Introduction statement at the panel discussion [contributed by James L. Snelgrove, ANL]

I just wanted to say a few words about the fuel demonstration part of the RERTR program, especially about our near-term activities. In the near-term we are considering three types of demonstrations. One of those is the irradiation of individual fuel elements in the ORR. These elements are to be essentially identical to the current ORR elements except that the enrichment will be 45% and the uranium density in the fuel meat will be somewhere in the neighborhood of 42 wt %. The exact loading hasn't been determined at this time, and preliminary work is underway to determine this. I think that I should point out that we really feel that the licensing basis is already available for what we are calling near-term fuel. That basis exists because of the extensive data obtained in the development of fuel for the ATR and for the HFIR. Those two reactors plus a few others are already making use of fuels loaded to this density at, of course, 93% enrichment. We don't see any real need from a metallurgical point of view to go through the complete demonstrations again. What we are doing though is loading this fuel with an enrichment and density which will be applicable in many cases in the near term, as it was mentioned yesterday, putting it in the ORR, burning it up, and doing a reasonable amount of post-irradiation examination. This data, we think, will be useful in licensing. Dr. Kanda mentioned in a comment earlier today that through the joint study with our program the Japanese are wanting to gain certain information from these experiments. We think that the experiments can satisfy those requirements. We would also be very interested to hear from any of the rest of you, the users, those of you who have to be concerned about gaining these licensing approvals, what sorts of information might really be needed here for any near-term work. We're planning to irradiate at least some elements to the order of, I would say, 90% burnup. Now, as it turns out in these elements at this enrichment and density, 90% burnup doesn't even surpass the burnup levels that have already been demonstrated as acceptable in the 93%-enriched fuels, and, therefore, one doesn't expect to find any swelling problems or any metallurgical-type problems at all. We will go through certain postirradiation examinations, however, but unless we see the need from some of the early parts of the examinations, we don't plan extensive metallurgical examinations. For these experiments, we would encourage participation by any fuel vendor who is interested in providing a few elements. We are not talking about many, maybe the order of two or three per vendor, which could be used in this program and would then, at least, provide a look to the rest of the world at elements made by particular vendors to particular enrichments and irradiated to good burnups. The other parts of our program in the near-term have to do with whole core demonstrations, and the one that is really planned at this time is the Ford Nuclear Reactor at the University of Michigan. The primary purpose of this demonstration is to have a core in which we can investigate the physics and safety characteristics of 20%-enriched fuel and compare them with those of the 93%-enriched cores. We will also use the data for verification of our analytical methods, so that when these methods are applied in evaluating the potential of other reactors for use of reduced-enrichment fuel, one will have increased confidence in the methods. I point out, of course, that there have been other reactors run with 20%-enriched fuel. However, that fuel was used quite awhile ago, and it's difficult today to obtain the sort of data in which we are really interested. The measurements in the FNR are planned to look at detailed flux distributions in the core and reflectors and at the ends of beam ports. It is also planned to measure the typical sorts of reactor characteristics which are always necessary to show that you're meeting your technical specifications and safety requirements.. The Ford Nuclear Reactor is an NRC licensed facility, and, as such, the University of Michigan will have to obtain licensing approval from NRC. We're hoping for the full cooperation of NRC, and we've been basically told that that will be forthcoming. We certainly don't anticipate that there are any new safety issues involved. We say te basis is there, and it's up to us, in cooperation with the University of Michigan, to demonstrate to the NRC that this is true. Out of this work will then come some documents that will be very useful as reference documents for any further near-term conversions using this type of fuel. I conclude my remarks here, and I will be happy to answer any questions