Sure, if you want 16 to add that --
17 DR. LAHEY: Okay.
18 ADMIN. JUDGE WARDWELL: -- into it, you can 19 talk about that also.
20 DR. LAHEY: Those steps, I believe, are 21 quite prudent. We particularly like the replacement 22 of a degraded component and would highly encourage the 23 replacement of other degraded components, in 24 particular the baffle bolts, which are the most 25 vulnerable. The other part of your question, I guess, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5132 1 relates to the loads and are the loads which are 2 applied to determine the integrity or the time it will 3 take for a component to failure, if they are 4 appropriate loads, if they are impulsive loads, which 5 would have the maximum effect on causing a failure.
6 If they're absolutely correct, then I have no problem 7 with it. So far, I haven't seen that.
8 ADMIN. JUDGE WARDWELL: I think I was more 9 concerned with the bounding level of embrittlement 10 that was selected for this analysis and the 11 conservative assumption employed in evaluation of the 12 inspection findings.
13 DR. LAHEY: Okay. Maybe you can -- one of 14 the exhibits we have, New York State 000495, could we 15 bring that up and we talk about it?
16 ADMIN. JUDGE WARDWELL: What page number 17 would you like?
18 DR. LAHEY: It's Page 3.
19 ADMIN. JUDGE WARDWELL: Okay.
20 DR. LAHEY: Right. So this is a --
21 ADMIN. JUDGE WARDWELL: Well, wait just a 22 second until we get that --
23 DR. LAHEY: Oh, it's not up yet?
24 ADMIN. JUDGE WARDWELL: It's not up in 25 front of me at least.
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5133 1 DR. LAHEY: Oh, it's up here.
2 ADMIN. JUDGE WARDWELL: The ON/OFF button 3 being in the wrong position. We all -- none of us 4 have it, do we?
5 MR. HARRIS: Your honor, this is Brian 6 Harris for the Staff. We don't have it over here 7 either.
8 ADMIN. JUDGE WARDWELL: You don't have it 9 either?
10 MS. SUTTON: Neither does Entergy, your 11 honor.
12 MR. SIPOS: Your honor, for New York, we 13 have it up over here.
14 (Laughter.)
15 ADMIN. JUDGE WARDWELL: You seem pretty 16 smug about that, Mr. Sipos.
17 MR. SIPOS: It's a rare feeling.
18 (Laughter.)
19 ADMIN. JUDGE WARDWELL: You going to rent 20 us out a little sneak views of this or --
21 MR. SIPOS: I could turn the monitor 22 towards your honors direction if you like.
23 ADMIN. JUDGE WARDWELL: For the right 24 price?
25 (Laughter.)
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5134 1 ADMIN. JUDGE WARDWELL: These aren't 2 connected in series are they, like my Christmas 3 lights? Here we go. We've got a winner.
4 (Laughter.)
5 DR. LAHEY: Can you see it? Is it up on 6 your screen now?
7 ADMIN. JUDGE WARDWELL: It is, thank you.
8 DR. LAHEY: Okay.
9 ADMIN. JUDGE WARDWELL: Thank you for 10 waiting.
11 DR. LAHEY: Over there is it okay?
12 ADMIN. JUDGE WARDWELL: Is everyone happy?
13 MR. HARRIS: Yes.
14 MR. SIPOS: Yes, your honor.
15 DR. LAHEY: Okay. So this is a plot that 16 keeps getting passed around. It's in Gary Was's 17 classic book on nuclear metallurgy, it's in a lot of 18 EPRI reports, a lot of U.S. NRC reports. It came from 19 an individual and there's certain assumptions made in 20 it in terms of how you calculate the fluxes, the 21 neutron fluxes. And, in particular, what components 22 you're choosing. But it gives you a pretty good 23 estimate of the type of fluences you're going to see.
24 So on the upper curve is the fluence for greater than 25 one million electron volt neutrons, which are the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5135 1 damaging ones. And on the lower scale is the 2 corresponding displacements per atom, how many times 3 each atom is knocked out of its lattice. So you can 4 see, if you look at the end of PWR life extension, 5 it's expected to be greater than 100 dpa or greater 6 than ten to the 23 --
7 ADMIN. JUDGE WARDWELL: And what are the 8 years for, do you know, that they assume for this life 9 extension?
10 DR. LAHEY: This is the 20 year life 11 extension.
12 ADMIN. JUDGE WARDWELL: Okay. So the 60 13 years total?
14 DR. LAHEY: It's the type we're talking 15 about.
16 ADMIN. JUDGE WARDWELL: Okay, thank you.
17 DR. LAHEY: So you can see there's 18 significant, absolutely significant fluences and 19 damage at that point. And of course, that is where 20 you have to be really concerned because if you look at 21 the data on when bad things happen in terms of 22 embrittlement and all these other issues we've been 23 talking about, you have to get to a displacement per 24 atom of about one or so before bad things start to 25 happen. Then it drops off pretty fast, it really goes NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5136 1 down very fast.
2 So once you start getting out here, you're 3 really in a region where you have to be very careful 4 because you're on what I call the bathtub curve, if 5 you know what I mean by that. You're way out on the 6 other part of the bathtub curve where you're starting 7 to wear things out, you're really beating them up and 8 they're failing. So, hopefully this gives you some 9 insight into what the concerns are, just as a way to 10 benchmark yourself.
11 The NRC uses the criterion for significant 12 embrittlement, as I understand it anyway, you folks 13 can correct me if I'm wrong, but about one times ten 14 to the 21 for fluence, when for stainless steel you 15 start to get significant embrittlement. And other 16 people use like 6.7 times ten to the 20, but it's of 17 that order of magnitude. So you can see, by the end 18 of this thing we're a thousand times greater than the 19 onset of that kind of damage.
20 DR. LOTT: Your honor, may I just -- a 21 moment ago, I had trouble coming up with a reference 22 for you in terms of limiting fracture toughness 23 values. I wanted to point that actually those 24 limiting values, when I think about it, are in MRP 25 227, Section 6, as well as in the other documents we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5137 1 described. I'd also like to perhaps amend a little 2 bit to my statement and say that, one of the other 3 conservatisms in our analysis has been throughout, we 4 tend to apply peak fluence values to entire 5 components. So we don't necessarily -- and when we're 6 doing evaluations, when we've identified components, 7 we have looked at 60 year fluences and looked at the 8 peak location on that and there could be a large 9 gradient in fluence across the component. So I would 10 just suggest that, that's an additionally conservative 11 assumption we've made about determining the fluence on 12 a component.
13 ADMIN. JUDGE WARDWELL: Okay, thank you.
14 Do you have any other comments that you'd like to add 15 in regards to this figure that Dr. Lahey has presented 16 and his comments associated with it?
17 DR. LOTT: I mean, I think I've seen this 18 document many times before as well. It's just a 19 schematic that I think is consistent with actually 20 many of the assumptions we've made here in terms of --
21 I think Dr. Lahey was talking about the threshold 22 values for irradiation embrittlement of seven times 23 ten to the 20, that's exactly the value we used in our 24 evaluations, that's approximately one dpa. So, I 25 think that we're on the same -- we've dealt with and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5138 1 understand these concerns and that our testimony has 2 demonstrated that.
3 ADMIN. JUDGE WARDWELL: Clarify for me, 4 will you, if you are using seven times ten to the 20 5 or about one dpa, as what you assume as a conservative 6 fluence value, how does that relate to what you may 7 actually be experiencing, which is two orders of 8 magnitude higher than that by the end of the life 9 extension?
10 DR. LOTT: Well, first of all, what we're 11 looking at is the threshold for what's the lowest 12 value. I mean, so basically what that determines is 13 whether a component has any irradiation embrittlement 14 concern at all. Above that, all the way up to a 15 million, presumably, it has susceptibility. Our first 16 goal of the threshold values was to determine the 17 lower limit.
18 ADMIN. JUDGE WARDWELL: So it's 19 conservative because it's on the low side, 20 encompassing much more numbers of --
21 DR. LOTT: Right.
22 ADMIN. JUDGE WARDWELL: -- potential of --
23 DR. LOTT: Right, so yes. That's --
24 ADMIN. JUDGE WARDWELL: -- internals.
25 DR. LOTT: -- what the concern is and we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5139 1 identified components with potential irradiation 2 embrittlement. And our basis for that was a 3 relatively low value, the onset of embrittlement, 4 consistent with the curve I showed you earlier this 5 morning on yield stress.
6 ADMIN. JUDGE WARDWELL: Yes. Okay. Thank 7 you.
8 ADMIN. JUDGE KENNEDY: Dr. Lott, this is 9 Judge Kennedy. The baffle, let's see if I'm getting 10 this right, the baffle bolts are, at least components 11 I identify on this chart, at somewhere around ten to 12 the 22, PWR baffle bolt failures?
13 DR. LOTT: Yes. Some --
14 ADMIN. JUDGE KENNEDY: Are there --
15 DR. LOTT: Yes.
16 ADMIN. JUDGE KENNEDY: Are there reactor 17 vessel internal components that experience fluences 18 beyond that value?
19 DR. LOTT: Yes, there are. I mean, it's 20 not like that's a fall off the cliff value. When we 21 have -- and have reported seeing IASCC and baffle 22 bolts in operating plants, I think that, that value is 23 basically trying to demonstrate that it happened 24 roughly at that fluence.
25 ADMIN. JUDGE KENNEDY: Okay. All right, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5140 1 thanks.
2 CHAIRMAN MCDADE: Is this the baffle former 3 bolt or the baffle edge bolt or is this all baffle 4 bolts that they're talking about here?
5 DR. LOTT: Okay. There are basically three 6 different kinds of baffle bolts. There's the baffle 7 -- if you're familiar with the baffle structure, 8 they're plates that surround the core and there's 9 plates behind them, horizontal plates, that hold them 10 in place. The bolt between the baffle plate and the 11 horizontal plate is a baffle bolt. Where the two 12 plates come together, along the edge, along the seam, 13 there may be bolts that go from baffle plate to baffle 14 plate, not baffle plate to the former. Those are the 15 baffle edge bolts. They basically seal up the gap 16 between the two plates to keep water from jetting 17 through there. Not all plants even have baffle edge 18 bolts.
19 And we've done evaluations to determine 20 that in terms of holding the baffle together, in terms 21 of an accident type scenario, we don't take credit for 22 the baffle edge bolts at all. So it's, again, a 23 conservatism in our assumptions when we do these 24 acceptable bolting patterns. So baffle edge bolts.
25 And then there's barrel former bolts, which are from NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5141 1 the round barrel, which is the outside container of 2 the entire internals, that the horizontal plates, the 3 other side of them, attach to the barrel. So barrel 4 former bolts, baffle former bolts, and baffle edge 5 bolts.
6 MR. KUYLER: Your honor, there's a diagram 7 in Entergy's testimony, Exhibit 616, Page 54.
8 CHAIRMAN MCDADE: Okay, thank you. And 9 thank you, Dr. Lott.
10 ADMIN. JUDGE WARDWELL: NRC's testimony, 11 197, Answer 179, Pages 102 to 103, the reactor vessel 12 internals AMP is structured to managing the aging 13 effects such that the intended function of the 14 components will be preserved during the period of 15 extended operation from 40 to 60 years. It 16 accomplishes this task by establishing an Inspection 17 Plan for the relevant components that it's able to 18 identify potential aging effects prior to any loss of 19 function through appropriate schedules and 20 conservative acceptance criteria. And so I'll ask 21 Entergy, does Table 5-5 of the Applicant's Inspection 22 Plan, and that's New York State 496, Attachment 2, 23 contain these acceptance criteria?
24 MR. DOLANSKY: This is Bob Dolansky with 25 Entergy. Yes.
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5142 1 ADMIN. JUDGE WARDWELL: Okay. And does it 2 not state that the examination acceptance criteria for 3 visual examination is the absence of the specific 4 relevant condition?
5 MR. DOLANSKY: Yes.
6 ADMIN. JUDGE WARDWELL: And now the tougher 7 part, what are examples of this relevant condition?
8 And where would one find that?
9 MR. DOLANSKY: For an EVT1, it would be a 10 crack-like indication.
11 ADMIN. JUDGE WARDWELL: So it varies by 12 your inspection technique generally, rather than by 13 component?
14 MR. DOLANSKY: And by the component. In 15 other words, EVT1 is typically looking for cracking, 16 but a VT3 -- this is Bob Dolansky for Entergy. VT3 17 could be looking for either wear or it could be 18 looking for a dimensional change, like void swelling, 19 something like that. So the acceptance criteria would 20 depend on what you're looking for and the method that 21 you're using.
22 ADMIN. JUDGE WARDWELL: And where is that 23 documented anywhere in regards to the various 24 components so that we could turn to that and it would 25 say, yes, it's any indication of cracking or it's any NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5143 1 indication of a dimensional change or --
2 MR. DOLANSKY: Give me one moment, please.
3 ADMIN. JUDGE WARDWELL: If you want to get 4 back to us --
5 MR. DOLANSKY: Okay. If you -- I'll give 6 a cite, one second. New York State 496, Letter 12-7 037.
8 ADMIN. JUDGE WARDWELL: Okay.
9 MR. DOLANSKY: Table 5-5.
10 ADMIN. JUDGE WARDWELL Yes, that's what I 11 was referring to.
12 MR. DOLANSKY: That gives -- you'll see the 13 actual -- under examination acceptance criteria in the 14 table. For instance, for the upper core barrel 15 cylinder girth welds, the specific relevant condition 16 is a detectable crack-like surface indication.
17 ADMIN. JUDGE WARDWELL: Okay. So where 18 ever we see that, that's what it means. As soon as 19 you detect cracking, you're going to take some 20 corrective action.
21 MR. DOLANSKY: Right. So our inspection 22 procedure, that's the people actually doing the 23 inspection use a procedure, their acceptance criteria 24 for a recordable indication would be that.
25 ADMIN. JUDGE WARDWELL: Okay. Thank you.
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5144 1 MR. DOLANSKY: You're welcome.
2 ADMIN. JUDGE WARDWELL: That helps. Just 3 for completeness, Dr. Lahey, any comments in regards 4 to that acceptance criteria and where it's found and 5 the adequacy of it?
6 DR. LAHEY: This is Richard Lahey. I'm 7 sorry, I don't have that document in front of me and 8 I don't recall it.
9 ADMIN. JUDGE WARDWELL: Okay. You have no 10 comment?
11 DR. LAHEY: So I can't really answer right 12 now.
13 ADMIN. JUDGE WARDWELL: If you do later on, 14 if you do want to look at it later on and have some 15 comments --
16 DR. LAHEY: Okay.
17 ADMIN. JUDGE WARDWELL: -- remind me of it 18 and we'll be glad to. I want to make sure you have a 19 chance to anyhow.
20 DR. LAHEY: Thank you.
21 ADMIN. JUDGE WARDWELL: New York State's 22 testimony, 482, Page 79, states that, I believe that 23 the most vulnerable reactor pressure vessel internals 24 need to be carefully identified and repaired or 25 replaced prior to the extended operation since it is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5145 1 beyond the current state of the art to perform 2 realistic and accurate calculations on relocation of 3 failed RVP internals and the resultant potential for 4 core blockages or degraded core cooling. And I guess 5 I'll ask you, Dr. Lahey, that while I understand the 6 basis for your position in regards to this 7 replacement, does not this run counter to the whole 8 idea of managing aging? And doesn't it to a large 9 degree try to circumvent all of the regulations that 10 are geared towards aging management as opposed to 11 prescriptive replacements?
12 DR. LAHEY: That's a very interesting 13 question. In my opinion, aging management is indeed 14 important. But when you get components that look like 15 they're vulnerable and can fail and you're not able to 16 determine with any precision what the effect of that 17 might be, then I think the prudent thing to do is to 18 replace those components. I've spent ten years of my 19 life trying to calculate where things go and it's very 20 hard to do, very difficult to do, there's too many 21 possibilities.
22 But the one thing I know for sure is, once 23 you lose an intact geometry, you've got big problems.
24 So, anything that will preserve that, I think we're 25 way ahead of the game and, to me -- do you know what NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5146 1 a make-buy decision is? You decide is it cheaper to 2 make something or buy something? This is like a make-3 buy decision. Is it cheaper to calculate and inspect 4 and go through litigation and all that or is it 5 cheaper just to replace it and have the problem go 6 away? And, for many of these things, I think it's 7 cheaper and much more prudent to just replace it.
8 ADMIN. JUDGE WARDWELL: The -- I forgot my 9 question now because I was going to change it in 10 regards to your last comment. But with your 11 description of the details which basically are in the 12 AMP and the inspection program that's in there and 13 your applauding of it as not the complete what's 14 needed, but not really a lot of substantive 15 disagreements with the approaches and the extents of 16 what they're doing, given what we've heard in regards 17 to the conservative assumptions, why isn't there a 18 fair degree of reasonable assurance that something, if 19 it does go amiss, would be detected before that 20 intended functionality of the RVIs was lost?
21 DR. LAHEY: This is again Richard Lahey 22 from New York State. When you say conservative 23 assumptions, are you referring to embrittlement alone 24 now or --
25 ADMIN. JUDGE WARDWELL: I'm talking about NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5147 1 all that we've heard about --
2 DR. LAHEY: All of the above?
3 ADMIN. JUDGE WARDWELL: -- to date in 4 regards to our questioning and what's in their 5 testimony in regards to the decisions they make as 6 they prepared their AMP and then how they're 7 implementing their AMP.
8 DR. LAHEY: Right.
9 ADMIN. JUDGE WARDWELL: Does that not give 10 one reasonable assurance that if something does go 11 awry, it's not going to try to stop anything that 12 might go awry, but if it does, isn't there a 13 reasonable assurance that it would be detected prior 14 to the RVI losing all of its intended function, even 15 though it may crack or even do some other things?
16 DR. LAHEY: Not in my opinion. If you have 17 sufficient degraded components that can lead to a 18 destruction of intact geometry, there's a reasonable 19 chance that you can have an unexpected accident.
20 Accidents, by definition, you don't expect them, but 21 they happen, or an earthquake happens just when you 22 don't want it. And then what will that lead to? If 23 it can lead to degraded or materials which fail and 24 relocate, then I'm very concerned about it. Because 25 I just know you can't calculate the consequence of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5148 1 that.
2 So if you can identify materials that are 3 highly irradiated, and we can, and highly fatigued, 4 and we can, and they are things that could lead to 5 destruction of a coolable geometry, then you should 6 take action, the sooner the better. Don't wait for 7 something bad to happen. So that's why I'm very keen 8 on do what you're doing with the -- have Entergy do 9 what they're doing with the split pin, replace it 10 because it's degrading. Replace the things that have 11 a significant effect on the safety of the reactor.
12 There's not that many. I mean, we haven't talked 13 about pressure boundary components yet, but we will, 14 tomorrow I guess. And there's only a few that are 15 really, really crucial. And there's only a few things 16 in core that are really, really crucial. And, 17 happily, these are things that aren't that hard to 18 replace.
19 ADMIN. JUDGE WARDWELL: Thank you, Dr.
20 Lahey.
21 DR. LAHEY: That's why --
22 ADMIN. JUDGE WARDWELL: Thank you.
23 DR. LAHEY: -- I feel that way.
24 ADMIN. JUDGE WARDWELL: Entergy, have you 25 addressed anywhere or evaluated or how did you handle NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5149 1 this potential relocation of failed RVIs and the 2 resultant potential for core blockages and degraded 3 core cooling?
4 MR. DOLANSKY: This is Bob Dolansky with 5 Entergy. When we do an analysis that looks at 6 components, one of the requirements is that we 7 maintain core coolability and core geometry. I mean, 8 that's ultimately what we're trying to do. We want to 9 make sure that, that core -- that's the basis of the 10 whole thing is that the core stays coolable and the 11 geometry stays -- that it maintains core geometry. I 12 mean, that is what we do, that's the exact 13 requirements that we analyze for.
14 DR. LOTT: Yes, this is Randy Lott. And I 15 think, again, we kept coming back to it, but the 16 acceptable baffle bolting pattern analysis is a good 17 example of what we're doing. In that analysis, 18 effectively, you're looking to see that there are 19 enough baffle bolts to keep the baffles from moving, 20 interacting with the fuel, crushing fuel grids, or 21 interfering with the ability to drop control rods.
22 Which is really always our concern, it's maintaining 23 core coolability and being able to insert the control 24 rods, shut down the reaction. That's basically the 25 definition of the safety requirement.
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5150 1 MR. DOLANSKY: Right. And I just want to 2 clarify one thing that Dr. Lahey said. We are 3 replacing split pins. We're not planning to replace 4 baffle bolts. Replacing baffle bolts is not some no 5 nevermind, easy thing to do. They have lock bars 6 welded on, it's an extremely difficult area to get to.
7 Split pins were more designed that they might have to 8 be replaced. So the replacement of the split pins is 9 a little easier, it's more straightforward, something 10 that can be done much more easily.
11 It's easy to sit here and say, just 12 replace the baffle bolts. Actually replacing baffle 13 bolts, although it can be done, is not a simple 14 things. There's a lot of dose involved with that, 15 there's a lot of possibility of loose parts, and 16 there's consequences to replacing things that aren't 17 bad. So, the way we look at is, we will go out and do 18 these inspections on the baffle bolts using a 19 technique that's very good, difficult tooling that was 20 developed just to get us a better UT exam will 21 additionally -- when we do that inspection, as Dr.
22 Lott said, we're going to use an acceptable bolting 23 pattern analysis that ensures that if we find degraded 24 bolts that we can maintain core coolability, core 25 geometry. If we do that inspection and we find that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5151 1 there's very, very few bolts that are degraded, to me, 2 it doesn't make any sense to go out and just wholesale 3 replace baffle bolts. There's too much danger and 4 risk in doing that. If the bolts are good, I don't 5 see any reason to do that.
6 ADMIN. JUDGE WARDWELL: Okay, thank you.
7 CHAIRMAN MCDADE: Okay. And also, just as 8 our discussion moves on for the rest of the week, I 9 mean, we neither have the authority or interest in 10 micromanaging the way that Entergy does its 11 operations. You have to make certain business 12 decisions. Our function is just whether or not the 13 plans that you have put forward provide reasonable 14 assurance that these items will maintain their 15 intended function for the period of extended 16 operation.
17 So, certain issues of whether or not as a 18 matter of policy you replace, that's outside the scope 19 of what we're looking at. Again, the scope of what 20 we're looking at is whether or not the plans that you 21 have put forward provide that reasonable assurance 22 with regard to intended function, period. So, it's 23 not as wide -- it's not just an open-ended discussion.
24 We're trying to focus in on what we have to decide 25 here. Judge Wardwell?
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5152 1 DR. LAHEY: Yes, and your honor, can I say 2 something? What we just heard from Entergy is the 3 most encouraging thing I've heard in the last eight 4 years. I'm very happy to hear it because it sounded 5 like they're going to do something that makes a lot of 6 sense, look at the integrity of it, and if it's not a 7 big issue, then don't do it, if it is a big issue, 8 presumably they will entertain that as a possibility 9 to replace. Up until now, all I heard was, question, 10 do you agree with Dr. Lahey on anything and the answer 11 is, no. Everything, no, no, no.
12 (Laughter.)
13 DR. LAHEY: And so this was encouraging to 14 me.
15 ADMIN. JUDGE WARDWELL: Thank you, Dr.
16 Lahey. NRC's Exhibit 197, Answer 134, Page 82, states 17 that the Staff found Entergy's AMP met the Staff's 18 guidance for corrective actions because detected 19 conditions not satisfied in the examination acceptance 20 criteria will be processed through the plant's 21 corrective action programs. And I guess I'll start 22 with the Staff considering it's your exhibit. What's 23 your understanding of how the plant's corrective 24 action program works and interacts with the AMP for 25 reactor vessel internals?
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5153 1 DR. HISER: This is Allen Hiser of the 2 Staff. The corrective action program would be 3 initiated if there's an inspection finding that 4 exceeds the inspection acceptance criteria, which 5 we've discussed previously. I believe it's Table 5-5 6 of the RVI Inspection Plan from Indian Point. The 7 corrective action program would require that the 8 condition be assessed, I guess in maybe two or three 9 different ways.
10 One is, is the condition that was 11 identified, that condition needs to be resolved, is it 12 acceptable maybe through engineering evaluation, is 13 repair required, is replacement required? So that 14 would be one path that would be followed by the 15 corrective action program. A second would be a 16 consideration of expansion of the inspections 17 consistent with the AMP. So that would follow.
18 Consideration of reinspection interval would be a part 19 of that evaluation as well. So the corrective action 20 program should consider pretty much any aspect that's 21 relevant to the finding itself, be it for that finding 22 or other components.
23 ADMIN. JUDGE WARDWELL: Okay, thank you.
24 DR. HISER: And that is required in 25 Appendix B to 10 CFR Part 50, so it is a regulated NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5154 1 activity, a regulated program of the plant that the 2 NRC inspects periodically.
3 ADMIN. JUDGE WARDWELL: Thank you for your 4 understanding of the plant's corrective action. I'll 5 turn to Entergy and can you now explain how your 6 plant's corrective action program actually works and 7 interacts with the AMP or just state what differences 8 you might have with what Dr. Hiser just presented?
9 MR. AZEVEDO: Yes, this is Nelson Azevedo 10 for Entergy. What was just discussed is correct, 11 that's what we do. So we find an indication then we 12 put in our corrective action program to determine 13 whether we have to repair or replace it or whether 14 it's acceptable. We also do extended condition 15 inspections if it's warranted. So those are the 16 things we look at.
17 MR. DOLANSKY: I just want to -- this is 18 Bob Dolansky with Entergy. Just to add on something 19 I said earlier where I said Westinghouse was 20 developing acceptance criteria for us. The acceptance 21 criteria in the plan are the just evidence of a crack-22 like indication. That would be in the inspection 23 procedure. We're additionally, right now, getting 24 additional acceptance criteria that if there was a 25 crack and it could be a certain size, that's what's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5155 1 being developed. So I misspoke a little bit before, 2 that's not the acceptance criteria in here. That's 3 the --
4 ADMIN. JUDGE WARDWELL: As it exists now.
5 MR. DOLANSKY: -- acceptance criteria after 6 it enters our corrective action program.
7 ADMIN. JUDGE WARDWELL: Okay, thank you.
8 Ready to rock and roll.
9 CHAIRMAN MCDADE: Okay. This may be a good 10 time to break for lunch. Do you have anything before 11 we break?
12 ADMIN. JUDGE KENNEDY: No, I do not.
13 CHAIRMAN MCDADE: Okay. The question is, 14 how long we break for lunch? And I'm not sure if 15 there's any difficulty in people getting lunch within 16 a relatively short period of time. I would propose to 17 come back at 1:30. Is that going to give enough time 18 for people to get lunch and take care of what they 19 need to? NRC?
20 MR. HARRIS: Yes, your honor.
21 CHAIRMAN MCDADE: Entergy?
22 MS. SUTTON: Yes, your honor.
23 CHAIRMAN MCDADE: New York?
24 MR. SIPOS: New York, yes, your honor.
25 CHAIRMAN MCDADE: Riverkeeper?
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5156 1 MS. BRANCATO: Yes, thank you.
2 CHAIRMAN MCDADE: Okay. Do any of the 3 witnesses perceive a problem with restarting at 1:30?
4 Apparently not, we're in recess. We'll come back --
5 DR. LAHEY: No, your honor.
6 CHAIRMAN MCDADE: -- at 1:30. Thank you.
7 (Whereupon, the above-entitled matter went 8 off the record at 12:19 p.m. and resumed at 1:34 p.m.)
9 CHAIRMAN MCDADE: The hearing will come to 10 order. There are a couple of, or a few matters I 11 guess, out there that I just wanted to address before 12 we move forward. I believe there was a question, Mr.
13 Dolansky, to you, as to whether or not you could 14 provide other examples of monitoring of aging effects 15 not observable by inspection. Were you --
16 MR. DOLANSKY: Just the one. Just the one 17 that I gave.
18 CHAIRMAN MCDADE: Okay. And there was a 19 question to Mr. Poehler about an EPRI report and 20 whether or not that was a publically available report.
21 MR. POEHLER: Yes, this is Jeffrey Poehler 22 of the Staff. That report is publically available and 23 it's Exhibit NRC 000208, A through E, there's five 24 parts.
25 CHAIRMAN MCDADE: Okay. And what is the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5157 1 EPRI report number?
2 MR. POEHLER: It's an MRP letter actually, 3 MRP --
4 CHAIRMAN MCDADE: Okay.
5 MR. POEHLER: -- Letter 2014-09.
6 CHAIRMAN MCDADE: Okay. Thank you.
7 MR. POEHLER: You're welcome.
8 CHAIRMAN MCDADE: And, Dr. Lahey, you were 9 asked with regard to questions on Table 5-5 of New 10 York 496. Did you have an opportunity to review that?
11 DR. LAHEY: No, your honor, I haven't been 12 able to get access to it yet.
13 CHAIRMAN MCDADE: Okay. At our next break, 14 we will make sure that you get access to that.
15 DR. LAHEY: Okay. Thank you.
16 CHAIRMAN MCDADE: Mr. Sipos, it looked like 17 you had a question.
18 MR. SIPOS: I had a procedural question, 19 your honor. And it concerns scheduling of New York's 20 second expert, Dr. David Duquette, who is an expert on 21 Contention 38. And I was wondering if the Board could 22 provide any guidance as to when you would like to --
23 or when the Board might start Contention 38? And it's 24 purely a logistical question as to when --
25 CHAIRMAN MCDADE: No, I understand.
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5158 1 MR. SIPOS: -- to tell him to get in the 2 car and --
3 CHAIRMAN MCDADE: And my best estimate at 4 this point in time would be our start time on Thursday 5 morning. That we're still, it's Tuesday afternoon, 6 we're still working on 25, 26 we should get started 7 maybe later today, should run most if not all of 8 tomorrow. And a lot of what had been in 38, is in 38, 9 we have touched in on the testimony on 25 and will 10 touch on some more in the testimony of 26. So, from 11 my standpoint, and I haven't, you just asked the 12 question, I haven't discussed it with my colleagues, 13 I would, for planning purposes, plan on starting on 38 14 on Thursday morning at 8:00.
15 ADMIN. JUDGE WARDWELL: Yes, I guess we 16 just talk out loud. Yes, I would concur, in fact, to 17 the point that even if we got done 26 early tomorrow, 18 it would probably still be later in the day, that we 19 could just not plan on starting 38 until Thursday 20 morning.
21 MR. SIPOS: That's very helpful, thank you.
22 CHAIRMAN MCDADE: Dr. Kennedy?
23 ADMIN. JUDGE KENNEDY: That's fine with me.
24 ADMIN. JUDGE WARDWELL: And do you agree 25 with that, that we just wouldn't --
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5159 1 CHAIRMAN MCDADE: Yes. I mean, I think for 2 planning purposes, we're certainly not going to get 3 done with 26 early tomorrow. So, if we did finish 4 with 26 before 6:00, I don't think anybody would 5 complain if we leave it at 5:50 instead of 6:00. And 6 having Dr. Duquette get here for a relatively short 7 period of time --
8 ADMIN. JUDGE WARDWELL: You want to check 9 with the other parties to make sure they're 10 comfortable with that too? If we waste a couple hours 11 tomorrow afternoon by not starting 38, is that --
12 CHAIRMAN MCDADE: I mean, I don't think 13 it'll be wasting a couple of hours, I think it'll be 14 a few minutes, if anything. And we may still be on 15 26. You all can gauge how well we are keeping a 16 schedule. Ms. Sutton?
17 MS. SUTTON: Yes, your honor. That's fine, 18 we don't have any issues with that.
19 MR. HARRIS: The Staff has no issues with 20 that, your honor.
21 CHAIRMAN MCDADE: Okay.
22 MS. BRANCATO: Riverkeeper has no issues, 23 thanks.
24 MR. SIPOS: Thank you.
25 CHAIRMAN MCDADE: Okay.
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5160 1 ADMIN. JUDGE WARDWELL: Well, I have no 2 more questions on 25.
3 MR. SIPOS: We'll start 38?
4 CHAIRMAN MCDADE: There's no questions on 5 26?
6 MR. SIPOS: All right, let's start 38.
7 CHAIRMAN MCDADE: We're in recess until 8 Thursday morning.
9 (Laughter.)
10 ADMIN. JUDGE WARDWELL: Now, we're going to 11 start looking at some specific materials and 12 components associated with the RVI and starting off 13 with control rods and J-groove welds. Entergy 14 testimony, 616, Answer 98, Page 56, the Inspection 15 Plan NL12-037 Attachment 2 at 62 through 64, provides 16 a complete and corrected list of the RVI subassemblies 17 at Indian Point and breaks those subassemblies down to 18 their constituent components. New York State's 19 testimony, 482, Page 13, states the control rods and 20 the associated components are very important RPV 21 internals and their integrity is an extremely 22 important safety concern. In my opinion, omitting the 23 control rod assemblies and associated fittings from an 24 RPV internals Aging Management Program is a serious 25 and indefensible omission.
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5161 1 Entergy's testimony, 616, Answer 99, Page 2 56, states that control rods are not subject to aging 3 management review for two reasons. First, they 4 perform their intended function with moving parts or 5 a change in configuration. Thus, as the NRC Staff 6 concluded, the control rods are active components not 7 subject to aging management review. Second, control 8 rods are considered consumables and the NRC has 9 excluded from the license renewal review process those 10 components that are subject to replacement based on a 11 qualified life or a specified time period. And I 12 guess I'd ask Dr. Lahey, do you now agree that control 13 rods are not subject to aging management review?
14 DR. LAHEY: I agree that's the rule that 15 has been put in place. I still have the concern that, 16 because I'm looking at everything through the prism of 17 reactor safety, so I have the concern that if you have 18 a significant shock load, you can fracture these 19 highly, highly embrittled structures and they will 20 relocate in some way and you don't know how. And they 21 don't care if they're moving or not, I mean, they're 22 going to be at risk. So I think the reason for the 23 rule is because of what I call silos. Things are 24 being thought of quite separately, rather than in an 25 integrated way.
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5162 1 ADMIN. JUDGE WARDWELL: Thank you.
2 CHAIRMAN MCDADE: But, Dr. Lahey, as I 3 understand the position of Entergy and the NRC Staff, 4 these are consumables. Therefore, they don't need an 5 aging management plan, you are of necessity going to 6 be replacing them at set intervals. So, what would an 7 aging management plan consist of since you're already 8 going to replace them?
9 DR. LAHEY: No, I understand that's why 10 they view it the way they do. What I'm saying though 11 is, let's say a week before they're going to replace 12 them and they're in really bad shape in terms of 13 embrittlement, you have an event which causes them to 14 fail. This causes a big problem. So, I personally 15 believe that when you look at aging management, you 16 ought to look at all the things that affect the safety 17 of the plant, the safety of the plant during the 18 extended operation.
19 CHAIRMAN MCDADE: But the handling of those 20 control rods would be no different during the period 21 of extended operation than during the original 22 license. They'd be subject to the same replacement 23 criteria, would they not? So why does this have to do 24 with the aging management and whether or not there 25 should be a period of extended operation as opposed to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5163 1 just the current operating license of the plant?
2 DR. LAHEY: Well, I'm concerned with what 3 happens during the period of extended operation. And 4 this is a possible event that can happen, it's no 5 different than some of the static components. And 6 just because you can replace them doesn't mean that 7 they couldn't be at risk if you have an event which 8 causes a significant shock load.
9 CHAIRMAN MCDADE: Okay. But my question 10 is, why would that be any different during the period 11 of extended operation than it is during the original 12 licensing period?
13 DR. LAHEY: Oh, I misinterpreted your 14 question. It wouldn't, not for the control rods 15 themselves.
16 CHAIRMAN MCDADE: Okay. Thank you. Dr.
17 Wardwell?
18 ADMIN. JUDGE WARDWELL: Entergy's 19 testimony, 616, Answer 100, Page 57, "the control rod 20 guide tube assemblies, including the guide plates 21 (CRGTs) and the lower flange welds are subject to 22 aging management review and the effects of aging on 23 these components are managed through the RVI AMP."
24 And they're citing again NL12-037 Attachment 2 at 4 to 25 5 and Attachment 1 at 6 to 8. And those two NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5164 1 attachments, if we remember, are the Inspection Plan 2 and the AMP, respectively and that's New York State 3 496. Dr. Lahey, therefore, is incorrect when he 4 asserts that Entergy has claimed that the guide tubes, 5 plates, pins, and welds associated with control rods 6 are not RVIs. Dr. Lahey, do you now agree that the 7 control rod guide tube assemblies, including the guide 8 plates and the lower flange welds are in fact subject 9 to aging management review and part of the AMP for 10 RVIs?
11 DR. LAHEY: Yes, I believe they are.
12 ADMIN. JUDGE WARDWELL: Okay.
13 DR. LAHEY: And they should be.
14 ADMIN. JUDGE WARDWELL: Thanks. New York 15 State's testimony, 482, Page 45, Lines 3 through 9, 16 because of geometric considerations, many pressure 17 water reactors, including IP2 and IP3, cannot meet the 18 U.S. NRC's required minimum coverage for the non-19 destructive testing of so-called J-groove welds. And, 20 thus, the integrity of these important CRD stub tube 21 welds cannot be directly confirmed by inspection.
22 I'll start with you, Dr. Lahey, and what do you mean 23 by the CRD stub tube welds?
24 DR. LAHEY: Well, this is one of the issues 25 we talked about this morning, when we were talking NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5165 1 about the possibility of a leakage forming and boron 2 and water and forming boric acid and that sort of 3 thing. So this is the weld on the inside. And as I 4 understand it, they cannot get a complete inspection 5 of this.
6 ADMIN. JUDGE WARDWELL: Okay. And what 7 does that CRD stand for or --
8 DR. LAHEY: Oh, I'm sorry.
9 ADMIN. JUDGE WARDWELL: No, that's true, I 10 appreciate what you've answered so far, that helped.
11 But I also want to know what is the --
12 DR. LAHEY: Control rod drive.
13 ADMIN. JUDGE WARDWELL: Got you. Got you, 14 okay. Thank you. Entergy's testimony on 616, Answer 15 101, Page 57 and 58, the reactor pressure vessel head 16 penetration nozzle welds, sometimes referred to as J-17 groove welds, are not RVIs or even part of the reactor 18 vessel pressure, but instead part of the reactor 19 pressure vessel head. Aging effects applicable to the 20 J-groove welds on the, and this must mean control rod 21 drive M, head penetrations are managed under the 22 reactor vessel head penetration inspection AMP. So, 23 Entergy, what is the M of the control rod drive stand 24 for?
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5166 1 Entergy. The M stands for mechanism.
2 ADMIN. JUDGE WARDWELL: Okay, mechanism.
3 And, so, back to you, Dr. Lahey, and I think we 4 covered this morning, do you now agree that these are 5 managed under a different Aging Management Program and 6 are not really part of this Contention?
7 DR. LAHEY: Well, they are reactor vessel 8 internals. I mean, they're in the vessel, they're 9 subject to aging. How they want to deal with it is 10 not a great concern to me, as long as it's dealt with.
11 ADMIN. JUDGE WARDWELL: Good, thank you.
12 Let's move on to a couple of materials we're dealing 13 with. The first one we'll deal with is wrought 14 Austenitic stainless steel. Entergy's testimony on 15 Exhibit 616, Answer 105, Page 61, says that the 16 majority of the Indian Point reactor vessel internal 17 components are fabricated from wrought Austenitic 18 stainless steels. Even though an increase in strength 19 and decrease in toughness do occur when exposed to 20 neutron irradiation, these materials retain their 21 resistance to fast fracture within the operating 22 temperature of interest for PWRs. The only exceptions 23 are wrought Austenitic stainless steel materials with 24 high amounts of cold working, and that's greater than 25 20 percent of cold working, but these materials are NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5167 1 not present at Indian Point. And I think my first 2 question for Entergy would be that, what do you mean 3 by this cold working and what percentage do you have 4 here at Indian Point?
5 DR. LOTT: Cold working is basically the 6 rolling of the material or the stretching of the 7 material, the working of the material to increase its 8 yield strength. The more you -- for instance, you 9 would take a plate and reduce the area by rolling it, 10 the higher the strength becomes. In general, the 11 reactor internals are not cold worked beyond about 15 12 percent, they're controlled. That's one of our 13 screening criteria that we have been in fact 14 evaluating for.
15 ADMIN. JUDGE WARDWELL: Okay. Thank you.
16 And, Dr. Lahey, do you have any issues with those 17 statements made in their original testimony or what 18 was -- as amplified here?
19 DR. LAHEY: Are you asking me if I agree 20 with the percentage of cold work they have or --
21 ADMIN. JUDGE WARDWELL: Yes. And with 22 that, why they don't see the issue with the Austenitic 23 steel that is exhibited when cold working is greater 24 than 20 percent.
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5168 1 I misunderstood the statement you read, it was focused 2 on that stainless steel is not subject to stress 3 corrosion or embrittlement if it's not cold worked.
4 Did I misunderstand that?
5 ADMIN. JUDGE WARDWELL: No, this dealt 6 mostly -- well, let me repeat it again. The majority 7 of the reactor vessel components are fabricated from 8 wrought Austenitic stainless steel. Do you agree with 9 that?
10 DR. LAHEY: I agree with that, absolutely.
11 ADMIN. JUDGE WARDWELL: And even though an 12 increase in strength and decrease in toughness do 13 occur when exposed to neutron irradiation --
14 DR. LAHEY: Okay.
15 ADMIN. JUDGE WARDWELL: -- these materials 16 retain their resistance to fast fracture within the 17 operating range of interest for PWRs.
18 DR. LAHEY: I don't agree with that. I 19 mean, certainly, there are components that we've 20 talked about that are highly embrittled and do not 21 satisfy that.
22 ADMIN. JUDGE WARDWELL: Okay. And they say 23 that the only exception are those wrought Austenitic 24 stainless steels with high amounts of cold working, 25 greater than 20 percent. And would you agree that if NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5169 1 you had cold rolling steels greater than 20 percent, 2 they wouldn't necessarily show an increase in strength 3 or decrease in toughness and retain their resistance 4 to fast fractures within the operating temperatures?
5 DR. LAHEY: No, I would not.
6 ADMIN. JUDGE WARDWELL: You would not? So 7 you think they still would retain theirs even if you 8 did have cold working --
9 DR. LAHEY: Yes, I think if they're highly 10 embrittled, they are subject to fracture.
11 ADMIN. JUDGE WARDWELL: Right. So then you 12 would agree that, that is an exception from their 13 abilities to resist fast fracturing?
14 DR. LAHEY: I guess. I mean, I think if 15 you have a material that's stainless steel and it has 16 fluence above a certain level, and we talked about 17 that level this morning, it becomes embrittled and 18 then if subjected to the right kind of load, it can in 19 fact fail. And we can talk about dimple failure, 20 fracture versus --
21 ADMIN. JUDGE WARDWELL: We will in a 22 minute.
23 DR. LAHEY: -- to me that's semantics, it 24 failed.
25 ADMIN. JUDGE WARDWELL: Thank you.
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5170 1 CHAIRMAN MCDADE: A quick question, this is 2 probably going to seem very basic to you all. We've 3 been talking about Austenitic stainless steel, what 4 does that mean? What is Austenitic stainless steel?
5 Dr. Lahey or Dr. Lott or anybody can offer what that 6 definition is as opposed to just stainless steel.
7 MR. GORDON: This is Barry Gordon from 8 Entergy. Austenitic stainless steel is a -- it's 9 named after the fellow who discovered the micro-10 structure. It's a face centered cubic structure.
11 Austenitic stainless steel is not magnetic, a magnet 12 will not stick to it, and it can only be strengthened 13 by cold working. You can't heat treat it like you can 14 some other alloys, like low alloy steel, you can only 15 strengthen it by cold working it to some extent.
16 CHAIRMAN MCDADE: What are the different 17 characteristics of Austenitic stainless steel as 18 opposed to just your typical stainless steel? Are 19 there significant differences?
20 MR. GORDON: Well, there is a ferritic 21 stainless steel and there's a martensitic stainless 22 steel. Martensitic stainless steels can be hardened 23 and that's what -- like you have a stainless steel 24 cutlery at home, those are martensitic stainless 25 steels. And they'll stick, you have a magnetic board NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5171 1 and you can put your knives on it. But Austenitic 2 stainless steel is not a magnetic material.
3 CHAIRMAN MCDADE: Okay. For our purposes 4 right here in the way that it reacts, not how you put 5 it together, not how you work it, but how it reacts 6 over time, is there any specific distinction with it 7 being Austenitic stainless steel?
8 MR. GORDON: It's a very ductile material, 9 it can be embrittled, like Dr. Lahey said, by 10 irradiation, and it can be strengthened by cold 11 working also, and also by irradiation.
12 CHAIRMAN MCDADE: Okay, thank you.
13 ADMIN. JUDGE WARDWELL: As a follow-up --
14 MR. GRIESBACH: Your honor, this is Tim 15 Griesbach --
16 CHAIRMAN MCDADE: I'm sorry, who?
17 ADMIN. JUDGE WARDWELL: A follow-up 18 question on that, which is extremely important to 19 cover. When I buy some stainless steel fittings and 20 stuff, sometimes a magnet will stick, to say, washer 21 fenders and sometimes it won't. Now, it won't be 22 strong, but some of them there's nothing and other 23 times there is and, yet, it's supposed to be a 24 stainless steel. What am I getting?
25 MR. GORDON: It really depends. There's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5172 1 six families of stainless steels and there's also a 2 new family called the super stainless steels that have 3 more alloying elements in it. If you go to like a 4 Best Buy --
5 ADMIN. JUDGE WARDWELL: Yes, I'm going to 6 an Ace Hardware.
7 MR. GORDON: -- that's supposed to be a 8 stainless steel refrigerator and you can bring a 9 magnet if you want -- you really probably want 10 Austenitic stainless steel and that's probably what 11 you want in your sink also. But if you can bring a 12 magnet along, you can see if it's really Austenitic or 13 maybe ferritic stainless steel.
14 ADMIN. JUDGE WARDWELL: But, yet, I've 15 bought these fender washers --
16 MR. GORDON: Yes.
17 ADMIN. JUDGE WARDWELL: -- and some of them 18 it won't stick and other times it'll stick a little 19 bit, it won't stick heavy, but it's not going to be 20 like a --
21 MR. GORDON: Well, some of them, if you 22 cold work Austenitic stainless steel, you get this 23 diffuseness reaction, which you get martensite in the 24 Austenitic stainless steel. Part of it will be 25 transfer -- and you can measure this magnetically.
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5173 1 And this means you're -- it's a way of even measuring 2 the amount of cold working you have by how much 3 martensite you have in the Austenite.
4 ADMIN. JUDGE WARDWELL: So when I bring my 5 magnet to my local Ace Hardware dealer, I can impress 6 him, finally --
7 MR. GORDON: Yes. I mean --
8 ADMIN. JUDGE WARDWELL: -- instead of 9 looking like a fool in the hardware store with my 10 magnet go, oh, well, that one's had some cold working 11 and this is pure --
12 MR. GORDON: Right.
13 ADMIN. JUDGE WARDWELL: -- Austenitic 14 steel.
15 MR. GORDON: That's right. And steel comes 16 --
17 ADMIN. JUDGE WARDWELL: Thank you.
18 MR. GORDON: -- these manufacturers make a 19 stainless steel looking thing, but it's actually made 20 out of carbon steel, but they just put a finish on it.
21 So bring your magnet.
22 ADMIN. JUDGE WARDWELL: I thought we 23 weren't going to get anything out of this hearing and 24 I have been proven wrong.
25 (Laughter.)
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5174 1 ADMIN. JUDGE WARDWELL: Especially when I 2 tell him, gee, I wonder if this is a cast Austenitic 3 stainless steel or a CASS. Which brings us a nice 4 segue into NRC's testimony, 197, Answer 161, Page 92, 5 where the A/LAI, Action Item 7 requires an applicant 6 or a licensee to perform a plant specific analysis of 7 cast RVI components to demonstrate the components will 8 remain capable of performing their intended functions 9 during the period of extended operation. I guess my 10 question for Entergy is that citing your testimony on 11 Answer 175, Page 114, does not the Action Level 7 12 require that this analysis account for the potential 13 loss of fracture toughness of the components due to 14 both thermal embrittlement and irradiation 15 embrittlement?
16 DR. LOTT: Give me a minute to get 17 organized here. Can you give me the citation again?
18 ADMIN. JUDGE WARDWELL: Sure, the citation 19 that it was citing was your testimony, Answer 175, 20 Page 114 in response to that. I'm just asking, does 21 not the A/LAI Number 7, which again you see in MRP 227 22 or in the Aging Management Plan, I may be able to find 23 that, but does not that require, A/LAI 7, does that 24 not require that the analysis account for the 25 potential loss of fracture toughness of the components NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5175 1 due to both thermal embrittlement, or TE as we'll call 2 it, and irradiation embrittlement, IE as we'll call 3 it?
4 MR. DOLANSKY: This is Bob Dolansky for 5 Entergy. The answer's yes.
6 ADMIN. JUDGE WARDWELL: Say again?
7 MR. DOLANSKY: The answer is yes.
8 ADMIN. JUDGE WARDWELL: Okay. It does --
9 MR. DOLANSKY: Yes.
10 ADMIN. JUDGE WARDWELL: -- require that, 11 okay. New York State in their testimony, 576, Page 5, 12 Lines 1 through 17, and through Page 6, Lines 11 13 through 20, that in regards to NUREG/CR-7184, New York 14 State contends that the following observations support 15 previous opinions and testimony. And one of those, 16 the first observation is that for cast materials, 17 synergies may exist between TE, the thermal 18 embrittlement, and the irradiation embrittlement.
19 And, two, embrittled cast materials were observed to 20 experience transgranular brittle cleavage and ductile 21 tearing. And I guess the first question I will have 22 for Entergy is, what does transgranular brittle 23 cleavage and ductile tearing look like and how are 24 they formed and when are they formed and when are they 25 an issue?
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5176 1 DR. LOTT: Okay. We talked earlier about 2 Austenitic stainless steels and ferritic stainless 3 steels, or ferritic steels in general. For instance, 4 a reactor pressure vessel steel is a ferritic steel.
5 It's subject to a brittle fracture at low temperatures 6 that cuts across the grain because it's a very flat 7 surface, it's essentially a cleaving of the crystal 8 structure in the grain, a very brittle failure. That 9 does not tend to happen in the Austenitic stainless 10 steels where the failures remain ductile, they fail by 11 stretching of the material and eventually finding 12 small dimples or ruptures in the material and pulling 13 it apart. So there's a difference in the fracture 14 process between a cleavage failure, which is a very 15 low ductility, lower shelf reactor pressure vessel 16 kind of failure, and the failures we see normally in 17 stainless steels.
18 ADMIN. JUDGE WARDWELL: But say again, what 19 is that transgranular brittle cleavage look like? It 20 looks like a cut surface or a sheared surface or --
21 DR. LOTT: Yes. There's no deformation on 22 the surface, it's very clearly flat and sometimes 23 stepped.
24 ADMIN. JUDGE WARDWELL: So it just looks 25 like a break in the surface?
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5177 1 DR. LOTT: Yes, it's a very -- it's a clean 2 break, I guess is the best way to say it.
3 ADMIN. JUDGE WARDWELL: A crevice in a rock 4 type of thing?
5 DR. LOTT: What's the best way to describe 6 this?
7 ADMIN. JUDGE WARDWELL: Or a crack in a 8 rock?
9 DR. LOTT: Yes. Certain rocks you would 10 fail by cleavage, that's true.
11 ADMIN. JUDGE WARDWELL: And so what is 12 ductile tearing? Is that what the --
13 DR. LOTT: Ductile tearing is the manner in 14 which a crack in a ductile material would advance. So 15 when this begins to fail, you begin to slowly -- it 16 would continue to deform or continue to show some --
17 reconnected and you'd form small voids that would 18 grown into little pockets that would give you these 19 dimpled rupture effects on the surface of the 20 specimen.
21 ADMIN. JUDGE WARDWELL: And so usually 22 you'd say this ductile tearing is associated with 23 dimples at the surface?
24 DR. LOTT: Yes. You can tell by looking at 25 the surface that it's failed in this manner, under a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5178 1 microscope, not necessarily -- well, you might be able 2 to tell by eye.
3 ADMIN. JUDGE WARDWELL: Okay. And this 4 dimpled surface will appear before or after the 5 tearing? I mean, can't you just tell by the split in 6 the tearing or do you see the tear or --
7 DR. LOTT: Well, you can tell, obviously, 8 by the load -- if you're tearing this part and you're 9 measuring the load displacement, you can see that it's 10 continually taking new force to pull it apart.
11 Whereas a brittle fracture, a cleavage fracture, will 12 obviously be real sudden.
13 ADMIN. JUDGE WARDWELL: So a transgranular 14 brittle cleavage would be a definitive break, is that 15 a fair assessment? Where the ductile tearing, you may 16 not see a separation of the material, it just may be 17 a necking down and possibly this dimpling on the 18 surface that you talk about?
19 DR. LOTT: Well, eventually you will form, 20 after it's necked down, you'll begin eventually to 21 form cracks and those cracks will have these dimpled 22 ruptured surface. We're talking about the fracture 23 surface of the specimen, when it separates and you 24 look at it, that's where you'd see the dimples.
25 ADMIN. JUDGE WARDWELL: Okay, thank you.
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5179 1 MR. STROSNIDER: Your honor, this is Jack 2 Strosnider from Entergy. And maybe this will help you 3 visualize it. When you have a cleavage fracture, it's 4 flat because when it's transgranular, it's going along 5 the atoms. Basically it's a shiny, when you look at 6 that surface when it fails, it's shiny because it's 7 very flat. When you have the ductile failure, because 8 of the dimpled, because you have to break little 9 pieces of material, it's got a duller surface and you 10 can see it's not the flat sort of surface that you 11 would see in a cleavage fracture.
12 ADMIN. JUDGE WARDWELL: But both of them 13 are failure surfaces though that you're looking at?
14 MR. STROSNIDER: Oh, yes.
15 ADMIN. JUDGE WARDWELL: Okay.
16 MR. STROSNIDER: It comes apart.
17 ADMIN. JUDGE WARDWELL: Great. Thank you.
18 Now, do you agree with New York's statements in 19 regards to the observations from this NUREG 7184 that 20 the Cast materials synergies may exist between thermal 21 embrittlement and irradiation embrittlement and, two, 22 the embrittled Cast materials were observed to 23 experience these two types of failures?
24 DR. LOTT: I'm not sure that I -- I don't 25 think I do agree with the term synergism. I think NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5180 1 that has a meaning to me that I don't see proven in 2 any of the data. I do agree that irradiation 3 embrittlement and thermal embrittlement may both --
4 either one may happen in a Cast material.
5 ADMIN. JUDGE WARDWELL: Either one or both?
6 DR. LOTT: Well, certainly a material can 7 be subject to both these conditions at the same time.
8 It's just not clear that they interact to me 9 synergistically.
10 ADMIN. JUDGE WARDWELL: Okay. Dr. Lahey, 11 I want to ask you, did the NUREG 7181 use the term 12 synergy between them or how did you -- what did they 13 say in regards to the reference, I guess, to your 14 first statement that for Cast materials synergies may 15 exist between TE and IE?
16 DR. LAHEY: Your honor, this is Richard 17 Lahey, New York. I would actually have to go back and 18 look at it to see if it used that word. This word is 19 used often, in fact it was one of the things that 20 Judge McDade gave me as a homework to read something 21 and it asked about the synergy between these two 22 effects for cast stainless steel. And so it's not 23 something I made up. It may or may not be in NUREG 24 7184, I'd have to check.
25 ADMIN. JUDGE WARDWELL: So that could be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5181 1 your wording in regards to interpreting what they said 2 --
3 DR. LAHEY: Yes. I mean, my --
4 ADMIN. JUDGE WARDWELL: -- in regards to 5 the interaction between the TE and the IE?
6 DR. LAHEY: My personal belief is that you 7 certainly can have both of them and the effect can be 8 larger than each one separate. That's what I mean by 9 synergy.
10 ADMIN. JUDGE WARDWELL: And say that again, 11 what you mean by synergy?
12 DR. LAHEY: Synergy means that they 13 reinforce each other and the combination is more than 14 each effect separate.
15 ADMIN. JUDGE WARDWELL: But wouldn't it be 16 more than just the sum also?
17 DR. LAHEY: Yes.
18 ADMIN. JUDGE WARDWELL: Because that would 19 be the synergy --
20 DR. LAHEY: That's what I meant, I'm sorry.
21 ADMIN. JUDGE WARDWELL: -- you take this 22 TE, you take IE and yes it's more than either one of 23 them --
24 DR. LAHEY: Right.
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5182 1 it's more than the sum of the whole.
2 DR. LAHEY: Exactly.
3 ADMIN. JUDGE WARDWELL: Okay. Thank you.
4 DR. LOTT: Your honor --
5 ADMIN. JUDGE WARDWELL: Yes.
6 DR. LOTT: -- I think in that document, 7 7184, I have some notes on it here. It basically did 8 show that irradiation, relatively low doses 0.08 dpa, 9 can trigger a response in the material that similar to 10 thermal embrittlement, these materials that are 11 normally subject to thermal embrittlement. And what 12 happened effectively was that thermal embrittlement by 13 itself would produce an effect, the irradiation 14 produced a similar effect without the thermal 15 embrittlement, materials that were combined showed 16 about the same. So that's why I'm concerned that they 17 were not necessarily a synergistic effect. Yes, there 18 was an effect of thermal embrittlement, yes, there was 19 an effect of irradiation embrittlement, but to me, the 20 suggestion that it's a synergistic effect would say 21 that it could be larger combined than it is 22 individually.
23 ADMIN. JUDGE WARDWELL: Okay thank you.
24 CHAIRMAN MCDADE: Are you saying that it 25 isn't or you don't know?
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5183 1 DR. LOTT: I just don't -- it doesn't meet 2 my definition for what I would call synergistic.
3 CHAIRMAN MCDADE: Okay. As I understood 4 that, the definition you would use is that it's larger 5 than the sum of the individual parts.
6 DR. LOTT: Right. If I had both, in the 7 specimen if I had both irradiation and thermal 8 embrittlement, the decrease was about the same as it 9 was just due to irradiation and thermal embrittlement 10 alone. I think that's Figure 142 in the document.
11 That's probably 14.2, my handwriting is very bad, I'm 12 sorry.
13 ADMIN. JUDGE WARDWELL: And, Dr. Lahey, I 14 will leave you with this comment, that if you would 15 like to explore 7184 and come up with the wording that 16 they use that would support the synergistic effects of 17 this, meaning that there's some words in there that 18 state that the TE and the IE amplify one another to 19 the point that the effect is greater than the sum of 20 the two, then feel free to present that to us --
21 DR. LAHEY: Yes, sir.
22 ADMIN. JUDGE WARDWELL: -- whenever you 23 have a chance to review that further within the 24 context of why we're here this week.
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5184 1 though, I've read that in a number of Argonne reports 2 from researchers at Argonne.
3 ADMIN. JUDGE WARDWELL: Well, I'm mostly 4 interested in -- I'm only interested in this regard in 5 --
6 DR. LAHEY: Okay.
7 ADMIN. JUDGE WARDWELL: -- regards to this 8 as applied to 7184. Thank you.
9 DR. LAHEY: Thank you.
10 CHAIRMAN MCDADE: Okay. And, Dr. Lott, you 11 referenced Figure 142? Is that what you said?
12 DR. LOTT: We're looking it up right now.
13 Can we get back to perhaps in a --
14 CHAIRMAN MCDADE: Sure.
15 ADMIN. JUDGE WARDWELL: Entergy's 16 testimony, 616 again, Answer 106, Page 64, although 17 stainless steel and nickel alloy RVI materials are 18 also subject to irradiation embrittlement, they do not 19 undergo a ductile to brittle transition or fail by 20 brittle cleavage, even though the neutron exposure 21 levels are much higher than those of the vessel.
22 However, at fluences above the MRPA 175 screening 23 threshold, it is recognized that these Austenitic 24 stainless steels will experience decreases in fracture 25 initiation toughness and in the resistance to ductile NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5185 1 tearing. These effects have been explicitly 2 considered in the MRP 227-A guidelines and in the RVI 3 AMP implementation at Indian Point. And I guess my 4 question for Entergy, I just want to confirm that your 5 use of the word brittle cleavage is the same as this 6 transgranular brittle cleavage that we were talking 7 about earlier?
8 MR. STROSNIDER: Your honor, this is Jack 9 Strosnider for Entergy. That's correct, but --
10 ADMIN. JUDGE WARDWELL: Thank you.
11 MR. STROSNIDER: -- I think the point here 12 that needs to be understood, the point that's being 13 made, we continue to discuss about highly embrittled 14 materials. These Austenitic materials are embrittled 15 and there is a reduction in fracture toughness. The 16 thing you need to understand is when they test that 17 fracture toughness, it still shows ductility, it's not 18 failing by cleavage fracture. They're developing 19 what's called a J-R curve, which requires some 20 ductility in order to fail it. So, I just want to put 21 this highly embrittled in context. There's 22 embrittlement, but there's still ductility in these 23 specimens, they're not failing by cleavage fracture, 24 they don't go through the transition that the ferritic 25 materials go through. It's very important in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5186 1 understanding how these materials can respond to 2 various loads, including the accident loads.
3 ADMIN. JUDGE WARDWELL: And in this Answer 4 106 on Page 64, you say that they do not undergo 5 ductile to brittle transition or failure by, I'm 6 inserting now, the transgranular brittle cleavage, 7 correct?
8 MR. STROSNIDER: Right.
9 ADMIN. JUDGE WARDWELL: So they don't fail 10 by brittle cleavage, but they do still have resistance 11 to ductile tearing, is that correct?
12 MR. STROSNIDER: Jack Strosnider for 13 Entergy. Yes, that is correct.
14 ADMIN. JUDGE WARDWELL: Thank you. So in 15 summary, the aging effects of RVIs do not include 16 transgranular brittle cleavage, but do include ductile 17 tearing. But how about some of the other effects that 18 we've talked about? And I think they include 19 cracking, dimensional changes, wearing, dimpling, 20 stress relaxation, and void swelling cracking. Do 21 they exhibit those types of effects? And we can stay 22 with you, Mr. Jack, I can't see your name so I can't 23 --
24 MR. STROSNIDER: Jack Strosnider for 25 Entergy. Yes, they do. Those are the type of effects NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5187 1 that are identified in the program and managed by the 2 program.
3 ADMIN. JUDGE WARDWELL: Okay, thank you.
4 Entergy's testimony, 722, Answer 8, Page 5, says the 5 existing research also suggests that combined thermal 6 aging and irradiation of representative CASS materials 7 does not appear to lower toughness below what is 8 expected for thermal embrittlement alone. And I guess 9 I'd ask, what's the basis for this statement? Of 10 anyone from Entergy that would like to respond.
11 DR. LOTT: Let me first respond by 12 correcting my previous figure number. It's Figure 98 13 on Page 142. I had them switched in my notes.
14 ADMIN. JUDGE WARDWELL: Okay. Thank you.
15 DR. LOTT: And in fact, that figure I think 16 actually makes the same point that he just said and 17 that I said earlier, which was that the effect of 18 thermal embrittlement and irradiation embrittlement in 19 these high ferrite materials, and we haven't talked 20 about necessarily high ferrite and low ferrite 21 materials, these relatively high ferrite materials do 22 show thermal aging and irradiation caused a similar 23 effect.
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5188 1 should point out is that the materials we're talking 2 about at Indian Point in the lower support columns are 3 known to be low in delta ferrite. It's that delta 4 ferrite content that causes the thermal embrittlement 5 that we keep referring to and the materials in these 6 plants are known to not be susceptible to thermal 7 embrittlement based on their low delta ferrite. So, 8 by saying that, they should also not be affected by 9 the synergistic effects of thermal and irradiation 10 embrittlement.
11 ADMIN. JUDGE WARDWELL: Okay. Let me, I 12 guess -- I'm getting swamped by too much information 13 here, I think. And then diversion over to one chart 14 and then I didn't know whether they answered my 15 question or not is the way I interpreted it. You're 16 testimony, Entergy's testimony, 722, Answer 8, Page 5, 17 says the existing research also suggests that the 18 combined thermal aging and irradiation of 19 representative CASS materials does not appear to lower 20 toughness below what is expected for thermal 21 embrittlement alone. And my question to you is, what 22 is the basis for that statement?
23 DR. LOTT: And, again, I think it's 24 illustrated -- I was trying to answer your question 25 with my previous answer. In the data that was shown, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5189 1 it shows exactly what --
2 ADMIN. JUDGE WARDWELL: And I'm sorry, I 3 can't --
4 DR. LOTT: The data that's shown in that 5 figure shows that you start -- and again, for three 6 different materials, all high ferrite materials, where 7 thermal embrittlement would potentially happen, they 8 were either irradiated or thermally treated and the 9 decrease in toughness in the J-integral toughness was 10 basically the same in both cases. In other words, the 11 thermal embrittlement and the irradiation 12 embrittlement and the three of them together all gave 13 very similar results.
14 ADMIN. JUDGE WARDWELL: But the cast 15 materials that we're dealing with aren't high ferric 16 are they?
17 DR. LOTT: The cast materials we're dealing 18 are low ferrite, so we would not expect to see thermal 19 embrittlement, nor would we expect to see any large 20 amount of irradiation embrittlement by the same 21 mechanism.
22 ADMIN. JUDGE WARDWELL: So why does that 23 figure, which deals with high ferric, provide a basis 24 for your statement that the existing research suggests 25 that combined thermal aging and irradiation of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5190 1 representative CASS materials did not appear to lower 2 toughness below what is expected for thermal 3 embrittlement alone?
4 MR. STROSNIDER: This is Jack Strosnider 5 for Entergy and let me see if I can clarify this.
6 ADMIN. JUDGE WARDWELL: Sure.
7 MR. STROSNIDER: I think the first part of 8 the answer that Dr. Lott gave is a generic discussion 9 about the laboratory data --
10 ADMIN. JUDGE WARDWELL: Fine.
11 MR. STROSNIDER: -- which includes high 12 delta ferrite materials. When we talk about the 13 material at Indian Point, in particular the lower 14 columns, they are low delta ferrite not susceptible to 15 thermal aging, and, therefore, this -- what you see in 16 the generic data is really not applicable at Indian 17 Point because of the specific material that's at 18 Indian Point. So a generic part of the response and 19 a plant specific part of the response.
20 ADMIN. JUDGE WARDWELL: And so the plant 21 specific, you've made a statement that, again, I'll 22 read for the third time. The existing research also 23 suggests that combined thermal aging and irradiation 24 of representative CASS materials, which is what you 25 have at Indian Point, correct?
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5191 1 MR. STROSNIDER: Read that again, please?
2 MR. GRIESBACH: This is Tim Griesbach. The 3 --
4 ADMIN. JUDGE WARDWELL: CASS material, you 5 have -- you're dealing with CASS material. For those 6 CASS materials at Indian Point, do not appear to lower 7 toughness below what is expected for thermal 8 embrittlement alone. That was your statement. My 9 question is, where did that come from?
10 MR. DOLANSKY: This is Bob --
11 ADMIN. JUDGE WARDWELL: What is the basis 12 for it? Point me to a graph that demonstrates not 13 what it isn't, I want to see a graph for what it is or 14 some other discussion of why you came up with that 15 statement or how you came up with that statement.
16 MR. DOLANSKY: This is Bob Dolansky with 17 Entergy. I believe the graph that Dr. Lott pointed to 18 contains CF-8 material and that is --
19 ADMIN. JUDGE WARDWELL: Whoa, what's CF-8 20 material? Now we've got another material.
21 (Laughter.)
22 MR. DOLANSKY: -- which that is what we 23 have at Indian Point.
24 ADMIN. JUDGE WARDWELL: That's CASS?
25 MR. DOLANSKY: That's CASS and it's --
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5192 1 ADMIN. JUDGE WARDWELL: Great.
2 MR. DOLANSKY: -- low delta ferrite.
3 ADMIN. JUDGE WARDWELL: Yes, great.
4 MR. DOLANSKY: That is, I believe, and Dr.
5 Lott can correct me if I'm wrong, but that figure that 6 he gave you contains the low delta ferrite CF-8 7 material that's CASS that we have at Indian Point.
8 ADMIN. JUDGE WARDWELL: Okay. And you're 9 referring to Figure 98 on Page 142?
10 MR. DOLANSKY: Yes.
11 ADMIN. JUDGE WARDWELL: Okay. Because by 12 the time we get with putting the record together and 13 look at the transcript, we want to make sure we have 14 it.
15 MR. DOLANSKY: This figure here.
16 ADMIN. JUDGE WARDWELL: And so it's on our 17 screen now?
18 DR. LOTT: Yes.
19 ADMIN. JUDGE WARDWELL: Okay. Right? Is 20 that the one you're referring to?
21 MR. DOLANSKY: Yes.
22 DR. LOTT: Yes.
23 ADMIN. JUDGE WARDWELL: Okay. So how does 24 that support this statement that CASS materials do not 25 appear to lower toughness below what is expected for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5193 1 thermal embrittlement alone?
2 DR. LOTT: Well, I'm not sure what part of 3 your question I'm missing. I mean, if you look at the 4 data here, both irradiation and aging produce a 5 similar decrease in toughness and the material that is 6 both irradiated and aged has a very similar, 7 particularly in the CF-8 material, is a similar 8 toughness. Now, in the CF-8 materials in Indian 9 Point, we don't expect to see these decreases due to 10 thermal aging because the CF-8 material there is low 11 ferrite.
12 MR. COX: This is Alan Cox with Entergy.
13 And let me give you my layman's interpretation of what 14 this drawing shows. It shows --
15 ADMIN. JUDGE WARDWELL: Sure, just get a 16 little closer to your mic though because I can hear 17 you better.
18 MR. COX: The tall bar on the graph is the 19 material at the beginning and then the arrows show --
20 ADMIN. JUDGE WARDWELL: And we're looking 21 at the blue tall bar, which we're going to focus only 22 on the CF-8, is that correct?
23 MR. COX: Sure. I'm just -- the response 24 is similar in all three cases. But if you --
25 ADMIN. JUDGE WARDWELL: Sure.
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5194 1 MR. COX: -- the center one, the blue tall 2 bar is the starting condition. The dotted line that 3 says irradiation shows the decrease in the fracture 4 toughness as you irradiate the material. The dotted 5 line that says aging is the thermal aging. And you 6 can see it lowers -- both of those effects cause a 7 drop in the fracture toughness. The third bar, you 8 see another irradiation on the dotted line that goes 9 from the shorter blue bar to the pink/red colored bar 10 in front, that's the additional effect, that's 11 basically the combined effect of the --
12 ADMIN. JUDGE WARDWELL: Got you.
13 MR. COX: -- thermal aging and then the 14 irradiation. And you see it has a slightly lower 15 value than the irradiation alone, a little bit lower 16 than the thermal alone, but it's certainly not greater 17 than the sum of the two effects.
18 ADMIN. JUDGE WARDWELL: And what is CF-3 19 material?
20 DR. LOTT: It's just a different 21 specification of cast stainless steel.
22 ADMIN. JUDGE WARDWELL: Okay. So all of 23 these are cast stainless steel materials?
24 DR. LOTT: Yes.
25 ADMIN. JUDGE WARDWELL: Dr. Lahey, any NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5195 1 comments on how these graphs apply to what CASS 2 materials appear to do in regards to lowering 3 toughness below what is expected for thermal 4 embrittlement alone when you combine thermal aging 5 with irradiation of those?
6 DR. LAHEY: Just a request for a little bit 7 more information from Entergy. My understanding is 8 the delta ferrite in IP2 is like 14.6 percent, is that 9 correct?
10 ADMIN. JUDGE WARDWELL: Well, why don't I 11 ask the question. You tell me what you're interested 12 in and -- I want to make sure you're not --
13 DR. LAHEY: Well, the screening criteria is 14 15 percent and I'm trying to understand if that's what 15 they mean by low delta ferrite, the 14.6.
16 ADMIN. JUDGE WARDWELL: What do you mean by 17 low delta ferrite?
18 MR. AZEVEDO: Yes, your honor. This is 19 Nelson Azevedo for Entergy. It is true that both 20 Units 2 and 3 have low delta ferrite as measured by 21 the screening criteria 15 percent. So they're both 22 below 15 percent.
23 ADMIN. JUDGE WARDWELL: And do you have any 24 idea how far below?
25 MR. AZEVEDO: Yes. Unit 2 is 14 and NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5196 1 change, Unit 3 is a little bit lower, like 11, 12.
2 ADMIN. JUDGE WARDWELL: Okay. Thank you.
3 Dr. Lahey, any other comments or --
4 DR. LAHEY: Yes. My comment is I wish my 5 colleague Dr. Duquette was here because this is his 6 field, he's a world class metallurgist. And he would 7 have a lot of comment on this and all I can give you 8 is secondhand information because this is not my 9 field.
10 ADMIN. JUDGE WARDWELL: But he's not even 11 a witness for this Contention, so even if he was here, 12 he would be a spectator.
13 DR. LAHEY: Okay. Well, I can tell you 14 what he would say if you want to hear that.
15 ADMIN. JUDGE WARDWELL: No. No, I'd like 16 to hear --
17 CHAIRMAN MCDADE: Just tell us what you 18 would say.
19 ADMIN. JUDGE WARDWELL: -- what your 20 professional interpretation might be of anything, but 21 only from what you're comfortable saying in regards to 22 your professional background.
23 DR. LAHEY: Well, concerning the data that 24 we have on the screen, the data speaks for itself. If 25 this is good data, that's what it is.
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5197 1 ADMIN. JUDGE WARDWELL: Okay. Thank you.
2 CHAIRMAN MCDADE: Okay. It may speak for 3 itself, but it doesn't speak loud enough for me to 4 understand it. Okay. We start off --
5 ADMIN. JUDGE WARDWELL: You're never going 6 to understand this chart.
7 (Laughter.)
8 CHAIRMAN MCDADE: Okay. The low delta 9 ferrite, start from the premise that it's below 15 10 percent, it's 14 or 11 percent. How does that inform 11 your conclusion here?
12 DR. LAHEY: Well, if you want me to give 13 you some secondhand information, then I can do that.
14 But I can't give you any firsthand information other 15 than looking at the graph and what it shows to me is 16 there's obviously no synergism shown in this data.
17 CHAIRMAN MCDADE: Okay. There's no 18 synergism shown in this data, but from your 19 experience, do you believe when you have the low delta 20 ferrite that you would expect synergism?
21 DR. LAHEY: Not necessarily. I think it 22 depends on how the duplex structure is arranged. If 23 it's arranged in a certain way, it can have a 24 different effect than if it's arranged in a different 25 way. And unfortunately the only way you know that is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5198 1 with a destructive examination. So, I think you do 2 have to rely on data that shows, here -- when I am 3 talking about synergism, there's a number of other 4 experiments, not necessarily for these type of 5 percentages of delta ferrite, which do show synergism.
6 But it may be, apparently is, for the materials used 7 in Indian Point, this is the result you get.
8 CHAIRMAN MCDADE: So you're saying that 9 based on data that you have observed, if the 10 percentage was not 11 percent, but was 20 percent or 11 40 percent, it would have a different effect on your 12 conclusion with regard to synergism?
13 DR. LAHEY: That's my understanding of the 14 reading that I've done, yes.
15 CHAIRMAN MCDADE: Okay, thank you.
16 ADMIN. JUDGE KENNEDY: Dr. Lahey, this is 17 Judge Kennedy. If I looked at the material CF-3, look 18 at the data there, does that show more synergism than 19 CF-8 does?
20 DR. LAHEY: Well, if I understand the way 21 they're doing it, it would show less. The one on the 22 right, the little pink or red or whatever color that 23 is on the right is lower than the one on the left. So 24 it's a little odd that the thermal aging plus 25 irradiation would give you lower, but it depends on NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5199 1 the level of irradiation, I suppose.
2 ADMIN. JUDGE KENNEDY: All right, thank 3 you.
4 ADMIN. JUDGE WARDWELL: Yes. I guess that 5 raises a question I hadn't really come up with. What 6 is this J-integral that we're plotting on the Y axis?
7 DR. LOTT: It's a measure of the ductile 8 fracture toughness.
9 ADMIN. JUDGE WARDWELL: So that's the 10 measure of the toughness?
11 DR. LOTT: The toughness being the 12 resistance to crack initiation. And, again, I'll use 13 that word, let me qualify. In a J-integral test, you 14 would start with a crack specimen and you would 15 basically be measuring how much it would take to 16 reinitiate and grow that crack.
17 ADMIN. JUDGE WARDWELL: To reinitiate and 18 what that crack?
19 DR. LOTT: Grow it.
20 ADMIN. JUDGE WARDWELL: Grow that crack.
21 So what do low values of J mean? That it's not 22 susceptible to cracking or susceptible to crack or --
23 DR. LOTT: It's not the susceptibility to 24 cracking, it's susceptibility -- the impact of 25 cracking on the ability to maintain a load.
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5200 1 ADMIN. JUDGE WARDWELL: On the ability --
2 so what are --
3 MR. STROSNIDER: This is Jack --
4 ADMIN. JUDGE WARDWELL: -- the low values 5 mean?
6 MR. STROSNIDER: This is Jack Strosnider 7 for Entergy, let me see if -- first of all, in terms 8 of the J-integral and what it is, it can be related to 9 the amount of energy that's required to allow the 10 crack to tear through the material, the preexisting 11 crack. It can be related to the energy for that. So 12 the lower the J value, the less energy it takes. But 13 this J-integral approach was developed specifically 14 for ductile materials. If the material is not 15 ductile, if it's going to fail in the cleavage mode 16 that we talked about earlier, there's a different 17 measure that's used for that. So, in this case, it's 18 related to energy, the lower that value, the lower the 19 energy. But in every case, it's showing some 20 ductility.
21 ADMIN. JUDGE WARDWELL: And so, while the 22 CF-8 shows that we reach the same level of the J-23 integral with irradiation as we do with aging and 24 irradiation, the CF-3 material shows that we end up 25 with a lower value with the combination of the aging NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5201 1 and the irradiation --
2 MR. STROSNIDER: Right.
3 ADMIN. JUDGE WARDWELL: -- than just 4 irradiation, which means less energy is needed to grow 5 the crack. Am I interpreting that correctly?
6 DR. LOTT: I supposed you could interpret 7 it that way. I must admit, I think it's within the 8 scatter in the data, within the accuracy of the data.
9 To make that conclusion would be difficult in my mind.
10 CHAIRMAN MCDADE: Okay, Dr. Lott, excuse 11 me. Could you either move yourself closer to the 12 microphone or the microphone closer to you?
13 DR. LOTT: Okay, sorry.
14 CHAIRMAN MCDADE: Okay, thank you.
15 ADMIN. JUDGE WARDWELL: So you think that 16 the data is plus or minus almost 100 percent because 17 it seems to be about half of what it is before and if 18 that's the noise, then that's accuracy of what we're 19 dealing with here in this graph? Can you see how I 20 reached that conclusion?
21 DR. LOTT: Yes, I see how you reached that 22 conclusion. It's hard for me to answer that question.
23 ADMIN. JUDGE WARDWELL: Fine, thank you.
24 DR. LOTT: Again, I think it's important to 25 point out that basically the materials that we're NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5202 1 dealing with, we don't expect to see large amounts of 2 thermal embrittlement at all. Now, these materials 3 are higher in ferrite and we would expect to see that 4 happen. I think all materials measured at least at 5 levels greater than 20 percent.
6 ADMIN. JUDGE WARDWELL: Boy, you had me 7 right to the very end. These materials, what are you 8 referring to as these materials?
9 DR. LOTT: The materials in this graph, the 10 CF-3, the CF-8, and the CF-8M. The particular 11 materials that were tested.
12 ADMIN. JUDGE WARDWELL: But all of these --
13 now we're back to where we started I think.
14 DR. LOTT: I'm sorry.
15 ADMIN. JUDGE WARDWELL: What is the ferric 16 content of these materials on this graph?
17 DR. LOTT: The ferrite content in all three 18 cases I believe is greater than 20 percent, measured.
19 MR. DOLANSKY: Your honor, maybe I could 20 help a little bit. This is Bob Dolansky with Entergy.
21 CF-8 can have a range of delta ferrite. Indian Point 22 3 has CF-8 materials. We actually went and did the 23 research, pulled our CMTRs, our delta ferrite was 24 below 15 percent at both IPEC units. That doesn't 25 mean that all CF-8 is below 15 percent. Does that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5203 1 help?
2 ADMIN. JUDGE WARDWELL: And so, if I looked 3 at the rest of this --
4 MR. COX: This is Alan Cox --
5 ADMIN. JUDGE WARDWELL: -- Exhibit 488B, is 6 that what it is? And that is an 83 page report, it 7 would tell me what the ferric content of this is?
8 DR. LOTT: I believe so.
9 MR. COX: Judge Wardwell, this is Alan Cox 10 for Entergy. The paragraph immediately above this 11 graph says that all the samples tested were high delta 12 ferrite samples. And the report may say somewhere 13 what that is in terms of a number, but it does say 14 it's high delta ferrite.
15 DR. HISER: Dr. Wardwell, the information 16 you're looking for is in New York State 488A, Page 5, 17 Table 1, provides chemical composition of the 18 materials. And for the CF-3, it's measured 24 19 percent, CF-8 measured 23 percent, and the 8M is 28 20 percent.
21 ADMIN. JUDGE WARDWELL: So would -- now I'm 22 going to go back to Entergy. Thank you for that, Dr.
23 Hiser, from the Staff. Back to Entergy, you would 24 call these samples that were used to generate this 25 figure in 488 high ferric content samples?
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5204 1 DR. LOTT: And perhaps I could offer 2 further explanation. In the evaluation not in the 3 reactor internals, but in other reactor components, 4 particularly piping components, the guidelines are 5 effectively to consider thermal embrittlement in 6 materials containing more than 20 percent ferrite.
7 So, in our minds I guess, that is the threshold. I'm 8 not sure there's an absolute threshold, but certainly 9 there's no requirement in analyzing piping materials 10 to consider the loss of toughness due to thermal aging 11 for materials that are less than 20 percent ferrite.
12 ADMIN. JUDGE WARDWELL: So this graph is 13 not representative of the CASS materials we have in 14 the reactor vessel internals at Indian Point?
15 DR. LOTT: No, because we would not in that 16 case expect to see thermal embrittlement, or certainly 17 not this level of thermal embrittlement.
18 ADMIN. JUDGE WARDWELL: So, which gets me 19 back to the original question I had.
20 DR. LOTT: Well, I guess I could only 21 suggest that the only data had to offer one way or the 22 other on the original question, which was is there a 23 synergistic effect, was this data. I don't see, 24 again, in this data -- what it suggests to me quite 25 frankly is that the properties of the duplex steel, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5205 1 this material that has 80 percent Austenite and 20 2 percent ferrite or whatever, is affected by the 3 embrittlement of the ferrite phase. There are two 4 ways that you could end up embrittling the ferrite 5 phase. One is by thermal aging, the other is by 6 irradiation. This would indicate that it doesn't 7 really matter what you do to embrittle that phase, 8 once you have it embrittled, you have a similar effect 9 on the material itself. But that's an interpretation 10 --
11 CHAIRMAN MCDADE: Okay. Following --
12 DR. LOTT: -- the best I can give you.
13 CHAIRMAN MCDADE: Following up, Dr. Lott, 14 using this chart. The delta ferrite for the CF-8, 23 15 percent based on the testimony that Dr. Hiser just 16 gave, when you look at the difference between the 17 effect of irradiation and aging plus irradiation, you 18 have very little difference, minimal, drawing your 19 conclusion that there is no synergistic effect or 20 minimal synergistic effect. In the Indian Point 21 situation, we have, at IP3, the delta ferrite is 22 approximately 11 percent. So this chart suggests to 23 you there is no synergistic effect, but the fact that 24 there's a lower delta ferrite, meaning it would be 25 even less susceptible to the heat aging, would NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5206 1 indicate to you that there would be even a lesser 2 impact on the metal used at Indian Point 2 and 3 at 14 3 and 11 percent?
4 DR. LOTT: Yes.
5 CHAIRMAN MCDADE: Am I --
6 DR. LOTT: Yes. I --
7 CHAIRMAN MCDADE: I just want to -- I 8 repeat this just to make sure that I'm hearing what 9 you're saying and there's not something --
10 DR. LOTT: I believe you've well 11 interpreted what I'm trying to say.
12 CHAIRMAN MCDADE: Okay, thank you.
13 ADMIN. JUDGE WARDWELL: But then I'll go 14 back to my original question based on your original 15 statement in your testimony, which this is where I'm 16 trying to get to. This chart does not apply to Indian 17 Point materials, correct?
18 DR. LOTT: No, we have no measurements on 19 the Indian Point materials.
20 ADMIN. JUDGE WARDWELL: The statement that 21 was made on your testimony, 722, Answer 8, Page 5, 22 you've made the statement, the existing research also 23 suggests that combined thermal aging and irradiation 24 of representative CASS materials, and I assume you 25 mean representative of what's there at Indian Point, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5207 1 does not appear to lower toughness below what is 2 expected for thermal embrittlement alone. And so my 3 question is, what is the basis for that statement?
4 And if you're going to use this figure, I'd like to 5 know why you can use this figure to extrapolate for 6 something that isn't representative of what's at 7 Indian Point.
8 MR. COX: Let me take a shot at that and 9 then Dr. Lott can chime in. But I think the way that 10 I would look at that graph is that if you have little 11 thermal aging, I mean, you can imaging the aging line 12 on the graph would show much less of a decrease than 13 what you show there. So it seems very reasonable to 14 say if you have a lot of irradiation embrittlement, a 15 lot of thermal embrittlement, and you see very little 16 difference in the combined effect than you do from the 17 irradiation effect, if you reduce the thermal aging 18 embrittlement, like you would expect to see from 19 material with a low delta ferrite, you would have no 20 reason to expect a larger difference when you combine 21 irradiation with the thermal.
22 So, I mean, I think this is not the exact 23 same material, but knowing what we know about the 24 behavior of the material under irradiation and thermal 25 embrittlement, we would expect the same result. And NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5208 1 that is very little change when you combine the two 2 effects since the thermal embrittlement would be even 3 less of an effect, there would be very little change 4 from the combined effect and what you see from the 5 irradiation effect alone.
6 DR. LOTT: Can I request -- I offered this 7 slide because I thought it was going to quickly 8 address your issue, obviously it did not. If we could 9 have some time to look at the answer to your question 10 and come back to you with it, because I think it's 11 going to require us to go through more than one 12 document in order to put the answer together.
13 ADMIN. JUDGE WARDWELL: That's fine.
14 DR. LAHEY: Your honor --
15 ADMIN. JUDGE WARDWELL: Yes.
16 DR. LAHEY: -- could I say, if I understand 17 this graph correctly now, the CF-3 shows the synergism 18 that you are asking about, whereas the CF-8 does not.
19 And so I think that's what they're talking about, but 20 for a lower percentage. But over here you can clearly 21 see the synergism in terms of the integral or the J, 22 it puts the onset of the crack occurs.
23 ADMIN. JUDGE WARDWELL: In your opinion?
24 DR. LAHEY: Well, if I believe the heights, 25 you go down with irradiation, you go down with thermal NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5209 1 aging, and if you add on the irradiation, you're 2 significantly below than just each one.
3 CHAIRMAN MCDADE: But what you're saying, 4 and again, I'm repeating this to make sure I hear what 5 you're saying and understand it, is that the CF-3 6 shows the synergy even though the CF-8 does not on 7 this graph.
8 DR. LAHEY: It appears so.
9 CHAIRMAN MCDADE: Okay. And would you 10 explain the rationale for, theoretically, why the CF-3 11 would show apparently some synergistic effect whereas 12 the CF-8 did not? Dr. Lott?
13 DR. LOTT: Well, I think there's two issues 14 here. One, as I indicated before, there's a fair 15 amount of scatter in any measurement of toughness and 16 in CASS materials in particular. So I'm not sure 17 about the -- I'm not trying to cut these numbers that 18 fine. And, again, I'm not sure that -- and I don't 19 want to get into a semantic argument about the meaning 20 of synergism, but again, these effects are not larger 21 than the combined individual effects. The effects 22 seem to be similar for a large portion of the 23 behavior.
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5210 1 necessarily to do -- my question had nothing to do 2 with synergism. My question had to do with where is 3 the basis for your statement that you made in your 4 testimony? And if it's a professional estimate, 5 that's fine. If it's a wild guess, that's fine. If 6 it's based on something else, I would like to hear 7 about it. Thank you.
8 Entergy's testimony, 722, same Answer on 9 the same Page, because it goes on to state, we may 10 have to pull this back too, that given the ongoing 11 research in this area, the Electric Power Research 12 Institute (EPRI) Materials Reliability Program (MRP) 13 developed conservative screening criteria to identify 14 components that are potentially susceptible to the 15 effects of such mechanisms. And let me ask you, are 16 these such mechanisms the thermal embrittlement and 17 the irradiation embrittlement and the combination of 18 the two? Is that what is meant by the such 19 mechanisms?
20 DR. LOTT: Yes, I believe that's true.
21 ADMIN. JUDGE WARDWELL: And what is that 22 conservative screening criteria that EPRI developed?
23 Was that the 15 percent screening criteria or is it 24 some other criteria?
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5211 1 has proposed I believe is that 20 percent value with 2 a fluence level of about one dpa.
3 ADMIN. JUDGE WARDWELL: You tailed off at 4 the end and --
5 DR. LOTT: I'm sorry, my voice is tender.
6 No, I do not believe it's the 15 percent. I believe 7 the EPRI proposal is 20 percent. I might have to 8 check to see if that number is not in our file 9 testimony. I'll have to look and see.
10 ADMIN. JUDGE WARDWELL: What screening 11 criteria are we talking about that EPRI is proposing?
12 A screening for what?
13 DR. LOTT: Basically a screening for 14 thermal and irradiation embrittlement that's used to 15 identify components that would be -- again, we'd have 16 to look at the effect of -- when we go to the -- first 17 of all, when we identified in our screening process 18 originally for MRP 227, we identified all of the cast 19 materials as potentially susceptible to thermal 20 embrittlement because we did not have any ferrite 21 contents for them. And so we couldn't say if they 22 were even above or below 20 percent.
23 Since that time, we've done a large amount 24 of working with customers such as Entergy and we've 25 identified that no materials that are greater than 20 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5212 1 percent in the Westinghouse internals components that 2 are cast stainless steel. So, that would be -- but 3 that's not in our -- it's a conservatism, because 4 we've never undid the questions about thermal 5 embrittlement. And, in fact, that's what the EPRI 6 concerns are about is the, what are the threshold 7 values for identifying thermal and irradiation 8 embrittlement? Should there be a different fluence 9 level for determining irradiation embrittlement 10 susceptibility in cast stainless steels?
11 ADMIN. JUDGE WARDWELL: And do you need to 12 time to also --
13 DR. LOTT: Let me locate that --
14 ADMIN. JUDGE WARDWELL: -- determine what 15 that EPRI screening criteria is?
16 DR. LOTT: Yes, right. Yes, let me --
17 ADMIN. JUDGE WARDWELL: Can I ask this 18 question of, I don't know who answered it before, who 19 stated it before, but about a half hour ago when we 20 started on these two simple questions that I thought 21 we were going to zoom by --
22 DR. LOTT: Yes.
23 ADMIN. JUDGE WARDWELL: -- this 15 percent 24 of ferric content was brought up as a screening 25 criteria. Is that a different screening criteria or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5213 1 did I hear that wrong or where did that come from?
2 MR. GRIESBACH: This is Tim Griesbach from 3 Entergy. Molybdenum also plays a big part in this 4 thermal embrittlement of cast materials. So there is 5 --
6 ADMIN. JUDGE WARDWELL: How are you 7 simplifying this discussion?
8 (Laughter.)
9 MR. GRIESBACH: There are two different 10 screening criteria depending on the molybdenum 11 content. If it's less than one half percent or 12 greater than one half percent. If it's greater than 13 one half percent --
14 ADMIN. JUDGE WARDWELL: Of molybdenum?
15 MR. GRIESBACH: -- of molybdenum and high 16 or low delta ferrite, there are different criteria.
17 And then there's a separate criteria based on fluence 18 also. So we will produce that for you and hopefully 19 that will simplify things, especially as it applies to 20 the Indian Point materials.
21 ADMIN. JUDGE WARDWELL: Can I get one of my 22 questions out of the way beforehand, and that is, what 23 was this 15 percent we brought and should I just 24 forget about that and you'll trump it with more 25 specific --
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5214 1 MR. AZEVEDO: Yes, your honor --
2 ADMIN. JUDGE WARDWELL: I see Mr. Azevedo 3 warming up to the old mic.
4 MR. AZEVEDO: This is Nelson Azevedo for 5 Entergy. Yes, so let me see if I can explain real 6 shortly. The 20 percent came from EPRI, that's EPRI 7 developed. The 15 percent came from the NRC.
8 ADMIN. JUDGE WARDWELL: And the 20 percent 9 is ferric content?
10 MR. AZEVEDO: Ferrite content, right. So 11 EPRI proposed 20 percent, the NRC used 15 percent.
12 ADMIN. JUDGE WARDWELL: Dr. Hiser, do you 13 agree that you used 15 percent and EPRI used 20 14 percent for a screening criteria?
15 DR. HISER: Yes, that's correct.
16 ADMIN. JUDGE WARDWELL: Okay.
17 DR. HISER: Fifteen percent is for 18 irradiated cast stainless steel.
19 ADMIN. JUDGE WARDWELL: Okay. Great. And 20 so you will still get back to us with the molybdenum 21 and anything else you want to add to that, for 22 whatever you want to? Great. Moving on to other 23 welds, New York State testimony, 576, Page 4, Line 8 24 through 23, and moving through to Page 5, Lines 1 25 through 17, states that in regard to NUREG/CR--7185, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5215 1 New York State contends that, one, cast and Austenitic 2 stainless steel welds have a duplex structure and may 3 experience thermal embrittlement, which may increase 4 the hardness and tensile strength of a material, but 5 decrease ductility, fracture toughness, and impact 6 strength of cast materials and Austenitic stainless 7 steel welds.
8 Two, IE is a concern for CASS components 9 for fluences greater than two times ten to the 20, and 10 that's N per square centimeters, so I believe that's 11 neutrons per square centimeter, allegedly equivalent 12 to ten displacements per atom, or dpa, and irradiation 13 makes cast materials and Austenitic stainless steel 14 welds more susceptible to irradiation assisted stress 15 corrosion cracking, or the IASCC. And, three, that 16 IASCC increases the crack growth rate of cracks 17 induced by stress corrosion cracking, but there is 18 allegedly virtually no data above the ten dpa, 19 although some reactor vessel internal components may 20 experience several hundred dpa.
21 And, fourth, that there is possibly 22 synergy between TE and IE, although the report 23 stresses the need for more information to develop 24 reliable failure curves. And, five, that TE could 25 make the welds associated with the pressurizer spray NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5216 1 nozzles vulnerable to seismic and thermal pressure 2 shock loads. Let me start with Dr. Lahey who made 3 this statement. To your knowledge, is a pressurizer 4 spray nozzle an RVI component covered by the RVI AMP 5 within the list of those components?
6 DR. LAHEY: Not --
7 ADMIN. JUDGE WARDWELL: Or you don't know 8 for sure?
9 DR. LAHEY: It's not a reactor vessel 10 internal at all. It's an external pressure boundary.
11 ADMIN. JUDGE WARDWELL: Okay.
12 DR. LAHEY: And I never called it an RVI.
13 ADMIN. JUDGE WARDWELL: Okay.
14 DR. LAHEY: Also, there was a typo in my 15 thing, it's not ten dpa. If you look at that chart we 16 showed before, it's more like 0.3 and that was pointed 17 out by Entergy --
18 ADMIN. JUDGE WARDWELL: Okay.
19 DR. LAHEY: -- and rightly so. So, other 20 than that, the new information to me that I didn't 21 have when I expressed this concern was the composition 22 of the weld rods that they actually used. They used 23 308 weld rods for stainless steel 304 and 309 weld 24 rods for stainless steel 316. And this does give a 25 duplex structure, but the net effect is more like less NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5217 1 than ten percent. So it's likely not subject to 2 thermal embrittlement. And what I was made aware of 3 based on the feedback from Entergy or NRC, I don't 4 recall which one, was that Dr. Chopra had a very high 5 delta ferrite content in his welds. And that made the 6 difference.
7 ADMIN. JUDGE WARDWELL: So is your net 8 conclusion that you don't believe this is an issue 9 now?
10 DR. LAHEY: I don't believe the thermal 11 embrittlement of those particular welds is an issue, 12 if in fact that is how they did their welds.
13 ADMIN. JUDGE WARDWELL: Okay. And I'll 14 turn to Entergy, as just represented by Dr. Lahey, is 15 that how the welds were done? Or do you have some 16 other nuances to discuss in regards to how the welds 17 were performed?
18 MR. AZEVEDO: This is Nelson Azevedo for 19 Entergy. No, your honor, what's been said is correct.
20 ADMIN. JUDGE WARDWELL: Okay. Thank you.
21 So there was a list that I have here, without reading 22 through it I guess -- if you don't believe that welds 23 are now an issue, you're comfortable with the welds 24 that they were performed as represented by Entergy?
25 DR. LAHEY: Well, the welds we were just NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5218 1 talking about were the welds associated with the 2 pressurizer spray nozzle. There may be concern with 3 the welds inside of the reactor vessel internal --
4 ADMIN. JUDGE WARDWELL: Okay.
5 DR. LAHEY: -- component welds, like the 6 core barrel.
7 ADMIN. JUDGE WARDWELL: Okay, good. So 8 let's continue then. I guess I'll go with Staff now.
9 Are the pressurized spray nozzles to safe end welds 10 considered part of RVIs?
11 MR. POEHLER: This is Jeffrey Poehler for 12 the Staff. No, they are not considered part of RVI.
13 ADMIN. JUDGE WARDWELL: And are they 14 handled under some other Aging Management Program?
15 MR. POEHLER: Yes.
16 ADMIN. JUDGE WARDWELL: Do you agree that 17 there is a decrease in ductility, fracture toughness, 18 and impact strength of cast materials and Austenitic 19 stainless steel weld? And, if so, does this drive any 20 changes need to thermal embrittlement screening 21 criteria or other aging management procedures?
22 MR. POEHLER: Can you clarify -- under what 23 conditions are you referring to? Combined irradiation 24 and thermal exposure or one or the other?
25 ADMIN. JUDGE WARDWELL: One or the other.
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5219 1 MR. POEHLER: In cast Austenitic stainless 2 steel, yes, we agree there is a decrease in fracture 3 toughness that can occur to either irradiation or due 4 to thermal aging.
5 ADMIN. JUDGE WARDWELL: Maybe I should back 6 up a bit because I'm a little bit confused. What 7 welds are left that are part of the RVIs? Are there 8 a whole host of them or are there only a few isolated 9 ones that part of RVIs or --
10 MR. POEHLER: Are we discussing welds or 11 castings?
12 ADMIN. JUDGE WARDWELL: Well, I believe it 13 was welding of cast material.
14 MR. POEHLER: Yes, there are some welds in 15 cast materials, such as the lower support columns.
16 ADMIN. JUDGE WARDWELL: Such as the what?
17 MR. POEHLER: The lower core support 18 columns.
19 ADMIN. JUDGE WARDWELL: And are those -- do 20 you handle those as separate or do you handle them 21 within the component itself, the support column 22 itself?
23 MR. POEHLER: We handle those as part of 24 the component itself.
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5220 1 welds, though -- well, let me ask Dr. Lahey. In 2 regards to welds as cast material, given the 3 representation of Entergy on how those welds were 4 performed, is it your understand that, that is also 5 how the reactor vessel internals were welded, for any 6 welds that were needed for reactor vessel internals?
7 DR. LAHEY: Your honor, what we were 8 talking about before was welds outside of the pressure 9 vessel using wrought Austenitic stainless steel and 10 the type of weld rods that they used to perform that.
11 Which was one of my concerns until I understood what 12 they actually did. Inside the reactor pressure vessel 13 for some of these other components, it's a different 14 story.
15 ADMIN. JUDGE WARDWELL: And is that story 16 the same story you believe applies to the welds as 17 well as the -- that you also have for the internals 18 themselves?
19 DR. LAHEY: Well, depending on the delta 20 ferrite content, as long as it's above a certain 21 level, you heard 15 percent, then you could have this 22 thermal embrittlement as well as irradiation 23 embrittlement. And there's certainly welds that can 24 be subjected to both of those effects.
25 ADMIN. JUDGE WARDWELL: But if it's below NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5221 1 the -- if it's screened out below the 15 percent 2 ferric content, then you believe that thermal 3 embrittlement is not an issue, it's just the 4 irradiation embrittlement?
5 DR. LAHEY: Generally, yes, but there are 6 some exceptions if you have a linkage within the weld 7 material itself of the delta ferrite. But generally 8 you have to have more of it to have this effect.
9 ADMIN. JUDGE WARDWELL: Let me turn to 10 Staff, how would anyone handle leakage out of a weld 11 and the impacts of that? Is there a need to evaluate 12 potential leakage out of a weld?
13 MR. POEHLER: This is Jeffrey Poehler from 14 the Staff. Are you referring to pressure boundary 15 welds or --
16 ADMIN. JUDGE WARDWELL: I'm referring to 17 welds of the RVIs that we're dealing with under the 18 RVI AMP, whatever they are.
19 MR. POEHLER: Since the welds in the RVI do 20 not serve a pressure boundary function, then leakage 21 is not a failure criteria for those welds.
22 ADMIN. JUDGE WARDWELL: And in your review 23 of the AMP and the components that are in the RVI AMP, 24 have any of the components been welded such that the 25 weld and the component itself have a higher ferric NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5222 1 content than 15 percent? Or were they all screened 2 out?
3 MR. POEHLER: We didn't look at the ferrite 4 content of the welds. So weld filler metals used for 5 Austenitic stainless steel welds, they tend to have 6 lower ferrite content in general than cast stainless 7 steels. So thermal aging is not generally an issue, 8 has not generally been considered an issue for those 9 welds. So in the screening criteria used to develop 10 the AMP at Indian Point and MRP 227-A, they didn't 11 screen-in thermal embrittlement for the welds.
12 ADMIN. JUDGE WARDWELL: How many RVIs are 13 made out of -- are composed of cast materials?
14 MR. POEHLER: Basically, there's a handful 15 of components, the lower core support columns and then 16 at Indian Point it's only the upper portion of those 17 columns, which they call the column cap, is cast 18 stainless steel. You also have the lower core support 19 forging at the very bottom of the core barrel. There 20 may be a couple other ones, but those are the main 21 ones.
22 ADMIN. JUDGE WARDWELL: Let me turn to 23 Entergy and see if they want to clarify. Well, let's 24 start off with, how many of the RVIs are, to your 25 knowledge, cast materials versus the wrought NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5223 1 Austenitic?
2 DR. LOTT: I believe that we identified 3 that in our filed testimony. I'm looking at Page 62, 4 I think it's Question 105, and it identifies the upper 5 instrumentation conduit supports --
6 ADMIN. JUDGE WARDWELL: Sorry, you're going 7 to have speak into the mic and a little slower.
8 DR. LOTT: Okay. It identifies the upper 9 instrumentation conduits and supports, upper support 10 column assemblies, the lower support casting. That 11 should be lower support column assemblies. Oh, no, 12 I'm sorry, that's right. The upper instrumentation 13 conduits, upper support column assemblies, and the 14 lower support casting.
15 ADMIN. JUDGE WARDWELL: And how many of 16 those were screened out due to ferric content? Or 17 moly content or whatever?
18 DR. LOTT: In the original screening 19 process, none of them were screened out because 20 everything was screened in based on lack of knowledge 21 at that time of the ferric content.
22 ADMIN. JUDGE WARDWELL: Okay. I'll wait, 23 I guess, until I get to those individual components.
24 Because I thought somewhere, it was my impression that 25 only those upper caps of the lower supports were cast NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5224 1 materials that were part of the aging management plan, 2 but I don't know that. I'll have to --
3 DR. LOTT: But they are cast materials part 4 of the aging management plan, but they were not 5 necessarily -- well, first of all, most of them are 6 not irradiated at all. So this question of 7 irradiation embrittlement doesn't come in. And I 8 think there was also -- some of them were screened out 9 based on lack of structural requirement. In other 10 words, there were just no --
11 ADMIN. JUDGE WARDWELL: Well, I'll be able 12 to get to that quote once I get to it. I just can't 13 find it and I don't want to spend time looking for it.
14 I'll come upon it as I work my way through. But I 15 want to get back to welds, I think. How many -- how 16 do you handle the welds in your AMP, for both the 17 wrought and the cast materials? Are they part of the 18 individual component or are they considered 19 separately?
20 DR. LOTT: There are some key welds that 21 are considered separately. And those welds were, for 22 instance, the core barrel welds, we knew we were 23 concerned about them, they're identified separate.
24 There are other components that contain welds and in 25 the screening process, one of the things we identified NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5225 1 was which components contained welds and which did 2 not. And then the primary concern with welds was 3 large structural welds where concern for stress 4 corrosion cracking, not necessarily embrittlement.
5 ADMIN. JUDGE WARDWELL: And, so, how many 6 of these special welds were designated as individual 7 components when you split this out?
8 DR. LOTT: I'd have to go back and count, 9 I --
10 ADMIN. JUDGE WARDWELL: Okay.
11 DR. LOTT: -- couldn't answer that.
12 MR. DOLANSKY: This is Bob Dolansky from 13 Entergy. The Table 1 on Page 87 of our testimony 14 lists the primary components and the expansion 15 components. The primary components lists if it's a 16 weld or not. Does that answer your question?
17 ADMIN. JUDGE WARDWELL: To a certain 18 degree. I was hoping you had an approximate number of 19 those that were there that related to split out welds, 20 but that's fine. I don't need that right now.
21 MR. DOLANSKY: Okay.
22 ADMIN. JUDGE WARDWELL: I would like that 23 number to get a feeling for how many there are and 24 what's being -- so I can then refer to that table to 25 see what's being handled for them.
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5226 1 MR. DOLANSKY: I could go through this 2 table if you wanted and --
3 ADMIN. JUDGE WARDWELL: Yes, at some point 4 and then just come up with a number and that will be 5 sufficient so I make sure I see the same number that 6 you do when I -- if it does come up in our decision 7 writing, I'm able to refer to that and handle any 8 discussion associated with it.
9 MR. DOLANSKY: Just to clarify, what I'll 10 do the research on is for each primary component, if 11 it's a weld, I'll tell you how many there are.
12 ADMIN. JUDGE WARDWELL: Okay.
13 MR. DOLANSKY: That's basically what you're 14 looking for --
15 ADMIN. JUDGE WARDWELL: That would be 16 great.
17 MR. DOLANSKY: -- to understand?
18 ADMIN. JUDGE WARDWELL: Thank you.
19 MR. DOLANSKY: Okay.
20 ADMIN. JUDGE WARDWELL: Yes, let's do it 21 now.
22 CHAIRMAN MCDADE: Okay. It might be an 23 appropriate time to take a short ten minute break. A 24 couple of things. First of all, Mr. Sipos, we had a 25 discussion with Dr. Lahey about the Table 5-5 and he NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5227 1 hadn't had an opportunity to review it. Do you have 2 handy New York 496, which is the document that Table 3 5-5 is in? If not, I've got it right here and I can 4 provide it to you for Dr. Lahey to look at.
5 MR. SIPOS: Your honor, I believe we have 6 it with us. Thank you.
7 CHAIRMAN MCDADE: Okay. If you don't, then 8 at the next break I can provide it to you and --
9 MR. SIPOS: Thank you.
10 CHAIRMAN MCDADE: -- you can give that to 11 Dr. Lahey. Another thing that I actually should have 12 mentioned earlier, in Exhibit 616, which Entergy 13 provided, at the beginning of it, there's a table of 14 abbreviations, which is extremely helpful. Whoever 15 prepared it, I really want to thank them. I just 16 wanted to mention it, as the court reporter is going 17 through this, if you have not reviewed it, Exhibit 18 616, the table of abbreviations I think is going to be 19 very helpful to you in make sure that you -- well, 20 okay. Then that isn't necessary, apparently he found 21 it without my assistance. Do we have anything else we 22 need to take up before we take a short break?
23 ADMIN. JUDGE WARDWELL: But I think ISR was 24 missing from that table though, but I echo how much I 25 did use that for both other testimony, I would go back NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5228 1 to the Entergy abbreviations or the NRC ones, whoever 2 had them --
3 CHAIRMAN MCDADE: Okay. So it's ten 4 minutes after 3:00, we'll break until 20 minutes after 5 3:00. Thank you.
6 (Whereupon, the above-entitled matter went 7 off the record at 3:09 p.m. and resumed at 3:23 p.m.)
8 CHAIRMAN MCDADE: Let's get started. The 9 hearing will come to order.
10 ADMIN. JUDGE WARDWELL: Okay. Moving on to 11 baffle former bolts, the NRC testimony, 197, Answer 12 80, Page 59, says, an example of an augmented 13 inspection is the baseline volumetric examination 14 using ultrasound testing, UT, of the baffle former 15 bolts between 25 and 35 effective full power years, 16 which you use the acronym with the letters E-F-P-Y, 17 there's probably some fancy way to say that, that I 18 don't know, within a subset examination on a ten year 19 interval, as specified in Table 4.3 of MRP 227-A.
20 CHAIRMAN MCDADE: Wait, whose testimony is 21 this?
22 ADMIN. JUDGE WARDWELL: I'm sorry?
23 CHAIRMAN MCDADE: Which exhibit are you 24 reading from?
25 ADMIN. JUDGE WARDWELL: This is NRC's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5229 1 Exhibit 197, A 80, Page 59.
2 CHAIRMAN MCDADE: NRC's?
3 ADMIN. JUDGE WARDWELL: Yes. And I guess 4 I'd ask Entergy, how long -- have you initiated any UT 5 testing of the baffle former bolts at Indian Point?
6 MR. DOLANSKY: This is Bob Dolansky for 7 Entergy. No.
8 ADMIN. JUDGE WARDWELL: And has the 9 industry itself and do you have data on that?
10 MR. DOLANSKY: The industry has performed 11 inspections on baffle former bolts. They have -- when 12 you say, do we have data on that, what kind of data 13 are you looking for?
14 ADMIN. JUDGE WARDWELL: I'm just curious on 15 what the experience is with any of the failure rates 16 associated with these in other Westinghouse reactors.
17 MR. DOLANSKY: I would characterize it as 18 most people who have inspected have found some 19 degraded bolts, but not enough that they were required 20 to replace any bolts.
21 ADMIN. JUDGE WARDWELL: Okay, thank you.
22 DR. LOTT: May I suggest, there is some 23 data on that in Entergy Exhibit 650, which summarizes 24 the experience to date with various reactor internal 25 exams, including the baffle bolt exams.
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5230 1 ADMIN. JUDGE WARDWELL: Okay, thank you.
2 NRC's testimony on 197, Question 222, Page 124, 3 states, what does this operating experience tell us 4 about the probability of cracks existing in the PWR 5 RVIs? And the Answer says, this result summarized in 6 the presentation indicate no cracking has been found 7 with the exception of some cracking of bolts, about 8 1.5 percent of Westinghouse baffle former bolts. And 9 there I might ask Staff, what might be the cite for 10 this 1.5 percent cracking rate for the baffle former 11 bolts that you state in your testimony?
12 DR. HISER: This is Allen Hiser of the 13 Staff. I believe it's NRC 207.
14 ADMIN. JUDGE WARDWELL: Okay. And does 15 that also give the total number of baffle bolts that 16 they looked at in regards to this, that generated the 17 1.5 percent figure?
18 DR. HISER: If you'll indulge me, I can 19 pull it up and review it real quickly.
20 ADMIN. JUDGE WARDWELL: Okay. And what are 21 you pulling up now?
22 DR. HISER: From Exhibit 207.
23 ADMIN. JUDGE WARDWELL: Okay. As you're 24 pulling that up, let me ask this question also, is it 25 your understanding or do you know if this industry NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5231 1 experience with the inspection of baffle former bolts 2 is also documented in Appendix A of MRP 227, which is 3 NRC 114C?
4 DR. HISER: It is document in the MRP 227, 5 but that was as of several years ago. The Exhibit 207 6 represents data through the fall of 2014.
7 ADMIN. JUDGE WARDWELL: Okay.
8 DR. HISER: So it's probably the most 9 recent information. And this indicates that 8,887 10 baffle bolts have been UT inspected.
11 ADMIN. JUDGE WARDWELL: And that's 12 generated the 1.5 percent in regards to the cracking 13 rate?
14 DR. HISER: Yes, that's correct.
15 ADMIN. JUDGE WARDWELL: And do you know 16 what was the criteria used to say that a bolt had been 17 cracked? Is there any parameters that was given?
18 Does it have to be separated, loose? Is there just a 19 crack detected? Or a visual crack would apply to 20 that?
21 DR. HISER: In this case, I would expect it 22 at least was any indication of a crack from the UT 23 exam.
24 ADMIN. JUDGE WARDWELL: Dr. Lahey, it seems 25 like this experience has shown that very few cracked NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5232 1 or failed baffle former bolts have been detected 2 during these examinations and in most cases no cracked 3 or failed bolts were detected at all. Do you have any 4 other experience that dictates that they are more of 5 an issue --
6 DR. LAHEY: Well, there have been --
7 ADMIN. JUDGE WARDWELL: -- of these than 8 there seem to be?
9 DR. LAHEY: There have been failures 10 reported overseas as well. So, it's not been massive 11 failures, but there have definitely been failures.
12 While we're talking about baffle bolts, I was asked to 13 look at Table 5-5 and if you want me to -- which 14 includes the acceptance criteria for such things as 15 baffle bolts. Do you want me to report on that now or 16 not?
17 ADMIN. JUDGE WARDWELL: That was one of 18 your homework assignments wasn't it?
19 DR. LAHEY: During the break, I thought.
20 ADMIN. JUDGE WARDWELL: Yes. Well, we 21 allow homework to be done during a break --
22 DR. LAHEY: Right.
23 ADMIN. JUDGE WARDWELL: -- we don't 24 restrict that.
25 DR. LAHEY: Shall I report on that?
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5233 1 ADMIN. JUDGE WARDWELL: Sure. This is 2 probably as good an opportunity.
3 DR. LAHEY: Okay. And it has to do -- I 4 looked at the various items in the Table 5-5 and the 5 one that I have some concern about is the baffle 6 bolts. And, in particular, when it talks about 7 additional examination acceptance criteria, it's not 8 yet specified. It says, the examination acceptance 9 criteria for the UT of the bolts shall be established 10 as part of the examination technical justification.
11 So it's still sort of open-ended and since it's such 12 an important component and has safety significance.
13 That's a little unsettling.
14 ADMIN. JUDGE WARDWELL: And that's the 15 examination criteria, is that correct?
16 DR. LAHEY: Yes.
17 ADMIN. JUDGE WARDWELL: And may I turn --
18 DR. LAHEY: Table 5-5.
19 ADMIN. JUDGE WARDWELL: Yes. Turn to 20 Entergy in regards to that examination criteria. What 21 does that mean in regards to the statement made in 22 Table 5-5 for the baffle former bolts? For Entergy, 23 anyone at Entergy who wants to answer.
24 MR. DOLANSKY: The examination acceptance 25 criteria for baffle former bolts would be any NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5234 1 indication of cracking.
2 ADMIN. JUDGE WARDWELL: And what is the --
3 is there a difference between examination criteria and 4 inspection criteria or acceptance -- I guess we have 5 acceptance criteria and examination criteria. Are 6 they different or are those two sayings for the same 7 thing?
8 MR. AZEVEDO: This is Nelson Azevedo for 9 Entergy. The examination criteria is the criteria for 10 the inspectors. So anything that exceeds the 11 examination criteria, they must report it and then we 12 enter into our corrective action process. The 13 acceptance criteria or the engineering acceptance 14 criteria, if you will, it's how many bolts, for 15 example, can we afford to lose without impacting the 16 ability of the structure to perform its intended 17 safety function?
18 ADMIN. JUDGE WARDWELL: Great. Thank you 19 for that clarification. And did you have anything 20 more, Dr. Lahey, that you wanted to talk about in 21 regards to Table 5-5?
22 DR. LAHEY: No. With regards to that 23 table, that was our specific concern, the number of 24 bolts, because that is significant.
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5235 1 close the other item, to make sure -- you don't have 2 any hard numbers in regards to percent of these baffle 3 bolts that failed overseas or anywhere else that would 4 contradict the 1.5 percent that was observed by 5 Westinghouse?
6 DR. LAHEY: I've never tabulated it, no.
7 ADMIN. JUDGE WARDWELL: Okay. Thank you.
8 NRC's testimony, 147, Page 47, in order to maintain 9 the intended function, only about 20 to 30 percent of 10 the baffle former bolts need to remain intact. And I 11 guess I'll ask Entergy -- well, it's an NRC statement, 12 so I'll ask NRC. What's the basis for this 20 to 30 13 percent figure? And if it's just that it came from 14 the Applicant, so say, or is it something that you are 15 familiar with in regards to the generation of these 16 particular values?
17 MR. POEHLER: Right. This is Jeffrey 18 Poehler from the Staff. It did not come from the 19 Applicant. It is based on some -- well, there's a 20 couple of Topical Reports, WCAP reports, where they 21 provide a methodology for performing these types of 22 minimum pattern analyses for the baffle former bolts.
23 Several plants have actually used that methodology and 24 the results came out in the 20 to 30 percent range of 25 intact bolts needed.
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5236 1 ADMIN. JUDGE WARDWELL: What do you mean 2 by, they used that and came out with a 20 to 30 3 percent?
4 MR. POEHLER: They used a preapproved --
5 the NRC had approved the methodology for doing these 6 analyses and it was submitted as a Topical Report --
7 ADMIN. JUDGE WARDWELL: And this is --
8 MR. POEHLER: -- to the NRC.
9 ADMIN. JUDGE WARDWELL: This is an 10 engineering analysis, it has nothing to do with 11 inspection or anything or number of bolts, it's just 12 -- number of bolts that have cracked or anything, this 13 has strictly an analysis of how many are needed to 14 maintain the intended function of the baffle, is that 15 correct?
16 MR. POEHLER: Correct. And how many and in 17 what positions, what type of patterns --
18 ADMIN. JUDGE WARDWELL: Okay.
19 MR. POEHLER: -- that would be needed to 20 maintain the design function of the baffle former 21 assembly.
22 ADMIN. JUDGE WARDWELL: And that analysis 23 resulted in the 20 to 30 percent needed?
24 MR. POEHLER: When that methodology was 25 applied for specific plants.
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5237 1 ADMIN. JUDGE WARDWELL: Has that -- was 2 Indian Point one of those specific plants?
3 MR. POEHLER: I don't -- to my knowledge, 4 they weren't. If they've done that type of analysis, 5 I'm not aware that their results have been submitted.
6 ADMIN. JUDGE WARDWELL: Okay. Well, let's 7 turn to Entergy. Has such an analysis been done at 8 Entergy or do you just use that 20 to 30 percent as an 9 accepted figure based on the analysis that has been 10 done by others?
11 MR. DOLANSKY: This is Bob Dolansky from 12 Entergy. We're having a plant specific acceptable 13 bolting pattern analysis performed for us right now 14 that will determine plant specific. So the 15 methodology that -- the Topical Report is a general 16 methodology. We then took that general methodology 17 and put in our specific accident loads, LOCA loads, 18 all those things, and did the analysis. Well, it's 19 ongoing. It'll be ready before we perform the 20 inspections in the spring.
21 ADMIN. JUDGE WARDWELL: Do the --
22 MR. STROSNIDER: This Jack Strosnider for 23 Entergy. If I just could, this question has come up 24 a number of time about Topical Reports. I just want 25 to make sure people understand how that process works, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5238 1 if I could take just a second? When the industry sees 2 a generic issue, such as this issue, and they know 3 that a lot of plants are going to have to deal with 4 it, they'll go contract the vendor, like Westinghouse 5 or somebody, to develop a methodology to do some pilot 6 plants to do a case study. And then they'll submit 7 that to the NRC.
8 The NRC reviews that and, if they approve 9 it with whatever approvals they make, then a specific 10 plant, a utility, can come in and reference that 11 report when they do their plant specific analysis.
12 And the important part of that is that the methodology 13 has been approved by NRC. So as was just indicated, 14 that includes all the loading and how you do the 15 calculations and how you demonstrate functionality.
16 So what's happening is, at the plant specific level 17 then is just to do the plant specific evaluation.
18 When the NRC issues their safety evaluation, they will 19 include in it any specific action items that have to 20 be addressed on a plant specific basis. So I hope 21 that's of help to you because there's a number of 22 questions that have come up on various Topical Reports 23 and that's how that whole process works.
24 ADMIN. JUDGE WARDWELL: Thank you. And, 25 Dr. Hiser, has NRC approved this methodology for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5239 1 determining the number of bolts needed and the 2 patterns for the baffle former bolts?
3 DR. HISER: Yes, we have. And the exhibits 4 are ENT 654 and 655.
5 ADMIN. JUDGE WARDWELL: Thank you. And 6 previously, if Indian Point is consistent with what 7 was done before, they should arrive at -- if it's 8 consistent with what our understanding is now, it 9 would be 20 to 30 percent of the bolts are needed, is 10 that correct?
11 DR. HISER: I would expect that to be the 12 case.
13 ADMIN. JUDGE WARDWELL: I think I'll go 14 back to Entergy though to ask this question. Do you 15 know if that 20 to 30 percent -- does not the intended 16 function of a component, like the baffles and the 17 former bolts that are attached to it, include some 18 type of, in my field I'd call it a safety factor, 19 other people will consider it an error margin or 20 whatever else? Isn't there some sort of safety factor 21 or error margin applied when you estimate how many of 22 the bolts are needed?
23 MR. DOLANSKY: Yes. If I could just take 24 one minute and explain? We're getting a site specific 25 acceptable bolting pattern analysis performed. That's NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5240 1 going to tell us some very small number of bolts that 2 are allowed to be degraded without requiring 3 additional analysis. So let's just say one -- there's 4 832 baffle former bolts. Let's just say one, so we 5 could find one and we would be okay. But if we find 6 more than one, then we will have to do a thing called 7 a real time analysis and that basically is, they take 8 the actual place that we found the degraded bolts and 9 they put them into the analysis.
10 A big part of -- it's taken roughly a year 11 to do this analysis. They develop a computer program, 12 they'll then take the specific bolts that we actually 13 found degraded, put those into the computer program, 14 run it with our specific loads and site specific 15 requirements, and come out with, is that actual what 16 we found acceptable or not?
17 ADMIN. JUDGE WARDWELL: And by acceptable, 18 does that have an error margin or a safety margin 19 around it?
20 MR. DOLANSKY: Yes, it does.
21 ADMIN. JUDGE WARDWELL: And my question is, 22 do you know if the 20 to 30 percent figure of the 23 acceptable bolts that have been done in the past, that 24 number generated in the past, does that include that 25 safety factor or could it possibly be that, that is NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5241 1 what needed with a safety factor of only one?
2 MR. DOLANSKY: I can't answer that.
3 ADMIN. JUDGE WARDWELL: Okay. Staff, would 4 you be able to answer that question?
5 DR. HISER: This is Allen Hiser of the 6 Staff. In the generic methodology, Topical Report, 7 there would be safety factors included in that.
8 ADMIN. JUDGE WARDWELL: So you believe that 9 --
10 DR. HISER: It includes safety factors.
11 ADMIN. JUDGE WARDWELL: -- the 20 to 30 12 percent includes some sort of safety factor such that 13 it could be even less and the thing would still hold 14 together?
15 DR. HISER: If you use a safety factor of 16 one, yes, I expect you could go to fewer bolts than 17 the 20 to 30 percent.
18 ADMIN. JUDGE WARDWELL: Right. Or 19 conversely, if the safety factor was two that they 20 actually used, you could really go down to 10 to 15 21 percent?
22 DR. HISER: Well, safety factor is not on 23 the number of bolts --
24 ADMIN. JUDGE WARDWELL: Just in general 25 terms, just to put a handle on --
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5242 1 DR. HISER: Yes, that's correct. It would 2 be a lower number.
3 ADMIN. JUDGE WARDWELL: The point I'm 4 trying to get is I want to make sure that, that 20 to 5 30 percent doesn't pertain to a safety factor of one.
6 And, likewise, when you do your analysis, I'm asking 7 Entergy, will a number you have be the number that is 8 absolutely needed in order to maintain safety with no 9 extra margin or will it include some margin associated 10 with it?
11 MR. DOLANSKY: This is Bob Dolansky with 12 Entergy. First, when we're doing all this, we're 13 really concerned about reactor safety, maintaining 14 core coolability, maintaining the ability to insert 15 the control rods. That's factored into the analysis 16 and there are safety margins on that safety analysis.
17 MR. STROSNIDER: This is Jack Strosnider --
18 ADMIN. JUDGE WARDWELL: So when you -- no, 19 I'd like to stay with my thoughts. When you come up 20 with an acceptable pattern, that pattern will have 21 some margin built into it, such that in actuality if 22 you knew truth, you could end up with lesser number of 23 bolts, but because we don't know truth, you're going 24 to have some margin in there that allows for the 25 uncertainty associated with not knowing truth in that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5243 1 pattern. Is that a fair assessment?
2 MR. DOLANSKY: Yes. And the reason why I 3 spoke before about the real time analysis is we don't 4 want to know, when we don't know what we actually 5 have, come up with some bolting pattern analysis. We 6 want to wait until we find out what we actually have 7 out there and then run it through the program and make 8 sure that it's robust and is perfectly acceptable. So 9 that's why we actually run the real time analysis.
10 MR. AZEVEDO: Your honor, this is Nelson 11 Azevedo. If I may add, we use the NRC approved 12 methodologies, so the same safety margins are going to 13 be used.
14 MR. STROSNIDER: And this is Jack 15 Strosnider. Just to expand on that, typically what 16 you're going to see in these Topical Reports when they 17 do this type of analysis, is they're going to work to 18 maintain the safety margins that were in the original 19 licensing basis. So in this case, if you're talking 20 about the structural criteria, it's going to be the 21 ASME code factors of safety during normal and accident 22 conditions. And when you look at the other aspect of 23 this, we get into some of the core cooling and 24 accident evaluations and they would need to maintain 25 the margins that are in that licensing basis in terms NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5244 1 of margin to core damage and that sort of thing. So 2 they work to maintain the current licensing basis and 3 the margins that were in those.
4 ADMIN. JUDGE WARDWELL: And so it's your 5 professional opinion that the 20 to 30 percent that 6 has historically been generated by Westinghouse at 7 other plants includes those safety factors associated 8 with the ASME code?
9 MR. STROSNIDER: Yes, it would have those 10 margins in it.
11 ADMIN. JUDGE WARDWELL: Thank you. Your 12 testimony, Entergy, 616, Answer 152, Page 99, says, to 13 prepare for these inspections, as explained in the 14 Supplemented SER Number 2, that the UT examination 15 acceptance criteria for the baffle former bolts will 16 be developed as part of the technical justification 17 for the inspections. Page 100, Answer 154, that the 18 examination acceptance criteria for the individual 19 baffle former bolts will be no defect that could be 20 detectable via UT inspections. And in parentheses, 21 you say that detectable versus UT inspections are a 22 defect exceeding 30 percent of the bolt cross-23 sectional area.
24 The TJ will merely demonstrate that the UT 25 inspections at Indian Point will be capable of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5245 1 detecting such cracking. And I guess my first 2 question to you is, from where did this 30 percent of 3 the cross-sectional area come from in regards to -- I 4 gather that is what is needed before you're able to 5 detect it with the UT inspections.
6 DR. LOTT: I believe the last thing you 7 said is basically true, that the UT inspections, a 8 standard UT inspection, very reliably detect 30 9 percent. I believe that the requirement is stated in 10 MRP 228, if I'm not mistaken, that, that minimum 30 11 percent is there based on the judgement of the 12 inspectors.
13 ADMIN. JUDGE WARDWELL: But that's a result 14 of what you can get out of it. That's what you need 15 before you're -- that's the sensitivity of your UT 16 device --
17 DR. LOTT: Right.
18 ADMIN. JUDGE WARDWELL: -- if you will. Is 19 that correct?
20 DR. LOTT: Yes.
21 ADMIN. JUDGE WARDWELL: And, so, that means 22 that anything under that won't be detected, is that 23 correct?
24 DR. LOTT: Well, maybe not necessarily 25 won't, but might not be.
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5246 1 ADMIN. JUDGE WARDWELL: The odds are it 2 won't. I mean, there will be some that needs more 3 than 30 percent that won't be detected also, using the 4 same phraseology?
5 DR. LOTT: Yes. It's a very high 6 probability of detection, I believe, at 30 percent.
7 ADMIN. JUDGE WARDWELL: And is this 8 acceptance criteria -- has the 30 percent figure shown 9 up anywhere in the Inspection Plan or anything else?
10 Or is that just part of the testimony that you've 11 created here for this hearing?
12 MR. DOLANSKY: This is Bob Dolansky with 13 Entergy. I believe that 30 percent will be in the 14 technical justification --
15 ADMIN. JUDGE WARDWELL: Okay.
16 MR. DOLANSKY: -- that will be used by --
17 that has to be documented and reviewed by our NDE 18 Level III at the site before they perform the 19 inspection.
20 ADMIN. JUDGE WARDWELL: This is all stuff 21 you're going to get prepared and done prior to 22 starting your inspection, I think you talked about 23 earlier today.
24 MR. DOLANSKY: Correct.
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5247 1 have any comments on what you heard in regards to the 2 baffle former bolts --
3 DR. LAHEY: Yes.
4 ADMIN. JUDGE WARDWELL: -- inspection and 5 --
6 DR. LAHEY: Yes, your honor. This was a 7 very interesting discussion for me. I have not seen 8 that report or the results, I'd be very interested to 9 see it. In particular, I heard nothing about the type 10 of loads that were applied to assure the integrity of 11 these patterns with reduced number of bolts and would 12 be very interested in hearing about what type of 13 impulsive loads were used, if they were used. And 14 what type of codes might have been used to generate 15 these.
16 ADMIN. JUDGE WARDWELL: Okay, thank you.
17 DR. LOTT: Some of that information, I will 18 suggest, would be available in ENT 0655, which is 19 basically the same document that Dr. Hiser just 20 referred to in terms of the methodology for 21 determining acceptable bolt patterns. It discusses 22 the types of loads that are included, particular the 23 multi flex code and how they're transferred to the --
24 ADMIN. JUDGE WARDWELL: Have those loads 25 been documented in the Topical Reports that the NRC NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5248 1 has approved?
2 DR. LOTT: Yes, in great detail.
3 CHAIRMAN MCDADE: Okay. Dr. Lott, what 4 exhibit did you just reference?
5 DR. LOTT: Entergy 655.
6 CHAIRMAN MCDADE: Thank you.
7 ADMIN. JUDGE WARDWELL: When did you start 8 this endeavor in regards to planning for the 9 inspections of the baffles and going through this site 10 specific process of determining your pattern for the 11 baffle bolts?
12 MR. DOLANSKY: Just the baffle former bolts 13 --
14 ADMIN. JUDGE WARDWELL: Yes.
15 MR. DOLANSKY: -- you're asking about?
16 ADMIN. JUDGE WARDWELL: Yes.
17 MR. DOLANSKY: This is Bob Dolansky for 18 Entergy. The contract for the inspection of the 19 baffle former bolts was issued, I'm going to say in 20 spring of 2015. The contract to develop the 21 acceptable bolting pattern analysis, I think, was 22 January. It takes a little over a year to develop 23 that acceptable bolting pattern analysis. So I'm 24 pretty sure it was January and the final report will 25 be ready for us in February before our inspections in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5249 1 March.
2 ADMIN. JUDGE WARDWELL: For Dr. Hiser of 3 the NRC, when was the Topical Report approved by you 4 that --
5 DR. HISER: I believe it was --
6 ADMIN. JUDGE WARDWELL: -- highlighted 7 these inspection procedures or the whole methodology, 8 is what I guess the best word is?
9 DR. HISER: This is Allen Hiser of the 10 Staff. Is the question the methodology used to 11 demonstrate the minimum bolt pattern?
12 ADMIN. JUDGE WARDWELL: Yes.
13 DR. HISER: That was I believe 1999.
14 ADMIN. JUDGE WARDWELL: So my question to 15 Entergy will be, why haven't you started earlier in 16 this process then in order to -- if the methodology 17 has been outlined in regards to coming up with this 18 pattern, why wasn't it done earlier considering your 19 License Renewal Application was submitted in 2007?
20 MR. DOLANSKY: We didn't feel that we 21 needed the acceptable bolting pattern analysis -- Bob 22 Dolansky with Entergy -- until we were going to 23 perform the inspection. Basically, until you perform 24 the inspection and determine if you have any degraded 25 bolts, the acceptable bolting pattern analysis doesn't NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5250 1 do anything for you. You don't -- there's no value to 2 it until you --
3 ADMIN. JUDGE WARDWELL: I guess that gets 4 back to the question I asked earlier, why didn't you 5 start to do the inspection on the bolts earlier?
6 Which I think we've already talked about, so, okay.
7 MR. AZEVEDO: Your honor --
8 ADMIN. JUDGE WARDWELL: Mr. Azevedo, it 9 looked like you were -- you weren't, you were just 10 stretching or something?
11 MR. AZEVEDO: No, I was going to say what 12 Mr. Dolansky just said.
13 ADMIN. JUDGE WARDWELL: Okay. That's a cop 14 out. No, I believe you.
15 MR. KUYLER: Your honor, if I may? I 16 believe a moment ago Dr. Lott may have misspoke. The 17 Exhibit that I think he might have been referring to 18 is Entergy 654, that provides the accident loads.
19 DR. LOTT: Yes, 655 is the non-proprietary 20 version.
21 ADMIN. JUDGE WARDWELL: Okay.
22 DR. LOTT: Either way it works.
23 CHAIRMAN MCDADE: Dr. Lott had mentioned 24 655 and counsel is suggesting it was actually Entergy 25 654, is that --
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5251 1 MR. KUYLER: Yes, your honor.
2 CHAIRMAN MCDADE: Okay, thank you.
3 ADMIN. JUDGE WARDWELL: Moving on to the 4 clevis bolt --
5 CHAIRMAN MCDADE: Actually, could I clear 6 something up before --
7 ADMIN. JUDGE WARDWELL: Sure.
8 CHAIRMAN MCDADE: -- you move on to --
9 ADMIN. JUDGE WARDWELL: Yes.
10 CHAIRMAN MCDADE: -- a new topic, just 11 really quickly. Mr. Dolansky, before the break, you 12 were asked with regard -- questioned on Table 1, Page 13 87, regarding welds. And that particular document, 14 it's Entergy Exhibit 616, it identifies four different 15 kinds of welds. The upper core baffle flange, the 16 barrel cylinder girth welds, the lower core barrel to 17 lower support casting welds, and the core barrel 18 outlet nozzle welds. But it doesn't have any 19 indication as to for each of these categories, how 20 many welds we're talking about. Are we talking about 21 a single weld each time or are we talking about scores 22 or hundreds or thousands?
23 MR. DOLANSKY: This is Bob Dolansky with 24 Entergy. Looking at Table 87, the upper core barrel 25 flange weld is a single weld. The upper and lower NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5252 1 core barrel cylinder girth welds are two individual 2 welds, an upper and a lower core barrel cylinder girth 3 weld. The lower core barrel to support casting weld 4 is a single weld. The only ones that are multiple 5 welds are the lower flange welds on the control rod 6 guide tubes, which are at the top there, the second 7 box. That -- give me one second and if we could bring 8 up on the screen, let's see what, number seven -- NRC 9 114A through C and we're actually going to be looking 10 at 114B, Bravo, I believe. And we want Page 4-60.
11 It'll help clarify.
12 CHAIRMAN MCDADE: 4-60?
13 MR. DOLANSKY: 4-60, yes, your honor. Yes.
14 Stop right there, that's fine. So, this is a 15 depiction of the control rod guide tube assembly. If 16 you look on the bottom, so what we call the lower 17 flange welds in the table on Page 87, if you look on 18 the bottom of that, it shows two arrows pointing to 19 lower flange welds, the bottom one, let's say, there's 20 discrete, very small ribs and each of those ribs has 21 welds. So these components have multiple welds on 22 each component. But that's the only one that really 23 has multiple welds on each component.
24 CHAIRMAN MCDADE: Okay, thank you. And I 25 wondered, since we're at a break right here, pose a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5253 1 question to -- I'm sorry, Mr. Cox, did you have a --
2 MR. COX: Yes, I was just going to clarify 3 that the next page of that exhibit shows the other 4 welds that you had asked about. Mr. Dolansky 5 indicated there were one or two welds in those cases.
6 I just wanted to point out, those welds are around the 7 entire circumference of the core barrel, so it's -- I 8 mean, some of these guys could probably tell you how 9 many inches of weld that is, but it's several hundred 10 probably.
11 MR. DOLANSKY: Right. They're large -- I'm 12 sorry, I didn't mean to imply that they were small 13 welds. They're very large, long welds around a big 14 robust component.
15 CHAIRMAN MCDADE: Okay.
16 ADMIN. JUDGE WARDWELL: Does one guy do 17 them? Or do they have multiple guys?
18 MR. DOLANSKY: Well, if they're EVT1, they 19 have to -- because there's criteria on scanning speed 20 and so forth, it's actually done by special tooling.
21 But there are people watching screens for each of 22 them.
23 ADMIN. JUDGE WARDWELL: But the guy who 24 first built it, was it one guy doing one weld and he 25 can take ownership of that weld and say, there's my NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5254 1 weld?
2 MR. DOLANSKY: Probably. Back then, 3 probably true.
4 ADMIN. JUDGE WARDWELL: Okay.
5 CHAIRMAN MCDADE: Okay. And the other 6 thing I wanted to do is just pose a question, not be 7 answered right now, but to move on with Judge 8 Wardwell's question, but to raise it for Dr. Lott and 9 Dr. Lahey who had testified earlier, addressing New 10 York Exhibit 488, which is a NUREG/CR-7184, it's from 11 December of 2014. And, basically, my question is 12 this, we've had testimony that for these low ferrite 13 stainless steel material, they won't show a meaningful 14 combined effect from thermal aging and irradiation.
15 And in that NUREG it seems to suggest, 16 well, it states, the radiation and the reduction of 17 fracture toughness was more significant in the unaged 18 than in the thermally aged specimens. And it goes on 19 to have a further discussion of that. And I'd like 20 to, perhaps at the beginning of tomorrow, come back 21 and discuss whether or not this supports the 22 hypothesis of the -- Dr. Lahey's hypothesis or whether 23 or not there's an explanation explaining why this 24 doesn't support the synergistic effect of the thermal 25 aging and the neutron embrittlement. But I don't want NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5255 1 to take that up right now, I just wanted to address 2 that particular document to you so you would have a 3 chance to look at it this evening and perhaps you 4 could take it up briefly tomorrow. Dr. Lahey?
5 DR. LAHEY: Yes, this is Richard Lahey, New 6 York. You had asked me to look at this report, NUREG-7 7184, and to look at the word synergistic and I did 8 look at this report and it's very interesting, the 9 later version of it was redlined out. It was 10 synergistic, but it was redlined out and combined 11 effect was put it its place. So at one point it said 12 synergistic and I guess the final issue would have 13 said combined effect.
14 CHAIRMAN MCDADE: Okay. But you would view 15 those two words as synonyms, but I don't want to take 16 it up right now and I want to let --
17 DR. LAHEY: Right.
18 CHAIRMAN MCDADE: -- Judge Wardwell move 19 on. But what I'm looking for is to whether or not, 20 not just a particular word, but whether or not this 21 NUREG and the data presented there, what impact, if 22 any, does it have on the hypothesis of the synergistic 23 effects? And, again, I don't want to take it up right 24 now, we want to move on. Take a look at it, have Dr.
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5256 1 and we can discuss it very briefly in the morning.
2 DR. HISER: Judge McDade, quick question on 3 that. Allen Hiser of the Staff. What page number of 4 7185 were you quoting from?
5 CHAIRMAN MCDADE: Well, I was quoting from 6 the abstract --
7 DR. HISER: Okay.
8 CHAIRMAN MCDADE: -- which is I think 9 especially like a little I or a little double I --
10 DR. HISER: Okay.
11 CHAIRMAN MCDADE: -- I was just quoting 12 from the abstract, but it then goes on.
13 MR. SIPOS: And your honor, for New York, 14 I think --
15 CHAIRMAN MCDADE: I didn't say that, I said 16 I was quoting from the abstract.
17 MR. SIPOS: Your honor, John Sipos for the 18 record. I think you were referring to 7184. I just 19 want to make sure that's clear for the record, rather 20 than 7185, which I think Dr. Hiser just mentioned.
21 CHAIRMAN MCDADE: Wait, now --
22 ADMIN. JUDGE WARDWELL: You initially said 23 7184.
24 MR. SIPOS: And that is Exhibit New York 25 488 --
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5257 1 CHAIRMAN MCDADE: Hold on one second.
2 MR. SIPOS: -- I believe.
3 ADMIN. JUDGE WARDWELL: While you're doing 4 that, I want to correct something else that you put 5 words in Dr. Lahey's mouth and he said right and I 6 want to verify he meant to say right to what you said.
7 Judge McDade said that you believe that synergism and 8 combined are the same. That isn't what I heard you 9 say earlier today. I heard you say today synergism 10 was more than the combined.
11 DR. LAHEY: My understanding of what he 12 said and why I said right, was he indicated that they 13 were the same as far as I was concerned. What I 14 thought was synergism and what they called combined 15 was the same.
16 ADMIN. JUDGE WARDWELL: But it's not the 17 sum of the two components in your opinion is 18 synergism. Synergism is more than that, is it not?
19 DR. LAHEY: I think it can be, yes. But it 20 may also be the same.
21 ADMIN. JUDGE WARDWELL: Be what?
22 DR. LAHEY: Well, no, I think it can be 23 either one.
24 ADMIN. JUDGE WARDWELL: You think synergism 25 can be just the sum of the two?
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5258 1 DR. LAHEY: Can have two things --
2 synergism can mean two things going on at the same 3 time or --
4 ADMIN. JUDGE WARDWELL: And it's equal to 5 only the sum of the two individual contributions?
6 DR. LAHEY: It can be, I believe, and it 7 can be --
8 ADMIN. JUDGE WARDWELL: Greater?
9 DR. LAHEY: -- more.
10 ADMIN. JUDGE WARDWELL: Okay. So you 11 believe it's both the sum and/or?
12 DR. LAHEY: I believe so. I don't know 13 what the author or --
14 ADMIN. JUDGE WARDWELL: No, what I'm asking 15 you is what's your --
16 DR. LAHEY: Yes.
17 ADMIN. JUDGE WARDWELL: -- when you say the 18 word synergism --
19 DR. LAHEY: That's my view.
20 ADMIN. JUDGE WARDWELL: -- do you mean it 21 has to be more than the sum or can it be the sum or 22 more?
23 DR. LAHEY: Yes.
24 ADMIN. JUDGE WARDWELL: No, those are 25 choices.
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5259 1 (Laughter.)
2 DR. LAHEY: It can be either or.
3 ADMIN. JUDGE WARDWELL: Thank you. Moving 4 on to see if we can confuse some other stuff or I can 5 confuse some other stuff. I want to go back to the --
6 CHAIRMAN MCDADE: Well, let me clarify. I 7 was referring to New York Exhibit 488, NUREG/CR-7184.
8 ADMIN. JUDGE WARDWELL: Right.
9 CHAIRMAN MCDADE: If I said something 10 different, I just misspoke and I apologize.
11 ADMIN. JUDGE WARDWELL: Well, you said it 12 right, Dr. Hiser said 85 and we wanted to make sure --
13 DR. HISER: I apologize.
14 MR. COX: Could I add a clarification on 15 the report numbers?
16 (Laughter.)
17 CHAIRMAN MCDADE: Pardon?
18 MR. COX: I just -- this is Alan Cox with 19 Entergy. I just wanted to point out there's two 20 versions of the NUREG/CR-7184.
21 CHAIRMAN MCDADE: I was reading from the 22 December 2014. Is that the latest?
23 MR. COX: There is I believe a later 24 version. It is New York State, you said 574?
25 CHAIRMAN MCDADE: No, I said 488.
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5260 1 MR. COX: Okay. There's a 574, it's also, 2 it's a New York State 000574, it's NUREG/CR-7184. And 3 I think Dr. Lahey alluded to two versions of this 4 report, so those may be the two versions. One of 5 those may have the word synergy and the other may not.
6 CHAIRMAN MCDADE: Okay. Thank you, Mr.
7 Cox.
8 ADMIN. JUDGE WARDWELL: Now we get to move 9 to what we've already discussed, but I want to clarify 10 again in regards to the clevis bolts that New York 11 State testimony, 482, Page 56 to 57, and 56 it's Lines 12 20 to 23 and 57 it moves on to Lines 1 through 9. It 13 says that failures of the clevis insert bolts 14 apparently caused by PWSCC were detected at a 15 Westinghouse designed reactor in 2010. Out of the 48 16 bolts in this reactor, 29 were partially or completely 17 fractured, but only seven of those damaged bolts were 18 visually detected as having failed.
19 Despite this high rate of failure, about 20 60 percent of the total bolts were damaged, and a low 21 rate of visual detection, only 24 percent of the 22 damaged bolts were detected, the Applicant proposes to 23 managing the aging degradation of clevis insert bolts 24 with visual VT3 inspections rather than preemptive 25 replacement. And I guess I'll reaffirm again that, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5261 1 Entergy, do you agree with these numbers in regards to 2 what was previously reported from Westinghouse?
3 DR. LOTT: I agree with the scale of the 4 numbers, I haven't checked the actual --
5 ADMIN. JUDGE WARDWELL: There's no reason 6 not to believe those numbers?
7 DR. LOTT: No reason not to believe those 8 numbers.
9 ADMIN. JUDGE WARDWELL: And how does that 10 low percentage, again, support the use of VT3 as an 11 inspection, when in fact almost three-quarters of 12 damaged ones will go undetected?
13 DR. LOTT: It was never our intention to 14 inspect the bolts, it was our intention to verify the 15 stability and location of the clevis itself. As we 16 talked before, the clevis can function perfectly well 17 without the bolts if it's in place. So, again, our 18 inspection is basically -- and we made recommendations 19 in the Tech Bulletin that we issued on this subject to 20 basically inspect the clevis for its seating into the 21 lug on the vessel wall and to make sure that it had 22 not moved, that there was not undue wear on the 23 surfaces that would indicate that it had moved or was 24 free to move. But we did not anticipate that a visual 25 inspection would necessarily detect cracking of a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5262 1 bolt.
2 ADMIN. JUDGE WARDWELL: But did you not 3 also go on and testify that you're considering or 4 evaluating whether or not to replace any of the bolts 5 that may have been damaged?
6 DR. LOTT: Well, first of all, at that 7 particular plant a number of the bolts were replaced.
8 They were replaced, again, because of our concern 9 about the commercial aspects, the potential that those 10 clevises might become dislodged, which would make it 11 difficult for them to, as we said, reinsert the 12 barrel, restart the plant. So we certainly were 13 advising them, and I think it explains this in the 14 Technical Bulletin, that it might be wise to replace 15 the bolts, but not because of a safety concern.
16 ADMIN. JUDGE WARDWELL: Dr. Lahey, in your 17 testimony, 482, Page 58, Lines 5 through 9, you state 18 that rather than taking proactive steps to replace the 19 degraded clevis bolts prior to failure, the Applicant 20 proposes to wait for failures to occur before taking 21 steps to address the problem, an approach that is 22 totally unacceptable in my opinion. Hearing what you 23 heard now, do you still believe that your demand for 24 wholesale replacement of these bolts are a reasonable 25 expenditure of effort?
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5263 1 DR. LAHEY: Your honor, this is Richard 2 Lahey from New York. When we had this discussion 3 earlier today, it was my understanding that to replace 4 a few of the bolts is straightforward, if you have a 5 lot of the bolts, it requires a massive effort in 6 terms of realignment. And, so, I believe based on 7 that and the safety significance of it, it would 8 really depend on the degree of failure of the bolts.
9 What would make sense in any event, you don't want 10 loose parts rattling around.
11 ADMIN. JUDGE WARDWELL: In regards to that 12 last statement, Entergy, do these failures of the 13 bolts end up with loose parts? Or are they kind of 14 contained with the --
15 DR. LOTT: It's contained within the 16 system. It's difficult for the bolt head -- first of 17 all, there's locking bars. If there was wear through 18 or failure of the locking bars, still the bolt head 19 could not escape and the threaded part of the bolt 20 can't get past the head.
21 ADMIN. JUDGE WARDWELL: Thank you. Just 22 reading through here to -- Staff, is there any place 23 within your SER or SE that you evaluated the clevis 24 insert bolts and what is an appropriate action level, 25 if any, for these documented --
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5264 1 MR. POEHLER: This is Jeffrey Poehler for 2 the Staff. Yes, we do address that in our SER in the 3 discussion of operating experience, which is located 4 right after the ten element discussion. And we --
5 yes, we reviewed that quite extensively and we agreed 6 with continuing to do VT3 examination as an acceptable 7 means of managing the potential for bolt failures.
8 ADMIN. JUDGE WARDWELL: Okay, thank you.
9 New York State's testimony, 482, Page 57, Line 20 10 through 23, and then moving over to 58, Line 1, says, 11 the Applicant's analysis of the effects of clevis bolt 12 failures assumes that all of the components will be 13 functioning according to their design specification 14 and does not consider the fact that other components 15 may also be undergoing degradation from various 16 interacting mechanisms.
17 Entergy's testimony, Exhibit 616, Answer 18 166, Page 107, says, Entergy is not required to assume 19 without evidence that other components that are within 20 the scope of the reactor vessel internals AMP or any 21 other AMP are also degraded when it evaluates the 22 functionality of the clevis insert bolts. And I guess 23 considering we have the two competing statements, I'll 24 go to Staff and ask them. What do you look for and 25 what do you consider in regards to the potential for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5265 1 other components outside of the AMP, the RVI AMP, that 2 might not be functioning along with these clevis bolts 3 and how are they assessed? Or is there no requirement 4 to do that and no need for safety to do that?
5 MR. POEHLER: This is Jeffrey Poehler from 6 the Staff. We would not require them to assume that, 7 for instance, the part that interfaces with the clevis 8 insert is called the radial key and that's attached to 9 the bottom of the core barrel, and we would not 10 require them to assume that, that had failed or expect 11 them to do that when doing their functionality 12 evaluation of the clevis or the lower radial support 13 system. And in addition to that, they're both 14 redundant components, both the clevis inserts and the 15 radial keys, I believe there's six. So the likelihood 16 of a significant number of those failing at once is 17 low. And the radial keys were not even a -- they were 18 a no additional measure component, which means there 19 were no inspection requirements in MRP 227-A.
20 ADMIN. JUDGE WARDWELL: Are there any 21 components that do have inspection requirements that 22 are associated with one another where you have looked 23 at the potential failure of both in the reactor vessel 24 internals?
25 DR. HISER: In general --
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5266 1 ADMIN. JUDGE WARDWELL: Not necessarily 2 with the clevis system.
3 DR. HISER: This is Allen Hiser with the 4 Staff. We do not require postulation of failures of 5 other components when assessing a finding of a 6 degraded component. That's not a part of the 7 regulatory process.
8 ADMIN. JUDGE WARDWELL: And do you know any 9 technical basis for supporting that position?
10 DR. HISER: I think technical basis is just 11 Commission position that it's not required as a part 12 of the license renewal review and that annotation of 13 license renewal.
14 ADMIN. JUDGE WARDWELL: Thank you. Let's 15 move on to the lower support columns. NRC's 16 testimony, 197, Answer 27 on Pages 35 to 36, the fuel 17 assemblies are supported inside the lower internals 18 assembly on top of the lower core plate, that's LCP, 19 and the function of the LCP are to position and 20 support the core and provide a metered control of 21 reactor coolant flow into each fuel assembly. The 22 support columns transmit vertical fuel assembly loads 23 from the LCP to the much thicker lower support casing.
24 The function of the lower support casing is to provide 25 support for the core. My question to Staff, NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5267 1 considering it is your testimony, are these support 2 columns you're referring to the lower support columns?
3 Are they one and the same?
4 MR. POEHLER: This is Jeffrey Poehler of 5 the Staff. That's correct.
6 ADMIN. JUDGE WARDWELL: And for these 7 columns, what is the primary mechanism for developing 8 flaws from aging? Is it driven by normal operating 9 conditions?
10 MR. POEHLER: It would be normal operating 11 conditions.
12 ADMIN. JUDGE WARDWELL: And would other 13 operating conditions, such like seismic or LOCA 14 events, likely be the primary contributor to service 15 induced flaws and, if not, why not?
16 MR. POEHLER: No, those other events, like 17 seismic events would not be a significant contributor 18 because those events occur very infrequently. So the 19 --
20 ADMIN. JUDGE WARDWELL: Your testimony on 21 197, Answer 171, Page 92, says that the Action Level 22 7 requires an applicant or a licensee to perform a 23 plant specific analysis of the cast Austenitic 24 stainless steel RVI components to demonstrate that 25 components will remain capable of performing their NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5268 1 intended functions through the end of the plant life.
2 Your Answer on 163, Page 93, goes on to state that 3 Entergy identified the only components requiring a 4 response to Action Level 7 for IP2 and 2 are the lower 5 support columns.
6 Only the upper portion of the lower 7 support columns, known as the column cap, is made from 8 CASS. The lower support columns are an expansion 9 component of MRP 227-A and the associated linked 10 primary component is a control rod guide tube lower 11 flange welds. So, let me make sure I understand this 12 correctly. Is this not saying to me that the only 13 CASS materials in the population of reactor vessel 14 internals are the upper portion of the lower support 15 columns, i.e., what's called the column cap?
16 MR. POEHLER: It's not the only CASS 17 component in the internals, or the only type of CASS 18 component, I should say, at Indian Point. It is the 19 one that is of most concern to the Staff because of 20 the high irradiation levels that parts of it can 21 experience.
22 DR. HISER: Dr. Wardwell, this is Allen 23 Hiser of the Staff. Action Item 7 specifically 24 identifies the lower support column bodies as within 25 the scope of that Action Item. And it is due to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5269 1 potential for thermal embrittlement and irradiation 2 embrittlement of those. And those are the only ones 3 that were identified that would be within the scope of 4 irradiation embrittlement. So that's why it's limited 5 only to the lower core support columns.
6 ADMIN. JUDGE WARDWELL: But does the Action 7 Level 7 state that only that component is required or 8 is it more general that it requires an applicant or a 9 licensee to perform a plant specific analysis of CASS 10 components?
11 DR. HISER: Those are specified within the 12 Action Item, the lower core support columns for 13 Westinghouse plants. The intent of that Action Item 14 was to identify CASS that's within the high fluence 15 field that could lead to irradiation embrittlement.
16 Within the context of what was evaluated in MRP 227, 17 that was the only generic CASS component. Referring 18 back to Action Item 2, where plants are to evaluate 19 differences between the plant specific configuration 20 and MRP 227, if CASS was atypically used in a place at 21 Indian Point different from MRP 227, then that would 22 have been identified in Action Item 2 and also should 23 show up in Action Item 7.
24 ADMIN. JUDGE WARDWELL: And so what's the 25 criteria for being under 2 again? For an applicant to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5270 1 highlight whether a CASS composed material, an 2 internal composed of CASS, that's the way I want to 3 say it, I guess, would or would not require the 4 additional attention brought on by Action Level 7?
5 DR. HISER: It says that applicants should 6 review the information in Tables 4-12 and 4-2 in MRP 7 189 Rev 1, Tables 4-4 and 4-5 in MRP 191, and identify 8 whether these tables contain all the RVI components 9 that are within the scope of license renewal for their 10 facilities. I think as a part of that, because 11 material is critical, that, that would be identified 12 as well as a difference between the plant specific 13 configuration and the MRP 227 assumptions.
14 ADMIN. JUDGE WARDWELL: I guess I'm still 15 a little confused. Are the upper portion of the lower 16 support columns the only RVIs under license renewal 17 that are made of CASS? And, if not, how did the 18 others get screened out and what was the criteria for 19 screening them out, for not being highlighted under 20 Action Level 7?
21 DR. HISER: This is Allen Hiser of the 22 Staff. For Indian Point, our understanding is those 23 are the only CASS components that are subject to 24 irradiation embrittlement above the threshold level 25 used in the development of MRP 227. So they would be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5271 1 the only CASS components that would be subject to the 2 potential susceptibility of thermal embrittlement and 3 irradiation embrittlement.
4 ADMIN. JUDGE WARDWELL: And what is that 5 screening criteria that would separate out those that 6 were and were not susceptible to thermal embrittlement 7 and irradiation embrittlement?
8 MR. POEHLER: This is Jeffrey Poehler of 9 the Staff. The MRP screening criteria was one dpa.
10 ADMIN. JUDGE WARDWELL: Okay. And 11 Westinghouse, is everything that was stated consistent 12 with your approach and is, as a net result, the upper 13 portion of the lower column supports, that is made of 14 CASS material, the only RVI that does not meet the 15 screening criteria of one dpa of fluence and, 16 therefore, is part of and falls under Action Level 7?
17 DR. LOTT: For the most part, I believe 18 that's true. I'm recalling --
19 ADMIN. JUDGE WARDWELL: Well, then the 20 other parts I'm interested in.
21 DR. LOTT: I know and I'm trying to get 22 there. I'm not -- really I'm trying to get there.
23 And there's one component that I need to check on and 24 that might be the upper support column, which I 25 believe in my previous testimony indicated that it was NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5272 1 potentially CASS. I don't know if it is or not.
2 ADMIN. JUDGE WARDWELL: Let's put that on 3 your homework list for tomorrow morning then to 4 determine whether there are any RVIs --
5 DR. LOTT: Yes. I would point out that 6 there are some other CASS components in the system and 7 that part of the evaluation procedure that led to this 8 screening process in MRP 191 was an evaluation of the 9 impact of those components or of those degradation 10 mechanisms. Of course, they would have been 11 identified for thermal embrittlement anyway.
12 ADMIN. JUDGE WARDWELL: Do you agree that 13 the one dpa is the threshold screening criteria or is 14 it not?
15 DR. LOTT: Yes.
16 ADMIN. JUDGE WARDWELL: It is?
17 DR. LOTT: Yes, it is. But --
18 ADMIN. JUDGE WARDWELL: So we're back to --
19 DR. LOTT: -- what I would say, what the 20 classification process that identified components to 21 be in MRP 227 or not also had this evaluation of the 22 structural significance of the component. So it's 23 possible -- and as I understand the documentation in 24 Action Item 7, it basically said that the NRC had 25 reviewed those decisions in MRP 191 and said that if NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5273 1 things were appropriately dealt with there, there was 2 not a need for functionality analysis under Action 3 Item 7. So it grandfathered in things that were 4 identified within MRP 191 as potentially CASS and 5 evaluated and screened out, are not required to go 6 through this process. And that I'm believe is true of 7 the upper support columns.
8 ADMIN. JUDGE WARDWELL: Well, that's -- I'm 9 interested in whether or not all of those that haven't 10 either been grandfathered out by 191 or haven't been 11 screened out due to the criteria of the one dpa are 12 now present in or not highlighted by 7 now, the 13 reasons why they aren't? Well, I do not want to read 14 that question in the transcript at all.
15 (Laughter.)
16 DR. LOTT: I think I can -- I don't think 17 that, that description you -- there are any components 18 that meet that description that you just said.
19 ADMIN. JUDGE WARDWELL: I just want to have 20 you verify that you have captured all of those that 21 need to be captured and that the lower support column 22 is the only one that needs to be captured by 7.
23 DR. LOTT: Okay. Action Item 7, that I 24 believe is true.
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5274 1 for Entergy. I just want to say, one other parameter 2 that may come into consideration here is the 3 resistance of CASS material to any type of formation 4 of any type of corrosion cracks. And we haven't 5 talked about that, but this material, although it can 6 permeate and we talked about that earlier, depending 7 on the ferrite content, et cetera, but the good thing 8 about this material is that it's very hard to crack.
9 And I'm not sure that there's any operating experience 10 in which we actually have seen cracking in these CASS 11 materials. To my knowledge, we haven't, unless 12 there's something fairly recently. So, that may also 13 factor into how this is treated and we'll have to look 14 at that.
15 ADMIN. JUDGE WARDWELL: Thank you. Back to 16 Entergy again. The upper portion of the lower support 17 column is made of CASS, what's the lower portion made 18 of?
19 DR. LOTT: The wrought stainless steel.
20 ADMIN. JUDGE WARDWELL: The raw?
21 DR. LOTT: Wrought stainless steel. I'm 22 sorry, I'll get closer to the microphone again.
23 ADMIN. JUDGE WARDWELL: I thought I heard 24 you say raw. Okay, wrought stainless steel. Thank 25 you. Dr. Lahey, do you have any knowledge of any NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5275 1 potential RVIs that you believe should fall under 2 A/LAI 7 in regards to the Aging Management Program?
3 DR. LAHEY: Your honor, I have reviewed 4 EPRI 191 and I'll have to look back on it, it's in the 5 other room. But my recollection is there were core 6 plates, upper and lower, that were castings, but I'll 7 have to verify that.
8 ADMIN. JUDGE WARDWELL: Okay. So, Entergy, 9 you might note that also that those plates are of 10 interest and comment on that if you would.
11 DR. LOTT: Okay.
12 ADMIN. JUDGE WARDWELL: Thank you, Dr.
13 Lahey. Answer of NRC's 197, 163 on Page 94, states 14 that the lower support columns for IP2 and 3 are made 15 from type CF-8 stainless steel, which is a low 16 molybdenum contact grade of cast stainless steel. Low 17 molybdenum cast grades are less susceptible to thermal 18 embrittlement than the high moly cast grades, such as 19 types CF-8M. Entergy determined the lower support 20 columns at IP2 and 3 were not susceptible to thermal 21 embrittlement. And, again, do these statement refer 22 only to the cap as being not susceptible to thermal 23 embrittlement because the lower wrought iron has 24 already been screened out, it's not part of A/LAI 7?
25 DR. LOTT: Yes. The lower cast of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5276 1 Austenitic stainless steel, wrought stainless steel, 2 is not subject to thermal embrittlement. It would 3 only be the upper cast portion.
4 ADMIN. JUDGE WARDWELL: And so whenever we 5 talk about -- whenever you reference lower support 6 under this section of the CASS portion, you mean just 7 the upper part? Let me rephrase that question. When 8 you are referring to the cap of the lower support 9 system, that's what you're referring to is that upper 10 part that's made out of CASS, is that correct?
11 DR. LOTT: Yes.
12 ADMIN. JUDGE WARDWELL: And is that the top 13 half of the column? Is it at the top little piece of 14 the column? Or is it a top quarter of the column?
15 DR. LOTT: It's more like quarter to a 16 third of the column. I don't think it --
17 ADMIN. JUDGE WARDWELL: But it's not that 18 little, I saw a diagram of the column and there's a 19 little rectangular piece on the top, that's not the 20 cap, it's more than that?
21 DR. LOTT: No, it's more than that.
22 ADMIN. JUDGE WARDWELL: It's a --
23 DR. LOTT: Yes.
24 ADMIN. JUDGE WARDWELL: -- major portion --
25 DR. LOTT: It's --
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5277 1 ADMIN. JUDGE WARDWELL: -- a portion of 2 that --
3 DR. LOTT: Right.
4 ADMIN. JUDGE WARDWELL: -- entire column?
5 MR. DOLANSKY: It has some length to it.
6 ADMIN. JUDGE WARDWELL: Yes, okay. Just so 7 we have a feeling of it. And by screening it out as 8 being unsusceptible to thermal embrittlement, that by 9 necessity means that the caps cannot be susceptible to 10 the combined effects of irradiation embrittlement and 11 thermal embrittlement, it's just the irradiation 12 embrittlement, is that correct? Entergy?
13 MR. DOLANSKY: Yes, that's correct. This 14 is Bob Dolansky. Yes, that's correct.
15 ADMIN. JUDGE WARDWELL: And, Staff, as far 16 as you're concerned, you agree with that conclusion?
17 MR. POEHLER: This is Jeffrey Poehler with 18 the Staff. Yes, we agree with that conclusion.
19 ADMIN. JUDGE WARDWELL: Dr. Lahey, do you 20 agree that just the upper portion of this column cap 21 is --
22 DR. LAHEY: Yes, sir, that's my 23 understanding as well.
24 ADMIN. JUDGE WARDWELL: Thank you. The A 25 163, Page 95 goes on to say that screening criteria NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5278 1 would allow the low molybdenum cast with delta ferrite 2 content of 20 percent or less to be screened out, that 3 is not susceptible to thermal embrittlement. Since 4 portions of the lower support column will have neutron 5 fluences at the end of life greater than one times ten 6 to the 17 neutrons per square centimeter, the Staff 7 did not accept Entergy's screening and the Staff 8 updated the criteria of low molybdenum statically cast 9 CASS with a ferric content of less than 15 percent can 10 be screened out for thermal embrittlement and any 11 synergistic effects of TE and IE and as low molybdenum 12 statically cast CASS is only susceptible to 13 irradiation embrittlement at fluences greater than one 14 times ten to the 21 neutrons per square centimeter or 15 1.5 displacements per atom.
16 Staff, could you try to condense that 17 whole confusing thing I just read, or a bit confusing 18 to me, in regards to how these lower columns were 19 specifically, and specifically the lower portion of 20 this that's called the cap -- the upper portion of the 21 lower column, were screened out and assessed by you?
22 DR. HISER: Okay. This is Allen Hiser of 23 the Staff. The top of that paragraph talks about the 24 20 percent or less screening for thermal 25 embrittlement.
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5279 1 ADMIN. JUDGE WARDWELL: And that's for 2 delta ferrite content, correct?
3 DR. HISER: Right. That's correct. And 4 that was proposed for really CASS that was not subject 5 to neutron embrittlement. So it was subject to low 6 neutron fluence. With the possibility of higher 7 neutron fluence levels combined with the potential for 8 thermal embrittlement, we went back and looked at the 9 screening, the 20 percent, and concluded that we 10 should reduce that level by five percent to 15 11 percent. And we thought that, that was a reasonable 12 screen to preclude the potential synergism of thermal 13 embrittlement and irradiation embrittlement.
14 When Indian Point then came in with their 15 measured ferrite values, or their ferrite values for 16 their lower core support column caps, and it was below 17 15 percent, we thought it was reasonable to screen 18 that out then from potential synergistic effects of 19 thermal embrittlement and irradiation embrittlement.
20 So what that leaves then for those components is only 21 a concern with the irradiation embrittlement on the 22 fracture toughness for those materials.
23 ADMIN. JUDGE WARDWELL: And the fluence 24 screening and any molybdenum screening didn't play a 25 part in this?
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5280 1 DR. HISER: The molybdenum screening did 2 not because this was a low molybdenum material. The 3 fluence level did from the perspective that the 4 fluence exceeded one dpa and, therefore, the 15 5 percent delta ferrite screening level was implemented 6 instead of the 20 percent that would apply for lower 7 fluence CASS material.
8 ADMIN. JUDGE WARDWELL: Thank you. Dr.
9 Lahey, do you agree that this was an appropriate 10 screening process and that the lower column caps have 11 been successfully screened and are being assessed in 12 the way they should be for irradiation embrittlement 13 alone?
14 DR. LAHEY: Yes, your honor. Given the 15 criteria that's been established, it sounds like it 16 has been properly done.
17 ADMIN. JUDGE WARDWELL: Thank you, sir.
18 New York State's testimony on 482, Page 18, Lines 16 19 through 22, discusses a recent report prepared by 20 Argonne National Laboratory for the U.S. NRC and 21 acknowledges with respect to CASS materials that are 22 used for the lower support column caps that combined 23 effect of thermal aging and irradiation embrittlement 24 could reduce the fracture resistance even further to 25 a level neither of these degrading mechanisms can NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5281 1 impart alone.
2 And I guess I would just ask, and we've 3 already verified that it's been screened correctly, so 4 really what my question is here, I want to make sure 5 what Argonne National Laboratory work and report that 6 you were referring to in this statement. Do you 7 recall, Dr. Lahey? Is it NUREG/CR-7184 or is there 8 some other one? Because you go on, on Page 20, to 9 talk about NUREG-7184 and then even the Chopra report, 10 the degradation of light water reactor core internal 11 materials due to neutron irradiation, which is NUREG-12 7027. And I'm just trying to sort out all these 13 reports.
14 DR. LAHEY: Your honor, I'd have to go back 15 and look to know for sure.
16 ADMIN. JUDGE WARDWELL: Okay. Add that on 17 to your list.
18 DR. LAHEY: Yes.
19 ADMIN. JUDGE WARDWELL: And do try to get 20 to bed by at least 3:00 a.m. if you could. Because 21 we're starting here tomorrow at 4:00 aren't we?
22 (Laughter.)
23 CHAIRMAN MCDADE: Thirty.
24 ADMIN. JUDGE WARDWELL: 4:30, I'm sorry.
25 (Laughter.)
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5282 1 ADMIN. JUDGE WARDWELL: Yes, thank you.
2 And specifically to get back to that --
3 CHAIRMAN MCDADE: I digress for a second, 4 but Judge Wardwell has stayed on Central European Time 5 forever. So he insists that we do start early and 6 would prefer that we start at 4:00 Eastern.
7 ADMIN. JUDGE WARDWELL: No, because I 8 wouldn't be ready. I need that time for being ready.
9 (Laughter.)
10 ADMIN. JUDGE WARDWELL: It's your testimony 11 in 482 on Page 18, Line 16 through 22, and then it's 12 your testimony Page 20 on Lines 6 through 10 where 13 these various -- on 16 through 22, you talk about the 14 Argonne National Laboratory report and then you go on, 15 on Page 20 to cite both 7184 and 7027.
16 DR. LAHEY: Okay.
17 ADMIN. JUDGE WARDWELL: NRC's testimony on 18 197, Answer 164, Pages 96 to 97, states that the Staff 19 found that Entergy adequately addressed Action Level 20 7 based on the following. One, Entergy evaluated the 21 CASS components of the RVIs and that is limited to the 22 lower support column caps. Entergy screened the 23 column caps for TE and IE using plant specific 24 materials data and determined that the column caps are 25 not susceptible to TE. And Staff confirmed the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5283 1 results of the screening using its own screening 2 criteria.
3 Three, Entergy provided information on 4 fabrication, non-destructive NDE demonstrating that 5 preexisting flaws are unlikely to exist in the column 6 caps. Four, Entergy provided information on expected 7 stresses and neutron fluences for the column caps that 8 demonstrated the service induced cracking due to 9 irradiation assisted stress corrosion cracking is 10 unlikely. And, five, Entergy modified its reactor 11 vessel internals program to include a link to a lead 12 component that is an appropriate predictor of IASCC 13 and IE for column caps with an appropriate schedule 14 for performing the expansion inspection if necessary.
15 Therefore, the Staff found that the 16 formation provided by Entergy provides reasonable 17 assurances that the functionality in the column caps 18 will be maintained during this period of extended 19 operation. Entergy goes on in their testimony and 20 states in Exhibit 616 under Answer 197, Page 131, that 21 in regards to the inspection criteria, the lower core 22 barrel cylindrical girth weld reveals that it is more 23 likely to experience this irradiation assisted stress 24 corrosion cracking than the lower support column caps.
25 So it is appropriate to link the two with the weld as NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5284 1 a primary component and the LSCC as the expansion 2 component.
3 And I guess I'll start with Staff on this 4 and say that, in your approval of this as summarized 5 in that, that I just read, you mentioned about linking 6 to a component that is the primary component and I 7 thought earlier, which I think I read, you stated that 8 the control rod guide tubes lower flange welds were 9 the primary link to the lower support column caps.
10 And now Entergy is claiming that it's the lower core 11 barrel cylinder girth weld is the one. Which is it?
12 MR. POEHLER: At this time, it's both 13 because they added the girth weld as a plant specific 14 primary link. So in MRP 227-A, the generic primary 15 link for the lower core support columns is the control 16 rod guide tube lower flange. When the Staff was 17 reviewing the Action Item 7 information submitted by 18 Entergy, we were satisfied with what they had given 19 us, but we still had a lingering concern that for 20 irradiation embrittlement, because irradiation 21 embrittlement still applies even though thermal 22 embrittlement was screened out, we still had a 23 lingering concern that irradiation embrittlement 24 wasn't appropriately predicted by the standard MRP 25 primary link.
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5285 1 Because that control rod guide tube lower 2 flange weld is a relatively low fluence component, 3 didn't even screen-in for irradiation embrittlement or 4 irradiation assisted stress corrosion cracking, which 5 are two mechanisms that do apply to the lower support 6 columns. So we asked another RAI to Entergy regarding 7 an appropriate -- can you propose an appropriate 8 primary link for these mechanisms? And that's when 9 they proposed the lower core barrel girth weld. And 10 we reviewed that information that they submitted and 11 concluded that it was a more appropriate primary link 12 for those mechanisms of irradiation embrittlement and 13 irradiation assisted stress corrosion cracking.
14 ADMIN. JUDGE WARDWELL: Okay. Thank you.
15 Dr. Lahey, do you have any reason to not accept those 16 as appropriate primary links to these expansion 17 components?
18 DR. LAHEY: The use of the girth weld as a 19 proxy for the core cap, it seems an odd choice 20 actually because of the materials used and how they 21 would respond to thermal and irradiation effects. And 22 I know they needed to use something because they 23 weren't able to do an inspection of the core caps, but 24 it just seems it's a very odd choice. And I must say 25 when I was reading the ACRS testimony on this, they NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5286 1 had exactly the same opinion. So I know what they 2 did, it just seems odd.
3 ADMIN. JUDGE WARDWELL: And do you feel 4 it's also odd to use the control rod guide tube lower 5 flange welds as the other primary?
6 DR. LAHEY: It's not a casting. Yes.
7 ADMIN. JUDGE WARDWELL: So, Entergy, why 8 have you proposed the lower core barrel cylinder girth 9 weld as the primary?
10 DR. LOTT: Well, as Mr. Poehler explained, 11 it was basically a request coming from the NRC to 12 consider an alternative to the lower control rod 13 guide, I can't say it either myself, the CRGT lower 14 flange welds. And we looked at similarities in 15 materials and we understood that -- we looked at the 16 lower support column caps as a low ferrite material, 17 not susceptible to thermal embrittlement. And we 18 looked at the core barrel weld material and realized 19 it also would have a duplex structure, a low ferrite 20 content, and then some very similar characteristics.
21 Both components had similar fluences and in an 22 abundance of caution here, I think even though we 23 probably could have demonstrated that it would screen-24 out for irradiation assisted stress corrosion 25 cracking, I think we just wanted to make sure that we NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5287 1 were looking for the most likely locations for cracks.
2 ADMIN. JUDGE WARDWELL: Okay.
3 DR. LOTT: And let me just one further --
4 I think the tipping point that made the core barrel 5 the primary location for this particular situation was 6 that based on the potential for residual stresses in 7 the core barrel weld, we expect the stresses, 8 particularly on the surface, to be much higher in the 9 core barrel weld, where as we've just explained, we 10 looked through the normal operating stresses in the 11 lower support columns and found they're very limited.
12 So we would expect to see cracking in these kinds of 13 materials at these fluences first in the core barrel 14 and later in the lower support column, if ever.
15 ADMIN. JUDGE WARDWELL: Okay. What about 16 the control rod guide tubes lower flange welds, will 17 they also show up any cracking prior to the lower 18 support column caps?
19 DR. LOTT: Well, again, they have some 20 potential to have some residual stresses and some 21 other concerns that might put them ahead on the list.
22 But because they weren't irradiated, we felt that this 23 addition that Entergy has of using both as a primary 24 component made the most sense.
25 ADMIN. JUDGE WARDWELL: Now, then for Staff NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5288 1 --
2 CHAIRMAN MCDADE: If I could interrupt for 3 a second. Dr. Lott, when we come back to you, if you 4 could just state your name, otherwise --
5 DR. LAHEY: Okay.
6 CHAIRMAN MCDADE: -- Mr. Cox is going to 7 get --
8 DR. LAHEY: I'm sorry.
9 CHAIRMAN MCDADE: -- blamed in the 10 transcript for things that you say.
11 DR. LAHEY: Yes, I'm sorry.
12 CHAIRMAN MCDADE: Okay.
13 ADMIN. JUDGE WARDWELL: Take credit. Dr.
14 Hiser, this discussion raises a question in my mind, 15 why shouldn't these lower support column caps be 16 elevated to a primary component? And what is the 17 criteria that puts something as a primary component as 18 opposed to an expansion component, if I've used the 19 right two terminologies?
20 DR. HISER: This is Allen Hiser of the 21 Staff. One of the reasons is that the column caps are 22 a highly redundant structure. And, therefore, it 23 would require multiple failures for the functionality 24 of the lower core support system to be challenged. In 25 this case, as Dr. Lott mentioned, the residual NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5289 1 stresses in the core barrel weld create a situation 2 where we expect that cracking would initiate much 3 sooner than in the column caps. Therefore, we believe 4 it's a reasonable linkage with the column caps being 5 an expansion component predicated on inspection 6 findings from the core barrel girth welds.
7 ADMIN. JUDGE WARDWELL: And why still hold 8 with the control rod guide tubes lower flange welds?
9 Why should that still be a primary for this? It 10 sounds like you don't feel comfortable that it will 11 supply any advanced notice of what might happen or 12 substitute notice for what might happen at the lower 13 support column caps.
14 DR. HISER: This is Allen Hiser with the 15 Staff. The CRGT welds, I believe, are still a primary 16 link for the column caps. Again, because of weld 17 residual stresses, we would expect the likelihood of 18 stress corrosion cracking to be higher there. So 19 they, again, if you were to look at a hierarchy of 20 where you expect things to occur first, the CRGT welds 21 and the lower core barrel girth weld would be higher 22 than the column caps.
23 ADMIN. JUDGE WARDWELL: Is it possible to 24 inspect the lower support column caps?
25 DR. HISER: From discussions we've had with NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5290 1 Westinghouse, I believe it is possible, yes.
2 ADMIN. JUDGE WARDWELL: Okay.
3 DR. HISER: Our understanding is that it 4 would be an extremely difficult inspection to do, but 5 it is certainly possible and would be necessary 6 pending the results from the primary inspections, if 7 expansion was required then the plant would be 8 required to do those expansion inspections.
9 ADMIN. JUDGE WARDWELL: Dr. Lahey, after 10 hearing this additional explanation, do you have any 11 other comments in regards to the expansion monitoring 12 plan for the lower support column caps?
13 DR. LAHEY: No. I understand what they 14 said and I understand the rationale. Just personally, 15 I think if you're interested in core cap, it's better 16 to do the inspection on the core cap, even if it's a 17 more difficult inspection.
18 ADMIN. JUDGE WARDWELL: Okay, thank you.
19 New York State's testimony, 482, Page 59, Lines 1 20 through 3, states that the Applicant, based on a lack 21 of documented fractures of core support columns, 22 assumed that only a limited number of columns could 23 actually contain flaws of significant size. And I 24 guess I'll ask Entergy, did you assume only a limited 25 number of lower support column caps contained flaws or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5291 1 was this the result of any analysis or conclusion or 2 was it just an assumption?
3 DR. LOTT: I'm not sure exactly where we 4 think we -- my name is Randy Lott, I'm sorry, thank 5 you. I'm not sure exactly --
6 ADMIN. JUDGE WARDWELL: Alan doesn't want 7 to take credit for what you're saying or Mr. Cox 8 doesn't want to take credit for what you're saying.
9 DR. LOTT: I'm not sure exactly where we're 10 saying we made this assumption. We certainly did do 11 analyses as Dr. Hiser has suggested to look at 12 effectively the redundancy in the system, in the 13 system of calculations that are very similar to what's 14 done for the baffle former bolts to look at minimum 15 requirements, and found that there was a large degree 16 of redundancy in the system. That's all reported in 17 -- it's a proprietary report, but it's PWROG14-048, 18 which is ENT 667.
19 ADMIN. JUDGE WARDWELL: Did you have any 20 other knowledge of any potential flaws or lack thereof 21 in the lower support column caps?
22 DR. LOTT: We have no expectation of flaws.
23 MR. DOLANSKY: This is Bob Dolansky with 24 Entergy. I just want to add, we also -- for the lower 25 support column caps, when they were installed they had NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5292 1 detailed NDE radiography performed on them, which 2 would give very, very high confidence that there were 3 no flaws in those when they were installed.
4 ADMIN. JUDGE WARDWELL: And what is that 5 testing composed o?
6 MR. DOLANSKY: Basically similar to an x-7 ray on a person. It's an x-ray of the lower support 8 column cap, was performed on all column caps, and 9 determined that there was no flaws in -- it's the, I 10 don't want to say the best, but it's an extremely good 11 NDE technique used to verify that there's no flaws.
12 ADMIN. JUDGE WARDWELL: Thank you.
13 CHAIRMAN MCDADE: If I could jump in here 14 quick between questions and get something clarified.
15 A while ago, a few minutes, we were talking about, and 16 I believe that you said, Dr. Lahey, that you thought 17 it was an odd choice of using the core barrel girth 18 weld as a proxy for another item. Do you recall that 19 testimony?
20 DR. LAHEY: Yes.
21 CHAIRMAN MCDADE: Okay. And --
22 DR. LAHEY: For the column caps.
23 CHAIRMAN MCDADE: -- the control guide tube 24 flange, was it or --
25 DR. LAHEY: Well, I believe we were talking NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5293 1 about that and the cast column cap that was --
2 CHAIRMAN MCDADE: Okay. But what I'm 3 trying to get at is, what I thought I heard you say is 4 that you thought that it was an odd proxy because of 5 the differences in the materials that they were made 6 out of. And then I thought I heard Dr. Lott testify 7 that part of the reason that they were chosen is 8 because of the similarity of the materials. So, let 9 me go to Dr. Lott, would you state for the record what 10 components we're talking about and what they're made 11 of?
12 DR. LOTT: Okay. This is Randy Lott for 13 Entergy. We're talking about the large structural 14 weld basically around the middle, I'll call it the 15 belt line, of the core barrel and these column caps, 16 which obviously we've been discussing all along, which 17 are CF-8 material, but with a relatively low ferrite 18 content of about 15 percent. They're castings, which 19 means that effectively they're melt, cast, and cooled 20 in the system, which gives us this duplex grain 21 structure. In fact, we need some duplex structure in 22 order to get a successful casting, so that's -- in a 23 weld, you get a very similar micro-structure or 24 material structure because, again, it's a material 25 that's welded and solidified.
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5294 1 And you get very similar behavior in terms 2 of duplex construction of those material, albeit 3 sometimes a slightly lower ferrite content. So, in 4 the sense that they're both these duplex Austenitic 5 ferritic structures, there's some real similarities in 6 the contents. Particularly, it seemed to us, because 7 there was particularly low ferrite content at Indian 8 Point, it made sense to couple up these two materials.
9 Then, as I said, the controlling factor for us was the 10 large potential residual stresses, which would drive 11 much more cracking in the core barrel, much sooner, 12 than it would in the lower support columns.
13 CHAIRMAN MCDADE: Okay. Thank you, Dr.
14 Lott. And, Dr. Lahey, what were the dissimilarities 15 that you thought made it an odd proxy?
16 DR. LAHEY: This is Richard Lahey from New 17 York State. The dissimilarities are the metals 18 themselves. The weld has typically significantly 19 lower delta ferrite than would be in the casting. And 20 the stresses on the structure are significantly 21 different. As I understood it, one of the things they 22 liked about the two were the potential for corrosion 23 cracking. But it's a stretch, I just think it's quite 24 a stretch to look at one and say, it's going to happen 25 here before it happens there. So they're looking for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5295 1 the canary in the coal mine to decide what they might 2 want to do on the core cap. And I, frankly, I think 3 it's better to look at the coal mine, go down and 4 check the core cap.
5 CHAIRMAN MCDADE: Okay. Thank you, Dr.
6 Lahey.
7 ADMIN. JUDGE WARDWELL: Do you have any 8 reason to believe that the cracking would not occur 9 sooner in the lower core barrel cylinder girth weld 10 prior to anything happening with the lower support 11 column caps?
12 DR. LAHEY: Your honor, I think the 13 stresses on the things are -- there's no reason for 14 them to be similar. You have one loaded supporting 15 structure, you have the other one loaded in this way.
16 It's just -- to me it's a very odd choice. I heard 17 what they said, I do understand why they decided to do 18 it, but primarily I think it's driven by 19 accessibility.
20 ADMIN. JUDGE WARDWELL: Thank you.
21 DR. HISER: Judge Wardwell, could we 22 supplement that a little bit?
23 ADMIN. JUDGE WARDWELL: Yes, you may.
24 DR. HISER: I think you've heard a little 25 bit -- this is Allen Hiser of the Staff. You've heard NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5296 1 a little bit about the reasons that the Applicant 2 thought that was acceptable. The Staff basis is sort 3 of similar, but there may be at least two additional 4 factors. As Dr. Lahey mentioned, the column caps will 5 be in compression. The core barrel girth weld, the 6 CRGT welds will not. So it's much more likely that 7 they will exhibit stress corrosion cracking because of 8 the weld residual stresses first and the higher 9 membrane stresses from the operational loads.
10 In addition, the column caps are cast 11 Austenitic stainless steel, which has generally been 12 found to have a very, very low likelihood of 13 initiating stress corrosion cracks. So, therefore, 14 the material in the column caps is much better from a 15 likelihood of stress corrosion cracking. The other 16 materials have much higher stresses and so they will 17 be much more likely to have cracking prior to the 18 column caps. So that provides a more complete 19 picture, I think, of the basis for the Staff finding 20 that to be acceptable.
21 ADMIN. JUDGE WARDWELL: Thank you, Dr.
22 Hiser. New York State's testimony, 482, Page 61, 23 Lines 11 through 17, states that the reactor vessel 24 internal components made from non-cast stainless steel 25 will also experience the combined effects of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5297 1 irradiation induced embrittlement, corrosion, and 2 other aging mechanisms. The Applicant has failed to 3 evaluate the mechanisms that occur for many of the 4 other important and vulnerable RVI components, such as 5 the core baffles, the baffle bolts, and the formers.
6 Entergy's testimony, Exhibit 616, Answer 7 202, Page 135 to 136, states that these components are 8 not made of cast materials, so they are not 9 susceptible to thermal embrittlement. MRP 227-A and 10 the reactor vessel internals AMP identify irradiation 11 assisted stress corrosion cracking which, as the name 12 implies, is actually the result of multiple underlying 13 mechanisms itself as the aging mechanism of concern 14 for these components, again referring to the core 15 baffles, baffle bolts, and formers. And that these 16 components are all designated for primary inspections 17 under the RVI AMP and Inspection Plan. And I guess 18 I'll start with Entergy just quickly to confirm that 19 when you were referring to these components, you were 20 referring to the core baffles, baffle bolts, and 21 formers that you claim are designated as primary 22 inspections under the RVI AMP and Inspection Plan. Is 23 that correct?
24 DR. LOTT: Yes, I believe that's true.
25 ADMIN. JUDGE WARDWELL: Say --
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5298 1 DR. LOTT: Yes, I believe that's true.
2 This is Randy Lott.
3 ADMIN. JUDGE WARDWELL: Thank you. And, 4 Dr. Lahey, do you now agree that those core baffles 5 and the baffle bolts and the formers are handled by 6 the RVI AMP?
7 DR. LAHEY: It's clear that they are being 8 inspected using the techniques in MRP 227-A, if that's 9 what you mean.
10 ADMIN. JUDGE WARDWELL: Yes.
11 DR. LAHEY: Yes.
12 ADMIN. JUDGE WARDWELL: Okay. Do you have 13 any other comments on that? And they're being 14 inspected as the primary components, is that your 15 understanding also?
16 DR. LAHEY: Yes, that's my understanding.
17 ADMIN. JUDGE WARDWELL: Okay, thank you.
18 That's all I have.
19 ADMIN. JUDGE KENNEDY: This is Judge 20 Kennedy. Just going back to the baffle former bolts 21 and maybe this most recent discussion answered the 22 question, but earlier in today's testimony, you had a 23 colorful expression for the baffle former bolts being 24 subject to shock loads and you referred to unzipping 25 the rest of the bolting.
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5299 1 DR. LAHEY: Yes.
2 ADMIN. JUDGE KENNEDY: Is this still a 3 concern of yours given what we've gone through this 4 afternoon and all the other testimony on the baffle 5 former bolts?
6 DR. LAHEY: Yes, your honor, it is.
7 ADMIN. JUDGE KENNEDY: Within your 8 testimony, do you have any support for the assertion 9 that these bolts, when subject to a shock loading, 10 would fail catastrophically?
11 DR. LAHEY: If they are significantly 12 weakened, we're talking as we get out in time and 13 they've been irradiated significantly, they've been 14 subjected to fatigue, they've been subjected to 15 irradiation assisted stress corrosion cracking, and if 16 they're significantly weakened in that way and they're 17 subjected to a strong shock load, yes, I have a 18 serious concern about it.
19 ADMIN. JUDGE KENNEDY: Is this combination 20 of aging mechanisms all applicable to these bolts?
21 DR. LAHEY: I'm sorry, could you --
22 ADMIN. JUDGE KENNEDY: Is this combination 23 of aging mechanisms all applicable to these --
24 DR. LAHEY: I believe they are, yes.
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5300 1 that this combination of aging mechanisms would be 2 applicable to the baffle former bolts?
3 DR. LOTT: Yes. I mean, to the baffle 4 former bolts -- can you name the mechanisms again for 5 me -- I'm sorry, I'm -- they're subject to irradiation 6 embrittlement, they're subject to irradiation assisted 7 stress corrosion cracking, they're subject to 8 irradiation induced stress relaxation, they're 9 potentially subject to void swelling. So, yes, 10 they're -- and we model all of those components.
11 ADMIN. JUDGE KENNEDY: I guess the other 12 one I heard was metal fatigue. I don't know --
13 DR. LOTT: Metal fatigue is, yes, certainly 14 a possibility.
15 ADMIN. JUDGE KENNEDY: So is this now back 16 to the question of how much synergism there is between 17 all these aging mechanisms, if there is any?
18 DR. LOTT: Well, I mean, I will point out 19 that our experience with baffle -- we have experience 20 with failure rates in baffle former bolts, a fair 21 amount. We have a fair amount of data now on baffle 22 former bolts. So, we have -- I don't see a lot of 23 surprises coming forward in this testimony and because 24 we're monitoring, as we said, the effects and if we're 25 looking for the effects on cracking, whatever combined NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5301 1 effects are there, are there in that data base or in 2 that operating experience.
3 ADMIN. JUDGE KENNEDY: Does this data 4 include bolts subjected to higher fluences, towards 5 the end of design life? Maybe not extended life, but 6 design life at least?
7 DR. LOTT: Yes. I mean, some of these 8 examinations have been performed as part of similar 9 plant life license renewal applications, PEOs. So, 10 yes, they've been towards the end of plant life.
11 ADMIN. JUDGE KENNEDY: So it does include 12 some bolting that has been examined during the period 13 of extended operation for other plants?
14 DR. LOTT: At or near.
15 ADMIN. JUDGE KENNEDY: Okay. At or near.
16 I don't want to put words in your mouth.
17 MR. STROSNIDER: This is Jack Strosnider 18 for Entergy. I'd like to come back to the first part 19 of your question on loads for just a minute if I 20 could.
21 ADMIN. JUDGE KENNEDY: Okay, sure.
22 MR. STROSNIDER: We talked earlier about 23 the WCAP report that NRC reviewed and approved with 24 regard to the methodology for establishing the bolting 25 patterns, et cetera. That WCAP report evaluated the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5302 1 design basis dynamic loads associated with design 2 basis accidents, such as the loss of coolant, et 3 cetera. So to the extent that Dr. Lahey's concern 4 about dynamic loads, that those loads are within the 5 original design basis, they have been addressed. If 6 there's something over and above what was in the 7 original design basis, I'm not -- it's not clear, to 8 me at least, what they are and also I would suggest 9 that they're outside the space of license renewal.
10 But it's just not clear what loads he's talking about 11 that would be over and above what's in the licensing 12 basis. And if they're within the licensing basis, 13 they were evaluated.
14 ADMIN. JUDGE KENNEDY: And would those 15 bolts that were subjected to the design basis loads 16 include the effects of these various aging mechanisms?
17 Consideration of the effects of these various aging 18 mechanisms? Which seems to be the other part of Dr.
19 Lahey's concern.
20 MR. GRIESBACH: Your honor, this is Tim 21 Griesbach for Entergy. If you're asking whether those 22 multiple aging mechanisms have been considered as part 23 of the Aging Management Program for the baffle former 24 bolts, as Dr. Lott said, the answer's definitely yes.
25 They've also been identified as primary components for NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5303 1 inspections and, to the best of my knowledge, there's 2 been at least 19 PWR units that have performed those 3 inspections already for MRP 227 and very few failed 4 bolts have been found throughout those inspections.
5 ADMIN. JUDGE KENNEDY: Is the foundation of 6 the strength of this bolting subject to these aging 7 mechanisms the lack of cracking? I mean, if it hasn't 8 cracked, it still maintains its design fatigue life or 9 design life?
10 MR. GRIESBACH: This is Tim Griesbach again 11 for Entergy. The answer is yes. Those materials 12 undergo strengthening, in fact, due to the irradiation 13 effects. So they are stronger than they would have 14 been initially and still maintain the margins that 15 were there from the beginning for that reason, if they 16 are uncracked.
17 ADMIN. JUDGE KENNEDY: And we're back to 18 the beginning. All right. Thank you very much.
19 DR. LAHEY: Your honor, I think I've been 20 asked the question many times, are my loads different 21 than your loads? They're the same loads. The only 22 difference is how impulsive these loads are. So it 23 has to do with how they're calculated. I haven't seen 24 the document which gives the details on that. If they 25 have a shock capturing routine, like adaptive grid, or NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5304 1 they're done by method of characteristics or some way 2 where they don't smear out things, instead of hitting 3 it with a hammer, you're hitting it with a powder 4 puff, then I'm happy that everything's surviving. But 5 I haven't seen that. And I've done a lot of work in 6 the past on RELAP and TRAC and RETRAN, so if you're 7 using those kinds of codes, it's more like a powder 8 puff than it is a hammer. So that's the difference.
9 ADMIN. JUDGE KENNEDY: Would Entergy like 10 to respond to that? I mean, again, this is a question 11 of how you analyze the dynamic loads.
12 DR. LOTT: Well, I guess --
13 CHAIRMAN MCDADE: Dr. Lott?
14 DR. LOTT: I'm sorry. This is Randy Lott.
15 That information, I think, is dealt with in the 16 methodology documents that we talked about earlier, 17 and I'm sorry I don't have the number right in front 18 of me, but I think we discussed earlier the approved 19 methodology for analysis of the bolts, the computer 20 programs and the process are in that document. We 21 believe it's more than adequate and consistent with 22 the licensing basis.
23 MR. STROSNIDER: This is Jack Strosnider 24 for Entergy. I'd like to add to that, that the codes 25 that are used for doing these analyses are codes that NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5305 1 have been reviewed and approved by NRC as acceptable 2 for analyzing the design basis loads. So without 3 going into all the details of how those calculations 4 are done, which would require a code expert, you 5 should at least recognize that the NRC has done that 6 type of review and concluded that these codes are 7 acceptable for the application.
8 ADMIN. JUDGE KENNEDY: Thank you. Dr.
9 Lott, I believe earlier Entergy testified that the 10 Topical Report is an exhibit in this proceeding. Is 11 that true? The bolting methodology?
12 DR. LOTT: The approved -- yes.
13 ADMIN. JUDGE KENNEDY: Dr. Lahey, did you 14 get a chance to review this exhibit as part of your --
15 DR. LAHEY: You're talking about the WCAP 16 report that they --
17 ADMIN. JUDGE KENNEDY: Yes, sir.
18 DR. LAHEY: -- talked about? I haven't 19 seen it yet, I was planning to look at it this evening 20 if I can get my hands on it. I mean, one thing that 21 --
22 ADMIN. JUDGE KENNEDY: Did I miss one of 23 your homework assignments? Is this --
24 DR. LAHEY: Yes, I know, it's just adding 25 up. One thing you folks should know though is, I was NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5306 1 very involved with the development of RELAP and TRAC.
2 So I'm very familiar with that methodology, what it 3 does and what it doesn't do. And let me say one last 4 time, hopefully the last time, after we assured 5 ourselves that these significant shock loads did not 6 distort the core geometry, we then focused on, do you 7 have the right mass in the right place, is the heat 8 transfer right, could you cool the core adequately?
9 That's what these codes are intended for. It's true, 10 they calculate pressure versus time and temperature 11 versus time, but they're in no way sharp shock load 12 codes. They're not intended for that purpose.
13 ADMIN. JUDGE KENNEDY: All right, thank 14 you. Appreciate it. I have nothing further.
15 CHAIRMAN MCDADE: Okay. Very quickly, this 16 may have been covered, but it's sort of a question in 17 my mind and I just want to clarify it. You've done 18 the UT examination on baffle former bolts for the past 19 20 years. Yet, in the testimony, it talks about the 20 UT examination acceptance criteria for the baffle 21 former bolts will be developed. And the question is, 22 what are you using now?
23 MR. DOLANSKY: This is Bob Dolansky with 24 Entergy. The plants that have performed UT of baffle 25 bolts would have that technical justification. Indian NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5307 1 Point has not performed UT of baffle bolts yet. So 2 what we're getting is a plant specific technical 3 justification for the UT of the baffle bolts, that's 4 why we don't have it. So when we say that 8,000 bolts 5 have been inspected, the technical justifications for 6 those, I can't speak specifically for each plant, but 7 I would assume each plant has one. I know that when 8 we go to do our baffle bolts inspection, we will have 9 a plant specific technical justification. And that's 10 what's being developed for us now.
11 CHAIRMAN MCDADE: Okay. One of the 12 concerns that we have is the differentiation between 13 the development and implementation of a plan. That 14 the development of the plan being something to be done 15 that we can look at, the implementation is going to go 16 on and is going to be monitored by the NRC during the 17 period of extended operation if the license is 18 granted. But is there assurance that these acceptance 19 criteria that are to be developed will be adequate to 20 ensure the continued operation?
21 MR. DOLANSKY: I believe the acceptance 22 criteria is contained now in the AMP. This is Bob 23 Dolansky, I'm sorry. So if we go to our --
24 CHAIRMAN MCDADE: I mean, here's the thing, 25 and correct me if I'm wrong, any time you use the term NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5308 1 will be developed, it sort of is a red flag for us.
2 And as I understand, the term that you're using is 3 that you basically have a procedure in existence now, 4 but the plan is that you will be using perhaps more up 5 to date technology down the road and that the 6 acceptance criteria will recognize that increased 7 data, increased knowledge at the time. But it's not 8 that there are no acceptance criteria now, it's that 9 just simply they're going to be updated based on newer 10 technology, additional information. Am I incorrect in 11 that?
12 MR. DOLANSKY: I want -- at the first part 13 of your question, you said that we have a procedure 14 now to perform baffle bolt inspections. Did I 15 understand you correctly, is that what --
16 CHAIRMAN MCDADE: Let me ask the question.
17 In the AMP as it exists right now, are there 18 acceptance criteria for the UT examination of the 19 baffle former bolts?
20 MR. DOLANSKY: No. Right now, the 21 examination acceptance criteria for UT of the baffle 22 former bolts shall be established as part of the exam 23 technical justification.
24 CHAIRMAN MCDADE: Okay. So to Dr. Hiser, 25 how does the NRC determine now that these acceptance NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5309 1 criteria to be developed will be adequate to ensure 2 the proper functioning of the baffle former bolts over 3 the period of extended operation?
4 DR. HISER: This is Allen Hiser of the 5 Staff. Our understanding is that for individual 6 baffle bolt examinations, any indication of cracking 7 from the UT exam indicates that, that bolt is no 8 longer functional. So it represents a failure. I 9 believe the analysis that Mr. Dolansky is discussing 10 is post-inspection, when they have identified each 11 bolt, whether it is acceptable or unacceptable, and 12 they're looking more at the configuration whether it 13 is acceptable. So the acceptance criteria for the 14 final configuration post-inspection is, I believe, 15 what they are still working on. But for each 16 individual bolt, it's crack/no crack, 17 unacceptable/acceptable, is our understanding.
18 MR. DOLANSKY: That's correct. Any 19 indication of cracking will be considered a degraded 20 bolt and will be put into our corrective action 21 program.
22 DR. HISER: So I think -- this is Allen 23 Hiser again. I think there's just been a confusion 24 between acceptance criteria on a piece inspection 25 basis versus the assembly basis after one has NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5310 1 inspected all of the bolts. And I think that the 2 terminology has gotten a little bit mixed up. So the 3 inspection acceptance criteria is any crack is a 4 failed bolt. The assembly acceptability criteria is 5 different. That is to be determined. From the NRC 6 perspective, that's a corrective action that is within 7 the purview of the applicant. If they require 8 approval of any of that, then they would come to the 9 NRC for that.
10 CHAIRMAN MCDADE: Okay. But the way you've 11 just described it, that corrective action may or may 12 not be adequate to ensure the continued viability and 13 the continued operation, the utility of that 14 particular part. So how does the NRC assure itself 15 now, and New York and Riverkeeper assure themselves 16 now, that something that still has to be developed 17 will be adequate? Do you understand what my concern 18 is?
19 DR. HISER: I think so. The corrective 20 actions, there are three potential paths that they 21 could take. If they were to find degraded bolts, they 22 could replace the bolts. Repair really is not an 23 option in this case, but they could replace the bolts.
24 If they wanted to continue to operate with degraded 25 bolts, they would need to submit an analysis to the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5311 1 NRC that justified their configuration that they want 2 to continue operating with. From that perspective, we 3 still have approval of any operation with the degraded 4 condition. So I think from that perspective, I 5 believe that there are sufficient controls on this.
6 MR. STROSNIDER: This is Jack Strosnider 7 for Entergy. The one other thing I think we should 8 add here is what we talked about earlier. The 9 methodology that's being used to develop the plant 10 specific bolting pattern, that acceptable bolting 11 pattern, that methodology was reviewed in the WCAP 12 report that was submitted to NRC, it's an extensive 13 review with all the Requests for Additional 14 Information, the back and forth, the safety evaluation 15 written. So they're using what the NRC has looked at 16 and concluded that, that's an acceptable methodology.
17 You do it on a plant specific basis to make sure that 18 if anything plant specific comes up that you've 19 accounted for it. But you should be able to move 20 forward with that methodology and come up with an 21 acceptable technical evaluation for the acceptable 22 bolting pattern.
23 MR. COX: This is Alan Cox --
24 MR. STROSNIDER: That's the whole purpose 25 of, as I explained earlier, the Topical Report process NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5312 1 and reviewing these methodologies up front. The NRC 2 does not want to look at one of these -- do that same 3 review over and over and over again for every plant.
4 So they look at the generic methodology, if the plant 5 can apply it within the boundary conditions that, that 6 methodology is good for, then they can go do the plant 7 specific evaluation and you should have confidence in 8 it.
9 MR. COX: This is Alan Cox with Entergy.
10 Let me add one more thing. I believe the Standard 11 Review Plan provides for, when it talks about 12 acceptance criteria, it says the acceptance criteria 13 doesn't necessarily have to be a formal value. It can 14 be a description of the method that will be used to 15 establish that value. And I would say that it's not 16 practical to do all that in advance. You're talking 17 about finding a -- you're going to find a certain 18 pattern of bolts where you have a failed bolt, there's 19 832 bolts, you can't predict or it would be 20 impractical to try to do an analysis for every 21 combination of which bolt might be failed where.
22 So you have a method established, you take 23 the results of the inspection, you look at the bolts 24 that are failed, and you apply that data using the 25 method that's established and that's how you determine NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5313 1 your acceptance criteria. It would be impractical to 2 evaluate all of those combinations ahead of time that 3 you could have with the inspection results.
4 MR. DOLANSKY: This is Bob Dolansky. I 5 just want to add one other thing. The inspection of 6 baffle bolts, the fact that it's been going on for 7 years and the fact that other plants have done it 8 means that all this stuff has been looked at before.
9 So, I mean, it's not like a first of its kind. We're 10 not the first ones doing it, other plants within the 11 industry have performed these inspections, they've had 12 NRC reviews of both during the inspection and after 13 the outage. So not only has the Topical Report been 14 reviewed by the Staff, but the actual implementation 15 has been reviewed by the Staff at other plants.
16 CHAIRMAN MCDADE: And the preface to my 17 question assumed that and recognized that these 18 inspections have been going on for more than 20 years 19 and that there has been a method established. And 20 what I'm trying to just make sure that I want to 21 cement in my mind is what is currently established and 22 what, if anything, by way of methodology still needs 23 to be developed? As Mr. Cox pointed out, I think it's 24 relatively obvious that you can't say what you're 25 going to do as the result of an inspection until you NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5314 1 get the results of that inspection and then you have 2 to assess the meaning of that.
3 But in developing a plan that provides 4 assurances that you're going to have continued safe 5 operation, you look, is the methodology used adequate 6 to develop an appropriate response? And all I'm 7 trying to do is cement for the record of, what is the 8 current status of development? Again, the language 9 you used, will be developed, which sort of, well, more 10 than sort of, which suggests that there is still a 11 sort of, we'll figure this out in the future, we hope.
12 And from what Dr. Hiser has said, it isn't that and I 13 just wanted to get as clear as possible what we 14 currently have and what it is that still needs to be 15 done so we can determine whether or not there exists 16 now reasonable assurance of continued safe operation.
17 Can you elaborate on that?
18 MR. AZEVEDO: Yes, your honor. This is 19 Nelson Azevedo for Entergy. Maybe it's already clear, 20 but just to make sure it is clear. The methodology is 21 established, has been approved by the NRC, that's what 22 we're using. What is going on right now is taking 23 that methodology in developing a model specifically 24 for Indian Point. But the methodology itself, how it 25 is done, is established, it's been approved by the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5315 1 NRC, and it's what Indian Point is using.
2 CHAIRMAN MCDADE: And that's part of the 3 AMP for these baffle former bolts? Where do we look 4 to see where that methodology is and the fact that 5 it's been approved by the NRC?
6 MR. AZEVEDO: It's referenced in MRP 227, 7 that's what Mr. Cox was telling you.
8 MR. COX: This is Alan Cox. Chapter 6 of 9 MRP 227 has a discussion on evaluation of bolts and 10 pins. And basically it says that you don't have to do 11 the individual evaluation, you have to look at the 12 effect on the assembly. And then Section 6.4 gives 13 you the guidance on doing assembly level evaluations.
14 CHAIRMAN MCDADE: Okay. And, Dr. Hiser, if 15 that were to change, is that something that the 16 Applicant would be able to do and inform the NRC? Is 17 it something that would be subject to the 50.59 18 procedure? Or is it something that would require a 19 license amendment?
20 DR. HISER: This is Allen Hiser of the 21 Staff. My guess is it would be subject to 50.59.
22 Given the safety implications and -- my expectation 23 would be that it would not pass 50.59 and would 24 require a submittal to the NRC.
25 CHAIRMAN MCDADE: Okay. And once the plant NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5316 1 specific details are worked out, what is the vehicle 2 for NRC review of that plant specific? Do they submit 3 it to you, is it subject to any kind of a formal 4 review and approval? Or is it simply a situation 5 where you would have the ability to review and 6 comment?
7 DR. HISER: This is Allen Hiser of NRC. My 8 expectation is that, that would require at a minimum 9 a submittal to the NRC. And could involve a license 10 amendment. But I'm not certain of that.
11 CHAIRMAN MCDADE: Okay. Two separate 12 things. One, we were talking about the methodology 13 set out in Chapter 6 of MRP 227. And we were talking 14 about what would be necessary if that methodology were 15 changed. My question right now is not the methodology 16 having been changed, but rather, as Entergy explained, 17 they are taking that methodology and are currently 18 working to make a plant specific program. And my 19 question is, once they have completed the details of 20 that plant specific, how is that subject to review by 21 the NRC?
22 DR. HISER: This is Allen Hiser. My prior 23 statement really still holds. Whether it's an 24 approved generic methodology or a plant specific 25 methodology would require the same approval, if NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5317 1 approval is necessary. The only difference in that 2 case is for the plant specific methodology, we would 3 have to review the entire methodology instead of 4 relying on the generic approval from the generic 5 methodology that previously was approved.
6 CHAIRMAN MCDADE: And if you viewed it to 7 be unacceptable, would this be going at that point 8 through the 50.59 procedure or is this something less 9 than the 50.59 procedure?
10 DR. HISER: I'm sorry, if we found what to 11 be unacceptable?
12 CHAIRMAN MCDADE: Their plant specifics.
13 If you identified problems with the plant specifics, 14 how does that work through?
15 DR. HISER: If it is a license amendment 16 request, then they would need to modify their approach 17 to become acceptable. In the absence of approval of 18 the license amendment, they could replace the bolts 19 that they found to be degraded.
20 CHAIRMAN MCDADE: Okay. But I -- at least 21 I had not anticipated at the level of a -- if it's a 22 license amendment, then it's a situation where New 23 York gets a notice of an opportunity for a hearing and 24 has the opportunity to challenge the adequacy. What 25 I'm trying to get at is once they take the methodology NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5318 1 that is in existence, that the NRC has reviewed, has 2 approved, is satisfied with, and apply it specifically 3 to the plant, what does the NRC do then and what is 4 the nature of review and the public aspect of that, if 5 any?
6 DR. HISER: This is Allen Hiser again. If 7 we can just discuss for a moment here.
8 CHAIRMAN MCDADE: Well, we're basically at 9 the end of the day. Do you want to answer that at the 10 beginning --
11 DR. HISER: Sure.
12 CHAIRMAN MCDADE: -- of tomorrow rather 13 than -- I mean, to me this is something that's 14 important and I don't want an answer off the top of 15 your head that I think this is the way it should be.
16 So, why don't we leave that and take that up first 17 thing in the morning?
18 DR. HISER: Okay. That's acceptable.
19 CHAIRMAN MCDADE: Okay. Do you have 20 anything further for --
21 ADMIN. JUDGE WARDWELL: I do not.
22 CHAIRMAN MCDADE: Okay. I would propose 23 that we would start tomorrow at 8:30. Before we 24 break, NRC is there any administrative matters or 25 other matters that you want to take up before we break NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5319 1 for this evening?
2 MR. HARRIS: No, your honor.
3 CHAIRMAN MCDADE: Entergy?
4 MR. KUYLER: Yes, your honor. A few 5 minutes ago, we were discussing the Westinghouse 6 proprietary reports regarding the baffle former bolts 7 and the minimum bolting pattern analysis. And I 8 believe one of the things that Dr. Lahey said was that 9 he had not reviewed those reports. And we just wanted 10 to note that those have been in the record for several 11 months and they've been disclosed several years ago.
12 And, so, for us to look at that for the first time 13 afresh tomorrow would be challenging, your honor.
14 CHAIRMAN MCDADE: Well, I mean, these are 15 exhibits, are they not?
16 MR. KUYLER: Yes, they are.
17 CHAIRMAN MCDADE: So, I mean, if there's 18 something that's already admitted as an exhibit that 19 will help clarify Dr. Lahey's testimony, we will allow 20 him to refer to that and will of course give Entergy 21 the opportunity, if their experts need some additional 22 time to respond. Quite frankly, so far I've been 23 amazed at the capacity of all of our witnesses to know 24 a rather voluminous record and to be able to testify 25 with regard to the contents of the literally tens of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
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5320 1 thousands of pages of documents that we have. So I 2 think you may have underestimated your experts, but in 3 any event, we would certainly give them time and the 4 opportunity to refresh themselves with the documents, 5 with the reports in order to comment on Dr. Lahey's 6 testimony.
7 MR. KUYLER: Yes, your honor. But I would 8 just note that we have not seen any specific 9 criticisms of those documents from Dr. Lahey in the 10 past. So for him to introduce those on the very last 11 day of the hearing for the first time, we would object 12 to that, your honor.
13 CHAIRMAN MCDADE: We don't know whether 14 he's going to criticize them. He may look at them and 15 say, this is great. This allays my fears, if I had 16 read this before, if I had studied it harder, I 17 wouldn't be here. We don't know, we'll find out 18 tomorrow. Mr. Sipos, anything before we break for 19 tomorrow?
20 MR. SIPOS: Nothing further from New York 21 at this time.
22 CHAIRMAN MCDADE: Ms. Brancato?
23 MS. BRANCATO: No, your honor.
24 CHAIRMAN MCDADE: Okay. Does 8:30 tomorrow 25 pose a problem for anybody?
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5321 1 MS. SUTTON: No, your honor.
2 CHAIRMAN MCDADE: Okay, apparently not.
3 We'll see you all at 8:30 in the morning. Thank you.
4 (Whereupon, the above-entitled matter went 5 off the record at 5:36 p.m.)
6 7
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