- Email: [email protected]
Discussion on Session I
Nuclear Engineering and Design 124 (1990) 23-29 North-Holland
23
Discussion o n Session I Chairman Karl E. Stahlkopf:
I wish to introduce to you the first speaker of our plenary session. The author is Makio Amano who is executive vice president of IHI, one of the major fabricators of both pressure vessels and piping in the world. I had the pleasure of meeting Mr Amano some 15 years ago. For those of you who knew me, you know that at that time I was running the pressure vessel and piping group at EPRI. With both the Institute and myself being rather new to the area of heavy section fabrication, we put together a" list of very distinguished people from throughout the world to help us define what our program should be. We had such people as Roy Nichols from the UKAEA, Spencer Bush with many years experience on the Advisory Committee on Reactor Safeguards and Makio Amano from IHI. They did much to aid our program, to make it successful. Certainly many of the things which came out of EPRI were as a direct result of the help we received from Amanosan and others. So, it is gives me great pleasure today to open the plenary session with Mr Amano's paper:
P a l e r 1.1: M. Amano, The needs of the nuclear pcessuxe bomKIm3, industries in the 1990s Discussion S. Bhandari, Framatome, France - I would like to
know if you have the thermal aging problem of cast stainless steels, in your reactors for example in the hot leg of PWRs and, if so, what remedies are you seeking? That is, is there a reduction of toughness because of thermal aging? Could you say a little more if you have such problems? Bhandari - That is reduction of toughness because of thermal aging. Could you say a little more if you have such problems? Chairman S t a h i k o p f - T h e toughness problem to which he refers is due to the spinoidal decomposition of the ferritic stainless steels. With higher ferritic content particularly you see a decrease in the fracture toughness. This primarily has been a problem in European and American reactors which contain stainless steels
which have high ferritic contents, that is between 10 and 20~. The question from the Framatome representative is: have you experienced this sort of thing in Japan? Bhandari - I think that puts it more clearly. It may be best if I discuss this personally with you later. A m a n o - Sorry, we have no data on the thermal aging problem of cast stainless steels. S t a h l k o p f - I would suggest to the questioner that this is something we should take up after Professor Shewmon's paper. I think he does discuss this and I know that from the make-up of the audience that there is much experience of this subject in the room. D r W. O'Donnell, O'Donneil Associates, USA - T h a n k
you. I am chairman of the ASME subgroup on Fatigue Strength, which is responsible for the fatigue curves in the ASME code. I would like to support the remarks made by Mr Amano about the need for upgrading those curves to include environmental effects. I would mention that the International Cyclic Crack Growth Committee has made much progress on the latter. Of course, EPRI has done much on this - Joe Gilman, in particular, in the area of environmental effects on fatigue which are currently not in the ASME Code. I wanted to mention that the Code is very actively trying to do all this under the Subgroup on Fatigue, the Subgroup on the Strength of Weldments, the Subcommittee on Design and the Section XI Committee. These groups have come up with a program, a multi-year effort, that is aimed at correcting the life assessment curves so that one may more accurately quantify the safe residual life of nuclear components. This program was recently approved by the Subcommittee on Design of the ASME Code and they have asked the Pressure Vessel Research Committee (PVRC) to implement that program. Fig. 1 shows the fatigue failure curve for A533B pressure vessel steel which is obtained when the stainless cladding is assumed to protect the ferritic material from reactor water effects until an underclad crack is initiated. The integrity o~ the cladding is then assumed to be compromised, and reactor water effects on crack propagation were taken into account using J-integral technology using the reactor water crack growth rate curves from Section XI of the Code. The latter are of
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24
Discussion on Session I
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made. I note that in your paper you have a figure, fig. 19, which relates to environmentally corrected fatigue data for carbon and low-alloy steels. I am having trouble understanding that figure because, as I read it, the best fit curve for the room temperature air data goes fight through the middle of all data points. This is not a
course in the process of being improved to include frequency and stress ratio effects, as illustrated in fig. 2. Any such improvements could also be included in the S-N fatigue design curves. J. Hickling, Siemens / KWU, W. Germany - This is a follow-up question to the point which has just been
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Discussion on Session I
pattern of which I am aware. As has just been pointed out, there are a lot of data which fall well away from that line when environmental effects are considered. Could you please comment on that figure? A m a n o - A different diagram has been shown on the overhead. This is because it might be easier to explain environmental effects in general. (Showing the figure.) It is necessary to note that, on this figure, we have the inference of a different strain rate under which the fatigue testing was done. So, we must have a more accurate evaluation for the fatigue results. S t a h l k o p f - Would anyone from the fatigue crack international group wish to make any comment conceruing this matter? A m a n o - At any rate, the purpose of my talk is a general statement on the subject. These issues require discussion in more detail. Thank you. I(. Kussmmd, M P A , Stuttgart, Fed, Rep. Germany - I
think that all the data which have been reported up to now on crack growth rate under corrosive attack have to be reviewed again. It seems to me that it is too early to make such statements as we have it now. This is true for both the presentation of Mr Amano and the comments of Dr O'Donnell. Chairman S t a h l k o p f - What are the Japanese doing in the way of on-line monitoring of fatigue and, if this is not being done, how are you accounting for fatigue cycles in your current plants so that, if you decide to go beyond forty years you will know the fatigue state of the plant7 A m a n o - T h e on-line monitoring of fatigue is not applied on reactor plants in Japan. I have discussed just the target for which we are aiming. Technologies I have mentioned are to detect an amount of fatigue damage of operating plants in the future.
Paper 1.2: Paul ShewnHm, Ohio State University, USA, O m ~ m aml e m e ~ . ng~ l m m ~ e bomKlary ismes - a U S Regu~ pmpe~e Chairman - It is my pleasure next to introduce an old friend who is well known through the pressure boundary industry, Professor Paul Shewmon, who is Professor of Material Science and Engineering at Ohio State University, and a senior member of the Advisory Committee on Reactor Safeguards of the US Nuclear Regulatory Commission. I first became aware of Paul when I was a graduate student at Beridey and had the good fortune to pick up his classic textbook on diffusion in crystals. I must mention, however, Paul, that I did find a small error in
25
the mathematics in that book. You probably do not remember but I wrote you a letter. However, I must say that has been some 20 years ago and that is the first and last time I have known Paul Shewmon to make a mistake. With that, I introduce Professor Paul Shewmon who will be speaking on Current a n d Emerging Pressure Boundary Issues: A Regulatory Perspective. Discussion S. Bhandari, Framatome, France - I am interested in
one of your statements regarding a 1983 incident which you told of a water hammer problem in the Maine Yankee reactor which opened up a leak. And, in fact, that this led to the exclusion of water hammer from leak-before-break relief. In fact, I would say that, if the water hammer did give us a leak, that means it is favoring leak, that this is favoring the leak-before-break demonstration. So this critical incident excludes your consideration that some other problems might be there. Could you say a little more about this? But, the fact that this particular incident which opened up a leak should cause the leak-before-break concept not to be applied seems to me to be somewhat controversial. Shewmon - Well, if you could show me or show the NRC that the water hammer that occurred there was the biggest and strongest and nastiest water hammer that could ever occur, then what you say is right. However, if it turns out that this is just the smaller brother of the bigger one that could come along, then there is no way of knowing if the next one to occur could rip the pipe off the wall and do more damage. So, the ability to predict and bound water hammers is not as good as we would like. Also, the NRC has been quite conservative in where they would allow leak-beforebreak to be applied. One decision that I am not in sympathy with but which the NRC has stayed with is to say that leak-before-break cannot be employed in lines which undergo stress corrosion cracking. You could say that there is a wonderful series of examples in the BWR history that leak-before-break works extremely well there. So, I think you could make a good case for using it there and an even better case than you could for water hammer but the NRC hasn't allowed it. Griesbach, E P R I Professor Shewmon, you described to us several possible solutions for resolving the reactor vessel embrittlement problems, including annealing and flux reduction. We are active in this area as well, and I wonder if you could share your opinion on why credit for improved inspection is not always talked about. As you know, Section XI (ASME Code) and parts of the NRC regulations are requiring more and
26
Discussion on Session I
better inspections and yet there is very little credit given for these. In performing the various fracture analyses, we are still having to use large reference flaws or flaw distributions. Could you tell us your perspective on this topic? S h e w m o n - A r e you interested in how this would impact IriS or in a more general vein? Griesbach - I refer to all vessel integrity questions including low upper shelf, PTS or plant operating criteria. S h e w m o n - I have not looked at the other plant operating criteria. The PTS situation I am familiar with. Certainly, if someone did come in with the analysis that would be required by the PTS rule, one of the things that I personally think they should put in that package would be how they have looked for flaws and that they have every reason to believe that the flaws there are smaller than some reasonable limit. But, the NRC has said the only time they will consider inspection, or surveillance data on your vessel instead of using Reg. Guide 1.99, is when you come in with this package to show why it is safe to go on beyond the PTS criteria. That package, will be thick enough that everyone, as you know, has been trying to do flux reductions and other things to avoid reaching the Reg. Guide criteria. Griesbach - Right. We are aware of all that, but how about the low upper shelf question? That happens to be an area where most of the emphasis is being placed on materials rather than flaws. S h e w m o n - Yes, I would agree with you. In a sense, that comes back to Section XI which says, "don't tell me about how good it is, just assume a quarter thickness flaw". I think there is a good case for what you say, but it might be played differently and, when you get your friends in Section XI to agree to that, we would be pleased to review it. Server, Tenera, USA - With regard to the low upper-shelf issue, you emphasized the Linde 80 welds as being a problem area. I think when we consider plant life extension, some of our oldest vessels which have A302B plate material in them can also have toughnesses approaching 50 ft lb after a good deal of irradiation. Do you have any comments relative to that? (As an aside, within Section XI of the Code, we have been looking at some of the A302B materials, relative to the welds, and there are some indications that for this steel (A302B), even at a toughness slightly above 50 ft lb, in terms of the integrity, the problem may actually be worse than for a Linde 80 weld at slightly below 50 ft lb.) S h e w m o n - No. I wish I did but I don't. I remember there was a study of some of the oldest plants and what their basis was for continued operation when they had
been licensed under a different set of rules. About 8 years ago, a report came out which studied these and pressure vessel integrity was one issue there. I guess I would wonder if, the licensees of these oldest plants would apply for life extension but, if they did, you raise an interesting question, and I have no particular basis for answering it. Mager, Westinghouse - I have a comment on thermal aging which goes back to Mr Amano's paper. I am very closely involved in plant life extension (PLEX) in Japan and to the best of my knowledge thermal aging of cast stainless piping is not a problem in Japan. The Japanese are being prudent in assessing thermal aging in thier PLEX program. S h e w m o n - How good is the database on where to find the high ferrite? Certainly a 20% ferrite is worse than 10%. Sometimes they are as high as 20% and my question is do we know where these high ferrite castings are located? M a g e r - Yes, we have a data base on all our centrifugal cast piping and static fittings. Now we are going back periodically to look at those materials to actually measure the ferrite. I guess the important point is that I do not want anyone to go away with the impression that there was a problem in Japan. S t a h l k o p f - Tom, when you have taken a look at the ferrite in your castings, have you done through-wall measurements and, do you see variations in ferrite as you move through wall? M a g e r - No, I have not done any of that work. StahlkopfIs there anyone in the audience who might have done some of this kind of work? Nanstad, O R N L - O n e more comment in relation to the cast stainless steel aging. I mention some work we have done recently on aging and that is on stainless steel weld metal. I realize that the cast stainless steels have received enhanced attention in the past ten years for good reason, but weld metals typically have ferrite contents as well in the range of about 8 to 10 ferrite number. In fact, this may go as high as 12-15 in some cases. As a result of this fact, we did a thermal aging study on some type 308 welds in the last five years. Aging was at 343°C for up to 20000 h, (about 2½ years) and the degradation effects on toughness axe fairly severe in that the Charpy impact energies have shown reductions of about half of the room temperature values. The reason I mention this as being significant, or potentially significant, is that the stainless steel welds have an upper-shelf Charpy energy of about 50-60 ft lb in the as-welded condition, as they begin life. So, we are talking about Charpy-V toughness levels of about 25-30 ft lb. I am not saying that this is necessarily a problem,
27
Discussion on Session I
and I think that Dr Shewmon's comment about the fact that we merely need to monitor this kind of behavior in light of the fact that we have seen no failures, is important. Nonetheless, I would encourage people to monitor the weld materials as well as the cast stainless. StahlkopfRandy, I have a couple of questions regarding the work that you did. First, were these accelerated tests and, if so, how accelerated were they7. N a n s t a d - They were slightly accelerated. The temperature of aging was 343°C, (650°F). People with whom I have talked in the industry have told me that temperatures typically above 600 ° F are experienced in the plants. The 6 5 0 ° F level is somewhat above the upper limit but, if you look at some of the work in the cast stainless steels at lower levels, the effect of temperature down to 5 5 0 ° F is still fairly strong. So, yes, they are slightly accelerated but we are definitely talking about low temperature aging effects. S t a h l k o p f - T h e second question is: we are all aware of the charpy energy degradation, but did we do any J-integral testing in these welds? Could you compare the degradation in J qualities to those you see in Charpy? N a n s t a d - No, we have not done that. We didn't have enough material but we are planning to do some of that if we can with some pre-cracked Charpy specimens. But, based on the results we saw in terms of tensile testing, it is my feeling that the static type of measurements are not going to show the same degradation that the dynamic tests show because the stainless steel weld metal is a strain rate sensitive material. I think it is the dynamic properties that are going to show that degree of degradation, however, experiments need to be conducted. StahlkopfYes, I think this is precisely the point and, if you look at the work that other investigators have done, that is one of the things that at least gives me some confidence that this problem may not be as bad as it may appear. I have a hard time envisaging accident conditions which would lead you to strain rates as high as those under Charpy tests. But, your points are very well taken and we'll be interested in following that work. H . Schulz, G R S , W. G e r m a n y - I am concerned about tube failure in PWR steam generators. I am especially interested in your judgement regarding the significance of the recent plug failure. S h e w m o n - By way of background, one tube failure is part of the plant's sign basis and, though it makes things exciting for the operator, it probably is not a safety issue. There is not a too well-defined number of tubes whose failure, would make the accident a real
safety concern. People have tried to estimate this number and it is as high as 10 tubes but maybe it is a lower number. So, once you get into the possibility of a mechanism that will fail several tubes, the severity of the accident and the concern for it goes up. Haggag~ O R N L - I wish to return the discussion back to the subject of pressure vessel annealing. We have heard that the USSR has been conducting annealing of reactor vessels. I wonder if you could tell us more about the annealing with mechanical properties testing that they have done and also about the reembrittlement rate? In the High Flux Isotope Reactor at Oak Ridge, people have been worried, for example, if they should have done or should do annealing, and how many years would we be gained in terms of actual operation after vessel annealing? S h e w m o n - Your questions were how fast would it reembrittle, and what was the first part? Haggag T h e first half is, we have heard of Russian vessel annealing and I wish to know the degree of success and what properties were measured and with what results? S h e w m o n - T h e i r program is not much different than what Tom Mager is going to talk about. They have measured Charpy toughness and, as I recall, as he will tell you and others have learned, the temperature at which one anneals is important. If you go up to 850 ° F or above, apparently the reembrittlement rate is about the same or less than it was originally. Also at 850 o F or a little bit higher, you can get 100% recovery. As I recall from the Russian work, if they had particularly high phosphorus heats, then they would not get 100% but about 80~ recovery. I do not know what the counterpart in our situation would be but the good news was if you went up to this temperature, then you would get a re-embrittlement rate similar to what it was originally while, if you used lower annealing temperatures, it re-embrittled faster. H a g g a g - Do you know of any available literature on this subject? At the Andover meeting (ASTM, Andover, Massachusetts), for example, we were expecting a USSR paper on annealing but it was never presented. Also, what type and size reactor have they annealed? S h e w m o n - This was on their first generation of PWRs. They are about 400 MW? S t a h l k o p f - They are 440 MW; the VVER 440s. S h e w m o n - That's 440 MW electric; not quite as big as we axe talking about but still a genuine commercial power plant. You might talk with Chuck Sexpan (NRC), who kept the minutes for that meeting. I do not see why -
28
Discussion on Session I
his minutes should not be available to you, but, I do not think anything has been officially released on it. He would be a better source on that than I am. J. Sievers, GRS, K~ln, W. Germany - My question concerns PTS. From a structural mechanics point of view, there is a (significant) difference between an axial symmetric and a non-axial symmetric cooling of the reactor pressure vessel. Therefore, is this taken into account from a safety point of view?. Shewmon - I do not understand what you are saying. If you cool down or change the water temperature of the water inside the vessel, what exactly do you ask7 Sievers - Cooling of the vessel by all nozzles (cold side) may be approximated by axisymmetric loading conditions but cooling by only one nozzle like in a strip shows a significantly different loading of the vessel wall. Is this taken into acx'.ount? Shewmon - The analysis and part of the argument between the industry and the NRC had to do with what you assume about how long the cold water stays next to the pressure vessel before vorficity removes it and pulls it out. And, since that was a factor in the discussions, I am sure they did this in a non-axisymmetric manner. To what extent they would allow them to give credit for axial welds that were away from nozzles, I am not as clear as I should be. Have I answered your question? Sievers - To a certain degree. Shewmon - I am afraid I have done the best I can. You might try Mike Mayfield (NRC, RES). Thermal hydraulics isn't his bag, but he might help. M a s e r - Paul, you mentioned that steam generator problems are not going away as opposed to the BWR piping question. How about lnconel 6907 Do you believe that this would more or less eliminate the problem? Shewmon - Well it certainly is an interesting effort. I was at a Jekyl Island meeting recently and there was someone there who pointed out that IN 690 was now in one of the replacement steam generators that D.C. Cook had installed. And, 690 would certainly be an attractive alternative but I do not need to point out to you that, while you are a young man, I may not live long enough to see all the stream generators with Inconel 600 removed from service. So, it gets back in a sense to the old steam generators and how you are going to continue to live with them? Though, if you are putting in a new steam generator, it looks like 690 is the way to go since it seems to have a lot fewer corrosion problems. P,. Smith, E P R I - N D E Center, Charlotte, N.C. - Paul, I noticed that we have been discussing a number of the problems that we have either experienced or perceived
to be potential problems. Looking in your crystal ball, what areas have we that you would see as future issues? Shewmon - I don't have a particularly clear crystal ball. I brought up fatigue issues as one of the problems that occurs quite frequently. John H i c k l i n g Siemens / K W U , Fed~ Rep. o f Germany
Professor Shewmon, I felt from a European standpoint that you were being perhaps a tittle unduly pessimistic about single phase erosion-corrosion. It is worth pointing out that in West Germany, conventional power plants with once-through Benson boilers often operate with oxygen added to the water. That has been water treatment practice now for nearly fifteen years and the advantages of that treatment in preventing erosion-corrosion in feedwater trains were pointed out as long as ten years ago. Much of the work to quantify the effects was also done at that time. Therefore, there is a very large database out there on erosion-corrosion in the conventional side of the industry, which to some extent only needs to be applied properly to the nuclear plant. don't feel there is any need for pessimism on the way to get a handle on that particular problem. Shewmon - I come from farm country and there is a story about the new fresh graduate from the university who came out as county agent and was trying to encourage the farmer to use the results of some of the new research. The farmer figuratively patted him on the head and said, "Son, I'm not farmin~ half as good as I know how now". It seems to me that, though you are right, I feel I am too, because there are a lot of reactors out there and a lot of piping systems where this hasn't been considered adequately. We know that the BWRs have been adding oxygen to their feedwater for years but up to now the PWR operators have not believed it was needed there. This activity has been going through a rapid transition, so, there is a transition phase as people decide where they have to look for thinning, what they have to do about it, and things of that sort. But, you are quite right and when we had a meeting on this soon after the Surry event, it was interesting to see that most of the research did come from Germany and erosion-corrosion seemed to be well understood though this knowledge had not come into the US nuclear PWR business the way it should have and will. Bhandar/- Regarding the question about other areas of concern, I would certainly mention the problems of the fracture behavior of piping under dynamic loading conditions. This is one of the top potential problems. I know there are international programs addressing this matter, but my first comment should have been to your comments regarding what I asked about leak-beforebreak under water hammer conditions. My comment -
D i s c u s s i o n on S e s s i o n I
concerns the drastic decision one takes whenever something occurs like that. Now, for example, leak-before, break is excluded under stress corrosion cracking. This is OK, but what about with fatigue problems? From my point of view, instead of taking drastic decisions like that, (refers to Maine Yankee case - Ed.) why don't you go ahead and give a chance to the owners and move ahead. Show it if you can. If the person is capable of demonstrating under given conditions, for example the water hammer, that the loading is such and such and can show his leak-before-break, then why do you exclude it by such a drastic decision. My point is, as a manufacturer - Framatome - we have to demonstrate integrity under the loading conditions which our PWRs are facing. But, whenever we face such drastic decisions, there are major difficulties. We feel, to close my comment, fatigue technology is as well developed as the fracture technology. So, why do we exclude under fa-
29
tigue conditions, if we can demonstrate, under such and such conditions, leak-before-break why not allow it. Shewmon - One of the favorite sayings of the people who work for the government is "we are from the government; we are here to help you". (Laughter.) It gets that sort of response often, but, I think, the NRC's response would be "fine, come in and show us your information". You must realize that where they were starting from was, no allowance for leak-before-break. So, we have now gone from never allowing it, to allowing on a variety of big pipes which has allowed people to make their plants easier and safer to operate by removing a bunch of things. Whether or not it can be pushed farther depends upon having a good technical case for it. They (NRC) listen to technical arguments quite regularly and they may well accept them. However, there must be a good argument there before they will.
23
Discussion o n Session I Chairman Karl E. Stahlkopf:
I wish to introduce to you the first speaker of our plenary session. The author is Makio Amano who is executive vice president of IHI, one of the major fabricators of both pressure vessels and piping in the world. I had the pleasure of meeting Mr Amano some 15 years ago. For those of you who knew me, you know that at that time I was running the pressure vessel and piping group at EPRI. With both the Institute and myself being rather new to the area of heavy section fabrication, we put together a" list of very distinguished people from throughout the world to help us define what our program should be. We had such people as Roy Nichols from the UKAEA, Spencer Bush with many years experience on the Advisory Committee on Reactor Safeguards and Makio Amano from IHI. They did much to aid our program, to make it successful. Certainly many of the things which came out of EPRI were as a direct result of the help we received from Amanosan and others. So, it is gives me great pleasure today to open the plenary session with Mr Amano's paper:
P a l e r 1.1: M. Amano, The needs of the nuclear pcessuxe bomKIm3, industries in the 1990s Discussion S. Bhandari, Framatome, France - I would like to
know if you have the thermal aging problem of cast stainless steels, in your reactors for example in the hot leg of PWRs and, if so, what remedies are you seeking? That is, is there a reduction of toughness because of thermal aging? Could you say a little more if you have such problems? Bhandari - That is reduction of toughness because of thermal aging. Could you say a little more if you have such problems? Chairman S t a h i k o p f - T h e toughness problem to which he refers is due to the spinoidal decomposition of the ferritic stainless steels. With higher ferritic content particularly you see a decrease in the fracture toughness. This primarily has been a problem in European and American reactors which contain stainless steels
which have high ferritic contents, that is between 10 and 20~. The question from the Framatome representative is: have you experienced this sort of thing in Japan? Bhandari - I think that puts it more clearly. It may be best if I discuss this personally with you later. A m a n o - Sorry, we have no data on the thermal aging problem of cast stainless steels. S t a h l k o p f - I would suggest to the questioner that this is something we should take up after Professor Shewmon's paper. I think he does discuss this and I know that from the make-up of the audience that there is much experience of this subject in the room. D r W. O'Donnell, O'Donneil Associates, USA - T h a n k
you. I am chairman of the ASME subgroup on Fatigue Strength, which is responsible for the fatigue curves in the ASME code. I would like to support the remarks made by Mr Amano about the need for upgrading those curves to include environmental effects. I would mention that the International Cyclic Crack Growth Committee has made much progress on the latter. Of course, EPRI has done much on this - Joe Gilman, in particular, in the area of environmental effects on fatigue which are currently not in the ASME Code. I wanted to mention that the Code is very actively trying to do all this under the Subgroup on Fatigue, the Subgroup on the Strength of Weldments, the Subcommittee on Design and the Section XI Committee. These groups have come up with a program, a multi-year effort, that is aimed at correcting the life assessment curves so that one may more accurately quantify the safe residual life of nuclear components. This program was recently approved by the Subcommittee on Design of the ASME Code and they have asked the Pressure Vessel Research Committee (PVRC) to implement that program. Fig. 1 shows the fatigue failure curve for A533B pressure vessel steel which is obtained when the stainless cladding is assumed to protect the ferritic material from reactor water effects until an underclad crack is initiated. The integrity o~ the cladding is then assumed to be compromised, and reactor water effects on crack propagation were taken into account using J-integral technology using the reactor water crack growth rate curves from Section XI of the Code. The latter are of
0 0 2 9 - 5 4 9 3 / 9 0 / $ 0 3 . 5 0 © 1990 - Elsevier Science Publishers B.V. ( N o r t h - H o l l a n d )
24
Discussion on Session I
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made. I note that in your paper you have a figure, fig. 19, which relates to environmentally corrected fatigue data for carbon and low-alloy steels. I am having trouble understanding that figure because, as I read it, the best fit curve for the room temperature air data goes fight through the middle of all data points. This is not a
course in the process of being improved to include frequency and stress ratio effects, as illustrated in fig. 2. Any such improvements could also be included in the S-N fatigue design curves. J. Hickling, Siemens / KWU, W. Germany - This is a follow-up question to the point which has just been
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Discussion on Session I
pattern of which I am aware. As has just been pointed out, there are a lot of data which fall well away from that line when environmental effects are considered. Could you please comment on that figure? A m a n o - A different diagram has been shown on the overhead. This is because it might be easier to explain environmental effects in general. (Showing the figure.) It is necessary to note that, on this figure, we have the inference of a different strain rate under which the fatigue testing was done. So, we must have a more accurate evaluation for the fatigue results. S t a h l k o p f - Would anyone from the fatigue crack international group wish to make any comment conceruing this matter? A m a n o - At any rate, the purpose of my talk is a general statement on the subject. These issues require discussion in more detail. Thank you. I(. Kussmmd, M P A , Stuttgart, Fed, Rep. Germany - I
think that all the data which have been reported up to now on crack growth rate under corrosive attack have to be reviewed again. It seems to me that it is too early to make such statements as we have it now. This is true for both the presentation of Mr Amano and the comments of Dr O'Donnell. Chairman S t a h l k o p f - What are the Japanese doing in the way of on-line monitoring of fatigue and, if this is not being done, how are you accounting for fatigue cycles in your current plants so that, if you decide to go beyond forty years you will know the fatigue state of the plant7 A m a n o - T h e on-line monitoring of fatigue is not applied on reactor plants in Japan. I have discussed just the target for which we are aiming. Technologies I have mentioned are to detect an amount of fatigue damage of operating plants in the future.
Paper 1.2: Paul ShewnHm, Ohio State University, USA, O m ~ m aml e m e ~ . ng~ l m m ~ e bomKlary ismes - a U S Regu~ pmpe~e Chairman - It is my pleasure next to introduce an old friend who is well known through the pressure boundary industry, Professor Paul Shewmon, who is Professor of Material Science and Engineering at Ohio State University, and a senior member of the Advisory Committee on Reactor Safeguards of the US Nuclear Regulatory Commission. I first became aware of Paul when I was a graduate student at Beridey and had the good fortune to pick up his classic textbook on diffusion in crystals. I must mention, however, Paul, that I did find a small error in
25
the mathematics in that book. You probably do not remember but I wrote you a letter. However, I must say that has been some 20 years ago and that is the first and last time I have known Paul Shewmon to make a mistake. With that, I introduce Professor Paul Shewmon who will be speaking on Current a n d Emerging Pressure Boundary Issues: A Regulatory Perspective. Discussion S. Bhandari, Framatome, France - I am interested in
one of your statements regarding a 1983 incident which you told of a water hammer problem in the Maine Yankee reactor which opened up a leak. And, in fact, that this led to the exclusion of water hammer from leak-before-break relief. In fact, I would say that, if the water hammer did give us a leak, that means it is favoring leak, that this is favoring the leak-before-break demonstration. So this critical incident excludes your consideration that some other problems might be there. Could you say a little more about this? But, the fact that this particular incident which opened up a leak should cause the leak-before-break concept not to be applied seems to me to be somewhat controversial. Shewmon - Well, if you could show me or show the NRC that the water hammer that occurred there was the biggest and strongest and nastiest water hammer that could ever occur, then what you say is right. However, if it turns out that this is just the smaller brother of the bigger one that could come along, then there is no way of knowing if the next one to occur could rip the pipe off the wall and do more damage. So, the ability to predict and bound water hammers is not as good as we would like. Also, the NRC has been quite conservative in where they would allow leak-beforebreak to be applied. One decision that I am not in sympathy with but which the NRC has stayed with is to say that leak-before-break cannot be employed in lines which undergo stress corrosion cracking. You could say that there is a wonderful series of examples in the BWR history that leak-before-break works extremely well there. So, I think you could make a good case for using it there and an even better case than you could for water hammer but the NRC hasn't allowed it. Griesbach, E P R I Professor Shewmon, you described to us several possible solutions for resolving the reactor vessel embrittlement problems, including annealing and flux reduction. We are active in this area as well, and I wonder if you could share your opinion on why credit for improved inspection is not always talked about. As you know, Section XI (ASME Code) and parts of the NRC regulations are requiring more and
26
Discussion on Session I
better inspections and yet there is very little credit given for these. In performing the various fracture analyses, we are still having to use large reference flaws or flaw distributions. Could you tell us your perspective on this topic? S h e w m o n - A r e you interested in how this would impact IriS or in a more general vein? Griesbach - I refer to all vessel integrity questions including low upper shelf, PTS or plant operating criteria. S h e w m o n - I have not looked at the other plant operating criteria. The PTS situation I am familiar with. Certainly, if someone did come in with the analysis that would be required by the PTS rule, one of the things that I personally think they should put in that package would be how they have looked for flaws and that they have every reason to believe that the flaws there are smaller than some reasonable limit. But, the NRC has said the only time they will consider inspection, or surveillance data on your vessel instead of using Reg. Guide 1.99, is when you come in with this package to show why it is safe to go on beyond the PTS criteria. That package, will be thick enough that everyone, as you know, has been trying to do flux reductions and other things to avoid reaching the Reg. Guide criteria. Griesbach - Right. We are aware of all that, but how about the low upper shelf question? That happens to be an area where most of the emphasis is being placed on materials rather than flaws. S h e w m o n - Yes, I would agree with you. In a sense, that comes back to Section XI which says, "don't tell me about how good it is, just assume a quarter thickness flaw". I think there is a good case for what you say, but it might be played differently and, when you get your friends in Section XI to agree to that, we would be pleased to review it. Server, Tenera, USA - With regard to the low upper-shelf issue, you emphasized the Linde 80 welds as being a problem area. I think when we consider plant life extension, some of our oldest vessels which have A302B plate material in them can also have toughnesses approaching 50 ft lb after a good deal of irradiation. Do you have any comments relative to that? (As an aside, within Section XI of the Code, we have been looking at some of the A302B materials, relative to the welds, and there are some indications that for this steel (A302B), even at a toughness slightly above 50 ft lb, in terms of the integrity, the problem may actually be worse than for a Linde 80 weld at slightly below 50 ft lb.) S h e w m o n - No. I wish I did but I don't. I remember there was a study of some of the oldest plants and what their basis was for continued operation when they had
been licensed under a different set of rules. About 8 years ago, a report came out which studied these and pressure vessel integrity was one issue there. I guess I would wonder if, the licensees of these oldest plants would apply for life extension but, if they did, you raise an interesting question, and I have no particular basis for answering it. Mager, Westinghouse - I have a comment on thermal aging which goes back to Mr Amano's paper. I am very closely involved in plant life extension (PLEX) in Japan and to the best of my knowledge thermal aging of cast stainless piping is not a problem in Japan. The Japanese are being prudent in assessing thermal aging in thier PLEX program. S h e w m o n - How good is the database on where to find the high ferrite? Certainly a 20% ferrite is worse than 10%. Sometimes they are as high as 20% and my question is do we know where these high ferrite castings are located? M a g e r - Yes, we have a data base on all our centrifugal cast piping and static fittings. Now we are going back periodically to look at those materials to actually measure the ferrite. I guess the important point is that I do not want anyone to go away with the impression that there was a problem in Japan. S t a h l k o p f - Tom, when you have taken a look at the ferrite in your castings, have you done through-wall measurements and, do you see variations in ferrite as you move through wall? M a g e r - No, I have not done any of that work. StahlkopfIs there anyone in the audience who might have done some of this kind of work? Nanstad, O R N L - O n e more comment in relation to the cast stainless steel aging. I mention some work we have done recently on aging and that is on stainless steel weld metal. I realize that the cast stainless steels have received enhanced attention in the past ten years for good reason, but weld metals typically have ferrite contents as well in the range of about 8 to 10 ferrite number. In fact, this may go as high as 12-15 in some cases. As a result of this fact, we did a thermal aging study on some type 308 welds in the last five years. Aging was at 343°C for up to 20000 h, (about 2½ years) and the degradation effects on toughness axe fairly severe in that the Charpy impact energies have shown reductions of about half of the room temperature values. The reason I mention this as being significant, or potentially significant, is that the stainless steel welds have an upper-shelf Charpy energy of about 50-60 ft lb in the as-welded condition, as they begin life. So, we are talking about Charpy-V toughness levels of about 25-30 ft lb. I am not saying that this is necessarily a problem,
27
Discussion on Session I
and I think that Dr Shewmon's comment about the fact that we merely need to monitor this kind of behavior in light of the fact that we have seen no failures, is important. Nonetheless, I would encourage people to monitor the weld materials as well as the cast stainless. StahlkopfRandy, I have a couple of questions regarding the work that you did. First, were these accelerated tests and, if so, how accelerated were they7. N a n s t a d - They were slightly accelerated. The temperature of aging was 343°C, (650°F). People with whom I have talked in the industry have told me that temperatures typically above 600 ° F are experienced in the plants. The 6 5 0 ° F level is somewhat above the upper limit but, if you look at some of the work in the cast stainless steels at lower levels, the effect of temperature down to 5 5 0 ° F is still fairly strong. So, yes, they are slightly accelerated but we are definitely talking about low temperature aging effects. S t a h l k o p f - T h e second question is: we are all aware of the charpy energy degradation, but did we do any J-integral testing in these welds? Could you compare the degradation in J qualities to those you see in Charpy? N a n s t a d - No, we have not done that. We didn't have enough material but we are planning to do some of that if we can with some pre-cracked Charpy specimens. But, based on the results we saw in terms of tensile testing, it is my feeling that the static type of measurements are not going to show the same degradation that the dynamic tests show because the stainless steel weld metal is a strain rate sensitive material. I think it is the dynamic properties that are going to show that degree of degradation, however, experiments need to be conducted. StahlkopfYes, I think this is precisely the point and, if you look at the work that other investigators have done, that is one of the things that at least gives me some confidence that this problem may not be as bad as it may appear. I have a hard time envisaging accident conditions which would lead you to strain rates as high as those under Charpy tests. But, your points are very well taken and we'll be interested in following that work. H . Schulz, G R S , W. G e r m a n y - I am concerned about tube failure in PWR steam generators. I am especially interested in your judgement regarding the significance of the recent plug failure. S h e w m o n - By way of background, one tube failure is part of the plant's sign basis and, though it makes things exciting for the operator, it probably is not a safety issue. There is not a too well-defined number of tubes whose failure, would make the accident a real
safety concern. People have tried to estimate this number and it is as high as 10 tubes but maybe it is a lower number. So, once you get into the possibility of a mechanism that will fail several tubes, the severity of the accident and the concern for it goes up. Haggag~ O R N L - I wish to return the discussion back to the subject of pressure vessel annealing. We have heard that the USSR has been conducting annealing of reactor vessels. I wonder if you could tell us more about the annealing with mechanical properties testing that they have done and also about the reembrittlement rate? In the High Flux Isotope Reactor at Oak Ridge, people have been worried, for example, if they should have done or should do annealing, and how many years would we be gained in terms of actual operation after vessel annealing? S h e w m o n - Your questions were how fast would it reembrittle, and what was the first part? Haggag T h e first half is, we have heard of Russian vessel annealing and I wish to know the degree of success and what properties were measured and with what results? S h e w m o n - T h e i r program is not much different than what Tom Mager is going to talk about. They have measured Charpy toughness and, as I recall, as he will tell you and others have learned, the temperature at which one anneals is important. If you go up to 850 ° F or above, apparently the reembrittlement rate is about the same or less than it was originally. Also at 850 o F or a little bit higher, you can get 100% recovery. As I recall from the Russian work, if they had particularly high phosphorus heats, then they would not get 100% but about 80~ recovery. I do not know what the counterpart in our situation would be but the good news was if you went up to this temperature, then you would get a re-embrittlement rate similar to what it was originally while, if you used lower annealing temperatures, it re-embrittled faster. H a g g a g - Do you know of any available literature on this subject? At the Andover meeting (ASTM, Andover, Massachusetts), for example, we were expecting a USSR paper on annealing but it was never presented. Also, what type and size reactor have they annealed? S h e w m o n - This was on their first generation of PWRs. They are about 400 MW? S t a h l k o p f - They are 440 MW; the VVER 440s. S h e w m o n - That's 440 MW electric; not quite as big as we axe talking about but still a genuine commercial power plant. You might talk with Chuck Sexpan (NRC), who kept the minutes for that meeting. I do not see why -
28
Discussion on Session I
his minutes should not be available to you, but, I do not think anything has been officially released on it. He would be a better source on that than I am. J. Sievers, GRS, K~ln, W. Germany - My question concerns PTS. From a structural mechanics point of view, there is a (significant) difference between an axial symmetric and a non-axial symmetric cooling of the reactor pressure vessel. Therefore, is this taken into account from a safety point of view?. Shewmon - I do not understand what you are saying. If you cool down or change the water temperature of the water inside the vessel, what exactly do you ask7 Sievers - Cooling of the vessel by all nozzles (cold side) may be approximated by axisymmetric loading conditions but cooling by only one nozzle like in a strip shows a significantly different loading of the vessel wall. Is this taken into acx'.ount? Shewmon - The analysis and part of the argument between the industry and the NRC had to do with what you assume about how long the cold water stays next to the pressure vessel before vorficity removes it and pulls it out. And, since that was a factor in the discussions, I am sure they did this in a non-axisymmetric manner. To what extent they would allow them to give credit for axial welds that were away from nozzles, I am not as clear as I should be. Have I answered your question? Sievers - To a certain degree. Shewmon - I am afraid I have done the best I can. You might try Mike Mayfield (NRC, RES). Thermal hydraulics isn't his bag, but he might help. M a s e r - Paul, you mentioned that steam generator problems are not going away as opposed to the BWR piping question. How about lnconel 6907 Do you believe that this would more or less eliminate the problem? Shewmon - Well it certainly is an interesting effort. I was at a Jekyl Island meeting recently and there was someone there who pointed out that IN 690 was now in one of the replacement steam generators that D.C. Cook had installed. And, 690 would certainly be an attractive alternative but I do not need to point out to you that, while you are a young man, I may not live long enough to see all the stream generators with Inconel 600 removed from service. So, it gets back in a sense to the old steam generators and how you are going to continue to live with them? Though, if you are putting in a new steam generator, it looks like 690 is the way to go since it seems to have a lot fewer corrosion problems. P,. Smith, E P R I - N D E Center, Charlotte, N.C. - Paul, I noticed that we have been discussing a number of the problems that we have either experienced or perceived
to be potential problems. Looking in your crystal ball, what areas have we that you would see as future issues? Shewmon - I don't have a particularly clear crystal ball. I brought up fatigue issues as one of the problems that occurs quite frequently. John H i c k l i n g Siemens / K W U , Fed~ Rep. o f Germany
Professor Shewmon, I felt from a European standpoint that you were being perhaps a tittle unduly pessimistic about single phase erosion-corrosion. It is worth pointing out that in West Germany, conventional power plants with once-through Benson boilers often operate with oxygen added to the water. That has been water treatment practice now for nearly fifteen years and the advantages of that treatment in preventing erosion-corrosion in feedwater trains were pointed out as long as ten years ago. Much of the work to quantify the effects was also done at that time. Therefore, there is a very large database out there on erosion-corrosion in the conventional side of the industry, which to some extent only needs to be applied properly to the nuclear plant. don't feel there is any need for pessimism on the way to get a handle on that particular problem. Shewmon - I come from farm country and there is a story about the new fresh graduate from the university who came out as county agent and was trying to encourage the farmer to use the results of some of the new research. The farmer figuratively patted him on the head and said, "Son, I'm not farmin~ half as good as I know how now". It seems to me that, though you are right, I feel I am too, because there are a lot of reactors out there and a lot of piping systems where this hasn't been considered adequately. We know that the BWRs have been adding oxygen to their feedwater for years but up to now the PWR operators have not believed it was needed there. This activity has been going through a rapid transition, so, there is a transition phase as people decide where they have to look for thinning, what they have to do about it, and things of that sort. But, you are quite right and when we had a meeting on this soon after the Surry event, it was interesting to see that most of the research did come from Germany and erosion-corrosion seemed to be well understood though this knowledge had not come into the US nuclear PWR business the way it should have and will. Bhandar/- Regarding the question about other areas of concern, I would certainly mention the problems of the fracture behavior of piping under dynamic loading conditions. This is one of the top potential problems. I know there are international programs addressing this matter, but my first comment should have been to your comments regarding what I asked about leak-beforebreak under water hammer conditions. My comment -
D i s c u s s i o n on S e s s i o n I
concerns the drastic decision one takes whenever something occurs like that. Now, for example, leak-before, break is excluded under stress corrosion cracking. This is OK, but what about with fatigue problems? From my point of view, instead of taking drastic decisions like that, (refers to Maine Yankee case - Ed.) why don't you go ahead and give a chance to the owners and move ahead. Show it if you can. If the person is capable of demonstrating under given conditions, for example the water hammer, that the loading is such and such and can show his leak-before-break, then why do you exclude it by such a drastic decision. My point is, as a manufacturer - Framatome - we have to demonstrate integrity under the loading conditions which our PWRs are facing. But, whenever we face such drastic decisions, there are major difficulties. We feel, to close my comment, fatigue technology is as well developed as the fracture technology. So, why do we exclude under fa-
29
tigue conditions, if we can demonstrate, under such and such conditions, leak-before-break why not allow it. Shewmon - One of the favorite sayings of the people who work for the government is "we are from the government; we are here to help you". (Laughter.) It gets that sort of response often, but, I think, the NRC's response would be "fine, come in and show us your information". You must realize that where they were starting from was, no allowance for leak-before-break. So, we have now gone from never allowing it, to allowing on a variety of big pipes which has allowed people to make their plants easier and safer to operate by removing a bunch of things. Whether or not it can be pushed farther depends upon having a good technical case for it. They (NRC) listen to technical arguments quite regularly and they may well accept them. However, there must be a good argument there before they will.