Appendix D
Consultant Reports

Note: The principal investigator (P.I.) consulted with several expert consultants in the course of this study. At the request of the P.I., the following reports were contributed by 11 consultants who were invited to attend the second information-gathering meeting and two consultants who did not attend that meeting (see Appendixes B and C). Biographical sketches of the consultants are included in Appendix E.

The consultants who attended the second information gathering meeting had access to many (but not all) of the reports cited in Appendix F. Most significantly, none of the consultants had access to the predecisional draft of the Highly Enriched Uranium Task Force report (DOE, 1992b) discussed in Chapter 5.

The opinions, findings, conclusions, and recommendations provided in these reports represent the views of the consultants and do not necessarily represent the views of the P.I. or the National Research Council.



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--> Appendix D Consultant Reports Note: The principal investigator (P.I.) consulted with several expert consultants in the course of this study. At the request of the P.I., the following reports were contributed by 11 consultants who were invited to attend the second information-gathering meeting and two consultants who did not attend that meeting (see Appendixes B and C). Biographical sketches of the consultants are included in Appendix E. The consultants who attended the second information gathering meeting had access to many (but not all) of the reports cited in Appendix F. Most significantly, none of the consultants had access to the predecisional draft of the Highly Enriched Uranium Task Force report (DOE, 1992b) discussed in Chapter 5. The opinions, findings, conclusions, and recommendations provided in these reports represent the views of the consultants and do not necessarily represent the views of the P.I. or the National Research Council.

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--> Topic: Proliferation Aspects of the Treatment Options Consultant: Harold Agnew Once the material is under U.S. or DOE custody under today's management, I see no proliferation risk from any of these spent or reprocessed fuel forms. Unless the enriched material is needed it will be a waste of money and resources to reprocess it under any proliferation scenario. Reprocessing will also add to our present waste disposal problems. Proper containment should be the only objective in any reprocessing or repackaging. For the longer term, if it were to be placed in the center of an overpack surrounded by five canisters of high-level vitrified waste and buried in the repository among thousands of overpacks of spent commercial reactor fuel and other canisters of vitrified waste I would not consider it a credible proliferation target. As an aside, if one worries about proliferation using enriched uranium, one should be concerned about Deputy Aleksadr Belosokov's statement reported in the 12/12/97 (p. A3) New York Times that Russia will scrap the contract to sell processed enriched uranium from weapons to the United States Enrichment Corporation and make the material available worldwide. With the USSR/Russia starting to renege on its sale of 500 metric tons of HEU235 (approximately ten times the amount of research reactor fuel) the question of final form for spent aluminum clad or alloyed research reactor fuel is moot. In my opinion, none of the proposed "recycled" forms are less or more susceptible to proliferation. In fact, I believe the less the fuel is "massaged" the better. Reprocessing will result in "muffs." It will be easier to account for the material if it is stored in its original form. It will save reprocessing costs, and accounting costs and produce no wastes, so it will be environmentally better. The real worry will be if Yeltsin leaves: commerce between Russia and Iran, Iraq, Pakistan, and others will be the real concern. HEU from Russia's reserves and from stockpile reductions are enormous. The issue of spent aluminum-clad or alloyed research reactor fuel is not worth considering with regard to changing its physical form. Leave it alone and account for it. The more you handle material, the greater the chance for mischief.

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--> Topic: Proliferation Aspects of the Treatment Options Consultant: John Ahearne The principal proliferation concern relating to the aluminum-based fuels is that many are highly enriched uranium (HEU). For example, according to a DOE document, one shipment has over 100 kg of 93% enrichment (Savannah River Site FY97 Spent Nuclear Fuel Interim Management Plan, WSRC-RP-96-530, 21 October 1996, p. C-2). The significance is that "typical weapons-grade uranium is more than 90 percent U-235" (Management and Disposition of Excess Weapons Plutonium, National Academy Press, 1994, p. 30). Although recent concerns have focused on plutonium, HEU may be of greater concern because "plutonium can only be used in implosion weapons." However, "[h]ighly enriched uranium (in weapons, typically 90 percent U-235 or more) can be used in either gun-type nuclear weapons designs like that used at Hiroshima or in the more efficient implosion design" (Management and Disposition of Excess Weapons Plutonium: Reactor-Related Options, National Academy Press, 1995, p. 43). Thus, the spent fuel from many of the reactors, with HEU still a large part of the fuel, is a serious proliferation risk—IF taken by a group that would be able to extract the uranium from the highly radioactive fission products. The protection provided by this radioactivity is the basis of what the National Academy has called "the spent fuel standard" (ibid., 1994, op. cit., p. 12). So long as the disposal option keeps the fission products with the HEU, this protective barrier remains. However, several of the options appear to include separation of the uranium, at least at one or another stage in the process. For example, in the electrometallurgical treatment, which some have recommended be retained as "a secondary and diverse backup" (Technical Strategy for the Treatment, Packaging, and Disposal of Aluminum-Based Spent Nuclear Fuel: A Report of the Research Reactor Spent Nuclear Fuel Task Team, Vol. 1, June 1996, p.78), the uranium is separated out (see flow diagram on p. 42, ibid., and several recent NRC reports focused on this process). Although the plan here apparently is to mix in depleted uranium to blend down the HEU to LEU, the process does permit separation of nearly pure weapons-grade uranium. This would at a minimum require substantially tighter safeguards than other processes under consideration. Unless one of the options is chosen that blends down the HEU with depleted uranium,

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--> preferably without separating the HEU at any stage of the process, the issues surrounding disposal are essentially the same as those for the weapons-grade plutonium treated in the referenced NAS reports. This includes recognition that some of the more intense radioactive materials relied upon for self-protection have short half-lives (decades) so that if the spent fuel is in storage for many decades, the self-protection is weakened, increasing the need for safeguards. Compared with commercial spent fuel, this fuel presents greater proliferation risks because of the HEU. If the fission product load is sufficient to match the spent-fuel standard for commercial fuel, then the risk for this research fuel would be basically the same as for the commercial fuel. (The proliferation concern with commercial fuel is the plutonium produced during power generation.) The safeguards proposed for commercial fuel would be necessary if the radioactive protective barrier is maintained for the aluminum-based research fuel. Diluting the HEU with depleted uranium would reduce the proliferation hazard and, depending on how the dilution was accomplished (i.e., actual mixing would be required), could reduce the safeguards required.

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--> Topic: Nuclear Criticality Safety Consultant: Francis M. Alcorn This report is tendered by Francis M. Alcorn based on the following: Participation in a technical meeting held in Augusta, Georgia on December 2 and 3, 1997. On Tuesday, December 2 there were ten technical presentations on various aspects of the project. Only one of these dealt directly with criticality issues. That presentation, by Peter Gottlieb of TRW, addressed phases 1 and 2 (out of 3) for a co-disposal waste package in a repository; Peter's work is sponsored by the Office of Civilian Radioactive Waste Management (OCRWM). On Wednesday there was additional dialog with Peter. Review of approximately 14 documents. Of these, four sets dealt with criticality issues: Volumes I and II of "Technical Strategy for the Treatment, Packaging and Disposal of Aluminum-Based Nuclear Fuel" (June 1996). "Alternative Aluminum Spent Nuclear Fuel Treatment Technology Development Status Report (U)," WSRC-TR-97-00345(U) (October 1997). Six documents by OCRWM with the same primary report number of WBS: 1.2.2/QA.L; one of the reports was dated August 1997, while the other five were dated September 1997. "Technical Strategy for the Management of INEEL Spent Nuclear Fuel" (March 1997) was reviewed; although it discussed criticality issues, it was of marginal value to this review. Instructions from the Principal Investigator. These instructions were: Focus only upon the processing or preparation options, the resulting waste form properties, the canister as it might be affected by the waste form (canister qualification for the repository is outside of this review), and interim storage plus any incremental effects that might occur in the repository due to the addition of the waste form. Respond to two specific questions: What are the significant criticality issues that must be considered during processing, interim storage after processing, and

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--> shipment to a repository? Has DOE adequately addressed these issues? Do any of the waste forms produced by the alternatives pose significant internal or external criticality hazards in a repository? My general observation, based upon Peter Gottlieb's presentations and review of the six OCRWM reports, is that the Operating Contractors for OCRWM are performing very detailed and thorough criticality safety evaluations. The Savannah River status report mentioned above (2.b) is, from a criticality perspective, primarily a status report of OCRWM activity to date (section 4.5). Although one cannot argue with the technical quality of this work, the following should be noted: This work primarily addresses canister performance in a repository (which is somewhat outside of the requested focus) and satisfying 10 CFR 60; the work claims to have completed only two of three phases for the canister if the co-disposal waste option is assumed; and it is obvious that some of this work must be repeated since both the Peter Gottlieb presentation and the Savannah River status report talk of investigations in progress to select an appropriate neutron poison material for the co-disposal canister. It also appeared that the OCRWM contractors have performed a significant amount of criticality evaluations, while DOE/Savannah River staff has done very little in criticality evaluations to support processing, interim storage, and shipping for each of the alternatives. For the scope of this review it would appear, in my opinion, that Savannah River staff should have made the presentation on criticality. In my opinion the Savannah River staff is technically capable to complete their part of the project. Because so little had been done, criticality wise, for that part of the project under review it was difficult to do an appropriate review. In response to the first specific question: Has DOE adequately addressed the significant criticality issues that must be considered during processing, interim storage after processing and shipment to a repository?

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--> There is no evidence that DOE has addressed the shipping question. Given that the packages must meet 10 CFR 71 requirements including accident testing, quality assurance plans, transport configurations and criticality evaluations that must be based on different assumptions than those required by 10 CFR 60, it is prudent that DOE consider the many aspects of shipping before canister designs are set. There is similarly no mention of interim storage after processing; however, if the waste is to be stored in the repository canisters and the canisters are safe in the repository then it is reasonable to assume that the canisters are acceptable for interim storage. DOE needs to articulate this assumption if it is the basis for interim storage. The Savannah River status report has a one-paragraph criticality statement (5.2.3.3), which acknowledges that criticality stability still must be explored for the melt-dilute option; this is a start but there is insufficient information to judge the adequacy of criticality considerations during processing. It is my judgment that the transportation issues and justification that the canisters with their contents meet shipping requirements are the most important issue not addressed by DOE at this point. Road accidents might be more limiting than long term survival in a repository. A criticality accident on the road would be much more visible and could potentially have a greater health/environmental impact than a criticality event in the repository. To these ends, DOE must assure that canisters are acceptable not for only the repository, but also for interim storage as well as for transportation. The Savannah River status report (2.b above) makes a somewhat disturbing statement on page 4.52 in section 4.5.1. That statement is: ''The computer codes that we use for criticality calculations for disposal will require benchmarking and/or validating the code." Without validated computer codes and the cross section sets used with those codes, there is no basis on which to proceed with defensible safety evaluations. Also it must be noted that the status report does not identify cross sections being used with the two identified codes. The validity of the cross section sets used may become an issue in dealing with some of the exotic materials being considered (especially the wide range of neutron poisons under study).

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--> In response to the second specific question: Do any of the waste forms produced by the alternatives pose significant internal or external criticality hazards in a repository? Information presented by Peter Gottlieb as well as information in the Savannah River status report indicates that the current envisioned canister designs will require neutron poisons for direct disposal as well as for co-disposal with both High Enriched U-Al fuel (e.g., Massachusetts Institute of Technology (MIT)) and for Low Enriched U-Si-Al fuel (e.g., ORR). The actual neutron poisons to be used are still to be determined; however, several candidate materials have been identified. It appears that a stronger neutron poison will be required (e.g., Gd for high enriched fuel) than for low enriched fuel (e.g., borated stainless steel). The relative dilution rates of fuel versus neutron poison over a long period pose a vexing problem, given that the actual neutron material is still to be selected. Adding to this, the potential cross section validation problems with certain neutron poison materials as well as the quality assurance problem of misloading a canister with the wrong poison or the wrong type of fuel in a given canister, raise the question of the desirability of this alternative over the melt-dilute alternative. The melt-dilute alternative can be designed to remove any requirement for neutron poison material and likewise render relative dilution rate problems easier to define and defend. In my opinion, required use of a neutron poison material with both the direct disposal and the co-disposal alternatives represents a significant criticality hazard for the repository—a hazard that can be eliminated by use of the melt-dilute alternative. The dilute-melt will pose additional consideration for criticality during processing; however, the processing system can be designed with positive and monitored control. The internal canister basket would lend itself to a design with easier quality assurance requirements. It is my opinion, based upon the studies completed to date, that the melt-dilute alternative poses less of a criticality hazard in both shipping and the repository than does either direct disposal or co-disposal. Press and dilute/poison was mentioned as another highly attractive alternative; however, almost no information was presented and intuitively this alternative appears to be less attractive from a criticality perspective

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--> than melt and dilute. Press and dilute would probably be preferred to either direct disposal or co-disposal because it could be carried forward without need of a neutron poison. None of the other nine alternatives were considered as part of this review.

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--> Topic: Cost and Schedule Consultant: M.W. Angvall The Principal Investigator posed six questions concerning cost and schedule. These questions and the corresponding responses follow. Information was gathered from various references as well as briefings and meetings held in Augusta on December 2 and 3, 1997, and from revised cost data provided after the December 2 and 3, 1997, meetings. Question 1: Are the cost data provided by DOE reasonably complete and transparent? Response: The revised costs provided by DOE appear to be complete. The cost estimates were constructed using the major cost drivers, which together make up the full costs of SNF handling, conditioning, packaging, storage, and disposal for each treatment technology. The estimates were logically constructed using scaling factors, inflation, and project contingencies in a reasonably judicious manner. Financing costs were added as were IAEA implementation costs and NRC licensing costs. Upon reviewing the backup provided in the revised cost study,1 we can say the costs are reasonably transparent. Question 2: Are the cost and schedule estimates developed by DOE for the alternative processing options suitable as a basis for comparison and selection of one or more preferred alternatives? Response: The cost estimates are suitable as a basis for comparison and selection of one or more alternatives. The schedules have been upgraded to reflect reasonably realistic dates and would appear to support the detail work required to further refine the selected technologies. 1   Krupa, J. F. and Carter, J. M., Savannah River Site Aluminum-Clad Spent Nuclear Fuel Alternative Cost Study, Rev. 1(U).

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--> Question 3: Are the cost and schedule estimates developed by DOE for the alternative processing options suitable for budget planning purposes? Response: The cost and schedule estimates developed for the alternative processing options are not suitable for budget planning purposes. The transfer, storage, and treatment facility cost estimates for the direct co-disposal facility and for the melt and dilute facility were prepared based upon a preconceptual design estimate. However, the estimates for the other technologies as well as the remaining cost factors are of parametric or rough order-of-magnitude quality and cannot be considered accurate forecasts of actual financial requirements. The schedule estimates are based upon assumptions as to delivery of aluminum-based SNF shipping casks and aluminum-based SNF assemblies to Savannah River; upon projected dates at which the various technologies could be available using a privatization approach (which to date has not been successful) for the transfer, storage, and treatment facility costs; and upon the date on which the repository will be ready to accept shipments. Any significant slippage in any one of the assumed dates could have major cost ramifications for all of the technologies and could affect some technologies more adversely than others. However, the revised schedules are probably sufficiently realistic to be used to develop estimates for budget planning purposes. Question 4: Has DOE considered the costs of program delays in its budget development or budget planning for this program? Response: It would appear that no costs for program delays have been included in the cost estimates. All of the estimates have been based upon the revised dates for shipment to Savannah River, transfer within Savannah River, funding, start-up of new treatment facility operations, and shipment to the repository. Delays in any of the assumed dates will have a negative cost effect on the estimates. Project contingencies were assigned to quantify the uncertainty associated with the implementation of each SNF Technology. This contingency addresses such things as

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--> released at the melting temperature (850-1000 °C) and collected. Melting the fuel gives a large volume reduction and the number of canisters required will be reduced appreciably compared with the D-D product (by factors of about 4 x depending on the fuel). The cast product will have a low surface/volume, and the microstructure of the ingots will be more uniform than the heterogeneous D-D fuel. Two principal concerns with the D-D option are proliferation and criticality. M-D processing can remove both of these by dilution with unenriched uranium and aluminum. From a corrosion viewpoint, the M-D gives a product whose behavior is much more predictable in considerations of long term corrosion and whose composition could be optimized for such stability. One of the current research efforts of the WSR materials staff is study of the long term integrity of the waste form in water. In this vein they may have as a goal varying the composition to optimize the long term integrity of the waste on long term (10,000 yr.) exposure to water. There are a variety of fuel types and geometries that must be handled in this program and these elements possess a limitless variety of histories. Characterizing these in enough detail to assure suitability for Direct Disposal can be time consuming, and melting is an excellent way to assure the uniformity of the product and reliability in processing. The WSRC people have melted very few, if any, irradiated fuel elements, but they and others have melted a great deal of fuel for manufacturing Al-U fuel elements. Also, fission product release has been measured in severe accident studies [3]. Thus it would appear that the information needed for designing and building a facility for the melting and casting of these fuel elements is in hand and that the process could be put into operation with few if any surprises. Dry Storage (Interim Storage). After the aluminum clad fuel has been processed for D-D or M-D it will be sealed in a canister with an inert atmosphere, which will probably be dried air. Aluminum forms a protective oxide film under these conditions and there would be virtually no measurable reaction of the waste form with the atmosphere or the canister for the years or decades that the waste may wait for placement in Yucca Mountain. The formation and stability of this oxide on various aluminum alloys is well established in the technical literature for temperatures near room temperature [4] and has been expanded to cover

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--> the product of the M-D process and higher temperatures by work at WSRC [5]. Answers to Assigned Questions 1.   Are DOE's plans for fuel handling, drying, etc., technically credible? Yes. DOE's plans for fuel handling, drying, and interim storage are technically credible and the process steps are adequate to prevent significant fuel corrosion. 2.   For each of the processing options evaluated by DOE, are processing steps technically credible? Yes, for the two processes under serious study, namely Direct Disposal and Melt Dilute. However, WSRC is not trying to prepare a basis for all of the processes considered in the report of the Research Reactor Task Team Study (Jack DeVine, Chairman). 3.   Would filling the canister void space with aluminum or some other sacrificial material make any difference in long-term corrosion of the waste form? No, but it is quite likely that the cast product resulting from the Melt-Dilute process would have materially better corrosion resistance than that of the Direct-Disposal product. This is a topic currently under study. 4.   Will any of the waste forms resulting from any of the alternative processing operations be likely to increase internal corrosion of a standard repository container? No. 5.   What is the status of R&D activities at Savannah River on the Melt-and-Dilute and Co-Disposal options? Are the R&D activities appropriately focused? The metallurgical process information needed for the M-D and D-D processes is well in hand. The research activities needed for this have been well focused.

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--> References 1. Sindelar, R.L., et al., "Acceptance Criteria for Interim Dry Storage of Aluminum-Alloy Clad Spent Nuclear Fuels," March 1966, WSRC-TR-95-0347. 2. "Alternative Aluminum Spent Nuclear Fuel Treatment Technology Development Status Report," October 1997, WSRC-TR-9700345(U), Sec. 3.2-3.3. 3. Howell, J.P., "Fission Product Release from Spent Nuclear Fuel During Melting (U)", WSRC-TR-97-0112 (U). 4. Godard, H.P., "Oxide Film Growth Over 5 Years on Some Aluminum Sheet Alloys in Air of Varying Humidity at Room Temperature," J. Electrochem. Soc., 1967, v. 10, p. 354. 5. Lam, P.S., R.L. Sindelar, H.B. Peacock, Jr., Vapor Corrosion of Aluminum Cladding Alloys and Aluminum-Uranium Fuel Materials in Storage Environments, WSRC-TC-97-0120.

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--> Topic: Cost and Schedule Consultant: Richard I. Smith In November of 1995, the Department of Energy (DOE) established the Research Reactor Spent Nuclear Fuel Task Team to assist in developing a technical strategy for interim management and ultimate disposition of the foreign and domestic aluminum-based research reactor spent nuclear fuel in DOE's jurisdiction, including both current inventory and expected receipts. The Task Team developed a two-volume report titled Technical Strategy for the Treatment, Packaging, and Disposal of Aluminum-Based Spent Nuclear Fuel [DeVine et al., 1996], issued in June of 1996. Subsequently, DOE contracted with the National Research Council (NRC) of the National Academy of Sciences to review the set of technologies evaluated in the Task Team report and suggest other alternatives that DOE might consider; to examine the waste package performance criteria developed by DOE for aluminum-based spent nuclear fuel and suggest other factors that DOE might consider; and to assess the cost and timing aspects of each of the disposition strategies proposed by DOE. To facilitate this review, the NRC assembled a team of experts in the fields of nuclear criticality, nuclear proliferation, cost and schedule, corrosion and metallurgy, processing and remote handling, and regulatory/waste acceptance. Copies of the Task Team report were provided to the experts selected to participate in the review, a two-day meeting was held in Augusta, Georgia on December 2 and 3, 1997, where the Task Team report was presented by its authors and additional presentations were made by various staff from the Savannah River Site (SRS) on progress toward implementation of the various strategies since the Task Team report was prepared. Each of the groups of experts assembled by the NRC was posed a set of questions about the proposed strategies specific to its areas of expertise, to be answered from the information contained in the Task Team report, gathered at the Augusta meeting, and from any other sources available. This appendix is focused on the cost and schedule aspects of the problem.

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--> General Comments It was quite apparent from the strategies presented in the Task Team report that DOE's intent was to find ways to dispose of the aluminum-based spent fuel without recovering any of the residual highly enriched uranium from the spent fuel assemblies, presumably for reasons of non-proliferation. Reprocessing was not evaluated in the report and compared with the other alternatives, despite the fact that reprocessing of this type of fuel was presently ongoing and any comparison of alternatives should (must) include the possibility of continuing the current method of dealing with the spent fuel. As a result, the strategy with the highest probability of success, with the best-defined costs, and with a resultant waste product that is assured of repository acceptance, was not evaluated in the analyses. In a subsequent report [Krupa 1997], several strategies have been devised that include reprocessing of the current inventory of spent fuel through about 2010 and applying some other treatment process to those fuels that enter the inventory in later years. As might be expected, those strategies result in completing the disposition of the bulk of the anticipated inventory of aluminum-based fuels in the least time, with the least cost, and the highest probability of success. Specific Comments Six questions about the Task Team report were posed to the cost and schedule experts by the Principal Investigator for this study. Each question is presented and this reviewer's responses and discussions are given in subsequent subsections. Question 1: Are the cost data provided by DOE reasonably complete and transparent? Response: In general, the answer is yes. The detailed costs associated with each strategy are presented in Appendix C of Volume II of the Task Team report. The costs are broken down sufficiently far to see which elements are important to the result and which elements are

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--> common to all strategies. Generally, the bases (sources) for the various cost elements are given, and the rationales for various assumptions made are also given. While one might disagree with some of the assumptions or cost values, those used are well documented. The data presented represent the state of knowledge at the time of the report. However, some of that data has been superseded by more recent cost evaluations [Krupa 1997]. Question 2: Are the cost and schedule estimates developed by DOE for the alternative processing options suitable as a basis for comparison and selection of one or more preferred alternatives? Response: Yes and No. For those alternatives considered in the report, the data presented are probably sufficient for comparisons to be made and to select one or more preferred alternatives. The cost bases are generally internally consistent across the alternatives, the processes of each alternative are examined in sufficient detail to assure that no major cost elements have been overlooked. However, because continued reprocessing was not included in the analyses, there is no real basis for comparison between current practice and future possibilities. There is an old axiom in the cost estimation business: ''The less you know about a given process, the cheaper and easier it appears." Some of that phenomenon has likely occurred in the estimates for those processes for which little or no development or demonstration work has been carried out. Some of the uncertainty estimates for certain aspects of some alternatives seem rather large, but they may only reflect the state of knowledge at the time of the report. For example, the uncertainties assigned to the electrometallurgical (EM) treatment process are much larger than all processes except the GMODS process. The basic EM process had been demonstrated for other types of spent uranium fuel. Since that time, lab-scale development testing for the more complicated aluminum-removal process has been completed, and the developers are ready to proceed to engineering-scale development testing [Slater 1997]. Thus, confidence in success in developing the Al-U process would appear to have increased and the uncertainty in project costs would appear to have decreased, relative to those processes in the Task Team report that have not been demonstrated. It is not obvious that the schedules contained in the Task Team report are achievable. In general, the key milestones were established by

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--> DOE, without a bottom-up examination of the program elements necessary to reach those milestones. No consideration was given to the time required to establish a line-item in the DOE budget for construction of any new facilities nor to the time required to select contractors for design and construction of those facilities. As a result, most of the schedules are optimistic by several years as a minimum. Since delays in construction and operation of the required new facilities will require extended utilization of currently used water basins, total program costs will increase for each year of delay. Similarly, some of the processes have had little or no development work done. Any delays in developing and implementing the processes will also delay the program, with attendant cost increases. These types of delays may affect some alternatives more than others. Question 3: Are the cost and schedule estimates developed by DOE for the alternative processing options suitable for budget planning purposes? Response: No. The milestones artificially imposed on the Task Team considerations preclude using those schedules for developing budget estimates. They ignore the time necessary to place a project into the DOE line-item budget and the time (and money) necessary to select an architect-engineer and a construction contractor. They also ignore the time (and money) required to prepare and issue an environmental impact statement, or an environmental assessment, if such are necessary for these projects, and ignore the time (and money) needed to deal with satisfying Nuclear Regulatory Commission reviews and possible licensing of any new facilities or processes. At least several years would be added to the schedules outlined in the Task Team report, and extending the period during which the wet basins are needed for storage and handling of spent fuel will also add significantly to the overall project life-cycle cost. While these schedule extensions will increase the cost of the proposed projects, they are generally common to all alternatives (except perhaps continued reprocessing) and would not significantly affect the comparison between the alternatives presented in the Task Team report.

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--> The most recent cost analyses for the proposed alternatives [Krupa 1997] do take into account at least most of the above-described schedule delays and are more nearly suitable for preparing long-range budget estimates. Question 4: Has DOE considered the costs of program delays in its budget development or budget planning for this program? Response: No and Yes. The effects of program delays were not seriously considered in the Task Team report [DeVine et al., 1996], since such delays were generally common to all alternatives and did not affect the comparisons. The extended schedules considered in the most recent program cost analyses [Krupa 1997] are reflected in the projected program costs. However, no costs are included to reflect further technical development efforts on undemonstrated technologies. Apparently, these types of activities are being funded from other sources. Also, no schedule allocations are made to accommodate such development efforts. Any technical difficulties in proving out a selected treatment process could result in additional schedule delays. Question 5: Are the cost and schedule estimates for implementing the alternative processing options consistent with DOE procedures and systems? If not, has DOE identified what changes must be made to achieve its cost and schedule targets? Response: The first part of this question is essentially a restatement of Question 3 and is discussed there. The response to the second part of the question is not clear. Apparently DOE has not yet decided how to fund the project, either by privatization or by the budget line-item project approach. Both approaches require a significant amount of lead time to establish the appropriate contractual arrangements with contractors. To achieve the rather short schedules currently proposed, the project will have to be highly organized and tightly controlled, with the authority to make necessary decisions held at the local (site) level. It is not clear that DOE has yet made the decisions necessary to allow the project to go forward in an optimum fashion nor that it will make those decisions any time soon.

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--> Question 6: Are the cost and schedule milestones that are laid out in the Research Reactor Task Force Report for selecting and implementing an alternative processing option being met? Response: Difficult to Predict. The schedule for alternative selection in the Task Team report called for a decision late in 1999. Some winnowing of the alternatives originally recommended for further study has already occurred in that development activities on the press and dilute option have essentially been suspended for lack of funding. Development on the melt and dilute option is progressing reasonably well. Lab-scale development for the EM process has been completed and engineering-scale development is scheduled to start soon. It is not yet clear that the various co-disposal approaches will be able to qualify for repository acceptance, so those alternatives may be in doubt. The initial 6-8 years or so of the three reprocessing alternatives suggested by Krupa [1997] obviously can be implemented as quickly as space is available in the H-Canyon reprocessing schedule, although exactly which process should be utilized for the low-throughput period following closure of the H-Canyon reprocessing facility in 2010 is not clear. One possibility not yet considered by DOE would be to install a relatively low-throughput aluminum-removal stage of the EM process in the same hot cell that is presently occupied by the EM process being used currently for EBR-II fuel at INEEL. The uranium feed stream from the aluminum-removal step would feed directly into the existing uranium refining process to complete the separation of the uranium from the residual fission products. This approach would avoid the construction of any new facilities at all and require only addition of the incremental equipment for the aluminum-removal step to the existing hot cell system. However, for best economics, an ongoing mission for the existing uranium EM process would be needed (e.g., treatment of the N-Reactor fuel from Hanford prior to repository disposal), because the cost per unit of fuel processed for facility operations might be rather high if only the aluminum-based fuel stream were being processed, because the facility would have to be maintained ready for service even when there was no aluminum-based fuel in inventory.

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--> Other Comments The most recent analyses [Krupa 1997] show that the continued reprocessing of the aluminum-based spent fuel in the H-Canyon facility at Savannah River Site is the most cost-effective approach and can eliminate the existing inventory from both SRS and INEEL in the least time, with the greatest certainty of success (i.e., a guaranteed repository-acceptable waste form) and recovery of a valuable resource (highly enriched 235U) to be blended down for future use in our domestic nuclear power industry. Unfortunately, DOE has shown a tendency in the past to prematurely close existing reprocessing facilities before their missions were complete, apparently for the purpose of satisfying some non-proliferation policy desires and to gain the approval of those members of the public who are opposed to nuclear power in general and to reprocessing in particular. As a result of such premature shutdowns at INEEL and at Hanford, DOE now has a large inventory of residual aluminum-based spent fuel stored at INEEL and a large inventory (about 2,300 tons) of spent metallic uranium fuel from the final years of N-Reactor operation stored at Hanford in wet pools where it is slowly corroding into sludge. Because of the safety implications of a pool leaking into the Columbia River, DOE has had to establish a major program, which has been underway for the past five years or so to remove this fuel from the pools and place it into dry storage away from the river. The most recent project cost estimate is now $1.08 billion, with completion still several years away. The final product of this project will be metallic fuel elements stored in steel canisters, a product unlikely to be acceptable to the repository without further treatment before disposal, so the total cost of preparing this material for disposal will certainly exceed the current estimate by a significant amount. Continuing to operate the Hanford reprocessing facility (PUREX) instead of shutting it down, and reprocessing all of that material into separated fuel material and fission product wastes would have cost about $300 million to $400 million and required about 3 years of operation. Contrasting those fairly well-known costs and schedule with the presently estimated (and still uncertain) cost of $1.08 billion over 7-8 years for the current project suggests that the decision to close PUREX before its mission was completed was a major mistake. DOE is again faced with making decisions related to the aluminum-based fuel disposition program that are similar to the PUREX

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--> and INEEL decisions (i.e., to shut down an existing reprocessing capability before the mission has been completed to satisfy some policy desires related to non-proliferation or to continue reprocessing until the inventory of aluminum-based spent fuel has been eliminated). All of the analyses to date show that continued reprocessing is the best, fastest, cheapest, and most certain of success of all of the alternatives considered. I trust DOE will not allow the somewhat tenuous non-proliferation policy considerations to reject the path forward that is technically and economically the best. References Bailey, R.W., and M.S. Gerber, Purex/UO3 Facilities Deactivation Lessons Learned History, HNP-SP-1147, Rev 2, Fluor Daniel Hanford Company, Richland, Washington, 1997. DeVine, J. C., et al., Technical Strategy for the Treatment, Packaging, and Disposal of Aluminum-Based Spent Nuclear Fuel, a report of the Research Reactor Spent Nuclear Fuel Task Team. U. S. Department of Energy, Washington, D.C., June 1996. Krupa, J. L., Savannah River Site Aluminum-Clad Spent Nuclear Fuel Alternative Cost Study , Rev 1(U), WSRP-RP-97-299 REV. 1, Westinghouse Savannah River Company, Aiken, South Carolina, December 1997. Slater, S. A., and J. L. Willit. Electrometallurgical Treatment of Aluminum-Based Fuel, presented at the Augusta, Georgia review meeting, December 2-3, 1997. Westinghouse Hanford Company, PUREX/UO3 Standby Management Plan, WHC-SP-0631, Rev. 1, 1991.