ML20069C902
| ML20069C902 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 07/07/1982 |
| From: | Murphy E, Rajan J DEKALB AREA ALLIANCE FOR RESPONSIBLE ENERGY, SINNISSIPPI ALLIANCE FOR THE ENVIRONMENT (SAFE) |
| To: | |
| Shared Package | |
| ML20069C887 | List: |
| References | |
| NUDOCS 8207210324 | |
| Download: ML20069C902 (173) | |
Text
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) g UNITED STATES OF AMERICA, ," , , (j l
NUCLEAR REGULATORY COMMISSION ~~ l i
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _x 4 ,
In the matter of: :
COMMONWEALTH EDISON COMPANY : Docket Nos.:
0
- 50-454 (Byron Nuclear Station, : 50-455 7
Units 1 and 2) :
8 __ __________________
) 9 10 10th Floor Conference Room Nuclear Regulatory Commission II
} Maryland National Bank Building Bethesda, Maryland 12 Wednesday, July 7, 1982 -
The deposition of EMMETT MURPHY and JAI RAJ RAJAN, witnesses called by counsel for the Intervenors, convened at 9 : 0 7 a.m. , pursuant to agreement, taken 3 before Ann Riley, a Notary Public in and for Montgomery Coun ty , Mary land .
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ow 8207210324 G20719 PDR ADOCK 05000454 PDR.
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1 Appe aran ce s:
2 For the Applicant:
., 3 JOSEPH GALLO, ESQUIRE
,() Isham, Lincoln & Beale 4 1120 Connecticut Avenue, Northwest, Suite 840 Washington , D.C. 20036 5
l() 6 For the Staff:
7 STEVEN GOLDBERG, ESQUIRE, U.S. Nuclear Regulatory Commission 8 Office of Executive Legal Director Washington , D.C. 20555
.)
10 For Intervenors DAARE/ SAFE:
11 MICHAEL JENKINS, ESQUIRE
'(3 326 North Avon 12 Rockford, Illinois 61103 13 Also Appearing:
.O 14 ,
Leslie Bowen , Commonwealth Edison 15 John Connor, Westinghouse 16 Steven Chesnut, NRC 3
Richard Bunch, DAARE/ SAFE 18 Richard Udell, DAARE/ SAFE 19 O
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PROCEEDIEGS 2 MR. GOLDBERG: As I indicated in my July 2nd es v 3 letter to Diane Chavez of DAARE/ SAFE, this deposition of 4
Emme tt Murphy and Jai Raj Rajan is being conducted pursuant
, 5 to Ms. Chavez' June 2 9 th telephone request.
i3 6 I understand from Ms. Chavez that she has 7 contacted the local Illinois parties concerning this 8 deposition, including the legal representative for the n
9 Rockford League of Women Voters , who is not in attendance 10 at this deposition.
11 I agreed to make Mr. Murphy and Dr. Rajan 9
12 available for deposition concerning their June affidavit 13 on DAARE/ SAFE Contention 9-C, despite the f act that the C) 34 formal discovery period has closed.
15 There is obviously greater latitude normally 16 accorded the topics for deposition, but I would expect O
17 tha t the deposition will focus on the affidavit for which 18 these gentlemen are prepared to respond.
19 Any evidentiary objections I may make are for g
20 the record. I will not instruct the witnesses to decline 21 to answer on the grounds that I have interposed an
'O' 22 evidentiary objection, and I would not encourage discussion 1
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I on those objections, nor do I anticipate a necessity for 2 inviting any ruling by the presiding officer.
3 I understand also under the rules and practice O
4 that objections are not waived if they are not made, and 5 I will try not to disrupt the orderly flow of the deposition O 6 unne ces s arily.
7 And with that, I have nothing further.
8 MR. JENKINS: Gentle men , the purpose for our C) ~ gathering here today is to just get some information. We 9
10 are not going to be asking any trick questions or anything 11 like that. All our questions are going to be in the area 0
12 of'your expertise.
13 MR. GALLO: You have to talk louder.
() 14 MR. JENKINS: I'm sorry.
15 If there are any questions I ask that you don' t 16 understand, please just tell me and I will rephrase the
() 17 question or ask it in a different manner.
18 Now I've never asked questions of two witnesses 19 at once , and I'm not sure where your areas of expertise O
20 overlap. So if either of you feels more qualified to 21 answer than the other, then, please, that person answer the question. And if the other person has some information
) 22 O
O 6
.O g to add to that, please go ahead and add that information at 2 the end of the previous person's discussion.
3 Whereupon,
..,0 4 EMMETT MURPHY and JAI RAJ RAJAN O 6 7
were called as witnesses by counsel for the Intervenors 8
and, having been first duly sworn, were examined and
- O testified is follows
10 E,f A M 1 N.,& T_ 1 Q g 11 BY MR. JENKINS:
- O 12 0 Gentlemen, a couple of brief questions relating 13 to technical expertise in the licensing of these plants.
14 First f all, what areas of technical expertise LO 15 are required for a thorough understanding of steam 16 generator problems?
O 37 A (Witness Murphy) I'll do this one first, Jai, 18 and then you add to it.
g9 For a thorough understanding of steam lO
,' 20 generator problems, it requires a knowledge of a variety 21 of technical disciplines; mechanical, materials, chemistry, 22 and systems. Understanding of how the reactor plant works.
20 O
k C) 7 3 g Q Which of these areas do you individually have 2 expertise in?
3 A I personally am knowledgeable regarding the O
4 history of steam generators to date, surveillance programs 5 to monitor steam generator performance and inspection techniques. I have considerable background in structural
'() 6 7
mechanics, and I have been involved in the review of 8 steam generators, primarily operating steam generators for O 9 the past three years.
10 Q Dr. Raj an?
11 A (Witness Raj an) My involvement has been primarily O
12 in the area of mechanical engineering aspects of the steam 13 generator internals modifications, and I have been involved in development of plugging criteria for degraded steam
() 14 15 generator tubes. And my involvement has been for a 16 number of years in this area, off and on.
C) g7 Recently, the last two years, has been more 18 than the previous years.
19 Q What is you gentlemen's specific' authority and O
20 expertise with reference to Byron's SER?
21 A (Witness Murphy) In terms of authority -- I'm 22 not sure I understand the question. Speaking for myself,
)
O 8 1
I felt qualified to respond to the contentions set forth 2 and there were others that could have also responded, were 3 I not involved. I'm not quite sure I understand the full
)
4 breadth of the question.
5 MR. GALLO: Can we go off the record?
,(Discussion of f the record. )
O 6 7
WITNESS RAJAN: I don' t know if I can call it g an area of expertise, but I did review the steam generator n' ' design for the Byron-Bradewood plant, as I did for several 9
10 other plants, and our review was primarily within the 11 mechanical engineering aspects of the steam generator side.
O 12 This includes the internals, primarily, and does 13 not get involved with the systems aspects.
g) -g4 BY MR. JENKINS:
15 Q .Could you define for me what is an unresolved 16 safety issue?
A (Witness Murphy) One, an issue for which 17 18 Perhaps the safety significance is not fully understood, 19 and one where perhaps some question exists as regards iO 20 to whe ther additional regulatory action or study, perhaps, 21 should be devoted to this particular issue.
22 Basically I think an unresolved safety question 13 O
I l
?C) 9 l
i 13 g is one of whether or not the Staff is -- we have suf ficient 2 -- suf ficient regulatory control over the situation, over 3 the issue.
O
- 4 0 r e there any -- I'm sorry.
A' 5 A (Witness Rajan) I have nothing to add to that.
9, 6 0 Are there any safety issues which might be v
7 described as unresolved that might be unacceptable? I'm 8 not asking for specific safety issues, but is that contem-O , plated within the definition of a safety issue?
10 A (Witness Murphy) In terms of -- I'll go on and 11 I'll answer your question in a moment.
- O 12 Q Okay.
13 A But let me mention the f act that at the NRC, 14 we have a specific organization assigned to perform j) 15 unresolved safety issue studies, and I'm sure they can 16 give you a much more polished and perhaps a better i
17 philosophical feel for what these -- how these issues
!(3 l
18 in terrelate to existing regulatory approaches to reactor 19 plants, or nuclear plants.
O 20 Would you mind repeating the question? I'll 21 answer your question.
l 22 Q Right. Are there any unresolved safety issues
'O O
l . - _ . _ . . - - - ..- ~- . .
I C) 10 C) I which might be deemed unacceptable over the life of the 2 plant?
3 A With regards to steam generators, in my opinion, O
4 there are areas that certainly deserve further study by 5 the Staf f. These studies are in progress, but in my
,y 6 opinion, pending the outcome of these studies, we have 7 sufficient -- we have sufficient requirements and checks 8 and balances in place to ensure the safe operation of the
() 9 unit.
10 In my opinion, it is important that we not 11 sit on our present position indefinitely. I think it 's O
12 important that we continue to study the matter and identify 13 any of the areas where improvements should and can be made.
cs 14 0 .Okay. I want to dis ~ cuss briefly some of the v
is history of tube degradation problems.
16 How many Westinghouse reactors are currently O . i7 operational, approximately?
18 A On the order of 40.
19 Q How many of that number have had tube problems O
20 of some sort?
21 A I think I have seen -- somebody has actually 22 gone down a list and counted it, a number of approximately O
l
!(3 11 1 l
(3 g 40, so therefore my first answer to -- maybe change my 2 first answer a little bit. We must have a bit more than 3 40 operating Westinghouse -- no, wait a minute.
'O 4 MR. UDELL: There are 49 PWRs.
5 WITNESS MURPHY: I think you are correct. I 6 don' t think we have that many Westinghouse plants. I
$3 7 did bring a document, if it were necessary I could 8 physically run down a list of plants and count the number
- (3 9 involved.
10 , Let me j us t s ay that , one, the number of PWR --
11 operating PWR plants is well known, and the number of
- O
~
12 plants that have reported anywhere from minor to severe 13 degrees of tube degradation has also been reported in NRC 14 reports. I think we are talking on the order of in excess q) 15 of 40 PWR systems, total.
16 A major share of those, off the top of my head,
- (3 37 60 percent, might involve Westinghouse plants , and the vast 18 majority of PWR plants have reported anywhere from very 19 minor degrees of tube degradation ranging to severe degrees O
20 of tube degradation.
21 BY MR. JENKINS:
22 Q What proportion is that? Would you repeat that?
13 12 13 g Approximately?
2 A (Witness Murphy) Does 60 percent sound reasonable 3 to you, Jai?
O 4 Let me refer you to NUREG 0886. I don ' t think 5 it serves any purpose for me to speculate.
13 6 0 Please do.
7 A NUREG 0886 provides a listing of all operating 8 PWRs, and a summary of what kinds of tube degradation have O 9 been experienced in each of these f acilities, ranging from 10 very minor degrees of tube degradation to the most severe 11 forms of tube degradation. ,
- O 12 Q Would you refer now to that document and tell me 13 the number of Westinghouse plants that have had tube 14 degradation problems?
y) 15 A I count 29.
16 0 Okay. And how many Westinghouse reactors are (3 37 there?
18 A Operating reactors?
19 Q Yes.
O 20 A Nine, eleven -- I count 31. That doesn ' t say 21 I may not have missed one or two.
22 Q Are you aware of any problems since the
~
- O w,. , . . , - - - . . - - - _ - - . . - , - - - , . - _ - - - . . -
f O
13
) I publication of that document?
2 A Yes.
3 Q And how many?
6 4 A I'd have to give some thought to that answer.
5 Westinghouse plants -- I'm not aware of any plants
() 6 which have not experienced -- previously experienced 7 l degradation, now having experienced degradation, with 8 the exception of McGuire. McGuire, I thi'k, n since the O 9 publication of this document, has reported some very minor to degrees of -- some degree of tube degradation as a result 11 of the preheater vibration problem.
O 12 O So virtually all of the plants have experienced 13 some sort of problem; is that fair to say?
14 A Yes, ranging from very minor degrees to very
() ,
15 severe degrees.
16 Q One of the measures that your affidavit suggests
() 17 to minimize tube degradation problems is AVT water 18 chemistry. How many of the Westinghouse reactors employ 19 AVT?
O 20 A All but two.
21 Q How many 'of this number have had some type of
) 22 tube degradation problem, to the best of your knowledge?
53
_ - _ _ - - - - - - - - - - - - - - 1
c Y3 14 b g A We ll, to the best of my knowledge, the vast 2 majority have had tube degradation problems.
. 3 Q Your affidavit lists, along with AVT, three
- O 4 other measures to deal with tube integrity problems. Have 5 these measures been employed at the other Westinghouse 6 units as well?
-(3 7 A Let me refer to my affidavit.
8 Regarding item 2, improved controls and 3 9 monitoring secondary water chemistry, I'm really not the 10 one to address what steps, what specific steps the gg industry in general has taken in regards to operating iO 12 plants to improve secondary water chemistry control.
13 Plants that have -- well, in-service inspection y requirements in accordance with various criteria, plants
)
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15 which have been licensed in the past several years, have 16 been subje:t to the requirements of the standard technical f3 g7 specifications, in addition to the requirements of --
18 the criteria of Regulatory Guide 1.83 and the ASME Code.
19 The standard tech specs represent a general O
20 upgrading of steam generator tube surveillance requirements 21 with respect to Regulatory Guide 1.83 and the Code.
22 Q And the fourth item there, limiting allowable
- O
iG 15 I primary to secondary leakage rate?
2 A Leakage rate limits are specified in technical 3 specifications. The newer plants , subject to the standard 4 technical spe'cifications, have the most stringent limits 3 on -- tech spec limits on primary to secondary leakage.
(O 6 Let me just add there that where a plant runs 7
into severe difficulties with corrosion and so forth, it g is not at all uncommon for the Staff to require much more n'
stringent limits on allowable primary to secondary leakage, 9
10 as a special license condition, which are lower than the 11 actual tech spec limits.
O 12 Q Are there any plants that you are aware of 13 that have employed all four of these measures?
A Yes. In saying yes, it is based upon my O 14 15 understanding that many of the newer plants have implemented 16 improved controls and monitoring of secondary water C) chemistry. So my response to that would be that some of 37 18 our more recently licensed plants would have implemented gg all four of these items listed.
O ..
20 0 What is the record of tube problems in these 21 plants ?
A Mixed. One thing to keep in mind is that the
() 22 O -
) 16 C) g newer plants have had the least amount of operating 2 expe rience uud short-term trends may or may not be 3 indicative of long-term trends. Just looking over a list O
4 of plants which we have licensed in the past -- oh, since 5 1980, none of these plants have experienced any extensive o
degradati n, tube degradation problem.
O 6 7 Q But they have still experienced some degradation 8 Problems?
9 A Some have experienced some degradation problems.
10 Sequoyah 1, for example , has experienced minor amounts of 11 denting. Salem Unit 2 has not reported -- Sequoyah 1 was O
12 issued a license in September 1980. Salem Unit 2, it was 13 issued a license in April 1980, and has not reported any 14 d*9"*d^*i " t d****
O 15 That is not to say that they don't have any 16 degradation. It just isn't significant enough to be C 17 reportable to us. Neither of these two plants have plugged 18 any tubes as of last February.
19 0 Are these plants at 100 percent operation?
O 20 A There is nothing in the steam generators that 21 would preclude their operating at 100 percent. I c an ' t 22 speak to other constraints.
O 9
O
C) 17 C) g Q You earlier mentioned the imposition of license 2 conditions where problems have emerged. Instead of 3 imposing license conditions, why didn ' t you simply revise O .
4 the technical specifications?
5 A That is sometimes done, too, where we feel that a common vehicle for changing the requirements is
() 6 7 to revise the tech specs.
8 Another vehicle that has been used in the past O 9 has been to place special license conditions on the plant to which govern it, which, you know, are the governing 11 requiremen ts , as long as the conditions are in effect.
O 12 0 In your affidavit you state, and I quote:
13 "Several steam generator design features 14 are empi yed at Byr n t limit the regions O
15 where deposits could tend to accumulate 16 and possibly cause corrosion. "
() 37 What are these design features?
18 A One that immediately comes to mind is the g9 elimination of the tube to tube sheet crevices. These O
20 crevices -- they've added a design feature to enhance 21 the coolant flow, secondary coolant flow across the tube 22 sheet to minimize the buildup of sludge deposits on the O
O' 18 9
I tube sheet. A traditional location of corrosion problems 2 in steam generators has been within the area of sludge 3 accumulation.
O 4 Those are two that come to mind immediately.
5 0 Are these features employed in any other plants?
() 6 A Tube sheet crevice -- tube sheet crevices have 7 been eliminated at a handful of domestic plants, operating 8 domestic plants. None of these plants have, to my knowledge, O experienced any tube degradation, because the crevice has 9
go been eliminated. They are not subject to the crevice 11 corrosion that earlier plants had been. There have been O
12 no reported difficulties for domestic units as a result 13 of eliminating the tube sheet crevice.
O 14 0 That was my next question.
15 A With regards to the second part of the question, 16 I'm not sure that we have any significant or any significant O 17 amount of operating experience yet.
Ju2 18 0 In February 1982, the NRC issued a report 19 entitled " Steam Generator Status Report." This report iO 20 chronicled some of the problems of tube integrity, and l
21 in section 3 it states, and I quote:
22 "Short-term solutions to one problem may
,3
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- O l
L
'O 19 1 create other problems. Conversion from 2 phosphate to AVT water chemistry, which 3 minimized wastage and stress corrosion O
4 cracking but was followed by denting, 5 is a case in point.
'() 6 " Finally" -- I'm still quoting - "it 7 should be noted that the majority of plants 8 under review for operating licenses have O SGs of similar design to those currently 9
to in operation, so that the potential for.
11 SG tube degradation exists in these plants O
12 as well."
13 Is this steam generator at Byron similar.to those currently in operation?
() 14 15 A It's basically similar. I believe one of the 16 Byron units. employs a D-5 steam generator.
() 17 Q We 'll go into that. I just want to go into 18 the history first.
19 A They are basically similar.
O 20 Q Based on the history of tube degradation 21 problems, what is the probability of a tube degradation g 22 problem emerging at Byron within the operational life of O
(3 20
'O 1 the plant?
2 A I would say based upon previous experience, 3 there is a high probability that they will encounter i C) 4 somewhere between minor and significant amounts of 5 corrosion.
'(3 6 It's very difficult to project whether these 7 will be minor problems or severe problems.
8 Q What is the probability of tubes needing to be
- g
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9 repaired? Can you speculate on that?
10 A There is a possibility -- oh, repaired? You 11 mean plugging?
'O 12 Q Right.
13 A I would say there is a likelihood that Byrcn
!(3 14 might expect to have to. plug tubes during the life of its 15 plan t .
16 Q Is this a high likelihood?
(
lC) 17 A Based upon previous experience, it would seem
! 18 yes.
, 19 Q Af ter how many years is this likely to occur?
!O 20 A Five years. Just a ball-parkish number.
21 Q Is it possib.e that a tube problem could emerge
- n 22 in the first year of operation?
nJ f
- O l
O 21 O , A Possible.
Q How long is the vendor's warranty period?
A I don't know.
O A (Witness Rajan) To this I would like to add 4
that plugging of a few tubes should not be put out of perspective. It really does not affect the operation of the steam generator, and also it does not, in my judgment, constitute a safety problem. So if a few tubes need to be O ,
plugged during the early life of a plant, it should not be blown out of perspective.
O Mr. Rajan, does plugging or sleeving the tubes O
require any worker expo'sure as a result of the work 12 exposure?
A I do not have the numbers with me, but obviously during the early stages the exposure is less. As it goes to -- af ter the plant has been in operation for a longer
-O g period of time, and also the techniques have now been continuously refined to the point -- the plugging techniques, that is, that exposure is limited to a minimum, and O
plugging of a few tubes is within acceptable limits.
20 MR. JENKINS: Just a moment, please.
(Pause.)
2 Q I want to go on and talk about a comparison 3 between the different models of steam generators. I have O
4 reviewed the schematics of the D-2, D-3, and the D-4, D-5 5 .model, and other than the size of the preheater, I couldn't 6
note any differences. Could you describe how the models
-(3 7 are different?
8 A (Witness Rajan) Well, primarily the D-4, D-5
'O 9 preheater section has a flow pattern in it, in which the 10 flow, as it enters from the main feedline, hits an 11 impingement plate and is directed downwards, and then it
- O 12 curves upwards again. This is referred to as a counterflow 13 type.
n The reason for the preheater is to -- the
() 14 15 object of a preheater is to increase the flow within that 16 region so as to increase the efficiency of the steam
,0 generator, and this is one way of doing it.
37 l
l 18 In the D-2 and D-3 type, primarily the flow 19 hits an impingement plate and then is split almost half
- O 20 and half upwards and downwards, but the basic idea in 21 both the preheater designs is to create a region of flow in which there is mixing and high rate of heat
) 22
'O l
23 1 transfer.
l 2 Q So, th en , other than the design of the preheater, 3 there is no significant difference between the designs?
O 4 A I would say that it is a significant dif ference 5 in one sense: In the D-4, D-5 case, the impingement plate
[3 6 directs the flow downwards, and in that sense its effects 7 on the steam generator tubes would be different from those 8 that the tubes would' experience in the D-2 and D-3 model, O 9 because in that case there is a high cross-flow comnonent.
10 The flow, af ter hitting the impingement plate , goes upwards 11 and towards the tubes.
O 12 A (Witness Murphy) It's my recollection that D-5 13 steam generators employ stainless steel support plates as (3 14 opposed to carbon steel support plates of earlier designs.
15 With stainless steel support plates, we would expect the 16 D-5 steam generators to be less susceptible to denting.
O 17 Q Your affidavit discusses flow-induced vibration 18 and subsequent wear of tubes and preheater. You state, 19 mnd I quote:
O 20 "The tube excitation mechanism appears to 21 be a combination of a threshold type of
) 22 fluid elastic instability and turbulent buf feting. "
O I
C) 24 C) 3 Would you please define those two terms?
2 A (Witness Rajan) In fluid elastic instability 3 uYPe of phenomenon, if a tube is excited by a range of O
4 f requen cie s , it responds to a very narrow band of 3 f requencies . And when those frequencies are -- when 6 those frequencies excite the tube, this narrow band 'of vO 7 frequencies, the tube goes into a rather violent mode of g vibration.
() 9 In turbulent. buffeting, on the other hand, go the tube respond's to the exciting forces at all frequencies 11 and as the power level increases , the buffeting forces O
12 increase, and therefore' the tube vibrations increase , and 13 in such a case the vibration of the tube increases with the
, g4 power level.
y 15 Whereas, if the tube is excited by turbulent --
16 by fluid elastic instability, it is possible it may not O g7 experience violent oscillations and vibrations at other 18 power levels, except at which it goes into resonance.
?. 2 19 0 What happens if there is vibration and subsequent O
20 wear of tubes in the preheater?
l 21 A Some vibration is obviously acceptable. It is 22 only when the vibrations of the tubes -- as a matter of l
l i
e
25 C) g fact, all tubes in the steam generator vibrate to cross-1 2 flow velocities and axial velocities. So that is some 3 level of vibration which is acceptable and will not create
- O 4 any damage over the life of the plant.
3 However, if the tube, for some reason, goes into f(3 6 a violent state of vibration, only then the wear rates --
7 the wear rates as a result of the tube hitting the support 8 plates-- the wear rates increase to such levels that
!')
I 9 degradation is rapid and can lea'd to a tube f ailure in a 10 shorter period of time than it is designed for.
11 Q Has this ever happened in an emergent situation?
O 12 A (Witness Murphy) What?
- 13. Q An emergent situation.
() 14 A Emergen t?
15 Q Creating an emergency.
16 A We have not, to my knowledge, had a tube failure
. C) rupture, or a rupture event, ra ther , as a result of tube 37 18 vibration. There have been some large leaks as a result
! 39 of tube vibration, but not one which we formally classified 1
0 20 as a tube rupture event.
21 Q Are you aware of any that may have occurred in 22 foreign facilities?
3 i
l l .
1
'O
.C3 26 i A Not as a result of vibrations, that I'm aware of.
2 A (Witness Rajan) Are you talking of a leak, or 3 are you talking of a severance?
- D 4 A (Witness Murphy) Emergency -- the emergency --
5 we're speaking of rupture events.
Q3 6 A (Witness Raj an) You're talking of severance of 7 a tube. I agree with Emmett.
8 MR. JENKINS: Just a moment.
.C3 9 (Pause . )
go BY MR. JENKINS:
11 Q Mr. Murphy, you can go ahead. I was going to
- g 12 ask you about Ginna. You can go ahead and make your statemerit 13 on the record.
14 A (Witness Murphy) Okay, we'd like to amend our c) 15 previous answer to say at Ginna the tube that ruptured 16 at Ginna did not rupture directly as a result of vibration.
.C3 17 Vibration did play a role.
18 A (Witness Rajan) But the primary mode of failure 19 was another mechanism.
O 20 A (Witness Murphy) Vibration did play a role in 21 transferring -- in the overall failure scenario which was 22 i nitiated by foreign objects and through a complicated
- O
- C
() 27 1
I C) g sequence of events , led to excessive wear and pressure 2 burst of the f ailed tube.
3 A (Witness Rajan) We were responding, or I was O
4 responding primarily in terms of the flow-induced 3 vibration phenomenon and the fluid elastic and .the t'urbulent I) 6 buffeting type of phenomena. But as Emmett pointed out, 7 at Ginna, vibrations had a role in tube failures , although g not the primary cause of failure.
O 9 Q What was the nature of the problem with the 10 Ringhals plant in Sweden, if you're familiar with that?
11 A Primarily flow-induced vibrations acting on O
12 the tubes in the vicinity of the preheater section. And g3 as a result the tubes impacted against the support plate, gy 14 the holes in the support plates, and there was a high 15 rate of wear in those regions. And I believe one tube 16 was degraded to the point, or worn down to the point that O 17 it started to leak.
18 Q Was there any warning of this leakage?
19 A As far as I know, it was not a f ailure that l
O
'O resulted in the severance of the tube. When the tube wall i
21 wore down to the point that there was a throughwall area 22 of leakage, that obviously provided a warning to shut the p
E) 28 D 1 plan t down , because of the high leakage. And in that sense, 2 there was warning. -
3 0 But there was no warning of the problem emerging 4 or arising?
5 A (Witness Murphy) not prepared to answer Z) 6 whether or not the leak may hava grown from very
- 7 insignificant amounts slowly to something that was 8 detectable at plant shutdown, or whether the tube was 9 penetrated in one instant leading to a sudden small leak, i
)
10 if you will, we can' t answer that question without checking.
i 11 It was a relatively small leak, though, when
- O 12 you are comparing it to Ginna or Point Beach or something 13 like that.
i (3 14 Q Has there been a design fix for this problem 15 in the D-2, D-3 model steam generators?
i 16 A (Witness Rajan) In my judgment, there is a R3 17 design fix which is in an advanced stage of test and 18 evaluation, and from what results that I have seen, I am 19 convinced that this problem will be taken care of.
O 20 ,
Q Has this design fix been implemented in any 21 plants in operation?
y) 22 A No. It has -- the first domestic plants during I
I
.O
C) 29
) I which it will be implemented are scheduled in September or 2 October.
3 Q Describe the nature of this design fix.
O 4 A There are a number of modifications in the s preheat section that au being -- they have been finalized,
() 6 and these involve -- the basic change is the replacement 7 of the impingement plates by a maaifold, internal manifold, a which has two double walls and a large number of holes in O 9 both the plates, and the object of this impingement --
10 internal manifold is to produce a uniform flow velocity 11 at the first row of tubes beyond the inlet region, which O
12 is the 4 9 th row.
13 And as a result of this uniform velocity, the c) 14 flow-induced vibrations and the turbulent buf feting 'on -
15 these tubes, it has been shown has dramatically decreased.
t 16 Q How long would it take to retrofit this to O 17 existing operational plants?
i 18 A I do not remember the exact periods involved, 1
19 but the time required for tooling and installation is O
20 within weeks.
21 Q Will this design fix totally eliminate all 22 problems with tube degradation?
(O G
-3
- - , - - - - - , , + , - - - - - , -
X3 . 30
.+
I) 1 A No, this fi'x' is aimed at the flow-induced 2 vibration phenomenon that is experienced in the D-2, D-3 3
models only. It is not expected to, for example, have
- (3 4 any effect on corrosion -- stress corrosion cracking or 3
other denting' or other problems.
6 0 I see. Okay. That's a good clarification.
, '0' 7
Will-these design fixes be applied to the steam 8
generators used at Byron?
l(3 9 A No. As I pointed out,this fix is directed at 10 the model -- is designed only for the D-2 and D-3 steam i
generator design -- steam , generator models and will therefore
'O not be applicable for Byron.
12 13 MR. JENKINS: Excuse me just a moment.
14 (Paus e. )
g
- ]u3 33 BY'MR. JENKINS:
16 Q As the denting and corrosion problems
, (] 17 occur in plants , as they get worse in the plant, is it 18 necessary to reduce the operating capacity of the plant g, to lessen the possibility and the probability of new
'!f'7 ruptures occurring?
20 21 A (Witness Murphy) Would you repeat the last
. 22 part of the question, please.
- O
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l Q '
Yes. Is it necessary then bo decr,e@as,e;tNhe ,
2 operational ~ 1evel of operating of the pilnt in order to l'
3 reduce the possiEility of rupturing? b* -
O , __
~
tg u ,.
4 A Is it necessary to reduce the rated sower 3<
Yi s 4 s of the plant to reduce the potential for accidenhs? O l '
s, O 6 Q Right. f w, i.
t w ,
a\[^ '
I x
N <s .
s Once you 'geti into large-scale, plugging of .
7 A No. J w i ,y ,
%. <y 8 steam generators, ultimately you may reach th'e , point where '
y d, x ,-
l h - m 9 you have physically removed su\ch a 'substantisl part of" '
10 your heat transfer area, ,your available heat transfer; , ,
s s , - s s .
.s - '\ .
$ k..'
11 area, that you do affect your capsbilitg to produce heat '
O , ,
_., , s,' --
m ,
12 and power, but does not affect the. propensity for -- the l
l 13 level of plugging, as Jai Raj an , pointed o'uti earlier, d'oes O 14 not affect the potential for a rupture or' h failure.
Is Q How did We stinghouse . test its D-2', D-3 steam l 16 generators prior to its use in a' nuclear facility?
17 A (Witness Rajan) Basically, the design is i .
18 verified on a computer model which simulates the flow 19 phenomenon and the thermal hydraulics within the tube --
Q l~ within the steam generator, and this is done with the 20 i 21 aid of several highly sophisticated computer codes.
22 In addition to this, some verification is I9 l .
l l
l a
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,' N ) 32 1 t.
.m .
~O "
obtained with scale models.
'2 '"
2 Q How did those test results correlate with
~_
" i \'
3 in-plant experience?
OM 3J '
x 4 A This particular phenomenon, which occurred at s
3 Ringhals and McGuire, was obviously not predicted by the g , 6 computer codes and the scale modeling tests.
7 Q In your expert opinion, why do you think there
,1 Y 8 is a disparity between the predicted results and the actual
- 0_ , results?
10 A Well, for one thing, we have to recognize that .
11 this phenomenon is occurring only in a limited area of O
12 the preheat section, and is not something that the 13 conventional thermal hydraulic codes used in the design f the steam gene rator.
.O 14 gg They would predict -- as I pointed out e'arlier, 16 there is a high degree of turbulence and this was C 37 purposely created to extract more heat in that region.
gg So, with this high degree of turbulence, some 19 tubes were excited into resonant modes which was not --
.O 20 which had not been anticipated.
23 Q Your affidavit states that Wes'tinghouse will g 22 do " extensive analyses and tests, including large scale O
'O 33
) I model tests" on the D-4, D-5 generators.
2 How large a scale?
.. 3 A I believe it's the -- the preheat section is
'O 4 being -- it's a two-thirds scale model, the largest one.
s Although it's entirely possible to obtain meaningful data i
- (3 6 from smaller models.
7 Q Does NRC plan to empirically verify these tests?
s A What do you mean by empirically, empirically '
43 9 verify? -
to Q Well, are you planning to run your own tests, or 11 are you planning to look at the data and put it into your
- O 12 own model?
l
- 13 A As far as I know, there are no such plans.
) 14 Q What is your level of confidence, based on your l
Is expert ~ opinion, that Westinghouse's results can be 16 extrapolated to actual use experience?
l(3 17 A Now are we talking of which?
18 Q D-4, D-5.
19 A D-4, D-5? I have not seen any results of (3-20 model test data on D-4, D-5 yet, although I understand 21 that Westinghouse is in the process of conducting analyses 22 and test data for the D-4, D-5 models in a similar f ashion
)
O
C) 34 (3 g as they have done for the D-2 and D-3.
2 My understanding is that these will be available 3 by the end of the year.
C) .
4 A (Witness Murphy) I think it's important to 5 note that once it was recognized they had a problem in 6 the preheater, that they were able to determine -- they
-3 7 were able, through analysis and tests, to demonstrate how 8 the wear mechanism worked. With the ability of hindsight C) , they were able to demonstrate, yes, you know, you certainly 10 would expect a wear process to take place as a result of I 11 vibration.
'O 12 So they have been able, by analysis and test, 13 to say, yes, under these conditions you will get wear and 14 vibration, and it explains the wear patterns that we are O
15 actually observing in the field, and being able to do this 16 gives you confidence that they have a model where , if they
() g7 adjust certain parameters, they have the tools necessary gg to evaluate the ef fect of these parameters on the overall 19 performance of the preheater section.
O (Witness Rajan) To this I would add that one
'O A 21 nondomestic plant has been instrumented which has a model 22 D-4 design, and data obtained from that is providing useful
()
43 35 C) I information as to which modifications would be effective,
, 2 and which may not be of that great utility.
3 0 Is this domestic or foreign?
lO 4 A One foreign plant.
3 Q Let's jump ahead to that. Are you aware --
() 6 you're referring to the KRSKO plant in Yugoslavia?
7 A Yes.
3 Q Are you aware of the current operating status
!O 9 of that plant? -
to A I am aware that it has had about 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> of 11 operation, approximately, and there was negligible -- there
- O 12 was no detectable degradation found as a result of measure-13 ments in the tubes that normally would be affected.
g) 14 The instrumented data from that plant indicated 15 that there is a higher level of tube vibration than is 16 e XPe cted . So while there seems to be evidence that this (3 17 problem is there , there is no evidence of degradation of 18 the tubes as such -- detectable level of degradation of 19 the tubes as such, so far.
O 20 Q Eddy current te s ting , I understand, is capable 21 of determining whether thinning has occurred only past 20 22 percent; is that correct?
O
36 C) b) 1 A (Witness Murphy) That's not exactly correct.
2 At the support locations where the degradation is occurring, 3 we would be a little hard-pressed to say what the threshold O
4 of detectability is, but I think it is something less than s 20 percent. Certainly 20 percent, I believe, should be
() 6 detectable.
7 Q Is 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> of operation in the KRSKO plant 8 sufficient to establish whether you have reached that I) 9 threshold of being able to determine if thinning has occurred?
10 A (Witness Rajan) Well, definitely it indicates 11 one thing, that the mechanism of degradation is not so
- O ,
12 severe that it would manifest itself in a short while. But 13 it certainly does not preclude degradation if the operation
- g 14 were continued.
Is Q If repairs are an eventuality at Byron, which I l 16 believe is a fair assessment of your earlier statement --
l O 17 A (Witness Murphy) Plugging? Plugging repairs?
18 Q Right, plugging or sleeving repairs. -- would 19 their re trofit be a radioactive task?
O 20 MR. UDELL: I'm sorry, excuse me for a minute.
21 (Discussion of f the record. )
MR. JENKINS: Okay, I retract that question. We
) 22 l
4 l
l
() 37 4
'O 1 got an answer to it previously.
2 BY MR. JENKINS:
3 Q Let me ask the question this way:
4 Is a design fix inevitable at Byron, do you 5 think?
() 6 A (Witness Rajan) Are we talking -- a dasign 7 fix -- okay, now we are talking of the tube vibration 3 problem?
.O , Q Right.
I 10 A I believe there would be some fix. Now what 11 that exact fix would be, it's not -- there are several O
12 options available, and I don' t know exactly what form that 13 fix will take.
() 14 Q Okay. Now if the plant starts up before you are 15 able to establish that design fix, would that then be a 16 radioactive task?
C) 17 MR. GOLDBERG: Excuse me. I don't know if 18 the witnesses understand the question, but what would be 19 a radioactive task? Starting up the plant without O
20 implementing a vibration fix?
21 MR. JENKINS: Doing the lesign fix after the 22 start-up of the plant.
()
C) 38 1 MR. GOLDBERG: Okay.
2 WITNESS MURPHY: It would seem reasonable to 3 assume that installation of a fix after start-up would O
4 involve some amount of occupational exposure.
5 BY MR. JENKINS:
() 6 0 Okay, as a summary question here, in light of 7 the variety of tube integrity problems, in light of a Westinghouse's track record in testing, and in light of C) 9 NRC's own observation that solutions to one problem may p3 create other problems, why are.you recommending that Byron 11 be permitted to operate before ultimate resolution of these O
12 various issues?
13 A (Witness Murphy) I'm not sure that we've --
() 14 A (Witness Rajan) Let me say this: That it is 15 our -- that we anticipate that the fix will be in place 16 before Byron goes into operation.
17 A (Witness Murphy) But we may want to discuss 18 this . I understand there is some question on this point 19 which we may want to discuss later, but the Staff has not O
20 made its conclusions regarding the preheat -- regarding 21 the preheat problem and what an acceptable basis for l
j) 22 start-up of the plant will be.
l l
l
'O
1 l
O 39 C) 3 The Staff has committed in its SER to review it, 2 the generic problem, as it relates to Byron, and it has not 3 concluded -- and in the SER, that a fix is a necessary 0
4 condition for start-up. It has not made any conclusion 3 whatsoever as yet. The 'Staf f must make a conclusion and a finding before start-up. We would issue, to the extent
) 6 7 that -- if it were to turn out that a fix could not be g implemented prior to start-up, then the Staff would have
() , to evaluate the acceptability of a program for interim 10 operation pending a fix.
11 MR. GOLDBERG: Mike, I would interject, your O
12 question was a broad summary question, and if I un'derstood 13 the answer, when the term "fix" is used, are you talking about now a fix for the tube vibration phenomenon?
() 14 15 WITNESS RAJAN: That's precisely what we are 16 talking about.
() 37 MR. GOLDBERG: So I don't know if you got an 18 answer to your question. Their remarks were devoted to 19 a position of tube vibration. I don't want to unduly
- O 20 confuse the process, but I think they confined their 21 answer to vibration, their position on tube vibration.
22 MR. JENKINS: I'm satisfied with the response.
i 10
- O 4o i Ma. GoLDBERG
- okay. Fine. rine.
2 MR. JENKINS: Hold on just a moment, please.
3 (P ause . )
4 BY MR. JENKINS:
3 Q When a steam generator tube cracks, ruptures or lO 6 leaks, what is the potential for radiation to escape to 7 the environment?
g A (Witness Murphy) When a tube leaks or ruptures, O , there is a potential for radiation to escape to the environ-10 ment. I am not the one to -- I cannot provide any sort of 11 expert testimony regarding the amounts of doses, offsite iO 12 doses as a result of leakage or ruptures.
13 Q Let me ask it this' way:
How large a leak would be necessary before O 14 15 radiation could leak to the environment?
16 A I can only give you my nonexpert understanding, l
i lO 37 and that is that you will get some amount of radiological 18 release with leakage. But once the -- once radioactive 19 water gets into the secondary, the pathways are available 10 20 for the radioactivity to get into the environment.
21 Q Have there been any tube degradation problems 22 which resulted in radiation leaks?
O O
. -- , - - ~ n n, _ ,w ,
-- c - - - - - - .
O 41 0 3 A oh, I think -- again I think if you want to 2 discuss radiological releases to the environment, perhaps 3 I think you need a different --
0 4 Q Different set of experts?
5 A -- different set of experts.
A (Witness Rajan) The answer is yes , but we are --
() 6 7 I don ' t think we can quantify the releases.
8 -Q okay.
O 9 A The answer to your question is yes. Yes.
10 A (Witness Murphy) We 'll say that the Staf f does 11 have regulations regarding acceptable offsite releases, and
.O 12 tha t these are enforced.
13 Q Well, now, I have a number of other questions g 34 here relating to radiation leaks and safety design and so g3 forth, and I understand that you are not experts in this 16 area, so if you could just -- I want to run through them
() 37 and see if I can get an answer to the best of your knowledge ,
18 and just advise me if I'm going out of bounds here.
19 Are you familiar with the safety response systems 20 in the event of a ruptured steam generator tube? And if 21 so, could you please describe it?
A Jai, do you want to take a stab at that? I'm g 22 l
i ib L
C) 42 l(D 3 not, in terms of how the plant is brought to a cold shutdown ,
2 in event of a rupture event, I'm not prepared to comment.
3 I cannot comment on that.
'O 4 A (Witness Rajan) I am also not a systems man.
3 Q Is there a potential for any multiple f ailures, 6 f r example, the pilot operated safety valve sticking
()
7 open? Is there a potential for that to occur during a 8 tube rupture event?
l(O 9 A (Witness Murphy) Once you have a tube rupture 10 event, and you go into emergency shutdown situation, I gg would have to assume that there is something -- there is G"
12 no reason why something else couldn't necessarily go wrong 13 during the shutdown. ,
34 Q Do you know if this has ever happened in multiple 15 failures of some sort?
16 A I would assume that it has. I would refer you
' () 17 to an interesting document to review in that regard, 18 would either be the Ginna report put out by the Staff 39 that described the shutdown in excruciating detail. Any n
20 problems that were experienced with valves and the like 21 are all in there.
22 There have been -- we have had a number of these 10
I
() 43 C) ; rupture accidents. The first three are described in very --
2 in great detail in a NUREG report entitled " Evaluation 3 of Steam Generator Rupture Events." I don't know the O
4 NUREG number right off the bat. And the Ginna event was y analyzed separately in a more recent NUREG.
0 Do you know if there is any potential for o 6 7 secondary to primary tube leaks under certain accident g conditions?
() , A Under certain accident conditions , if you have a 10 tube failure, there is a potential for' secondary to primary 11 tube leaks.
O 12 0 What could that result in? What cbuld that 13 leak do, do you know?
A Ultimately if one were to have excessive g 14 15 secondary to primary leakage, you could affect your 16 capability to adequately cool the core. It 's an ultimate
() 37 consequence of excessive leakage.
18 Q Now my next question --
g9 A Well, not ultimate, but it's a consequence.
.O m 0 My next question I think you probably have a 21 little bit more expertise in. Is there a potential for i 22 more than one steam generator tube to fail at or about the 10 i
- 4 l
y
,r. _ _ .m 3 , ._ . , , . , . . - - . , , . . - . . .--,-.., . ,,- . . . - - . . - _ _ _ _ . - - . _ _ .,,....m., y, . - ,
O 44 O i same time?
2 A During normal operation, steady state normal 3 Operation, I cannot imagine that we would get more than a
.O 4 single tube failure. Ruptures obviously have happened, 5 tube ruptures have happened four times in this country.
They have been single failures. I cannot imagine -- to O 6 7 get more than one tube f ailure , you need a triggering event, 8 you need a transient of some sort.
- O Otherwise, under normal steady state conditilons ,
9 10 the rupture will occur, assuming the degradation is out 11 of control and not being adequately surveilled, it
- O 12 will occur in a random fashion.
13 0 aut doesn ' t flow-induced vibration af fect many
- g 14 ' tubes at once?
15 A Yes, but you would not expect that each tube 16 would be degraded to exactly the same degree, such that i
O 37 the f ailures would occur simultaneously.
18 (P ause . )
19 Q Okay. Again another summary question here O
20 relating to safety and so forth.
21 In your expert opinion, do you agree with a 22 statement of Mr. Harold Denton that there is no way to O
O 45 O
I ensure that tube leaks will never happen?
2 MR. GALLO: I'm going to object to that question.
O 3 Uid Y "
""S"*# it?
4 WITNESS MURPHY: No, I didn't answer it.
3- MR. GALLO: There is no foundation that Harold O 6 Denton made such a statement. Do you want to try to do i
7 that?
- 1 g MR. JENKINS: I think I'11 just withdraw the
- O 9 question. It was more of a fun question, anyway.
10 MR. GALLO: All right. Well, then, I have a fun objection.
- g !!
12 (Laughter.)
13 MR. JENKINS: Could I hear that?
- O 34 BY MR. JENKINS
15 Q Now I have a series of questions that are a 16 bit more technical than what we've gone into so far, to
! 17 Of f the record.
l 18 (Discussion off the record.)
39 BY MR. JENKINS:
-O 20 0 Okay, I have some questions about what to me 21 at least are technical questions relating to corrosion.
~
- O 22 What, besides deposition of corrosive products l
13 46 C3 3 in the tube and tube-supporting annulus, contributes to 2 tube denting?
3 A (Witness Murphy) What besides deposition of --
O 4
O Deposition of corrosion products in the annulus.
5 A Neither Jai nor myself is well versed in the 6
dynamics of corrosion. I can speak only in general terms.
- O 7
The dynamics of corrosion is not information I need to 8 have to do my job. But magnetite -- the corrosion O , products you are referring to is magnetite , and it's 10 the corrosion product that results from corrosion of the 11 carbon steel support plates. The corrosion products aren't O
12 carried to these crevices from elsewhere in these plants.
13 The corrosion product is a result of corrosion leaks in the support plate itself.
) 14 15 A (Witness Rajan) The denting phenomenon 16 essentially consists of these carbon steel support plates O 17 that react adversely in a certain environment, and as a 18 result of their interaction they put excessive stresses 19 on the tubes, and they also can cause cracking within the
~O 20 support plate itself. And if the denting progresses, it 21 progresses unchecked. Then the plate itself can be 22 broken into smaller pieces, and that's an advanced stage i
O
() 47 C) of denting.
2 Q Is it necessary for the annulus to be completely 3 filled with corrosive products for denting to occur?
O 4 A (Witness Murphy) Yes.
g Q Has tube denting occurred in any plants using 6
AVT water chemistry?
d) 7 A Yes.
8 Q What design features dealing with tube denting
() 9 have been used at other plants, and to what extent have 10 you evaluated their effectiveness?
11 A For new generation steam generators, the support
- O 12 plate designs will be used employing different materials, 13 stainless steel, different tube hole designs will be used 14 to reduce 'dae potential for denting.
43 I
g3 Unfortunately , to my knowledg e -- well, we don ' t 16 have any Westinghouse steam generators with these new 23 g7 features on line as yet, so we have no operating experience.
18 So that answers the question.
19 Q What, in your opinion, is the combined effect lO 20 of reduced water flow velocity and increased secondary 21 water temperature and pressure relative to promoting 22 corrosion at the annulus?
t O
i
48 I) 1 A I think to answer that question, you'd want 2 to refer to our corrosion specialists,
, 3 0 Okay. What, in your opinion, is the cause of
- O 4 the flow-induced vibration problem?
5 A (Witness Rajan) Flow-induced vibrations are a j) 6 result of high cross-flow velocities. Either the feedwater 7 or steam flow.
Q And this is true of all the D-4, D-5 models?
8
- Cu4 9 A Yes.
i 10 Q Have you been able to verify the Westinghouse
!! findings alluded to in your affidavit that quote:
-O 12 " Vibration response in the preheater g3 section is negligible for main feedwater 14 flow rate s , up to about 70 percent. "
j) 15 A This is primarily based on data obtained at i 16 KRSKO. In that plant, some tubes were instrumented and i
l3 37 their vibratory characteristics were obtained at different l 18 power levels , and the data from that plant seems to bear l 19 this out.
O (Witness Murphy) Let me add something else:
20 A 21 Even if the work Westinghouse has done --
22 and what they have recorded has provided us with a certain
!lO
- O
O 49 C) amount of confidence that they are pursuing the right 2 approach -- but even if they're wrong, if they developed a 3 problem, it will be discovered probably most likely through
'O 4 eddy current inspection, or perhaps small leaks in the g case of Ringhals, and if necessary, additional actions (3 6 can be taken as the need arises.
7 I'd just like to make that point.
3 0 Mr. Rajan, what is your definition of a f3 9 negligible vibration response?
10 A (Witness Raj an) Well, a negligible vibration response would be such that it would not cause -- it
. O-12 would not cause a wear of the tube as a result of its.
13 impacting with a support plate.
n 14 0 Is it possible that testing and powe'r escalation v
33 process at Byron might f ail to detect any vibration
[
l 16 Problems in the steam generators?
C) 37 A If the tubes are not instrumented, then 18 obviously there is no way to detect any flow-induced 19 vibration.
O 20 0 Are you going to require instrumentation?
21 A Well, we are assuming that their fix will be 22 available for Byron, and if we are convinced from the data O
L
'O so
,0 g that is provided to us that the fix is adequate, then we may 2 not require instrumentation.
3 But in answer to your question, we have not
- O 4 ruled the option out that it may be instrumented.
3 A (Witness Murphy) Let me add, one, we do have O 6 another vehicle, that is eddy current testing. What the 7 program will be for eddy current testing, of course, has g not been reviewed by the Staff as it applies to Byron.
O 9 But instrumentation, internal instrumentation, is one method to by which one might detect the onset of vibrations. Eddy 11 current testing is another.
~O 12 Q What is the significance of determining the i 13 optimum combination of main to auxiliary feedwater flow 34 rates, preheated water temperatures, tube support design, O
15 and tube' length between supports that would result in a 16 tube oscillation rate equal to the natural frequency of O g7 the tubes?
18 A (Witness Rajan) I'm not in a position to 19 respond to that. I can see what you are asking , b ut I
- O 20 don't have the answers to it.
! 21 Q Is this part of the SER?
l 22 A I'm sorry?
l o
X3 51 I3 1 Q Is this evaluation part of the SER?
2 A Not within the scope of the review that we s
3 conducted in the mechanical engineering branch.
13 :\
4 0 What is the significance of determining the 5 effects on dented steam generator tubes of a natural j) 6 frequency drop by a factor of four to eight?
- A (Witness Murphy) Would you mind repeating the a question?
l<3 9 Q Sure. What is the significance of determining
, 10 the effects on dented steam generator tubes of a natural
!! frequency drop by a f actor of four to eight?
- O 12 A Wha t is th e --
13 MR. CHESNUT: Are you talking about dropping 34 the frequency or the magnitude of the vibrations, or what?
- O 15 MR. GOLDBERG
- Excuse me , Steve . The witness 16 wants the question clarified.
l(3 37 WITNESS MURPH1': Why is natural frequency 18 dropping in a dented tube?
g9 BY MR. JENKINS:
O 20 0 If you do drop the natural frequency, what is 21 the effect on a dented tube?
22 A (Witness Murphy) How do we drop the natural
-O
,0 52 O i frequency? The natural frequency is a property of the tube 2 and its supports. It's a property of the tube system.
3 0 How is that affected by the denting?
O .
4 A (Witness Rajan) It would tend to make it 3 s ti f fe r , if anything.
O 6 Q Is this a part of the SER? Is this something 7 that is evaluated in the SER?
g A (Witness Murphy) For new plants, not typically.
O , Generally speaking, you know, denting per se does not to adversely affect the -- operating experience does not-11 indicate that denting per se adversely affects the lO 12 dynamic response of the tube. If you have very severe 13 denting, you get support plate cracking. For tubes near O g4 the periphery of the bundle you might effectively lose your 15 lateral support.
16 A (Witness Rajan) Let me add to this. In the
- O 37 denting phenomenon, the support plate tends to crimp the 18 tube and the supports and this results in a much stiffer 19 system than one normally would have, when the tube is 20 free to oscillate within the support plate holes.
21 So, as a result of dentinig, whatever happens to 22 the natural frequency of the tube is not likely to be a
.O
- C) 53 1 matter of concern.
2 A (Witness Murphy) That's generally -- for most 3 of the tube bundle , that is the situation. In row 1 and O
4 row 2 you have a peculiar situation where cracking can 5 lead to islanding, the islanding ef fect, whereby ef fectively 13 6 you are losing that lateral support.
7 A number of plants have run into this situation, 3 and have therefore found it necessary to reanalyze the O 9 dynamic response of the tubes. We are now making the 10 assumption 'of no lateral support at these support plates.
11 0 To what extent have you compared -- I'm sorry.
O 12 (Discussion of f the record. )
I 13 BY MR. JENKINS:
() 14 0 To what extent have you compared AVT , parameters, 15 monitoring and control systems at Byron with those in other 1
16 plan ts?
T3 37 A (Witness Murphy) We personally have not done 18 this. This again would be within the cognizance of our 19 Corrosion specialists who have responsibility for reviewing
-O 20 secondary water chemistry controls.
21 0 Are you aware if those other individuals have 22 recommended any changes that Commonwealth Edison should lO l
O 54
- O 1 incorporate in the Byron plant?
2 A They have made an evaluation which is described 3 in the SER of the Byron secondary water chemistry program.
As .
4 There is also generic activity ongoing, both on the part 5 of the industry and the NRC.
() 6 Q What has been, in your opinion, the significance 7 of condenser leakage as a contributing factor to tube 8 degradation?
O 9 A It's had a significant effect.
10 Q To what extent will this continue to be a factor?
11 A I believe it will be over the long term -- it
'O 12 will become a decreasing f actor, primarily for the reason' 13 that to implement the improved secondary water chemistry (9 14 controls and monitoring, it will be necessary to more 15 closely monitor and control the performance of condensers 16 to achieve the objectives.
) 17 Q Are you aware of any specific changes that 18 have been required of Commonwealth Edison in its condenser j 19 materials and designs in its condensate clean-up system?
10 20 A This particular area did not fall within our t
21 area of cognizance. The condenser materials and so forth l
q) 22 have been evaluated and are discussed in our SER.
l l
10 l
O 55 1 0 Do you know when you will be issuing NUREG 0844?
2 A Is that Task Action Plan A-3, A-4, A-5 ? Then 3 the answer is yes. No, I do not.
O 4 0 Do you have any prediction on the impact of any s requirements that may be required -- you will be issuing
() 6 it by January 1st of 1983; correct? I believe that's in 7 your affidavit?
3 A I don ' t think so.
9 Q I read that somewhere .
to A Le t me che ck . I don' t think I would have said a
11 that.
O 12 0 Well, th en , are you familiar with any proposed 13 requirements under that NUREG 0 844?
A Yes. But, you see, this is -- at this point --
O 14 15 an internal -- we have a draf t report that is being 16 intensely reviewed and critiqued at this very moment, and C) the sponsoring organization -- for you to understand --
37 18 if you wish information regarding the exact status of the 19 program and where it's going, I think you'd have to inter-0 20 view somebody from the sponsoring organization for the 21 TAPS report, the generic issues organization in NRR.
22 0 Well, it's hard for us to go a hearing in which c)
~
C) 56 E)
I one of the methods of evaluating the success of the steam 2 generator is its in-service inspection requirements, without 3 knowing what some of those requirements are.
O 4 Can you describe some of the proposed requirements 5 and what will be the impact?
() 6 A I'm not sure that it's really appropriate for me 7 to do so, because right now the various recommendations g are being proposed internally by the Staff and are being
() 9 debated internally and discussed internally, and I cannot to predict how this is going to come out, necessarily.
11 I can offer some judgment on that viewpoint, O
12 but I don ' t think it 's really appropriate for me in this 13 forum, because it might be prej.udicial to the proceeding.
O 14 So I think the Staf f is doing an intense review right now
! 15 and it is not for me -- I'm not the right person who should l
16 comment upon the status of the program. I think I ' d be (3 37 overstepping my areas.
18 MR. GOLDBERG: Let me try to ask a question, 19 because I think your question and answer were two different O
20 things .
21 I understood you first to want some kind of 22 broad indication of what the Staf f proposals were in the 3
lO
f3 57 O
1 prospective Staff document on Task A-3, 4 and 5. And then 2 you switched and confined your answers to, I gather, n 3 what the proposed in-service inspection program was for v
~
4 Byron.
5 And correct me if I'm wrong, Mr. Murphy, is not O 6 that program discussed in the Staff SER?
7 WITNESS MURPHY: I was not addressing myself 8 to in-service inspection requirements for Byron. I guess I O
9 was addressing what I thought was the question, what is 10 the status of the TAPS issue, and where are we going with 11 it, what will our recommendations be, and I can only respond 12 that it's in progress and being reviewed very intensely by 13 the Staff.
() 14 MR. GOLDBERG: Maybe we can get from the generic 15 to the specific, because I think the witness is a little l
! 16 confus ed . I was confused whether you were talking about O generic recommendations or specific plans for Byron.
37 18 BY MR. JENKINS:
[9 Q Well, let me ask it this way:
- O 20 Can you confirm for the record whether any l
21 generic requirements will have a.1 effect on the plant i(3 22 capacity performance or in any other respect of the proposed l
l0 1
() 58 U) ' requirements in this document on the Byron plant?
1 2 A (Witness Murphy) The Staff is considering -- has 3 under consideration a number of recommendations-- it is
.O 4 p' art of A-3 and it is part of other generic reviews -- a 5 number of recommendations that may have some effect on
- () 6 surveillance requirements, methods for improving the 7 performance of the steam generators, the corrosion g performance. We have a number o'f these things under C3 consideration.
9 10 Q Have you required, or is it possible that these 11 new requirements may require Commonwealth Edison to
- O 12 install radiation monitoring equipment at potential release l g3 points in the event of a steam generator tube rupture?
y A I d n ' t know the answer to that question .
O 15 Q What changes in Commonwealth Edison steam l 16 generator operational procedures or design in secondary 33 37 water chemistry monitoring and control systems will be 18 necessary to comply with this NUREG 0844?
19 A Well, I have no way to answer that question, i
- O 20 because .I don' t know how it's going to end up. The 21 generic recommendations are under -- being reviewed right 22 now. I can' t predict how it's going to end up.
)
io
g I.'
. ~ > N. 59
- O t i
'.s
,e v
'O
! MR. JENKINS: I have justf a couple ,more questions 2 more, but let me go off the record for a minute.
l .N 3 MR. MURPHY: Let me make one comment, just for '
C) 4 the record:
5 Task A-3, A-4 and A-5 is a generic on'gding
.(3 6 activity. It is not the-- there is ' a sepdrate generic 7 activity ongoing right now directly as a result of the' 8 Ginna incident, but it's getting into areas that' were
^' n 9 initially addressed .by A-3, A-4 and A-5. So wd have this 10 generic program ongoing, too. '
11 (Discussion off the record.)
- O 12 BY MR. JENKINS: ,
13 Q First of all, I'm going to refer to a telegram '
jg 14 from Mr. Goran Mandeus of the Swedish Nuclear Power 15 Inspectorate , and I have a copy of this, Mr. Gallo,'if 16 you'd like. It's a telegram addressed to Mr. Joseph I) 17 LaFleur of the Office of International Programs. It's i
18 titled " Urgent Telegram," and it says:
l 1 19 "This message should reach the persons
!O 20 who will be in telephone contact. . .
. conce rning 21 the Almarz plant, and so forth, and I quote hera 22 from the second to the last page:
j) i O
l n.
o 60 j s,
O ". . .it has been recognized that _ the 1
2 curve has been computed for undamaged tube 3 with adequate support in all baffle plates
)
4 and not damaged tubes as the ones in Ringhals 3 3 with substantially increased clearance in
.(3 6 the support plates and thus possibly 7 possessing a larger free oscillating length. "
8 MR. GOLDBERG: Mr. Jenkins, I'm not sure the
() witnesses have located your referen~ce here, but can I ask 9
10 that they be given a few minutes to f amiliarize thems elves 11 with this document?
- O 12 MR. JENKINS: Sure.
13 MR. GALLO: Why don't we take a recess so we n,
, 14 can read this thing?
15 MR. JENKINS: Fine.
16 (Recess.)
l 0. 4 sy MR. JENKINS:
37 l 18 0 You have had a chance to familiarize yourself
{
19 with the telegram from Mr. Mandeus. Could you describe,
- g 20 in your opinion, what is the significance of a drop in 21 the natural frequency by a factor or four to eight?
MR. GOLDBERG: Are you referring now to a
) 22 i
- O
.()
61
.O 1 statement in the document, Mr. Jenkins?
2 MR. JENKINS: Yes.
3 MR. GOLDBERG: Could you just refer to that for me ?
O 4 BY MR. JENKINS:
3 Q I will - ad the last three sentences prior to O 6 paragraph 4:
~
7 "If the support in one or two plates 8 is lost partly or in whole, the natural O
9 frequency may drop substantially. Approxi-10 mately by a factor of four to eight.
33 According to the theoretical model cited 12 by Westinghouse, the threshold fluid veloc'ity 13 for instability drops by the same f actor or
- O 14 down into the region of what can be described 15 as idling power for the plant. "
16 My question is: What is the significance of 17 a drop in the natural frequency of four to eight?
18 A (Witness Rajan) I think that's a very large g, drop. When we are talking of a drop of the natural frequency
- O 20 of four to eight, we have to assume that the support plates 21 are no longer ef fective at at least two locations , so that j,0 22 you have a much larger length of the tube now free to O
C)'
62 1
r3 I oscillate in its natural mode.
2 I cannot visualize losing support at least in 3 support plates in order for that to happen, but if one 4 were to make that assumption, then obviously the tubes s would then have a different natural frequency in response C) 6 to fluid elastic vibrations, but also be substantially 7 different.
8 Q Go ahead.
C) 9 A (Witness Murphy. ) The Staff has considered the 10 f act that with some wear of the tubes, there may be some 11 effect on the rate of wear, that the rate of wear may not O
12 be constant -- that may not remain constant, as you wear 13 away the surf ace of the tube.
() 14 In our monitoring and following of McGuire, 15 we have taken the consideration into account in reviewing l 16 and approving of their interim operating program.
() 17 Q If this scenario were to occur, what would 18 happen if the plant were not dropped to idling power?
g9 A (Witness Rajan) Are we assuming --
20 0 The drop in frequency rate by a factor of l 21 four to eight.
I 22 A Is this drop being assumed for just one or two 3
i O
O 63
'( tubes, or a who#1e bunch of tubes?
l 2 O Why don ' t you assess the response to that 3 question for both circumstances?
.O 4 A (Witness Murphy) Let me -- we have developed 5 between Ringhals and Almarz and McGuire a considerable
() 6 degree now of operating experience. We have a good idea 7 of qualitative -- an idea of the qualitative relationship 8 between operating at higher power levels and what effect 9 it has on the observed wear rates.
10 In the case of McGuire, we have been -- McGuire 11 has been performing steam generator inspections very 10 12 frequently, on the order of every couple' of months , s ome-13 thing of daat -- two or three months ', something of that j) g4 frequency. And based upon what is observed regarding the 15 amount of degradation or the incremental degradation that's 16 taken place since the last inspection, that experience is
,o as 17 f actored into our evaluation of the next short period of f 18 operation, in that we would not -- we do not predict that l
19 during each succeeding interval of operation, that the
- O 20 wear will be excessive or exceed allowable limits during i
21 that period. And even if it did -- which we don' t expect --
22 but even if it did, we would -- the likely consequence is a 9
L l
O l
l H
- (3 64 53 3 small le ak, but we don ' t expect that to be the situation.
Q Let me repeat the question, just so we can get 2
3 that on the record:
,Q If this scenario did occur, what would be the 4
ef fect if the plant were not dropped to idling power 5
f 11 wing that scenario?
- O 6 7
A I guess my response is that we'd eventually be 8
shutting down for steam generator inspections, we'd
- O ,
observe the degradation had proceeded beyond what we had to anticipated, and we'll take appropriate corrective action.
At worst, I would expect that we'd get a leak gg 10 and that would precipitate the corrective action.
12 Q Mr. Raj an?
f 13 l
gg A (Witness Rajan) My response would be that if i
15 the natural frequency of certain tubes were to change by this order of magnitude that has been postulated here, 16
!O 37 there would be excessive -- there would be excessive vibrations for those affected tubes, and the damage , if 18 g,
it were to occur in those tubes would be at the supports
'O which are affected, and as Emmett pointed out, there would l 20 l
be -- these would be detected by eddy current measurements.
f 21 g Q Earlier you gentlemen stated that you thought
,0 i
)
!O !
l
o 65 1 there was not a great probability of multiple tube leaks, 2 but would you say that. leaks --
3 A (Witness Murphy) Multiple tube failures, gross O
4 failures.
3 Q I'm sorry, I misunderstood that.
l(3 6 A We do have occasion to experience multiple tube 7
leaks, simultaneous leaks .
8 Q Is there any increased safety risk that occurs 9 from that?
10 A From multiple tube leaks?
11 Q Yes.
O 12 A No. The leak rate limit, the tech spec leak 13 rate limit for Byron, has been set such that if you have a n
v y throughwall crack or leaking crack which is leaking at 15 less than the technical specification leak rate limit, 16 that the length of the crack is smaller than the length of (3 g7 the crack that it would take to result in a tube rupture under postulated main steam line break conditions.
( 18 19 The f act that the leaks are occurring at less 10 20 t han the leak rate limit will provide assurance that these 21 leaks would not result in any rupture or gross leakage 22 under accident conditions .
3 I
'O ,
y - e , _ , - - - - -
-- w
C) 66
~)
1 Q For the record, would you -- excuse me just a 2 moment.
3 (P ause . )
O 4 BY MR. JENKINS:
$ Q Well, then, would you say that ruptures of
() 6 vibrations -- ruptures resulting from vibrations in the 7 D-4, D-5 model, the probability of those ruptures occurring 8 is greater than at plants without preheaters?
I) I'm sorry, let me withdraw that question. We're 9
10 having problems with it, auS 11 (Paus e . )
0 12 BY MR. JENKINS:
13 0 Two very quick questions here for the record:
- (3 14 Would you compare Westinghouse's record with is that of other vendors in prcblems of tube degradation?
16 A (Witness Murphy) They've all had tube degrada-13 17 tion. Sometimes the problems tend to be unique to a 18 particular vendor, but they have all had tube degradation.
19 A (Witness Rajan) The nature of the problems may
!O 1
20 be different, but degradation is not confined to just 21 Westinghouse steam generators.
22 Q Okay. Relative to one another, are any of the 3
O
) 67 1 vendors better or worse in terms of the seriousness of 2 the tube degradation problems?
. 3 A (Witness Murphy) Well, you know, one can make
'O
- 4 the qualitative observation that four plants -- steam 5 generators in four units have been replaced as a result M3 6 of tube degradation.
7 Q These are Westinghouse plants?-
8 A Westinghouse plants. And additional replacement 9 activities are scheduled for certain Westinghouse f acilities ,
10 A (Witness Rajan) One, and possibly two.
1
- 11 A (Witness Murphy) Pardon?
- O 12 A (Witness Raj an) One, and possibly two.
13 A (Witness Murphy) Yes.
lO 14 0 My very last question:
15 In your expert opinion, would you say that it 16 is possible that a steam generator tube will rupture at 33 37 an operating plant in the future?
18 A Possible.
l 19 (Discussion off the record.)
- CP 20 i
---==
- o 22 i
i f
.g 1 68
'o 1 EEEE1EE11SE 2 BY hR. CALLO:
3 Q Gentleren, do you have a copy of the contention 4
-in front of you? It's the steam generator tube integrity, 3
9-C. I take it you have it? Is that right?
A (witness Rajan) Yes.
O 6 7 Q Are you able to identify that the paper that 8
counsel gave you contains Contention 9-C?
O , A (Witness Murphy) Yes, the paper that we have 10 ust been given contains that particular contention.
11 C Is that the. contention , Mr. Murphy , that your O
12 affidavit addresses?
13 A Yes.
Q H w ab ut y u, Dr. Rajan?
- O 14 15 A ( ess Rajan) Yes.
16 Q The second sentence of the contention states lO 37 that -- refers to the previous sentence and indicates that 18 certain problems indicated in the first sentence , and new i.
l 39 I am quoting:
10 "
20
. . . constitutes a hazard, both during 21 normal operation and under accident conditions."
22 Do you see that sentence in the contention?
i l
l
'O
2
'O 69
- O A (Witness Murphy) Yes, sir.
A (Witness Rajan) Yes.
Q Mr. Murphy, what does the term "under accident
'O conditions" mean to you as used in this contention?
4 A (Witness Murphy) It means whether -- it means, for example, a main steam line break transient or LCCA transient could precipitate or initiate a tube failure.
Q It's not referring or -- strike that.
O ,
Is it referring to a situation where a steam generator break -- I'm sorry, steam generator tube break might cause an accident in and of itself ?
O A I think that the statement is general enough, 12 perhaps, to encompass that also.
Q But it's also including, if I understand your testimony, a situation where during the course of an accident, the accident phenomena, if I can use that phrase,
- g caused a steam generator tube rupture. Is that correct?
A Yes. The tube rupture, I guess, is an accident in and of itself. Secondly, there are other accidents
'O which one must be sure that you don' t run into a situation 20 where another type of accident could precipitate a failure.
, ,1 Q All right, fine.
- O
3 0 70 O Dr. Rajan, do you agree with that staterent --
i that interpretation of this part of the contention?
A (Witness Rajan) Yes, I agree with this, and O I would add to the accident scenario an earthquake event, 4
for example, a seismic earthquake , which can precipitate a steam line break.
'O 6 Q Okay. Now it's not clear to me which of you gentlemen is expert in the area ,of what I will call what g we've just been talking about, I'll call it accident analysis involving steam generator tubes.
Mr. Murphy, are you an expert in that area?
O -A (Witness Murphy) In ter:rs of accident analysis, it depends on what exactly what you mean. I have a solid background regarding what it takes to fail a tube. I know O
pretty well what it takes to fail a tube. Regarding the 15 systems aspects of the shutdown transient, I am not expert O n that matter.
g Q Well, how about the -- do I take it from the last statement you made that as f ar as the effects of O an accident involving --- let's use hypothetically a design basis loss-of-coolant accident, the effects of that accident on a steam generator tube rupture? Is that an
- O
4 O
71 O area of your expertise?
1 2 A It's an area I have knowledge regarding 3 the potential for a rupture event to aggravate certain 4 accident situations . For example, LOCA. I'm aware of 5 studies that have been done to assess these effects.
(3- 6 Q And would that include the effects on systems 7 within the reactor, including the ability to keep the core 8 cool?
-O A I'm, aware of some of the analyses that have 9
to been done and the conclusions which have been reached. I 11 am not an expert en how the analyses were conducted, what
.O 12 the assumptions were .
13 Q All right, I want to ask that question again.
- () 14 Do you consider yourself an expert in this area as we have is defined it here?
16 A I'm not quite sure exactly what we ' re -- I 3 17 know what it takes to treak a tube . I have a general 18 knowledge of what may constitute excessive leakage during 19 accident situations. This is information I must have
- O 20 in order to make a finding as to whether or not we have 21 adequate assurance that we are doing encugh to ensure 22 steam generator tube integrity.
(3 I
10
O 5 72 0 1 Q I guess that's the answer you gave me previously.
Really, what I wanted to know is whether or not you're an expert in terms of the ccnsequences of steam generator O tube failure'in a design basis LOCA and its effect on 4
reactor systems operation?
A No, but I am knowledgeable regarding some of the conclusions that have been derived from such studies.
7 Q Essentially you have acquainted yourself with O the werk of cthers; is that correct?
A That's correct.
10 Q What about you, Dr. Rajan? Are you an expert O in this area?
12 A (Witness Rajan) No, I am not. I an. aware of the forces and the stresses that would cause -- that would 14 O
act on steam generator tubes during a LOCA event, and a seismic event.
16 O I am not, however, an expert on how other n
systems would be affected during a LOCA event, and also how the LOCA would -- a LOCA event would affect the 19 O coolability of the core, for example .
Q Who is within the Staff? Either one of you can answer that.
O
6 73
.) MR. GOLDBERC: Mr. Gallo, let re suggest this:
If you have some questions which you want answered these witnesses haven ' t shown any reluctance to indicate where they are an expert or ill-equipped to provide the answers.
Why don't you ask the question?
G,.,
MR. GALLO: I am asking the questions. IIe is 6
doing a fine job of telling me the answers to the questions I'm asking. They have now established they are not O experts in the area I am inquiring in, and I am asking who is in the Staff.
10 MR. GOLDBERG: But you haven't made any particular inquiry. You have outlined some --
MR. GALLO: Do you have an objection? If so, state it. I want to get on with my cross-examination here.
14 All right?
WITNESS MURPFIY: Okay. S tudies of the e f f ects O of tube ruptures on accidents have been done under the heading of the Task Action Plan. The task manager for that program is Jack Strosnider.
q BY MR. GALLO:
20 Q Can you spell that-for me?
A (Witness Murphy) S-t-r-o-s-n-i-d-e-r.
g
7 C) 74 p* 1 Q He's the task action plan manager?
2 A Yes. ' Task manager, I think , was the formal 3 description. He managed and coordinated the studies done
- O 4 by a variety of organizations and outside consultants 5 having to do with the effects of ruptures on LOCAs and (3 6 main steam line breaks, et cetera, and so he could refer 7 you to the specific people who did the analysis.
8 Q I understand. All right. But ncw to your O 9 knowledge, do you know of any individual within the Staff 10 who is an expert on this accident analysis area that we
!! have been discussing here? I mean to your knowledge? I O
12 recognize that if Mr. Strosnider were here, perhaps he 13 could tell us , as well, but he's not.
14 A Yes, I think I know of guys that are fairly 93 15 knowledgeable in this area.
16 Q Can you name them for me?
'O 17 A Chris Parcheski.
la Q Can you spell that one?
19 A No, sir. It starts with a P.
'O 20 (Laughter.)
21 Q What's his first name?
7, 22 A Chris.
v O
K3 8- 75
() Q Can you pronounce his last name again?
A Parcheski.
2 Q All right, that 's close enough.
=O A I believe he was very involved in the LCCA 4
study, the one that was done to evaluate tube ruptures concurrent with LOCA.
6 i(3 Q Is that the one that was performed out at Idaho? -
A Yes.
8 l() Q Anybody else that comes to mind?
, A A person by the name of Akstulewicz.
1 (Laughter.)
() I think his first name is Frank, but that's 12 13 not for sure. He was involved in the -- was involved in the evaluation of tube ruptures concurrent with main steam line break.
15 l Q Could you try that name en me again, please?
16 c) A Akstulewicz.
MR. CHESNUT: I can give you the spelling of that name if you need to know the name.
-() A-k-s-t-u-1-e-w-i-c-z.
20 i BY MR. GALLO:
l 21 Q Is that gentleman a member of the NRC Staff ?
l 1
l
!O c .- .. . - . - . . . .- . . -.
9 -
I') 76 13 A (Witness Murphy) Yes.
Q Is that it?
A Those are people that I know for sure that were
!O involved. People I think -- a person I think had. considerab:.e 4
involvement is Pasedag, Walt Pasedag.
O That I can spell.
How about you, Dr. Rajan?
A (Witness Raj an) In my judgment, in answer to I) ,
your question, I don't have any other names besides these, but my feeling would be that there would be more than one person who would be involved in the kind of study that
!O
~
you are looking for, and these people probably would - be 12 from the Reactor Systems Branch and the Accident Analysis Br anch .
34 33 Q Let me tell you where I'm going with my questions. If I look at your affidavit, and I'm limiting
- (3 myself to just what's in your own affidavit, let me ask a preliminary question:
Am I correct in concluding that beginning
!O with page -- I'm sorry, beginning with paragraph 5, 20 i
through the end, which I believe is paragraph 12, that 21
- that represents a joint statement by both 2
- r. Murphy and 11 AJ 10
. . . - _ . . - _ . - . - . - . . - - . - - . ~ - --
10 O 77
- O , Dr. Rajan?
2 A (Witness Murphy) Dcn't we address that up. front?
8 through 12.
3
.O
~
Q Say again?
4 A 8 through 12.
Q Paragraphs 8 through 12 are your testimony; is that correct?
7 A It was sort of a joint prepared testimony.
8
!O ,
Q Oh, 8 through 12. And what about paragraphs 10 "9 .
g A That more or less represents mine. That ' doe s 10
.g represent my testimony.
Q I see. All right.
l 13 t
g Now in reviewing paragraphs -- well, strike that.
g Paragraphs 8 through 12 appear to be talking
'O 17 about a particular problem which has been referred to here as the flow-induced vibration problems; is that I
( ,
correct, Dr. Rajan?
- O A (Witness Rajan) That's correct.
l 20 .
l l Q Paragraphs 5 through 7 are addressing, I guess, i
l the generic question of steam generator tube integrity; is 22
!O
- O .
t 1
f __ _ _ _ -. . _ . . _ . _ _ _, _ -_ _ , _ -.,_ -
il 11 78 v that correct, Mr. Murphy?
1 2 A (Witness Murphy) Yes.
3 O And this is primarily your testimony; correct?
O 4 A Yes.
5 0 Where in these three paragraphs or any place
() 6 else in the affidavit, for that matter, do you discuss 7 the accident aspects of Contention 9-C?
8 A Item 5 is intended to address our assessment ID 9 of the requirements that have been imposed to prevent 10 tube failures.
11 G So paragraph 5 deals -- is intended to deal
'O 12 with routine operation and tube failure under accident 13 conditions, as well; is that correct?
7, 34 A The approach, the regulatory approach to date s
, 15 to preventing tube failures through normal operation or 16 accidents, is to surveil the tubes, inspect them regularly,
'(D 37 periodically, to remove those from service that are 18 excessively degraded within our acceptance criteria, and 19 to reinforce these requirerents with stringent leak rate
- O . .
20 limits during normal operation.
21 In addition, we -- the Staff has been requiring 22 plants during the licensing process to implement improved O
!O
O 12 79
.O controls in secondary water chemistry. This is, in a 3
nutshell, the regulatory approach to preventing f ailures during normal operation, and ruptures. When --
O 4
Q I'm sorry, go ahead. Go ahead A These are requirements that the plant starts up with. It's not at all unusual that as problems do-occur in service, for the Staff to impose additional requirements, with the express purpose of preventing or minimizing the O ,
p tential for tube ruptures during normal operation and accidents.
10 0 When you used the term " tube rupture," do you O mean an instantaneous failure?
12 A In the context I've been using, yes.
Q And is that also true when you use the term
" tube failure"?
15 A In that context, yes. But one of the things we 1 ked for, and the NRC does monitor the operating O n experience , you know, at operating facilities -- and one of the things we look for is leak experience. Also the O number of tubes found to be degraded during routine.
20 periodic inspections.
In other words, we don't wait for a tube 2'~
O
.O
13
.O 80 3 1 rupture to occur before we get involved and impose additional 2 requirements. So degradation -- if degradation occurs or 3 if leaks occur, of course, this then enhances any concern O
4 one might have about potential for excessive degradation.
5 Q Now I guess the confusion I have by reading 7() 6 this paragraph -- strike that.
7 By reading paragraph 5, how am I supposed to a know that it addresses' both accident conditions and routine 13 9 operation? What is in paragraph 5 that gives me that 10 clue?
11 A Well, my answer to that would be the second O
12 sentence of paragraph 5. I refer here to steam. generator 13 tube integrity problems. It's not explicitly stated here,
() 14 but it's assumed to be understood that we are concerned 15 with tube integrity during normal operation and during 16 accidents.
- (3 17 Q I understand that. Part of my confusion here i
18 is if I look at these four steps, if I can use that -- or 19 I guess that's the wrong phrase -- four factors or
- O 20 approaches that might be taken to deal with steam generator 21 tube integrity problems, I'm confused as to whether they 22 are mitigative measures or preventive measures. Can you
!O I
kb 81
() clarify that for me?
A Items 3 and 4 are intended to -- let me start from the beginning.
3 O
4 Items 1 and 2 are intended to reduce the potential for corrosion or degradation of the tubing.
Q Reduce, but not prevent?
-()i' 6 A Hopefully it is an ultimate objective to prevent.
In a practical sense, right now, certainly one seeks to
-(3 minimize any potential for corrosion.
Items 3 and 4 --
10 Q Wait a minute. Wait a minute. Let's get back 11
- O to 1 and 2. Are they preventive measures or simply 12 mitigative measures? I thought you were going to tell me i
they were mitigative measures , meaning they don ' t prevent corrosion, but neces,sarily try to control it. But you switched on me. Maybe I misunderstood you. Is my question (3 clear? I'll repeat it, if it's not.
A The question is whether these measures are intended to eliminate corrosion?
10 1 Q Items l'and 2, yes.
20 A As opposed to whether they are intended to minimize the corrosion?
!(3 22 i
n 15 82 O
1 Q Yes. What do you expect?
2 A I expect some amount of corrosion during the 3 life of the plant.
4 Q What do you expect in terms of results from s employing all volatile secondary water treatment and
() 6 improved controls and monitoring of secondary water chemistry?
7 A All volatile treatment, secondary water 8 tre a tmen t , should minimize, if not eliminate any concerns,
'O regarding phosphate wastage, corrosion of the steam 9
10 generator tubes.
11 Q Has that been the experience so f ar, that O
12 the NRC has seen at operating plants?
13 A AVT chemistry has been very successful in 0 14 arresting existing wastage problems and preventing.new 15 wastage problems from developing at plants which have not 16 operated --
O 17 Q Mr. Murphy, you and I are going to be here a 18 long time if we don't get a reconciliation. I'm trying 19 to get an answer to the question. You switched from
.O 20 " eliminate" to " arresting" to " mitigate" to " reducing,"
21 and I'm trying to segregate thos e te rms .
O 22 Now I thought you were telling me that all O
t .t O
16 83' 3 volatile secondary water treatment was successful -- well,
, 1 I won't try to characterize'what your testimony is. So-tell me.again whether or not all volatile secondary water treatment is considered, in your opinion, to be a mitigative measure or a preventive measure in terms of g steam generator tube integrity problems, as you use it in your testimony.
t
! A You're limiting the question to all volatile
. () treatment?
9 i
Q Yes. I'm taking it a piece at a time now.
! !.5 A With regard to phosphate wastage, I'm not a 11
'<0 corrosion specialist. I expect it is a preventive measure 12 which addresses that, particular phosphate wastage problem.
Obviously the treatment is intended to address 3
corrosion problems in general, and I would expect that in that sense it is a mitigative treatment.
!() Q Fair enough.
How about the next item, improved controls and i monitoring of secondary water chemistry? Is that, in your i 19 30 opinion, a mitigative or a preventive measure in terms of i 20 i
steam generator tube
- integrity problems, or perhaps both, t
21 as you previously testified?
.,()
l 22 i
iO i
- , _ . . _ . . . _ _ . - . _ . _ _ . . _.- m ..._ . . . _ . . . .____,m_ - . . - _ . . . . . _ _ _ _ . - , _ _ _ _ - . . _ . . , , , _ , , . . _ . _ . . . _ , _ . - . _ _ , - _ . , . . . . . _ - ,
lO 17 84
!O A nitigative.
O I'm sorry?
A In my opinion, this is a mitigative approach.
'O Q All right. Fine.
4 Now looking at items 3 and 4, I thought in one.
of your. previous answers you were putting items 3 and 4
,0 in a different category from items 1 and 2.
A They are.
- O Q Is that true?
Could you explain?
! A These items will neither mitigate nor prevent - - -
O well, let te withdraw that coment.
12 Well, these items will not directly mitigate
. 13 i
~
nor prevent corrosion problems. They will provide a l'
iO warning that you have prcblems and will warn the utility tha t i t mus t , if it wants to save the steam generators,
!O h""*""*"******
37 process or prevent it.
! Eut beyond that, these last two items, items 19
- O 3 and 4, are intended to detect a situation where the
. tubes have become excessively degraded, and for those 4 21
, tubes which are excessively degraded, they must be
- 0.
!O
- - - . _ _ _ - _ . _ _ _ , _ _ _ _ . . _ ~ . . _ _ . . . _ . . - . - - _ _ , . - . _ _ . . _ .
F 18 3 85 13 repaired, either plugged or sleeved, or whatnot.
Item 4 is -- we have a one-two punch approach here to ensuring tube integrity.
3
' C) 4 Item 3 deals with rea.ular .ceriodic in-service inspections.
C) 6 Item .4 is an . additional very important- me thod or -- not method, but provides considerable added assurance that on top of the periodic inspections, that
!O ,
the tube integrity is not becoming excessively degraded.
O Is item 4 pertinent to the question of tube rupture?
f C?
A Yes.
12 Q Can you explain how it's pertinent?
A Two ways:
One is the occurrence of leaks may be an indicator that corrosion is proceeding at a higher rate
,() than anticipated; that perhaps a sufficient allowance for additional incremental corrosion pre-inspection j bu6 hasn't been provided for in the plugging limits. Leakage 19
'O may be indicative of additional tubes which r'epresent --
20 which are incipient leakers.
O What do you rean by the t'erm " incipient le ake rs " ?
O
-s 19 3J 86 13 A About to le ak , 90 percent throughwall, or 95 percent. But in a practical sense, then, leakage in a sense is an early warning signal.
. 3
- D Secondly, any leakage rate, regardless of its number, provides -- regardless of what you set it at, provides some additional measure of assurance against
.a,) 6 tube failures. But the limit, which I understand will be 7
specified for Byron, has been set to assure that if a
- (3 given' tube is leaking at the leakage rate limit under normal operating conditions, that if you were to suddenly go into main steam line break, . that the crack length
.O involved would not be sufficient length to result in a 12 l tube rupture, or a gross tube failure, or a significant i 13 leakage during the accident condition.
Q How does a limit leakage' rate warn us of incipient leakers?
O A In general, I would say that where you have a g
tube with a defect that's gone all the way 100 percent throughwall --
- O I thought an incipient leaker 0 Wait a minute.
20 was one that wasn't all the way through.
t
! A Well, the answer to the question was why does 22
- O l
! C) -
I
O 20 87 O , the occurrence of a leaker tell me we have incipient _
leakers.
O I'm sorry I interrupted. Go a' head.
O A Based upon experience, whenever we have leakers, in all probability we have tube.s where simi'lar degradationi ,
has proceeded, at least part th: oughwall, of ten considerably part-throughwall. You generally have some secondary side 7 x i ,
3 corrosion. It typically affect:1 many tubes in [the',.spe'c'ific ,
S ; ,
O region of the bundle , not just one bundle. , 'Yh 9 .
, -c 10 So there will be many tubes behavikg s'imilarly.
? s.
11 A leaker will represent the tube. that 'has' been degra'ded '
- O the most. -
12 .
n Q What you are telling me 'is th'at if , there are t leakage symptoms, that this is an .'iridicator', ,per aps , that
- O
- 14 there may be other tube's in the steam generator; that mighti 15 l
be subject to bursting and testing ought to be dene,-
16 s . <
V / . , . . <
- O eddy current testing or other surveillance ought to.bc.done o -<
17 -
to check it out? Is that what you are. telling me?
18 A No. But as I indicated in ' the second aspect l
- O of the response, the leakage rate limit is intended to, assure that individual tubes won ' t rupture. Soifyou've/
got detectable leakage that's less than the tech spec limit.
, e
'O ,
h (
, . - - - . ~ . . . , ,
1 88
- O
- 6
, i Cys A I'm not suggesting that we're in a situation that you s
1 may f ail tubes even if you go into an accident.
It does suggest to anybcdy conitoring the 20 4
- performance that the steam generators are degrading, that
, a number of tubes may be involved, even beyond the leakers, and that certainly you want to keep on top of the situation.
IC) 6 In the case of the Regulatory Staff, we want to perhaps address ourselves to whether or not the plugging criteria remain adequate for the corrosion process that is
.(3 9 .
4 10 taking place; whether or not the frequency of inspections that are specified in the tech specs remain adequate for 9 '
/3
- , 12 the situation we are actually experiencing at the plant g in .que s tion .
i 13 i
Q Well, will a steam generator tube leak before it bursts? Aren't those two inconsistent phenomena?
i A The term -- the word " burs t" in the context of 16 j) steam generators generally refers to --
Q As you and I have defined it already, what l~s the word " burst" means.
i 19 O .
A- cross tube failure, you 're talking about 20
-/
i the gross tube failure. -
21 ja ' -
Leaks, when I speak of leaks, I'm generally 22
[$ -
i ,
d 3
I
~
F
. - , . . - - - , y - ._ , . . . _ . - - . , _ , . -- , . , , - . . ,
22 3
89 I speaking of local failure. Failure might even be the wrong 1
2 word. Local penetration of the tube wall, where you get
. 3 relatively small amounts of leakage, on the order of 1, 2 4 gpm or less.
5 Q All right. But will a steam generator tube jy 6 that is the subject of a gross failure, will that leak 7 before it incurs the gross failure?
i 8 A Cperating experience indicates that generally I)
, in the vast.' majority of the cases, that will be the case, go but not absolutely always.
11 Q All right. Are you f amiliar with the steam
- O i
12 generator tube failure incident at Point Beach back in 1975?
13 A Yes, but if you're going to ask me whether or j() 14 not -- yes -- well, I have some familiarity with it. It's 15 been a while since I reviewed the circumstances.
16 Q Do you know whether or not that involved a gross O 17 failure of the steam generator tube?
l 33 A It was on the order of 100 or 125 gym, or some-
[
19 thing like that.
iO 20 Q Do you know whether or not that leaked before 21 that happened -- the failure occurred, rather?
i j) 22 A I haven't reviewed it recently enough to say.
i u
O
L.- , a. --
. 23 90 O
1 I believe it was leaking at some rate, but I can't --
2 O I'm talking about the particular tube now, 3 not elsewhere in the system.
7 ~
4 A one -- well, I don't know for a fact without s checking the circumstances what .the prior leakage history
() 6 was for that unit prior to the rupture.
7 0 Are you familiar with the steam generator tube j g rupture that happened at Surrey in 19767 O
9 A It happened, yes, it was about 80 gpm.
10 Q Was that a gross steam generator tube f ailure 1; incident?
12 A, It's generally classified as one of the gross 13 rupture events , yes.
!() 14 Q Do you know whether or not that tube leaked is before it failed in a gross manner?
16 A As I recall it -- and I'd have to check the 17 f acts again -- there was some initial leakage prior to the l 13 failure. I don't recall how much. I don't recall the 19 specifics. I'd have to research that.
.O 20 0 Okay.
21 A There are two other rupture events, of course, l(3 22 which did not involve prior leakage. Prairie Island and l
l
I3 24 91
~6 1 Ginna. And the reasons these did not involve prior leakage 2 is because we were dealing with a wall thinning phenomenon
.. 3 and not cracking. If you have a general wall thinning,
.O 4 you can get -- you can lose enough wall thickness over s enough of an area of the tube, such that there will be
'(3 6 no tell-tale leakage prior to the event.
7 Q Was that wall thinning dus to corrosion or some 8 other problem at Ginna and at Prairie Island?
O 9 A Both occurrences involved mechanical wear or 10 abrasion of the outer surface.
11 Q There was some foreign material or something O
12 inside the steam generator that was wearing on these tubes; 13 is that correct?
() 14 A Yes.
Is Q But that wasn ' t the case at either~ Surrey or 16 Point Beach, was it?
17 A No.
18 Q Now as I understand this regime as you have l
19 described it in paragraph 5, that you have this leak rate
- O 20 limit and you have something called a plugging criteria, i
21 and you have an inspection interval, and the idea is to 22 coordinate all three so that you plug all tubes before you 3
l
'O .
,_y. _ . _ .,. _ ,.~.. , _ , _ , - - - - - . . -
25 f3 92 1 reach a point where the tube walls are so thin, they might 2 burst or leak or whatever; is that correct?
. 3 A That's correct.
O 4 Q And I'm interested to -- well, I guess let
, 3 me ask Dr. Raj an :
(J 6 Can you tell me briefly just what the plugging 7 criteria are?
i 8 A (Witness Rajan) The plugging criteria are --
I) 9 they are outlined in Regulatory Guide 1.12 1, and basically 1
go the . criteria -- there are three criteria:
11 One is, number one, that the tube will not
-O 12 reach the yield point, the tube material will not reach 13 the yield point 'during normal operating pressure differentia:.s .
g 14 The second criteria that has to be met.is that i 15 the margin to f ailure or margin to burst will have a f actor l 16 of safe ty of three against normal operating pressures.
l) -
37 In other words, if the burst pressure is 3 18 de lta P , then the normal cperating pressure should be l 19 no more than delta P.
O 20 Or, putting it the other way around, if the 21 normal operating differential is delta P, then the burst j) 22 pressure should not be more than 3 delta P -- no less than
.O
26 93 O
1 3 delta P.
2 And the third criteria is that there should be 3 an adequate margin to burst under accident conditions and O
4 pressure differentials and loads.
5 A (Witness Murphy) Also we built into the
() 6 plugging limit allowances to account for eddy current 7 error, and incremental corrosion between inspections.
3 Q Is there a generic yield for the first criterien O
9 which you characterized as the yield point of the material go itself, the steam generator tube material? Is there some throughwall thickness that establishes that yield point?
) !!
12 A (Witness Rajan) Yes. Based on a very large 13 number of tests with the dif ferent types of defects, I
() 14 believe for -- of course, this would depend on the 15 dimensions of the tube. Different steam generators have 16 different diameters and wall thicknesses. So this would 17 differ from tube to tube.
18 But in general it can be said that approximately 19 25 percent of the tube wall -- if there is a 25 percent i
20 of the tube wall remaining, the yield point would not be 21 reached under normal operating conditions.
() 22 Q All right, now, what about under accident lO
27 O 94 o I conditions?
2 A Under accident conditions , the minimum wall 3 thickness has been -- well, depending on the location of 4 the def ect in the steam generator.
5 For example, the tubes that are located near (J 6 the U-bend regions, they would be subjected to higher 7 loads than the tubes that are near the tube sheet and 8 the supports .
9 So if one were looking for defects near the 10 supports , near the tube sheet -- and here again the figures 11 differ from steam generator to steam generator --
O 12 approximately 25 percent of the tube wall would be adequate.
13 Q So basically the criteria are the same whether
- () 14 it's normal operation or -- at least for these particular is tubes you're describing, the criteria would be the same 16 whether it's normal operation or under accident conditions; 17 is that correct?
18 A (Witness Murphy) I can speak to a series of 19 D-1 steam generators.
- O 20 Q Wait a minute. I want to get an answer from 21 Dr. Rajan, and then you can add to it.
- () 22 A (Witness Raj an) The numbers work out to be
- O
.) 28 95 j(3 g about the same , but they are based on different analyses, 2 totally different analyses.
. 3 Q I see. It's just a coincidence?
.O 4 A It's just a coincidence. And here again, as I e 5 pointed out, we have to look at a specific model and a 4J 6 specific tube wall thickness to determine the minimum wall 7
requirement.
g Q Mr. Murphy, do you want to add to that?
13 A (Witness Murphy) No.
9 10 0 When is this analysis normally done , Dr. Rajan?
A (Witness Raj an) The Licensee makes a commitment O
12 to abide by the requirements of the reg guide prior to operation of the plant.
I 13
() 1-4 However, during operation, when the specific 15 tubes are being plugged, specific analyses may be done 16 for those tubes.
- () g.7 Q I see.
18 A (Witness Murphy) I'd like to expand on that, 19 perhaps.
lO 20 Q Sure.
l 21 A Standard technical specifications contain 22 in parentheses, plugging limits, which I think are I
r i
lO..
t l , ,_, . - - - - - _ - _ - -.
C) 29 .
96 C) g generally about 40 percent -- which are 40 percent for 2 Westinghouse steam generators. So this plugging limit is 3 shown on the standard tech specs with an asterisk. The O
4 asterisk provides guidance for the Applicant in terms 3 of how he might go about justifying a different limit.
) 6 The limit is -- the structural characteristics 7 of a tube is a function of its geometry, and there are 8 nly a few different tube geometries out there. We have (3' lots of plants, but we have just a few categories of 9
to dif ferent tube geometries. All Model Ds are the same, 11 all Model 51s are the same, and once a supporting O
12 structural analysis has been performed for a plant with 13 a given type of tube, that analysis is generally valid for other separate plants.
() 34 g3 Q I see.
16 A So, as far as I know, individual plants don ' t C) keep resubmitting the same analysis over and over again.
37 18 Q What is the inspection interval under the 19 tech specs, if you know? I mean, let me explain where O
20 I'm coming from.
21 Dr. Raj an has explained the tube plugging 22 criteria. So if I understand what he told me correctly,
)
.O
2 0 30 97
'O we have to inspect the steam generator tubes from time g
2 to time to make sure if there's any degradation, we catch 3
them before they reach the yield point; is that correct?
4 That's the objective; is that right?
5 A That's the objective.
O 6 Q That tells me as layman that we have to inspect 7
at some reasonable interval related to whatever the rate 8
of corrosion might be?
O A That's correct, yes.
10 Q Can you tell me what the inspection interval is?
11 A Typically plants are required to perform 12 under their tech specs steam generator inspections every 13 12 to 2 4 mon ths . There are provisions, depending upon the steam generator performance, how well they performed,
!O 14 ,
15 how free of problems they've been, for extending the 16 interval for inspections for longer periods.
O Q When those inspections are conducted, do they 37 18 sample by eddy current testing a segment of the steam 19 generator tubes? Do they do 100 percent testing?
'O 20 A Yes. The initial inspection sample is a 21 percentage of the tubes, depending upon the -- the results of this initial sample inspection can fall into one of three lO 22
't
- O
l C) 31 98 3 g categories, ranging from essentially good to bad.
2 If you are in the good category, no further 3 sampling is required. If you are in the bad category, O
4 additional sampling is required. You may go through several 3 sampling stages.
Eventually you may get thrown into 100 percent O- 6 7
inspection of steam generators, g Q How does the Staff know that a 12 to 2 4-mon th O , interval is sufficient, inspection interval?
10 Let me strike that and ask the question better.
11 How does the Staff know that a 12 to 24-month O
12 interval is sufficient to identify any steam generator 13 tubes that may be approaching the yield point, so that es 14 they might require plugging?
a 15 Does that clarify it for you?
16 A Yes. First, let me state that the Staf f is O g7 generally aware of the condition in terms of the general 18 condition of a plant, whether it's got an extensive 19 corrosion problem, whether it's occurring at a low or O
20 high rate.
21 Q It's an unf air question. Let's limit it to a 22 P l an t that is just beginning to operate, like Byron. Let's 0
33 32 99 I3 take that kind of clean plant.
g Can you answer the 2 question in that context?
3 A Yes.
'O 4 Q How do you know that 12 to 2 4 months is a 3 proper interval?
A Based on operating experience, with one O 6 7 exception, we don' t run into significant wall penetrations 8 by corrosion generally within the first -- well, one or.
'(3 9 two cycles of operation. If there is a corrosion process 10 taking place, you will see the early sta'ges of it during 33 your eddy current inspection.
.O 12 0 Well -- go ahead, I'm sorry.
g3 A There is at least one corrosion phenomenon, the primary side corrosion, stress corrosion cracking
- 4) 14 15 phenomenon, that can occur quickly. This particular I
16 corrosion problem could conceivably occur during the (3 37 first cycle of operation. It has been observed that way 18 for Model 51 steam generators.
19 0 What's the phenomenon?
O 20 A The so-called U-bend cracking phenomenon.
21 Non-denting-related U-bend cracks. It 's of ten called the 22 the tangent point cracking problem.
)
'O
33 O 100 1 Q This is a problem you say that is only 2 applicable to Model 51 steam generators?
3 A It's only occurred there, to date.
O 4 0 You said there was one exception to this. Is 5 this the exception you are talking about?
O 6 A Y'S- Ye8-7 Q Did this particular phenomenon result in a gross 8 tube f ailure at some plant?
() 9 A That particular phenomenon has not ever resuited go in a gross tube failure.
t 11 Q All right.
!O 12 A We have had dozens -- tens, or perhaps dozens, 13 of leaks as a result of this phenomenon and they_ have all .been very small leaks .
- 1) 14 15 Q Does the Staff change the inspection interval 16 depending on plant experience, in terms of corrosion 43 17 problems, or steam generator tube integrity problems that l
l gg might be identified during an eddy current inspection?
19 A Yes. Turkey Point 3 and 4 and Surrey Units 1 and
- O 20 2 ran into extensive and very severe denting. The Staff l
21 imposed requirements for performing -- see, first performing i
22 steam generator inspection every three months. This was (Gs
'O
m
'J 34 101 0
1 relaxed to every six months. For a number of years, these 2 four units were required nominally to operate for six 3 months between inspections, although the Staff did consider O
4 on a case basis extensions of two months or four months.
5 As I said, on a case basis.
(? 6 Q All right.
7 A There are other examples, as well, where we 8 have imposed additional inspections.
P#
9 Q Is it fair to say you start out 12 to 2 4 mon ths ,
10 and as experience dictates, you either lengthen it or 11 keep the same interval, or make it shorter? Is that it?
O
- 12 A The tech specs already m'ake provision for 13 lengthening the inspection interval, if you have real good
'(3 14 experience, and we generally don ' t -- we've never been l
15 requested to relax those criteria. But we have intervened i
(
16 to require more frequent inspections than required by the 17 tech specs.
i 18 Q I guess the only other question I have in this 19 particular area, is which do you select? Is it 12 or 24,
'O 20 or a range , or what would go in the Byron tech specs? Do 21 you know?
-)
( 22 A What would go in the Byron tech specs are what l
- (J -
l
< l
35
~h 102
() we have written into the standard tech specs, and as memory serves me, I believe they are required to do a periodic inspection every 12 to 2 4 mon ths . Something --
'C3 we are talking about -- and the precise length of the 4
interval is, you know, the Licensee will select that, depending upon his schedule.
.O s
6 Q All right. 4cw returning to paragraph 5, the reason I have asked these questions is that I would O ,
have expected to see in paragraph 5 an item 5 in parentheses dealing with some sort of analysis of steam generator tube failure during the course of an accident O
similar to the one that we briefly referred to, that was
'12 performed by Idaho with respect to steam generator tube rupture effects on a LOCA.
Can you explain to me why you don't deal with the consequences of that situation in your testimony?
O A With the requirements that we have --
g Q These four items.
A Yes. I believe that pending the outcome of
'O '
our ongoing generic programs, that with these programs ,
20 that we have reasonable assurance against --let ce remove 21 the term " reasonable assurance," because it applies to
.O 22 4
O vr -, - - - - , -
- - . - , , , w + -
() 36 103 I something else, bc: I believe that there is a --
2 Q Well, why don't you use the language in your 3 affidavit in paragraph 7?
O .
4 A It's used in a different context.
5 Q All right. Go ahead. Sorry.to interrupt.
O 6 A I believe that we have -- there is a very, very 7 low likelihood of gross tube failures during accidents.
8 Q That's because of the tube plugging criteria,
$b7 9 the in-service inspection and the in-service , inspection 10 interval, and the other controls for monitoring, and water 11 chemistry treatments; is that right?
O 12 A That's correct. That's correct, but let me 13 correct one thing. I don ' t wish to get unduly carried i
{<3 14 away. I said very, very unlikely. I mean to say I i
15 consider it very unlikely that we would have a rupture 16 during -- in the event that we did run into a major design I
33 17 basis accident. I do believe that there are a lot of --
18 with the extensive degradation that we have observed 19 throughout the industry, the tube ruptures that we have
!O 20 had during normal operating conditions, not during a 21 transient, that there is sufficient cause for the Staf f
) 22 to take a close look at the regulations to see that they i
!O o
O 37 104 1 adequately address the problems that do indeed provide 2 suf ficient assurance against a rupture event occurring, 3 both during normal operation and during accident conditions.
4 0 Is the Staff doing that?
3 A Yes.
O 6 Q Why isn't it in your testimony some place, then?
?. 6 7 A Well, it is. I looked upon my testimony as ..
8 sort of an expansion of testimony to the SER. The Task "o Action Plan, the generic safety issues, something that
- 10 is being reviewed by the Staff.
gj Q So in paragraph 2 of your affidavit, you 12 essentially -- I guess what you have done is adopted and 13 incorporated by reference the information and material
.O 14 that's in the Safety Evaluation Reports for Byron; is
~
15 that correct?
i 16 A Yes. And I think it so states. Yes.
,0 To your knowledge , has the Staf f determined 37 Q gg that it is necessary to -- strike that. Let me see how l
19 I want to phrase this .
O 20 We have been using the term " design basis
.21 acciden t'. " There are a number of accidents that the Staff 22 requires evaluation for before a nuclear power plant can
- O O
\
38 O los O g be licensed to operate; isn't that correct?
2 A I'm sorry, do you mind repeating the question?
3 0 Sure. We 've been using the term " design O
4 basis accidents" during our discussion here, and I just 3 wanted to establish that we are on the same wave length, that there are a numkr of so-called design basis accidents O '
7 which Applicants must analyze and the Staff must review 8 before a plant can be licensed to operate. Is that correct?
O 9 A That's correct.
10 0 I take it that the Staff has not determined 11 that an accident situation involving an established design
.O 12 basis accident, coupled with a concurrent tube failure, 13 should rise to the dignity of being called a design basis 34 a ident, in and f itself; is that correct?
O 15 A It his not, to my knowledge, issued a formal 16 conclusion to that effect. It is certainly something
- O g7 that is under discussion and consideration at this moment.
! 18 0 All right. I guess it's the lack of that 19 discussion and consideration that has triggered me to ask
- O 20 this line of questions. Perhaps it is in the Safety 21 Evaluation Report and I just didn ' t notice it. Could you 22 point me to where that discussion might be in the Safety
- O L
..O 39 106
- O 1 Evaluation Report?
- 2 A It's a lead sentence , in Appendix C.
O 3 Q Can I look over your shoulder? I'm sorry, it's 4 Appendix what?
5 A Appendix C, or whatever they call it.
() 6 Q Yes, I've got a copy of that. Go ahead.
7 A Addressing the primary concern of tube integrity.
3 The primary concern --
- O 9 Q Where are you reading, what page?
p3 A Page 9, page C-9.
.(3 11 Q Okay, give me a chance to catch up with you.
12 Okay.
13 A The section is entitled " Westinghouse Steam
,O Generator Tube Integrity," and we begin the discussion 34 .
i 15 by stating the primary concern is the capability of 16 steam generator tubes to maintain their integrity during O
17 normal operation and postulated accident conditions.
18 Q All right. Is that the extent of the discus-g 19 sion on the occurrence of steam generator tube f ailure 20 . under postulated accident conditions? I don' t see any 21 discussion on that subject on either page C-9 or page C-10.
!O 22 Maybe I just haven't seen it there. It might be there i
1 10
40 107 O
1 and I just missed it.
2 A No, it's the -- that sentence describes the
() 3 overall objective of our regulatory approach, you know, 4 the secondary water chemistry, the surveillance requirements ,
5 the plugging limits and the le~a k rate limits.
6 Q I know, but we got to this page through a 7 series of questions of which I ultimately asked you, and 8 I thought you said the SER might contain a discussion of c
9 the Staff's judgment in the consideration of the need to 10 deal with steam generator tube f ailure under accident 3 !! conditions, and we are now looking :tt page C-9 and C-10, 12 and I'm asking you where that is.
13 A Let me first refer again to the contention.
C) 34 As I say, the SER and my testimony describes what we 15 consider to be the rationale for --
16 Q We ll , I'm prompted to say that the answer to
%.rl 37 my question is no, there isn't any discussion in either 18 your testimony or the SER. But look at page C-9, and one, .
73 19 two -- the third paragraph, last sentence of the third 20 Paragraph. I'll read it:
21 "The tubes and tube sheet are analyzed in C 22 WCAP 78-32 and confirmed to withstand the q
O 41 108 O
1 maximum accident loading condition."
2 A Uh-huh.
3 0 ** th** **""*"
- 9't t th* P i"* I ""
O 4 trying to elicit here?
5 A I don't believe so.
O 6 Q It does not?
7 A You know, it goes without saying that the g Westinghouse and the Applicant are designed -- are required O
9 to design the plant, including the steam generators, to 10 meet all design loadings, including those which occur during a design basis. accident or a faulty condition.
() 11 12 There are requirements they must satisfy. They must 13 demonstrate , in accordance with established rules, that
() 34 they can sustain all normal operating or accident or 15 transient conditions.
16 Q Do those analyses assume a steam generator O
17 tube f ailure during the course of the design basis accident?
18 A No.
19 Q All right.
)
20 A But that wouldn ' t -- you are getting into a 21 dif ferent area, you are getting into in terms of systems,
() 22 does the systems respense consider that situation, and the O
42 O
109 O
1 answer is no. That is not a limited loading situation for 2 a given steam generator tube.
3 Q Mr. Jenkins was --
()
4 MR. GOLDBERG: Dr. Rajan had something he s wanted to add.
O 6 MR. GALLO: I'm sorry, go right ahead.
7 WITNESS RAJAN: In an appendix to this WCAP 3 78-32, they did consider the steam line break event also, C) .
9 BY MR. GALLO:
10 Q Say that again.
11 A (Witness Rajan) In an appendix to the WCAP
)
12 78-32, that we just referred to, the analysis to do a l 13 steam line break accident was also considered.
O 14 Q I see.
Is When you say that, do you mean a steam 16 generator tube failure was -- the consequences of that O
17 f ailure on the accident was considered in WCAP?
l j 18 A Essentially the WCAP considered the effects 19 of LOCA loads, loss-of-coolant accident loads on the 20 Steam generator tubes, and it determined to what extent 21 degraded tubes can withstand the dynamic LOCA loads, and
-(3 22 in an appendix they also considered these facts of steam i
O
- .~) 43 110 0
1 line break accident loads on steam generator tubes .
2 0 Well, then, Mr. Murphy, maybe I misled you or 3 wasn't clear with my question. Perhaps this sentence does J
4 address the very point that I was driving at. The 3 sentence on page C-9 that refers to 78-32.
O 6 A (Witness Murphy) You're --
7 Q I'm sorry?
8 A The'way I've been interpreting your questions, g
9 there were two situations one might want to consider:
10 One, whether or not we design the plants to g 11 prevent tube failures, and the answer to that questien is 12 yes. And there ar'e established rules for that.
13 This sentence alludes to the fact that these O g particular components discussed here have been designed 15 and analyzed to withs tand accident conditions .
16 But now we're going beyond that, and we're g7 saying let's assume that corrosion takes place and becomes 18 extensive, and let's assume that routine surveillance, 1
n 19 leak rate limits and so forth, in a particular instance u
20 didn't work. I mean it didn't successfully prevent the l
21 rupture. Okay? And what then? That's a different --
O 22 that's a little different aspect.
'O
_=
44 O -
111
- (3 Q That's the par't I'm trying to focus on.
3 A We designed the plants to avoid that situation.
Now, if it happens, okay, and then the question is --
3
- O the question then is so what? Is this a concern? This 4
is a different consideration entirely.
O In your judgment, in your opinion, is it unnecessary to consider those situations in circumstances that you just described, because of the four factors on
() ,
paragraph 5 of your testimony will essentially provide reasonable assurance that you're not going to have that kind of problem?
- O A Not exactly. I think, as I said before, we've 12 g got to look into this situation further, and we are as g part of the unresolved safety issue alluded to here, and it's part of another study which is also going on at the
! same time. It's a related study. I think we have to take l 16 g3 a close look at and understand the effects of ruptures i
and the consequences of LOCA, or main steam line break or g anything else we might want to postulate. We have to --
o O in view of these findings, we will have to reexamine, perhaps reevaluate our existing requirements in light of g these findings. But I know that these studies are ongoing.
iO I
i l
l
[
C)
iX 45 112
- O I You know, I'm familiar with the steps.that are 2 .being -- that this matter is being pursued, and pending 3 resolution of these items, of these issues, I believe that 4 the requirements that we have in place, will be putting 3 into place for Byron, are sufficient to provide reasonable lO 6 assurance of public health and safety.
7 MR. GOLDBERG: Can we go off the record for a 8 moment?
'O MR. GALLO: I just want to ask one question,.
9 10 and then we can go off the record.
11 BY MR. GALLO:
.to i
12 0 I guess maybe to some extent, Mr. Murphy, Ive 13 been unf air to you. If either Mr. Parcheski or Strosnider
- (D 14 or Pasedag were with you on the panel, perhaps they could 15 provide the insight that I'm striving for. Is that a fair 16 statement?
C) A (Witness Murphy) Well, the -- you haven't 37 18 really asked me about the mechanics of what specifically 19 is it about steam getting into the primary from the secondary O
20 that causes the fuel to heat up, and when do we have to 21 s tart worrying about the fuel melting and all this kind of 22 stuff. They would be able to address that and tell you how 13 1
O
46 113 O
1 they evaluated that.
2 Q You're correct, I haven't done tha t . I'm n
u 3 basically trying to determine in my own mind whether or 4 not your affidavit, as you have testified, is complete, 5 and I'm having a hard time reaching that judgment, because O 6 there is no discussion of the point that we have been 7 debating here for some time.
8 Let's go off the record.
n Y 9 (Discussion off the record.)
10 (Recess.)
A.7
%)
11 MR. GALLO: I've just got one question left 12 on the accident discussion, and then I'd like to go on to 13 a new subject.
C) 14 BY MR. GALLO:
15 Q Mr. Murphy, Mr. Jenkins gave me a document 16 prior to the start of this deposition enti~tled " Steam
'O 17 Generator Status Report, February 19 82, U.S . Nuclear 18 Regulatory Commission. "
l ,., 19 I understand it was obtained from the NRC v
20 under a Freedom of Information Act Request by the 21 Intervenors in this case.
CD 22 I am asking you if you recognize that document,
!n
(.)
!d
s m 47 114 a
1 and have you come across it in your work in dealing 2 with steam generator tube problems?
g 3 A (Witness Murphy) Yes.
4 Q Can you tell me who developed it, or wrote 5 the document, and the circumstances for its development?
o'
'" A Yes. Jack Strosnider. And the raw material 6
7 behind this report was prepared under the -- as part of 8 the Task Action activity.
o
%)
9 Q I see.
10 Do you work for Mr. Strosnider?
11 A No, Mr. Strosnider is in a different office of g
12 the NRC.
13 Q I see.
C' 14 A I used to work with him a couple of years ago.
15 Q Did you have any involvement in the preparation 16 of that document?
O 17 A Of this one? No, I did not.
18 Q How about you, Dr. Raj an ?
, 19 A (Witness Raj an) As far as formally, no. There
,,U l
20 may have been some input from me on the implementation of 21 Reg Guide 1.121, which I don ' t see here, so I would say the (J 22 answer is no.
10
gp 48 115 f
a 1 MR. GALLO: Can we go of f the record? ,
2 (Discussion of f the record. ) ,
g 3 MR. GALLO: Let's go on the record.
4 BY MR. GALLO:
$ Q Mr. Murphy, while we were off the record, I 6 asked you whether or not you recognized this document 7 as the complete report, and you indicated that perhaps 8 it might be a preliminary version. Would you clarify that 9 f or the record, please?
10 A (Witness Murphy) I cimply cannot say whether 11 or not it is a final document or not. I'd have to read 12 it in detail to know whether or not this was the one. Or 13 better yet, to check with the issuing organization to make
? 14 sure this is the proper -- this is the final report. I 15 would assume that it is, judging from the date, but --
16 MR. GALLO: All right. Subject to that check, D
17 I'd like to have this document marked as Applicant's -
18 Exhibit -- no, strike that, r~c.e3 as Murphy /Rajan
, 19 Deposition Exhibit Mc. .' .6 'll give it to the 20 reporter to mark for thc, purpu 2. But as I unde rs tand ,
21 during an of f-the-record discussion , that Mr. Goldberg, 7 22 on behalf of the Staff, will be kind enough to obtain a JE.
~. ,-
!/1 49 i 116
/ ,
O I copy of this document that does not have the markings
'2 and underscoring that is on this document, which I under-g3 / 3 stand had been performed during the review by DAARE/ SAFE
- 4 people.
l
! 5 (The document referred to was O 6 marked Murphy /Raj an Depo.
! 7 Exhibit No. 1, for identifica-ONGCC 8 tion.)
O 9 MR. GALLO: .That's all I have on this point,
- , 10 unless there is any other comment.-
h 'i 43 11 MR. GOLDBERG: My only comment is one of
- > 12 clarification. I assume you are making it a -- why 13 don't you just mark it for identification? I'm not sure C3 l '. how to make it an exhibit, unless these gentlemen are,
! 15 yaa know, responsible for adopting its contents or you
! 7 16 jut t want it --
O >
,3 17- MR. GALLO: That's the infirmity that exists.
13 If I or anybody should try to offer it into evidence, that
> 19 objection is there . I just want it a part of the deposition 3
20 record.
I 21 MR. GOLDBERG: Okay.
43 ' MR. GALLO:
22 That's all. But I'm not addressing I
O i
ll r I
D 117 O
1 myself to the question of whether or not it's an-2 admissible document in the present form.
3 MR. GOLDBERG: Right.
)
4 BY MR. GALLO:
5 Q All right, let's change the subject.
C) 6 Dr. Rajan -- let's see, I'll have to find 7 your affidavit -- I believe that you have described, in g answer to Mr. Jenkins ' questions, the flow-induced O vibration problem that you address in paragraphs 8, 9, 10, 9
10 11 and 12. Is this the problem that occurred at McGuire?
11 A (Witness Rajan) In paragraph 8 I do talk
)
12 about Model D-2 which is the steam generator model for 13 the McGuire Plant. Anc in paragraph 9 I describe -- I (3 g4 make a statement that D-4 and 5 were used -- are used at 15 Byron, and I go into some of the differences between 16 the Byron steam generator and the Model D-2 and D-3.
17 0 And did this phenomenon, flow-induced 18 vibration phenomenon, occur at a foreign plant, too?
19 A Yes, it did.
O 20 Q Was that KRSKO? Or some other plant?
21 A Well, KRSKO has the D-4 steam generator,
() 22 very similar to the one being used -- proposed for Byron.
CD .
O 1 118 O
1 Q Well, did the phenomenon occur there at that 2 plant?
3 A Yes.
()
4 Q And how is KRSKO s pe lled , for the reporter's 5 bene fit?
O 6 A K-R-S-K-0.
7 Q All capital letters?
8 A All capital.
O
, 9 Q Any other foreign plants besides KRSKO?
to A There are two other plants involved in this phenomenon. One is Ringhals in Sweden, and the other
() 11 12 one is Almarz in Spain. And both of these have D-2 and 13 D-3 type steam generators which are somewhat different l
!(D 14 in their preheat design than the Byron.
15 Q Now, as I understand it from your testimony, 1
l 16 Westinghouse has developed a generic program to deal
- O 17 with this problem; is that correct?
18 A Yes, sir.
19 Q Are you f amiliar with their program?
g I
20 A Yes, I am.
l 21 Q I believe you testified that you looked at l
,' (3 22 some data that was taken at the KRSKO plant with respect 1
,0
3 52 119 I to the evidence of vibration?
2 A Yes, sir.
. 3 Q And on the basis of that data, have you 4 determined that a 70 percent power level is about right 5 where the phenomenon might not be seen?
O 6 A , I would say that is a preliminary conclusion 7 I have reached.
8 Q All right. Now you say in paragraph 10 that
~
9 Westinghouse is evaluating modi ~fications to the auxiliary go feedwater system and you described one of those modifica-
,, 11 tions.
12 To your knowledge, is Westinghouse considering g3 modifications in addition to the one you described in
's 14 paragraph 10?
15 A That's correct, they are considering several 16 approaches and these may be used in combination or a
g7 individually.
18 Q Do you know whether or not Westinghouse is 39 recommending any of these approaches for implementation?
O 20 A No, they have not. They have not finalized 21 their recommendations as to which approach or combination c) 22 of approaches they will adopt for Byron, or for domestic O
C) 53 120 0
1 plan ts .
2 0 Do you know when they might do that?
A My understanding is that the' schedule for this
() 3 4 is they have made -- they have made a preliminary presenta-3 tion on what these options are.
C) Q To whom?
6 7 A To the NRC Staff, and they are currently doing 8 the analyses and testing and evaluation of the various
- O 9 options, and my understanding, is that by November. or 10 December they will have finalized the test results and
- 3 11 data for Staff review, and at that point we will proceed 12 on the acceptance or nonacceptance of those options. -
l 13 0 Will the Staff approve one or more of those 43 14 options? Is that what will happen?
15 A I can only predict at this point.
! 16 0 I'm not asking you if in fact you will, but lO 17 is an approval, up or down, down the road, is t'at what 18 you plan?
19 A Yes.
O I 20 0 So Westinghouse will come in with their program, l
l 21 Staf f will review it and approve those aspects that it 13 22 finds acceptable; is that a fair statement?
C) .
O 4 121
'O 1
A I think it should be clarified that we are 2 aware -- we are aware in more than a general way of what is being considered. All we are waiting for is hard data, g 3 4 and results of analyses. We are aware in a fairly --we 5 have a pretty good idea at this point as to how Westinghouse t
.(3 6 is approaching this problem and what the most likely fixes 7
are going to be. So we have a fairly good idea at this g poin t.
4 0
1 , Q What hard data do you need?
10 A Well, the hard data would consist of -- it 11 could consist of, for example, model test results and it j7 12 could also consist of stress analysis results of some of 13 the fixes that they are proposing. And it could also i
13 14 result -- well, thermal hydraulic analyses and the results.
i g3 So we are aware of the fixes in a general'way, 16 but we have not reviewed the documented information yet.
O j
37 Q I guess I neglected to ask you. What fixes l
18 are you aware of besides the one you described in your 39 paragraph 10?
20 A There are several fixes. One of them is the 21 addition of flow-straightening veins .that will be attached
() 22 to the impingement plate , and the effect of this would be O
15 8 122 4
3 to reduce the turbulence and make the flow uniform in that region.
2 Another is the change of the flow restricter 3
4 device from one which has three holes to one which has a larger number of holes.
And then they are also considering sleeving of the tubes and the supports to stif fen the tubes in that region, and I forget, but these are the major other options j ,
besides the change in the aux feed system.
Q Now is the purpose of these fixes to reduce 10 the flow of the water so that the vibrations don't occur?
Is that it?
A No, it's not -- the object is not necessarily g
to reduce it, but the object is to reduce the turbulence in the flow.
Q And one way to do that is as suggested in your 3 paragraph 10 in your testimony; is that correct?
17 A In 10, I talk about actual reduction of flow to the main feed.
O I see. Well, but you've got 30 percent coming from another source; is that correct?
A That's right.
m
. ,Y
.-6 C) 12 3
+
[3 3 Q The ef fect of the two is to provide as much 2
water as the current design has now?
3 A Yes. Yes.
O 4 O So by creating two sources, I assume we deal 5
with the turbulence problem you are talking about?
A That's right.
() 6 7 0 Now I think you have testified 'iduat you expect 8
the Westinghouse anclysis on their potential fixes and 43 , recommendations in October or November?
10 A That is the general timeframe that has been 11 discussed.
O 12 O Are you going to personally be involved in the 13 review of these analyses?
14 A Yes.
- g 15 Q Do you have any estimate of how long the Staff's l
16 review might take?
(3 g7 A Well, we have consultants assisting us in the Staff review, and generally we can complete this in short 18 19 order. I cannot give a timeframe.
to Q Well, what do you mean by short order? Don't l 20 i
l 21 give me a specific date, but just ball park.
22 A Within weeks.
l
O37 124 IJ g Q Within weeks?
2 A Within weeks, yes.
3 Q Now you answered Mr. Jenkins ' question that way.
O 4
Are we talking 52 weeks, 100 weeks , or can you do any better 5
than that? I don ' t want to press you unduly, but the reason I asked the question is on the bottom of page 4 of O 6 7
the testimony, you say it is anticipated that Westinghouse 8
will have completed its generic program to select the most O , effective combination of auxiliary feed and/or steam to generator modifications to enable . installation and Staff 11 review prior to start-up of Byren. And I am trying to O
12 pr be to find out the basis for that statement.
13 We now know that Westinghouse -- we expect something from Westinghouse in the October-November time-O 14 33 frame, and now how long is the Staff going to take?
16 A Let me cla d y d at again, dat what we are O expecting from Westinghouse is not going to be a surprise, 37 gg for example. It's something that has been discussed.
19 Q All right.
O 20 A And we are aware generally of what is there.
21 What they will come up with is field data and test data 22 from scale model testing. So essentially it will be a O
!C 38 125
!O 1 confirmation of what we feel are going to be the fixes ,
2 and the review for that should not take too long, if the 3 end results do indeed conform with what we expect from
)
4 them.
5 Q I see.
- C3 6 So these are in the nature of confirmatory 7 studies and analyses?
8 A- That's -- I think that would be correct.
O 9 Q So you feel that you know enough now that you 10 could draw the conclusion that this problem of flow-11 induced vibrction can be resolved prior to start-up of M,.
12 the Byron facility?
13 A. We have gone f arther along with the D-2 and D-3
() 14 fixes. We have reviewed the flow' model test data and the 15 ana lyse s , and based on what has been accomplished there, 16 we feel that these are very promising avenues, and these O
l 17 are very promising methods of approaching this problem.
l 18 So we do feel that an adequate fix will be available, 19 and we would like to see confirmatory results and analyses l
20 along these lines.
j 21 Q Is the supplement to the SER that -- strike i
$3 22 that.
lO 4
19 126 T) 1 Somewhere in the materials there is a statement 2 with respect to the flow vibration problem. The Staff
~
3 intends to address it further in the supplement in the
'O 4 SER; am I correct in that?
3 A (Witness Murphy) Yes.
6 0 Is the purpose of that to deal with the results 3
7 of these confirmatory studies?
8 A That particular part of the SER was meant to X3 9 _ address either one, the fix or any' other alternative 10 approach that the Applicant would propose before start-up.
gg The SER was written in a general way before, you know, we O
12 knew -- before we had much information from the Applicant 13 or from Westinghouse regarding where they were going with 14 this.
4,0 15 Cur anticipation, as expressed in the testimony l 16 here,was based upon our understanding of what Westinghouse's I) 17 schedule is for completing its design review and what they 18 call their generic modification selection program.
39 Q And is that the October-November timeframe that n
20 Dr. Rajan mentioned, or is that a different timeframe?
21 A Yes. Yes. Yes. This was a date they gave us 22 at a mee ting here in May.
O,
- o
s
. 30 n"
12 7 A^' Now we can ' t -- we certainly can ' t make any 1
1 2 conclusion regarding whether or not Westinghouse can meet 3 its schedule, or we cannot speak for the Applicant, who
- O 4 may not choose to buy this modification.
5 0 What happens if the unexpected occurs and the j) 6 analysis is either not completed by the time of start-up 7 of Byron or it shows it's unacceptable to the Staff?
g ' What happens then , in terms of licensing the Byron facility?
d) 9 Dr. Rajan?
10 A (Witness Rajan) It would seem to me that if 11 we do not find acceptable fix, we could limit the opera-
!O 12 tion of the Byron plant to somewhat less than 100 percent
[ 13 power. That is one of the options that is obviously j3 14 available.
15 Another option might be that it might be f
! 16 delayed in the extreme situation, the operation _might be C 37 de layed .
18 A (Witness Murphy) Well, you know, there are i
39 factors -- we have a lot of things to consider. I think
'O -- I believe that from a strictly technical standpoint, 20 21 forgetting about questions like ALARA and so forth, I 22 believe that a satisfactory technical basis could be
)
O
al
- O .
12 8 1 arrived at, in terms of justifying an operating program 2 for Byron. Based upon ene experience we have acquired 3 day to day at McGuire, and based upon the experience
!O 4 overseas and what our knowledge of the Westinghouse 3 analysis and test results regarding the causes of the l(3 6 problem are.
7 Q The Staff believes that -- I guess it is your 8 present understanding and belief that the vibration I) , problem doesn't occur below 70 percent. Would that be 10 the power level you'd select if you were going to limit 11 start-up of Byron, something less than 100 percent of full 12 power?
13 A (Witness Rajan) At this point it would be j) g4 conjecture, but that could be. That could be an option.
15 We will have to examine the KRSKO data in far greater 16 detail and make a determination as to what level of O 17 power operation would be safe.
18 Q But you're telling me at that time you'd have 19 to review the data to see if 70 percent was still the 20 20 correct number?
21 A That's correct.
r 22 Q What's happening at McGuire? Are they limited
)
1 to
52' v
129 1 in some fashion presently?
2 A They are, they are limited.
3 Q Can you tell me what it is?
)
4 A They are limited to 75 percent power.
5 Q At the present time?
() 6 A At the present time.
7 A (Witness Murphy) This is following their 8 start-up from the present outage?
- Q 9 Q Do you want to speak up?
10 A This was an aside. It should be off the record.
11 MR. GOLDBERG: If the witnesses want to 12 confer before giving an answer, they are~ entitled to.
13 (Discussion of f the record. )
- gp8 g.4 MR. GALLO
- All right, let's go on the record.
15 BY MR. GALLO:
16 Q Do you want to clarify, Mr. Murphy , Dr. Raj an 's D
g7 statement or testimony that McGuire is operating under a 18 75 percent power limitation?
19 A (Witness Murphy) McGuire has operated at lO 20 different times under either a 50 percent or 75 percent 21 power limitation. They are currently shut down, and
() 22 the Staff is evaluating what it considers to be an O
o
v33 130 O
1 acceptable program for future operation beyond this outage.
2 Q Do you know whether these limitations, these 3 p wer limitati ns, were imp sed by the Staff, or were O
4 they voluntarily assumed by the utility?
3 A (Witness Rajan) They made a recommendation
() and we reviewed the data and the analyses which formed 6
7 the basis for proposal, and then we allowed them to 8 continue for a certain period of time for 75 percent, up
-O 9 to 75 percent power, and as Mr. Murphy pointed out, they
[ 10 completed that period of operation recently, and now they 11 are shut down.
12 A (Witness Murphy) They originally at one point
[ 13 last spring proposed a period of operation at 75 percent
- () l'4 powe r . We found they provided insufficient justification 15 for that power level, and limited them to 50 percent 16 operation. They resubmitted their basis, and we bought O 17 off on it, based upon our review, the second time around.
18 0 You mean you agreed with the higher --
19 A We agreed with their justification, that they l0 20 had reasonable justification, based upon their resubmittal 21 on a technical basis.
() 22 O For what level?
'O L
14
'(3 131 O I A 75.
2 .
A (Witness Rajan) 75.
- 3 MR. GALLO: I just want a minute here to look.
- O 4 (Paus e . )
s MR. GALLO: Okay , I'm finished, Steve .
MR. GOLDBERG: Okay. I'd just like to note
() 6 7 for the record that it's 1:05 p.m. , and we have exceeded a the allotted time limit,. with the indulgence of. the
- 13. 9 witnesses , and at their further indulgence , I am going to ~
10 ask some further questions, but I would urge the parties I
11 to attempt to bring this to a rapid conclusion there-
.O 12 after.
i 13 l
is 16
! .O - 17 r
18 l
l 19 lO
! 20 21 10
.O
SI) 132 0
1 EXAMINATION 2 BY MR. GOLDBERG:
O 3 Q T the extent I address a question to Mr.
4 Murphy and, Dr. Rajan, you wish to add a comment, or vice
~
5 versa, please feel free to do so.
C# First, I wonder, Mr. Murphy, can you briefly 6
7 describe your relevant educational background regarding 8 your testimony or affidavit on Contention 9-C?
O 9 A (Witness Murphy) I will. My education, I to have an M.S. and a B . S . degree in the field of engineering.
I go t the B . S . in aeronautical engineering, and M.S. in g, 11 12 civil. My specialty being structural engineering.
13 I have worked for the Bettis-Tye Power. Lab C) 14 for six years, engaged in the analysis and design of 15 core s tructurals , the Naval Reactors Program. I was not 16 involved at that time specifically with steam generators.
O 17 This was primr.rily in the area of core structurals.
18 .#,ince joining the NRC in July of 19 79, I have 19 been involved exclusively in the review of steam generator O
20 operating experience , surveillance programs , repair 21 programs , e t ce te ra, anything having to do with the O 22 integrity of the steam generator tubes, particularly with O
l
. ri6
.o 133 O
1 regard to operating reactors, I have been involved in 2 those reviews.
3 With regards to my association with steam 4 generator problems, that association goes back three years.
3 Q I think you have combined an answer that I was
- (3 6 going to ask. I was also going to ask professional 7 background, but I think you have given that in addition g to educational background.
'O Let me ask you, what is your role in the 9
10 Staf f consideration of resolved safe ty issue A-3, which' if 11 I'm not mistaken, is the steam generator tube integrity 12 task?
13 A For the past several weeks, at least the past
() g two weeks , and probably for the next few days, we have 15 been commenting extensively on the current draf t --
16 Q I'm sorry, I'll take that answer, but I think 17 it's not to the question I asked. I said what is your 18 personal role in that effort?
, 19 A My personal role has been that for the past
- O 20 several weeks, has been to provide my comments to the 21 draft report.
19 22 Q Are you part of that Task A-3?
0 1
'7 i(3 134 C) g A I'm not part of the task, per se. I have made 2 contributions to the report.
3 Q In what areas?
O 4 A In the areas of steam generator inspection i
g programs, procedures, sleeving, and surveillance in
<, 6 general.
v 7 Q These are subjects in which you have been engaged 8 since you joined the NRC in, I believe, 1979, you say?
!() 9 A That's correct.
10 Q I wonder, Dr. Rajan, if you can give me your 11 relevant educational and professional background.
O 12 A (Witness Rajan) I have a B.S. in physics-13 chemistry , and an oth er B . S . in civil engineering, with a 34 major in hydraulics, and a Master's in structural
- Cr 15 mechanics, and a Ph.D. in fluid mechanics. I have worked l
l 16 f r six years with the Naval Research Laboratory in their
() 37 piping programs for nuclear submarines, and since '74, 18 I have been with the Nuclear Regulatory Commission, and 19 besides other things, been the principal. reviewer in dae 20 mechanical engineering branch for problems of steam i
! 21 generators related to the mechanical engineering branch i
22 scope of review.
"O lO i
8
() 135 C'> 1 Q You have a prominent Staff role in consideration 2 of this so-called flow-induced vibration phenomenon?
3 A I'm sorry, can you repeat that?
-O 4 0 Yes. Do you have a prominent Staff role in s terms of the consideration, further Staff consideration jg 6 of the developments in this flow-induced or mechanical 7 tube vibration problem?
8 A That's correct.
13 , Q Mr. Murphy, there were several questions that to Mr. Jenkins asked, all regarding past, present and 11 anticipated steam generator tube integrity concerns.
- O 12 I wonder if you'can tell me just generally 13 on the basis of your experience at what level of tube 14 degradation does it become significant from a public g
is safety standpoint?
16 A (Witness Murphy) Okay. That particular -- I C) 17 interpret your question to mean how much -- how much 18 leakage would it take during an accident before we got 19 . severe consequences . And that particular issue is lQ
! addressed in some de tail in the document -- the February 20 1 21 1982 document that has been made part of the record about 22 20 minutes ago. This is the one prepared by Jack L_
d9 136
'O g Strosnider in the Office of Research.
2 0 Okay, let me clarify. I'm not sure it's been 3
made part of the record. It's been identified, and O
4 will accompany the transcript of this deposition.
3 I guess what I'm saying is there was a line O 6 f questi ns ab ut what I think was described as overall 7 steam generator tube problems, and I'm trying to get some 3 kind of understanding of this, the magnitude of the CJ ,
, problem , from the standpoint of public safety.
10 In other words, does a steam generator tube 11 problem equate to a public safety problem, and where is O
12 the line drawn, based on your experience?
I 13 A That's a very complex issue. In my opinion, O 14 based upon what I know, I think that without proper g3 controls and regulation, that steam generator tube 16 degradation could-ultimately lead to severe problems.
O 37 The whole issue in terms of what the concerns are, and 18 how we should be approaching these concerns is under --
19 y u know, it's under study by the Staff, but I believe
- O I
20 that based upon everything known to me of the analyses --
21 based upon my understanding regarding preliminary analyses :
1 22 concerning the consequences of an accident with ruptures l l'
I i
lO L
0 C) 137 E
$) I and so forth, that our current regulatory approach is
- 2 adequate for this interim period before we issue our 3 final generic conclusions.
'O .
- 8 4 Q Let me ask it a little differently:
5 You indicated, I think, in response to some
- n 6 questions by Mr. Jenkins that you couldn 't preclude some M
7 steam generator tube integrity problems over the expected g lifetime of Byron; is that correct?
[3 9 A It _ was -- I couldn ' t preclude corrosion.
l to O Okay, we'll confine ourselves to corrosion.
I
!! A Or degradation in gencial.
!O 12 Q Okay. What kind of measures give-assurance 13 that this isn't going to be a public safety _ problem?
c3 14 Should we be concerned about this?
Is A In my opinion, the current regulatory approach, 16 which we have discussed quite extensively up to now, I I3 17 think provides reasonable assurance for the public health 18 and safe ty but, you know, I also believe that it's 19 necessary for the Staff to reexamine all the relevant l0 20 issues concerning the steam generators, both for what l
the safe ty concerns are , and I think we have to study and i 21 7
22 make the finding that the current regulatory approach is l
71
- O 138 jO satisfactory.
Q Let me get at it differently:
I gather that --
3 O
A Let me take away the word "satisf actory. "
Whether or not it should be approved further, g Q This corrosion-related steam generator tube integrity phenomenon is considered generally an unresolved 7
safety issue; is that correct?
- O A yes.
Q And that is part of an ongoing Task A-3; is that correct?
11
'O-A Yes.
g Q To which you referred.
Now you have also taken a position, I believe,
.O 14 in the SER and in your . testimony that notwithstanding those ongoing efforts to which you have continually alluded, O g that to quote you on paragraph 7, page 2, of the affidavit, that Byron can be operated before resolution of the above, l 18
- O' 19 O
undue risk to the health and safety of the public.
A That's correct.
,O 22 O Why don't you just, at the risk of redundancy, l
I
!Q
c3 72 a
139 O'
I then tell me why.
2 A It's not really on the basis of -- well, the ,
3 major basis for this finding is my assessment of operating O .
4 expe rience . The fact that surveillance requirements have 5 proven generally successful in preventing rupture ij (, occurrence s , though not absolutely. That where we have 7 had rupture occurrences , that these have not resulted in 8 unacceptable consequences. The results of the consequences
'O 9 have not been severe. That even if we were to have.a 10 tube rupture assumed to occur concurrently with the 11 design basis accident, the only Staf f studies -- the 0
12 conclusions of Staff studies that I'm aware of all 13 indicate that the results would not be unacceptable.
,g 14 That last item does not factor directly into 15 any safety evaluation which we customarily prepare. It 16 is something that I -- that provides me with some added I?> assurance.
17 18 0 Go ahead, if you want.
19 A That's enough for now.
O 2e Q By the way, Dr . Rajan, if you want to add 21 anything, go ahead at any time; not that I'm inviting a c, 22 response now.
%)
()
-n -w a -- , --- -
CJ 3 140
\
O I You indicated that as part of your answer to my 2 last question, you alluded to the steam generator tube g 3 ruptures that have occurred in the past; am I correct?
4 A I'm sorry?
5 0 In the past, there have been past instances.
O 6 I believe in response to questions by Mr. Jenkins, you 7 identified four instances in which there have been steam 8 generator tube ruptures on domestic reactors; is that C
9 correct?
10 A That's correct.
11 Q Did any of these ruptures result in impermissible g
12 rele as es to the environment, to your knowledge?
13 A To my knowledge, none of these rupture events
() g4 resulted in unacceptable releases to the public, or to the 15 environment.
16 Q By unacceptable, do you mean --
l'7 A To my knowledge, no requirements , 10 CFR 100 18 or otherwise, have been violated. The radiological 19 consequences of the rupture events which have occurred,
.O 20 have been evaluated in detail by the Staff, and have been 21 reported upon and documented.
{J 22 Q Has the Applicant, to your knowledge, done a O
,4 ,
C) 141 l(3 3 steam generator tube rupture accident analysis, as part of 2 its application?
3 A The answer to that is yes.
- O l 4 Q Was it done as a design basis accident?
g A The tube rupture event?
l <
ld, 6 Q Yes.
7 A It's my understanding that that is a design l
l 3 basis condition.
'CJ , Q Okay. I believe there were earlier questions 10 about the necessity,- to consider a steam generator tube 11 rupture coincident with some other significant design O
12 basis accident, such as a LOCA, and you ind'icated by --
13 well, I'm not sure what you -- ,
14 A That's not a design basis accident.
7 15 Q It's not presently then a Staff requirement 16 f r design basis accident, and you're not making any 40 37 recommendation today whether or not it should be; is that 18 correct?
g9 A No, I'm not. It's something under evaluation
(
i '3 20 by the Staff. ,
21 Q And it's under evaluation , I gather, as part 22 of the overall review of steam generator tube integrity 0
0
.i>
^'
l 142 P"
I problems generally; is that correct?
2 A Yes. But as this report that's been identified, 3 this February 1982 report, discusses, the analyses that I'm iO 4 aware of that have been performed to examine the 5 consequences of tube failure concurrent with an accident lJ 6 indicate that the consequences of a tube rupture, a single 7 tube rupture during accident conditions will not result 8 in unacceptable consequences , whether we're talking about 9 a LOCA or a main steam line break, or what-have-you.
10 0 The February 1982 document you are referring to 11 is the one that Mr. Gallo earlier marked as Deposition
.O 12 Exhibit l?
f i 13 A That's correct.
- (3 14 Q Let me talk about this mechanical tube vibration 15 problem for a moment. I guess I'll direct my comments,
! 16 then, to Dr. Rajan.
") 37 I wonder -- first of all, the Ginna accident 18 which was referred to earlier, I believe it was your l 19 testimony that that did not result from this flow-induced
- O I 20 problem; is that correct?
l 21 A (Witness Rajan) That is correct.
I
.(7 22 Q I wonder if you can distinguish for me the l
!O
/
+
16 C) 143~
C) I difference be tween the Model D-2, D-3, and Model D-4, 2 D-5 Westinghouse type steam generators from the standpoint 3 of both their susceptibility to tube vibration problems
.O 4 and, secondly, their amenability to corrective modifica-5 tions.
6 A The one thing in common with D-2 and D-3, on the
()
7 one hand, and D-4 and 5, is that both are preheat type a s te am generators . The difference is that in the D-2, D-3, C) , the flow is split upwards and downwards , as it emerges 10 from the feedwater nozzle.
11 In the D-4 and D-5, on the other hand, the flow n
12 is directed downwards, and this design is referred to as 13 the counterflow type , in which the flow is directed down-14 wards, and then it goes upwards again to a series of baffles.
1O 13 The f act that the flow does not impinge on the 16 tubes directly as it comes from the feedwater nozzle, in C 17 my judgment, somewhat reduces the possibility of flow-18 induced vibrations in the D-4, D-5 model. Although let me 19 emphasize that this whole area of preheat region is an O'
20 area of turbulence, and the first row of tubes has evidence 21 frcm data at KRSKO to experience unaccept'able flow-induced l
22 vibration.
lO l
l lO l
l l
i 17
,O 144
) So we do recognize that there is a problem, but 2 in my judgment, the magnitude of the problem is somewhat 3
less severe in D-4, D-5, as opposed to the D-2, D-3.
.O 4 0 Let's assume for a moment that contrary to 3
present expectation, the corrective modifications are not available and satisf actory at the time that Byron is ready 3 6 7
to operate.
8 In questioning by Mr. Gallo, I think you gave O , some alternative measures that were available.. I wonder 10 if y u can just briefly summarize what posture you think 11 y u'll find yourself in if that's not the case.
.O 12 A Well, at the end of the review of the options --
13 of the analyses and test results from Westinghouse , if 34 we do conclude that it has not been adequately demonstrated 15 that the flow-induced vibrations have been eliminated or 16 reduced to within acceptable limits, then at that point
- O 37 we will have, in my judgment, two options:
i One of them would be again based on the results 18 of the test data and operating plant experience at that l 19 20 point, we can limit the power at Byron to less than 100 21 percent. And until such time that a fix can be found.
22 The other option, of course, would be to delay,
,0 O
i
18 l O 145 0 3 delay th; s c.rt of the plant until a fix has been
-- acceptable fix has been found.
Q Okay. Both of these alternatives -- some kind 3
,0 of, power restriction and delay -- would you say it's fair
~
to characterize those as primarily imposing an economic burden on the Applicant, as distinct from representing a g
safety problem to the public?
Is my question unclear?
O A Our decisions would primarily be based on what the effect 'of the fix would have on the safety.
Q Okay. So clearly, then, the objective of
.O some kind of alternative plan, then, would be to ensure 12 public safety during some period of operation or defer operation if it was found that it could not be satisf actory?
A That's right.
Q So just to follow up, then, the absence of C final corrective modification at the time Byron may be g
prepared for power operation does not mean that there is not some alternative measure by which public safety can O be assured pending some ultimate corrective modification or fix; is that not correct?
A Yes.
O
19 M.
146
.O , MR. GOLDBERG: Hold on for one second.
2 (Pause.)
BY MR. GOLDBERG:
3
'O 4
Q Mr. Murphy, let me ask you one or two particu-lar questions about some areas in which you were examined g by Mr. Gallo.
One of the measures identified in your affidavit 7
on page 2, paragraph 5, to minimize the onset of steam O ,
generator tube integrity problems, is utilization of all v latile secondary water treatment; correct?
to g A (Witness Murphy nodding.)
O O Does all volatile treatment chemistry -- excuse me, hemistry control add phosphates to the steam generator?
U A (Witness Murphy) No.
g Q Does it then prevent phosphate wastage on new plants such as Byron?
!O A Well, you don' t have the ingredients for phosphate
, 17 wastage, as I understand, g Q So is it your belief and is it your testimony O
7, today that these -- I'm sorry, strike that.
g Okay, a second component of the second measure i
that you identify as designed to minimize steam generator lO l
22
.O
,m3 - -- r - - , - - + .
10
() .
147 C) g tube problems, is improved controls and monitoring of 2 secondary water chemistry; is that correct?
3 A Yes.
O 4 0 Is it your belief that improved chemistry 3
controls will reduce corrosion to at least controllable 6
levels?
7 A Yes. Yes.
g Q Okay, how does both the all volatile treatment
- (3 , chemistry control and other improved chemistry measures 10 protect the steam generator from corrosion?
11 A The chemistry AVT affects the -- AVT is a
'O 12 method for treating the secondary water, for scavenging the 13 oxygen, for gauge control, and its function is to minimize 14 corrosion problems relating to the secondary coolant.
4) 15 I cannot go into it any deeper than that, 16 because quite simply corrosion is not my specialty.
O 37 The dynamics.
gg Q Okay , Mr . Murphy . As I understood your answer 19 to questions by Mr. Gallo, you referred us to the statement O made in Appendix C, Section A-3, containing the Staff 20 21 review of the steam generator tube integrity problem 22 which you adopted as a principal response to questions l
l O
.-- ___ . _ _ _ = . _ __ .
]l 148 1 concerning, I guess, the hazard of steam generator tube 2 f ailure problems coincident with other accidents; am I 3 correct?
!O l 4 A Yes.
i 3 Q And that statement -- and let me just read you lO 6 the statement, I realize you may not have it in front of 7 you. The statement is that the primary concern is the 8 capability of steam generator tubes to maintain their I) , integrity during normal operation and postulated accident 10 conditions. Is that correct?
j gg A That's right. And therefore it pertains to iO i
12 the hazard presented by various degradation mechanisms 13 like stress corrosion cracking and so forth.
14 Q And if we look to Contention 9-C, it cites
'<3 15 the steam generator tube integrity problem stemming from 16 corrosion cracking and denting and fatigue, and goes on C) to submit that this constitutes a hazard, both during 37 33 the normal operation and under accident conditions.
3, A You're asking me to respond to that, and the O
20 response is indicated in item, I believe, 5 of the affidavit 21 -- of the testimony, which states that we have implemented i
j7 22 -- requirements have been established to keep these l
l0 l
2 C) 149 O 1 degradation problems from becoming a hazard to the public.
2 Q Okay. Realizing the limitations, perhaps, on 3 your knowledge and experience in this area -- and, Dr.
O 4 Rajan, if you have anything to add, I'm just going to 5 kind of ask you the question that maybe Mr. Gallo stopped g 6 short of, but in the context of that particular contention, 7 do you have an opinion about what kind of incremental 3 risk is posed by the addition of the steam generator tube O 9 rupture to a design basis accident?
u) Let's maybe stick with one of the worst, a 11 design basis LOCA.
o 12 A I'm not sure I understand the question. You
, 13 are saying -- does the question assume a tube rupture l
14 concurrent with an accident?
() .
15 0 Assure that you have a design basis accident, 16 let's say it's design basis LOCA, one of the most severe.
l I) 17 What incremental contribution would a steam generator l
l s tube rupture have to the severity or public risk stemming 19 from that kind of an accident?
O
! 20 MR. GALLO: I'm going to object to the question.
21 The witness has already tcstified he is not an expert in
,, 22 daat area. I just make that for the record. Go ahead.
!S 1
t
3 150 1
MR. GOLDBERG: That's your objection.
2- WITNESS MURPHY: Well, I think its safe to 3 say that tube rupture during a design basis accident
-O 4 aggravates the severity of the accident. I don' t think 5 there is any doubt about that.
() 6 The question is whether -- how much does it 7 aggravate the accident. The Staff is -- other organizations 8 are pursuing studies in this area. The document that has 9 been identified, the internal -- I guess it's classified 10 as an internal NRC document, addresses the preliminary 11 findings, or the findings of those analyses.
O 12 BY MR. GOLDBERG:
i 13 Q Unless you have anything to add, Dr. Rajan --
- (3 14 A (Witness Rajan) The only thing I wouId say, 15 that these studies have been pursued and we are aware of 16 some of the results of these studies.
Cd9 17 My understanding is that during a loss-of-coolant 18 accident, several -- a rupture of several tubes can be l 19 tolerated.
lO 20 Now, the exact number I'm not aware of.
21 A (Witness Murphy) Well, the exact number, 22 according to the existing analyses, is 1300 gpm. This is
-(3 i
i l
O
- 34 C) 151 C) 3 more than the le akage that you would expect during a LOCA, 2 even if you assumed the tube completely double-ended during 3 a LOCA, which is an unlikely f ailure mechanism. But this g
4 amount of leakage considerably exceeds the expected leakage 3 in the amount of the double-ended failure of a tebe during e3 6 LOCA conditions.
v 7 0 You are acquainted with existing Staff analyses 3 that document that point?
O 9 A Like I say, I have testified here that I am to acquainted and have read the results and conclusions, and 11 am f amiliar with the results and conclusions of the analysis, 12 MR. GOLDBERG: Okay. I have no further questions, T.9 g3 MR. JENKINS: I have three, very brief, that 14 were raised, but before we go into those questions, I would j) 15 like to note for the record Mr. Connell's presence here, 16 which I do not believe was noted in the record.
O 17 MR. GALLO: Who? Mr. Connor?
18 MR. JENKINS: Mr. Connell? Is that your name?
g9 MR. GALLO: Connor.
m
20 MR. CONNOR: Connor.
21 MR. JENKINS: Connor. I'm sorry.
22 MR. GALLO: I noted that when he entered the lO lO l
-.-. _ _ - _ -_ - _ - -. . - - .. . . = . . - - - - - . - . . . . _ . _ - . - - . . .- - - --
M i,
152
+
E
- O room. You'll find that in the transcript, I'm sure.
1 t
- 2 MR. JENKINS
- I'm sorry.
} 3 MR. GALLO: He's from Westinghouse, in any
- O
- 4 event.
i i
! 5 MR. JENKINS: All right. Very good.
I
!O 6 i
7 4
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15 3 O
1 R E - E X A f4 I N A T I O N 2 BY MR. JENKINS:
3 O The first question, you have noted that 4 the inspection frequency typically is 12 to 24 months.
s Why is McGuire inspected more of ten, every two months?
O 6 A (Witness Murphy) Because we are concerned 7 that with the vibration, or potential for vibration, daat a they have, that if we -- because of uncertainties that O
9 exist regarding the exact wear rates that take place at to 50 percent power, 75 percent power, what-have-you, we feel 11 that much more frequent inspections like are being performed n
v 12 are needed to ensure that the wear will not proceed beyond 13 acceptable limits , before these tubes are reinspected.
O 14 Q But I guess I don't understand; if you would is normally inspect every 12 to 24 months , why are you 16 inspecting in this circumstance every two months?
O 17 A (Witness Raj an) The rate of flow-induced 18 vibrations in certain groups is rather rapid, and if we 19 allow the plant to operate for the 12-month interval, which 20 it is normally scheduled for, it is quite likely that 21 several tubes would reach their plugging limit prior to
() 22 that, completion.of that operation period, and may begin O
.?j7 134 O
1 to leak, perhaps.
2 Q Are you saying that McGuire has some unique 3 problems?
g 4 A That's correct.
I s A (Witness Murphy) All Model Ds have unique l (3 6 Problems.
7 A (Witness Rajan) All Model Ds have unique i
3 problems.
O 9 Q You averted to U-bend cracking. Now that is <
10 something that I have never been f amiliar with before .
11 Would you just briefly describe what that is?
.O 12 A (Witness Murphy) It's described in exc'ruciating 13 detail in NUREG 0886. There are two U-bend cracking O 14 phenomena.
15 One is caused by -- is a direct result of l
l 16 denting in the upper support plate.
17 The other is not related to denting.
t 18 The denting-related U-bend cracking led to the l 19 Surrey rupture in 1976. The phenomenon is believed to be
!O r
- 20 well understood now. There are a lot of tell-tale --
l l 21 there are some tell-tale indicators that a 'p lant may be l
l
!(3 22 approaching the condition where one must be concerned about l
O .
i
93 155 0
1 such a failure, and suf ficient surveillance -- steam 2 generators can be surveilled to see that these conditions O 3 d n' t exist. or if they do exist, you can take preventive 4 measures.
5 The nondenting-related cracks have resulted in 6 a number of small leaks at some units. I'm not sure to y tell you just where to go. It appears to be related to 8 -- the occurrence of the cracks is related in part to the O
9 residual stresses in the U-bend as a result of the bending 10 Process.
O 11 0 What class of incident was given to the Ginna 12 accident? As I understand it, incidents are classified 13 somehow in a numerical basis.
14 A I can ' t answer that question.
15 A (Witness Rajan) I'm not aware of it.
16 MR. CEESNUT: We have two or three dif ferent O
17 types -- at least two different kinds of classifications.
18 I don' t know if you are f amiliar with the numbering system l' # --
O 20 MR. BUNCH: Yes.
21 MR. GOLDBERG: Mr. Chesnut is not sworn.
O 22 MR. JENKINS: But these gentlemen said they O
M 156 I don't know.
2 MR. JENKINS: Mr. Gallo?
,s 3 v
e 4
5 nv 6 7
8 n
us 9
10 7, 11
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12 -----
13 O 14 15 16 n
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(, - 19 r) 20 21 0 22 c.
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.CO- 157 O j gg;ggAMggAggqg 2 BY MR. GALLO:
3 Q Has the Staff, Dr. Raj an , conducted any analysis
'O 4 to determine the likelihood of a gross tube failure 5 occurring the course of a design basis accident?
() 6 A (Witness Rajan) Are you talking in terms of a 7 probability number now?
g Q Yes.
O 9 A No, sir, I have not.
to Q Are you aware of anyone on the Staff that might I 11 have performed such analysis?
C) i 12 A I am not aware of anyone.
13 Q How about you, Mr. Murphy?
14 A (Witness Murphy) The analyses which have been
()
15 done have determined the probability of tube ruptures 16 during accidents, assuming that you have -- that you have
() g7 degradation in the steam generator.
18 These analyses were done for the purposes of 19 evaluating various sampling plans versus the tolerable O
20 number of tube failures during an accident.
21 Q Is it possible -- I'm sorry, go ahead.
22 A Insofar as I know, there has been no risk O
15 8 C) assessment evaluating the probability of a tube rupture at 2 any plant if it had a main steam line break.
3 Q Do either of you know whether or not it is
.O 4 possible to perform such analysis? Just offhand,. the 3 uncertainties seem to be so great that it would seem to
() 6 be difficult to perform an analysis. Do either one of you 7 have an opinion on that point?
8 A (Witness Rajan) I would think that it is C3 -possible to make an analysis although the uncertainties 9
10 about the accuracy of such an analysis would be rather high.
11 A (Witness Murphy) Jai and I have been recently O
12 discussing such a risk assessment, but this is -- to our 13 knowledge, there is no -- there has been no -- well, maybe g 14 I'd better -- I'm not aware of a specific risk assessment 15 regarding the potential for steam generator tube ruptures l 16 during an accident.
C) 37 Q Okay. Two short questions :
l 18 You have been referring here today to design 19 basis accidents and the occurrence of a gross steam O
20 generator tube failure during the course of such an 21 accident. What ones are we talking about? We mentioned a l
design basis loss-of-coolant accident which I guess for a
}) 22 l
I l
P
12 0 159
- (3 1 PWR would be the double-ended guillotine break. Is that 2 correct?
3 A Yes.
- O 4 Q And the main steam line break, is that another 5 design basis accident that we would be concerned with?
O ' ^ ("it"*** **d"") Y***
7 A (Witness Murphy) Yes.
8 Q And I think, Dr. Raj an, you mentioned an C3 9 earthquake situation. Are you talking about SSE?
10 A (Witness Rajan) Yes, that would be a loss-of-11 coolant accident in conjunction with a safe shutdown
'O 12 e arthquake , in conjunction with an SSE.
13 Q How about a break to a feedwater line, or is that
- g 14 not a PWR problem?
15 A It is a PWR problem.
16 Q Would that be another design basis accident?
- <3 17 A That would be another design basis accident.
18 Q Are there any others?
19 A Well, for the steam generator, I would say that O
I 20 these are the main design basis accidents, or the bounding 21 design basis accidents.
22 Q Just one last question:
t O_
O
13 O 160 C) 1 Out of curiosity, if I wanted to go find a list 2 of these accidents, where would I find them? Are they 3 in a Standard Review Plan some place, identified as design O
4 basis accidents, or a reg guide, or regulation?
^
5 A (Witness Murphy) The -- maybe you can correct g 6 me if I'm wrong, Jai, but the regulatory guide provides 7 guidance.
8 For example, Regulatory Guide 1.121 provides O 9 guidance on this sort of --
10 A (Witness Rajan) That lists these accidents.
11 MR. GALLO: That's all I have. Thank you.
.O 12 13 O
15 16 O 17 18 19 O
20 21 22 O
O 1
M 161 O 1 RE-EXAMINATI ON 2 BY MR. GOLDBERG: .
l l
j 3 Q Dr. Rajan, the accidents you have just !
- O 4 identified for Mr. Gallo and called bounding, those are 5 all fairly low probability accidents, aren't they?
6 A (Witness Rajan) They are.
<3 7 Q Coupled with a steam generator tube rupture ,
a they would be even lower; is that correct?
33 9 A That's right. That's absolutely right.
10 MR. GOLDBERG: Okay. I have no further questions.
i 11 (Whereupon, at 1:55 p.m., the deposition
- O 12 was adjourned.)
13 l
() EMMETT MURPHY 15 16 O 17 JAI RAJAN
, 18 19 o 20 21 C) .
?
95 162 O
g STATE OF MARYLAND :
2 COUNTY OF MONTGOMERY :
O 3 I hereby certify that the above-named witnesses ,
4 EMMETT MURPHY and JAI RAJAN, personally appeared before 5
me and signed and subscribed to this deposition.
6 7
8 Notary Public,
,, Montgomery County, Maryland
's .
9 10 My Commission Expires:
O 11 12 13 O 14 i I 15 ll 16 lQ
( l'7 l
18 g 19 20 21 1 ,,
'd 22 l
s
\
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16 O 163 C) 1 STATE OF MARYLAND :
2 COUNTY OF MONTGOMERY :
3
() I, ANN RILEY, a Notary Public in and for 4
Montgomery County, Maryland, do hereby certify that I S
reported in Stenomask the deposition of EMMETT MURPHY O 6 and JAI RAJAN, the witnesses herein.
7 3 I further certify that the foregoing (3 , pages are a true and correct transcription of the 10 testimony given.
11 O I further certify that the deposition was 12 transcribed either by me or under my personal supervision.
13 g4 I further certify that I have no interest,
)
15 financial or otherwise, in the outcome of this proceeding.
16 Given under my hand and seal of office, this O 37 the 9th day of July, 1982.
18 Q' x 20 Ann Riley /
Notary Public, 21 Montgomery County, Maryland 22 My Commission Expires:
July 1, 1986 0
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S M G NERAE R STAWS REPO.RT O
O FEBRUARY 1982 O
U.S. NUCLEAR REGULATORY COP. MISSION l
O Q
O 1
0.
l
!. Problem Defintion 3 .
A. Summary of Tube Degradation Degradation of steam generators (SG) manufactured by each of the three pressuri;:ed water reactor (PWR) vendors has resulted due to a ccnbination of steam generator mechanical design, themal hydraulics, O materials selection, fabrication techniques, and secondary system design and operation. In the early and mid-1970s, Westinghouse (W) 5.G. experienced caustic stress corrosion cracking, and W and CombustierThese modesEngineering of degradation(CE) 5.G.s experienced tube thinning (wastage).
were due to difficulties encountered with pnosphate secondary. water O c hemi s try. Because of these difficulties, most W and all CE plants Al tnougn converted to an all volatile ( AVT) secondary water treat: ent.
this conversion greatly reduced the occurence of stress corrosion cracking and wastage, other degradation modes including denting (defomation of
- ne S.G. tubes due to corrosion of the caroon steel sup: ort plates) g, began to occur.
V Babcock and Wilcox 5.G., which have a significantly differen:
design fr:n W or CE and nave operated exclusively with AVT water chemistry, had relatively good operating experience in their early years TheOf operation.
crincipai Nevertheless, they have experienced numer0us tube leaks.
modes cf degracation in S&W units have been fatigue crack crew:h, confine:
n
" primarily to limited sets of tubes located on the cpen insiec ier lane, and more recently erosion-corrcsien and primary side intergranular attack.
To date, many different foms of steam generator degradation have been identified including: stress corrosion cracking, wastage, g intergranular attack, denting, erosion-corrosicn, fatigue cracking, j pitting, fretting, and suoport plate degradation. Jna or -c ra'~ma e *ase ~
4 foms of decradation have affected at leas 4 0 ~ = -" o'20 e SM sns w - crs, uce cluccinc. recair. ce rec! ace e-:
.pesu l ted in extens ivaxecentiy, toreign Westir.; house S.G.s of ne same experlerced tube wear associa:ed wi:n #10w induced vibra:icn cue :o a 0
i new in sgral preneater design. References 1, 2, and 3 present de:ai'e:
discussions or dcmestic S.G. ocerating experience.
1 The econcnic impact of steam generator degradation has been j 4:roxica tel v ? : of non-refuelina If significan:. Thecutace cos: of tire sucnhas ou: been ages in 0 a t ributed to stea- cenerator cecraca:1on. However, cernaps :ne tems of replacement power alone is very high.
g rea tes: financial costs incurred :o date are those associated wi:n Replacement of the Surry Uni 1 and Uni-
! s team generator replacment. cwer.
2 S.G.s cos: ac;roximately cS200 million, including cos: of makeuc 5.G.s. curren-Iv in crocress, will c s:
Feolacement of the Turkev o in
>@ an es t'ma tec $460 m1111on. NRC staff :1=e invol vec *1:n :nese activi-ies is estimatec a: 6000 manneurs for TurkeyLess Point (whicn included time for radical opera:icns also a nearing) and 3000 manhours for Surry.
incur significant costs. Recent tube sleeving c:erationsa3:cost San Cnofre i
1 involved repair of aoproximately 7000 degradec tuces a: of 570 I
l million. Procosed sleevinc of 3000 tubes at R.E. Ginna has an estimated
!4 cos t of $20 mil lion.
ll S. Safety Significance I!i . - .
The safety sign 1:1cance +., 3.g.
. . ube integrity can be divided 0
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ed g U'into three categories:
e tube failures under nomal operating conditions;
{ tube failures concurrent with postulated accident conditions; and personnel y exposure associated with S.G. inservice inspection (ISI), repair, and replacement.
4 e
ii The majority of the S.G. tube failures that have occurred under nomal operating conditions were small stable leaks sometimes e requiring plant shutdown, inspection, and corrective actions, but for
{4 ratemost the limit)part thatsmall enougn (e.g., below tecnnical specification leak operations continued until a scheduled shutdown.
( Mewevar, fou r sienificant S.G. tube ructures have occurred in domestic PWRs since 1975.
These events occurred on Feoruary 25, 1975, at Point
{
Seach Unit 1, September 15, 1976, at Surry Unit 2, Oc ober 2,1979, at Prairie Island Unit 1 and on January 25, 1982, at R. E. Ginna. The first three of these events were evaluated in NUREG-0651, " Evaluation cf Steam Generator Tube Rupture Events." The report includes an evaluation i of system response, operator action, and radiological consequences during the three events.
rangsd from about SG gpm toThe 390leak gpm.rateThe associated with these events j
j that no si gnificant offsite doses or systemsconclusion cf :ne report is inacequacies occurrec
- during tne tube rupture events analyzed. However, the cotential fe r noce sienificant consecuences was recocnized and a number of procedural 9)
' recmmendations were made to correct ne ceficiencies that were noted.
, The present disposition of each of the recommendations is ciscussed in a recent memo to Cmmissioner Bradford from W. Dircks (Pef. 4). The i present design basis for assuring that plants are acceptably protected against a single S.G.
S.G. tube tube.rupture events is a postulated double-ended rupture of g leak rate for a spectrum of rupture gemetries in a single tube an spectrum of smaller leaks in multiple tubes within a single S.G. The it4 n ave cc r ve r teenconsecuences cf multiole tube failures, in excess of the design ba rico rous i v s tudl ec . Racic cegraca:1on ce: ween inspecti:ns of a large nt=cer of tuces could create the cotential for multicle tuce f_ailures in the ever.: of a plant transient or failure of a single tute o anc :ne acc:.mpanying jet impingement and tube whip coulo cause failure of additional tubes.
involvinc multicle ecut orentFurthemore, the cotential for c:molicatin; circu s:ance
- ne Ginna incident and coss1ble failures sucn as :ne stuc< ocen FORY curanc a-aava ne* 9ea avaluarec. steam cucole fema:1on in tne or1 mary 4' 3 ve
<c s. in :nis event. L4n o t n e r concern i s ruc tu res in mu l ticle O
- breuen uniess :he clant can te racic1v decressur1:ec anc onto Res1cual Hea t 9emoval , there is the cotential to continuously lose emercency core cooline water outsice of containment. The acove concerns are ceing accressec as par: of :ne TN Ac:1on Plan.
i Item I.C.1 in :ne TMI failures (sucn Action Plan ocen as a stuck addresses saferv S.G. tube failures couclec wi:n c:ner relie# valve in :ne seconcary O system), ruptures of multiple tuces, anc simultaneous ruptures in multiple S.G.s. The ;:urpose of :nis effort is no: to expanc the plant design basis but to assure that operator emergency procedures provide proper guidance Althougn rien-e for safely controlling the plant during these types of events.
=a=19eae nava a- aaa ' m nv cf ne scenarios costula:ed above
~rala% ISI, leak rate limi:s, anc :ute plugging recuirements O are intenced summa ry, :h to guard agains t such cccurrences (See Section II). In condi tions hav. nsequences of 5.G. tube ruotures uns nomal operating been smalli hoever, such events challence to olan: coera:ces anc sa *etv sys tems. an cresent a sienifican:
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O Durino costulated accident conditiens. such as main steam line break (MSLB), feedwa ter line break, er LCCA, the S.G. tubes are sub4ect to increased cressure differentials and cessible cressure waves (e.g. ,
subcooled cectr.1pression pnenmena) anc vicrationai icacines u., l_T h e s e O loads increase the cotential for failure of decradec s.G. tuees whicn c'ouic exacerca:e ne accicent secuence. In :ne event of MSL3, fallec S.G. :ubes would crovice a leakage catn frcm :ne primary to seconcary system ano. severai co tential ieak ca:ns for racicactivity to :ne envi ronmen:
knuld then exts: In tne event of a LOCA, One core reficoc rate could
-#'be retarded by steam bindinc. This phencmenen is associatec with a colc Q leg break, in willen reflood cf the core requires displacing steam generated in :ne core through the hot leg, the affect d " eam generator, and cut of the cold leg break. S.G. tube failures woulc create a secondarv to o rka rv leak cath which aceravates the stearNndinc effect and coulc 1 c *nef'ective refloodino of "e core. Analytical anc experimental evaluations of :nis phena11enon are contained in References 4 and S.
O Large MSL3s and LOCAs are considered extremely low prcbability events, but are postulated as bounding concitions. More realistic events mign:
incluce small and intemediate size MSLSs or LOCAs. A1:heugh :nese postulated accidents pose a less severe challenge to S.G. tube integrity, tube rupture (s) leading to or following sucn events could have sericus
-o consecuences. This is particularly true if fuel damage has occurred as in tne case of Three Mile Island.
The final area' of concern is the radiation ex osure of cersonnel
g^ involved in 5.G. inspection, repair, and replacement. Reference 3 presents a summary cf data on S.G. related personnel exposure for selectec O plan:s feca 197c to 1980. In recen: years, as mucn as 25: cf scme plants annual occupational exposure has resulted frem reutine S.G.
inspection and maintenance and as high as 63 for S.G. replacement.
Recent tube sleeving cperations a: San Cncfre incurred 3500 man rem exposure and similar c;erations are planned fer 'other plants.
O II. Regulatory Aaproach The NRC approach to assuring S.G. tube integrity under all operating conditions is based on inse-vice inspection (:SI), primary ::
secencary leakage rate limits, and preven:ive tube plugging recuiremen:s.
Guicance for perfoming :S! is provided in R.G.1.83, " Inservice Inspection
,0 of S.G. Tubes," anc plan: technical s;eci'ica icns include requirements
. for IS:. Typical plant s;ecifica:icns recuire periccic inspections of
_ 3LoMe S.G. tu:es in :ne plant and augmented IS: in the event tube cegracation is cetectec. Repuired frecuency of inspecticn is generally flexible enougn to allow inspectices :: :e :erfemec c:ncurren wi :n refueling cutages. Certain incicents such as tune leakage recuire
- O unscneculed IS:s. Furtnemere, many plan:s witn extensive cegraca:icn problems have licensing an.endments im:csing higher frequency and larger si:e inspections. The :S: requirments wre developed largely :nrcugh a cmcination cf engineering jucgement and operating experience. More rigorcus statistically based ISI prcgrams have been developed as part of Unresol ved Safety Issues -3, A 4, and A-5 (see Section v). The purpose O of the required IS:s is Oc detemine df :ube degracation is occurring in the S.G. , assess :ne rate of tube ".agradaticn based on results of successive inspections, anc identify : nose tuces requiring plugging or repair.
O
_d-O p*a rv to secondarv leak rate limits 'are an extrenelv imoortant emui ra-as for ensurinc safe S.G. coeration. Sa e foms of tuce cecracanen l Eave been cbserved to ceorade tubes beyond ne prescribed plugging limit Tecnnicai speci fica tion primary durine the intervai re ween inscections.
to seccndary leak rate limits requiring shutdown, ISI, and corrective n* actions provide protection against unacce; table levels of degradation between inspections. Many serious conditions of tube degradation have been detected by monitoring of primary to secondary leakage and subsequen inspection. Primary to secondary leak rate limits exist in each plant's technical specifications. The bases for these limits are twofold.
p#
Fi rs t, the leak rate limit ensures that tne calculated dosage contribution fr:rn tube leakage will be limited to a small fraction Second,of :ne allowable the leak limits in the event of a S.G. tube rupture or MSLB.
rate limit is' intended to correspond to a defect si:e that would not be expected to result in tube rupture under nomal or postula:ed accident condi:t Finally, degradation lini:s for tube piagging exist in the O
plant Technical S:ecifications. Criteria for establisning the tube plugging limits are presented in R.G.1.121, "Sasis for Plugging Degracec Pressurizec Water Reactor S:eam Genera::r Tubes."
These criteria recuire that the plugging limit include margins for eddy current testing error and continued degradation between inspections. Thus, it is important to O have a good estimate of the rate of degradation based c'1 successive !S!
results and an understanding of the degradation pnenomena.
The primary focus of the current NRC cnilosconv is directed at
-aintainino crima rv svstem intecrity. Th1s 1s acccmplisnec primarily througn the recuirements :escribec above for ISI, leak rate monitoring, O and tube clugging. In a sense, it is directec a: treatine the sym : oms anc not the cause of 5.G. cecraca:1cn, wnicn lies Orlma ri' y in seconcary svsten ces1cn anc coerarons. In1s pnliosopny nas :een ceaatec ex:ensively, ou: the current pos1:1on regards eliminatine me Orcolem at its scu-ce as an indus trv rescons ibil i ty.
O III. Current Corrective Actions an e#fec-ive solution to 5.G. tube decracation :r:blems aculd recuire major chances in 5.G. mecnanical cesien , :ne -na;-nveraui l es ,
ma er,ais selec en. faer,can en tecnnicues. anc c"ances in :ne sec:cctcg D s vs te- cesien and ocera-ion. Elimina: ion of S.G. cegraca: ion recuires a "a ~'
I systems approacn integrating all of nese considerations. ha
< w1 a ~ -act va acci m . This is particularly true for : nose pian:s i
wnicn nave signiricant cperating time and have ex;erienced S.G. degracatier.
Design changes in operating S.G.s :nat woulo be necessary :: eliminate For exam:le, :ube ::
cegradation proclems are virtually impossible.
9 pbesheet crevices already contaminated with corrosive environ ents are virtually impossible to clean, carton steel succor: plates canne: :e replaced wi tn more corrosien resistant materials, and resicual faorica: ion s tresses cannot be removed. Thus, corrective actions may prolong S.G.
life, but tube decraca tion is exeected to con:inue in coe-atine :1 ants _.
Once ne secondary system is contaminated by an aggressive env1ror. men; o i t is di"icul t to reverse :ne adverse affects. For example, caustic stress corrosion cracking anc aastage, cue :o residual :nos;na:e water j chenistry conditions, still continue in some plants long after conversion to A'/7 wa ter ch eni s try .
Seve ral corrective actions, newever, have been proposed and b
5-
~O.
aro in use. These fixes include such actions as tube sleeving, sludge lancing, soaking and flushing, reduced operating temperatures to slow corrosion, boric acid injection to arrest denting, support plate modifications to retard denting, S.G. replacement, and improvements in secondary system design and operation. Secondary system improvements include O pr:ript correction of condenser in-leakage, condenser retubing, removal of copper based alloys frcr1 the secondary system, and addition of demineralizi systems. An industry constituted secondary water chemistry guidelines cer:mittee, under chaimanship of EPRI, is developing generic chemistry limits and operating guidelines. NRR has been in contact with this ccrimittee for the past year and will review a copy of tne draf t reports o . prior to issue. Chemical cleaning has also been proposed but has not been implemented due to uncertainties regarding its longer-tem affect on S.G. integrity. Industry efforts are currently underway to eliminate these uncertainties and chemical cleaning may become a viable option in ene near future. These fixes have met with varvino decrees' of success, Curremore, snort :em soluticas to one v nana n# % is a canacea.
'o Conversion from phospnate to AVT horcoiemcav create otner crcolems.
water chemistry, wnten minimizec wastage and stress corrosion cracking but was followed by denting, is a case in point.
Fira11v it should be noted that the maiority of the clants C [under review for coeratino licenses nave S.G,s of sim11ar cesien to tnose curren:1v in ooera: Ton, so tnat ne cotential for S.G. tuce decradation exists in these olants as wil.
IV. NRC, Industry, and Foreign Research and Cevelopment Activities to NRC's steam generator research prcgram accresses improved eddy current inspection tecnnicues for steam generator tubing, stre.ss corrosion cracking of steam generator tubing and evaluation cf tube in:egrity.
The objective of the eddy-current program is to upgrade and improve eddy-current inspection probes, technioues and associated instrumenta for inservice inspection of steam generator tubing to improve the abili:y to identify and cnaracteri:e tube defects. Specific objectives incluce improving defect detection and cnaracteri:atien as affected by tube
!l diameter anc tnickness variations, tube centing, probe wobble, tubeshee; g and tube support interference, anc defect location and type.
The stress corrosion cracking program is develocing data and models whicn will te usec to predic: ne stress corrosion cracking The initiation and service life of Inconel o00 steam generator tubing.
p .esting program incluces variables wnien influence stress corrosion cracking sucn as temperature, stress, strain anc strain ra e, re allurgical structures and processing, and ingredients in :ne crimary and seconcary c ool a nt.
A steam generator, with service induced cegracation will be
!q ,
used for the validation of :ne accuracy and confidence limits of nondestruct inspection instrumentation anc technicues; burst and collapse tests on field degraded tubes to validate tube integrity models; and for developing data for validation of previously develocec stress corrosion cracking precictive mecels, cnemical cleaning and decontamination, dose-rate reduction and secondary side characteri:ation. In addition, s:atistically gi based sa= cling mdels for inservice inspection programs will be confimed and/or improved utilizi'ng :ne firs: ever confimec data base.
O There are many ongoing programs addressing 5.G. issues at EPRI, most of which are sponsored by the S.G. Owner's Group, and the rest by EPRI i tsel f. The programs address the following areas: (1) chemistry and corrosion, (2) materials selection and testing, (3) ther. al O hydraulic and structural testing and analysis, and (4) nondestructive examination (NDE). Efforts in the chemistry and corr 0sien area are -
directed at examining the causes of corrosion related degradation such as denting, intergranular attack, and stress corrosien cracking, and identifying potential fixes such as alternative secondary water chemistry trea=ents. Materials selection and testing efforts are directed at O characteriz.ing and evaluating the suitability cf alternative tubing and S.G. materials. This includes consideration of new hea: trea=ents for tubing and compatability of S.G. tubing with structural materials.
Testing and analysis in themal hydraulics and structures is directed at secencary side S.G. design and perfomance and their effect on S.G. tube O integrity. The EPRI nondestructive examination pregrams focus on development of improved inspecticn techniques. These technicues include multiple frequency / ultiparameter eddy curren testing, at:ccatic edcy current signal analysis, profilcmetry for cuantifying den: ecnfiguration anc strain levels in dented tubes, and metheds for evaluating :ne condition of :ne tube support plates. In additien, EFRI has established the NDE O Center in sharlotte, NC, dedica:ed :c previding seed NDE techniques, anc effectively. transferring research anc development results to :ne indus:ry.
Research and develo:: ment activities underway on steam generators outside -he US A are ceine funcec at nion ieveis in severai countries.
"a haea are conducting a very large pregram w1:n emanasis on :nermal/
O hycraulics, and also on water eneM s:ry and tube :esting. To da:e, we have receivec little infomation on the progress or results of :neir p rec rams . De rench have work uncerway on addy current NDE, crevice chei-istry, anc decontamination. There is work uncersey in Swecer on water enemi stry. The Ge-ans have work underway in eddy curren: NDE, and at GU on crimary side cecontamina:icn anc secondary sice cleaning; O however, Geman s:eam genera:ces are tuced wi:n Incelley 300 sc much of
- neir research is less relevant to curs. Finally, :He Italians nave uncer ay a large program anich wii' allcw :nem to make new cesigns :c avoic current and possible future problems.
V. Long Tem Approacn O
A. Unr esolved Safety Issues A-3, A-a, and A-5 Regarcing Steam Generater Tube Integrity In .LI.L the NRC estaclisned Unresolvec Safety Issues A-3, A-g 4, and A-5 (USI) regarding degrada:icn in W, CE, and S&W steam genera:crs, respectively. A draf t recort, NUREG-0844, presenting the precosec NRC staff resolution of :nese generic safety issues nas been preparec anc is currently being reviewed cy NRR management prior to transmittal to :ne Ccmmi::ee for Review of Generic Recuirements and :ne Co mission and publication for public cxcent. The report integrates technical stucies O in the areas of systems analyses, inservice ins;:ection (ISI), and tube integrity to establisn improved criteria for ensuring adequate tube integrity and safe steam generator cperation under all conditions.
O
e .
In the systens analyses portion of the report, the consequences of steam generator tube failures during nomal cperation and postulated loss-of-coolant and main steam line break accidents are evaluated. ' The
- evaluation considers predicted fuel behavior, emergency core cooling systen perfomance, radiological consequences, and containment response.
The results of the syste"s analyses lead to proposed criteria for establishinc ~
a tolerable level of steam generator leakage during postulated accidents.
ISI techniques are then evaluated, and statistically based ISI precrams presented which, if inclemented, would provide adcitional assurance tna:
no more than the tolerable level of tube leakage, defined by the systems analyses, would occur during nomal or postulated accident conditions.
In the tube integrity portion of the report, the behavior of degraced tubes during nomal and postulated accident conditions and tube
, plugging criteria are evaluated. Proposed changes in operatinc ;rocedures and design changes to minimize tube degradation are also ident fied.
Implementation of the proposed requirements and criteria developed in the program for resolution of the USI are net expectec ::
totally eliminate S.G. degradaticn. The intenc of the progesec recuirements
^
~; is to estaclisn a logical approach to evaluating s:eam genera:ce ube integrity and ensuring safe steam genera:or operation. The draft NUREG-08a4 recmmends criteria and recuirenents : hat can be used to evaluate current and future degradation programs yi steam generators. The establisnmen:
of maximum allcwable steam generat:r tube leak' rates during postulated '
accicent condi ti ns and associated : lerable number of defective tubes D is a major contributien to the evaluation cf steam cenerator tube degrada:icr p roblems . I provides cbjective criteria agains hich steam ' generator tube integrl:y can be evaluated. Similarly, ne development of sta. tis:ica!
ISI programs provices a rational, scientific basis that can be used to establisn and evaluate ISI requirements :aa will ensure :ne above criteria are satisfied. Results fem NRC S.G. researcn :r: grams are _
? expecced to lay the experimental basis for many of these criteria. '
In keeping wi th the NRC's current and pas: philoscchy crf:nis issue, :ne proposed regulatory recuirements developed in the dra't repo r: fccus on !S' programs a1d technicues and tube pluggin; criteria, The reina rv -es cons ibil i ty for a ttackine ne :: nbiem a: its scur c and ,,
J e'i-inau nc 5.G. cecraca:1cn is :ne incustrv's: Mcwever, spera! t :ne requi rements prcposec in NUREG-034t are in:enced :: pe rote incust.ry efforts in this area. .~cr examcle, one recuiremen; is to ensure :na all o:erating plants nave 'molementec anza or0ved secondary water chemigtry moni:Oring and control program. This is a require-en: in :ne mos recen version of the NRR standarc review. plan for Licensing of ney plants. In addi tion, this type of c-cgram has been imolemented at scme bu: not all operating plants. Unde :nis recuirement, i t i s ta.e incus try's rescensibili ty to .:aclish specific -ater cheMs:ry limits and effective monitoring tecnniques. This will ensure na: egen utility at least considers the imccr:ance of seccccary system aa ter cne-istry and cuts in the effort to cevelop a cmprecensive water cnedstry program. Simil a rl y ,
ISI requirements for condensers are orcrosed. 9ese recuirements will s
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hccefully reduce the frecuency of conde'nser in-leakage and encourage
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utilities to improve condenser performance. Use of noncopper based alloys when retubing condensers and feedwater heaters is also a requirement.
~ Additional requirements are preposed for plants in the preoperating license stage and many recommencations for operating and future plants c) are made. The intent of the croposed requirements as stated in the report is to leave primary responsibility for correcting the S.G. preblem in the hands of the incus try, to allow the indus try flexibility in addressing the issue, but at the same time, to strongly encourage precer indus try actions.
O B. Canprenensive NRC/ Industry prog ram The preceeding review has attenpted to summarize the status of the S.G. issue at this time. As indicated, the NRC has many engoing
' efforts to address this multifaceted problem. However, to cate, fein
() N C and industry coccerative efforts on this issue have no: ceen extensive.
This is due larcel y to :ne c1fferent focuses en :ne issue. NRC 1s
>J je. m orica ril y concerned wi tn recuirinc acecuate ISI anc corrective actions fE c easu re cri. arv syste- inteer1:y, wntle ne incus try nas ceen ccncernec y 's AI- wi tn develocinc fixes to crolone S.G. service life anc rella:111:y. NRC
'y/
and incustry ef forts nave :een crimarily compienentary in nature.
. C) . Mcwever, to the extent that reliability implies safety and vice-versa the NRC and indus try efforts are synonomous. Therefo're, the staff is pursuing t.5e development of a joint NRC anc industry program to accress both near-term and long-term acticns required for continued safe operation of steam generators and ultimate resolution of :he S.G. degracation preciem. The.inten: is :o evaluate the cegree to wnich the NRC can
.: C) expand its role in crevention of tute degradation and work with the indus try :: solve :nis pecblem. Efforts to ce:emnine :ne feasibili y of
- nis ty;e of coc;erative program have been initiated and :re;csals for a l join: 1RC anc incus ry cr: gram w11' be crese'ntec in a '.ater cccumer. .
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REFERENCES
.. 1. Eisenhut, Lf aw, Strosnider, " Summary of Operating Experience
' C) witn Recirculating Steam Generators," NUREG-0523, January 1979.
2, Liaw, Strosnider, " Summary of Tube Integrity Operating Experience with Cnce-Through Steam Generator," NUREG-0571, March 1980.
- 3. S ECY-81 -6 64, "Infonnation Report - Steam Generator Tube Experience,"
'(3 fran W. J. Dircks to the Commissioners, November 24, 1981.
- 4. tienorandum for Commissioner Bradford fra W. J. Dircks, Status of Reca mendations Made in NUREG-0551, " Evaluation of Steen Generator Tube Rupture Events," to be transmitted.
i C) 5. EG&G Idaho, Inc. Report TREE-NUREG-1213 (NUREG/CR-0175), " Investigation of the Influence of Simulated Steam Generator Tube Ruptures Curing Lo s s-of-Coel s r.: Experiments in Semiscale 500-1 Systens," May 1972.
- 6. EG&G Idaho, Inc. Report CAAP-TR-022, " Steam Generator Tube Ru; ure -
Effects on a LOCA," November 1978.
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