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Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal (1998)

Chapter: Appendix D. Consultant Reports

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Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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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.

Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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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.

Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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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,

Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
×

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.

Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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Topic: Nuclear Criticality Safety

Consultant: Francis M. Alcorn

This report is tendered by Francis M. Alcorn based on the following:

  1. 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.
  2. Review of approximately 14 documents. Of these, four sets dealt with criticality issues:
    1. Volumes I and II of "Technical Strategy for the Treatment, Packaging and Disposal of Aluminum-Based Nuclear Fuel" (June 1996).
    2. "Alternative Aluminum Spent Nuclear Fuel Treatment Technology Development Status Report (U)," WSRC-TR-97-00345(U) (October 1997).
    3. 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.
    4. "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.
    1. Instructions from the Principal Investigator. These instructions were:
      1. 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.
      2. Respond to two specific questions:
        1. What are the significant criticality issues that must be considered during processing, interim storage after processing, and
  • Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×
    • 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:

    1. This work primarily addresses canister performance in a repository (which is somewhat outside of the requested focus) and satisfying 10 CFR 60;
    2. the work claims to have completed only two of three phases for the canister if the co-disposal waste option is assumed; and
    3. 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.
    4. 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?

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    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).

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    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

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    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.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    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).

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    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

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    equipment unknowns, complexity of process, and process integration.2 No program delay costs were included.

    Question 5: Are the costs 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 original schedule used in the Task Force Report was directed by DOE in the task force ground rules. This schedule did not permit the development of budget quality estimates. The revised schedules discussed at the Augusta meeting and those listed in the revised cost study would appear to allow development of estimates meeting DOE procedures. DOE appears to have made a policy decision concerning new Savannah River spent fuel management projects that has allowed development of schedules consistent with DOE procedures and systems.

    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: The schedule milestones presented in the Research Reactor Task Force Report are not being met. However, the schedules have been revised to more realistic dates, which have the possibility of being attained.3 The question remains as to the dates selected being sufficiently realistic considering all the basic assumptions made, and the number of policy-making parties involved.

    2  

    Krupa, J. F. and Carter, J. M., Savannah River Site Aluminum-Clad Spent Nuclear Fuel Alternative Cost Study, Rev. 1(U), p. 21.

    3  

    Krupa, J. F. and Carter, J. M., Savannah River Site Aluminum-Clad Spent Nuclear Fuel Alternative Cost Study, Rev. I (U), p. 3.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Topic: Regulatory and Waste Acceptance

    Consultant: Robert M. Bernero

    Introduction

    The Department of Energy (DOE) is considering alternatives for the final disposition of aluminum-based spent nuclear fuel recovered from reactors here and abroad. The fuel, which contains high-enriched uranium (i.e., approximately 20-93% 235U), is of particular interest because of the potential for diversion and the higher expected rate of corrosion for aluminum.

    The National Research Council (NRC) of the National Academy of Sciences (NAS) was asked by DOE to perform a review of DOE's aluminum-based spent nuclear fuel disposition technologies. The evaluation would include the following: (1) examination of the set of technologies chosen by DOE and identification of other alternatives that DOE might consider; (2) an examination of the waste-package performance criteria developed by DOE to meet anticipated waste acceptance criteria for disposal of aluminum-based spent fuel and identification of other factors that DOE might consider; and (3) to the extent possible given the schedule for this project, an assessment of the cost and timing aspects associated with implementation of each spent nuclear fuel disposition technology.

    This study was undertaken using the Principal Investigator project model. The analysis here is by one supporting contributor evaluating the regulatory and waste acceptance approaches for final disposition of the aluminum-based spent nuclear fuel.

    Background

    Until the late 1970s the prevailing concept for final disposal of all high-level nuclear waste was that all spent fuel would be reprocessed and the resulting high-level waste stream would be somehow solidified to prepare the waste for long term disposal. The solidified waste, after reprocessing, would contain only small amounts of fissile isotopes, leaving the way open for specification of waste form performance criteria as the first line of protection against release and transport of the waste isotopes. In 1970 the Code of Federal Regulations, in 10 CFR Part 50,

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Appendix F, defined high-level liquid radioactive wastes as "those aqueous wastes resulting from the operation of the first cycle solvent extraction system, or equivalent, and the concentrated wastes from subsequent extraction cycles, or equivalent, in a facility for reprocessing irradiated reactor fuels." That regulation also limited a reprocessing plant's inventory of high-level liquid radioactive waste to that produced in the prior 5 years. That regulation also required shipment of the high-level radioactive waste to a Federal repository no more than 10 years after the waste is separated from the fuel.

    In the winter of 1976/1977, Presidents Ford and Carter decided that commercial reactor spent fuel would not be reprocessed, but would be disposed of directly as high-level waste. That decision for the once-through fuel cycle immediately changed the standard form of high-level waste, from waste solidified to meet a performance specification to the as removed form of commercial spent fuel, covering a wide range of sizes, burnup levels, etc.

    Later, the U.S. Nuclear Regulatory Commission (USNRC) promulgated regulations, 10 CFR Part 60, to require that these high-level wastes be emplaced in waste packages providing "substantially complete containment" for 300-1,000 years, in an engineered barrier system that releases less than 1 part in 100,000 after 1,000 years, in a geologic medium with a groundwater travel time of at least 1,000 years (10 CFR Part 60.113). In addition, in Part 60.112, the USNRC regulation requires the repository to meet the performance standards of the generally applicable environmental standard set by the Environmental Protection Agency (EPA).

    The original final disposal standard from the EPA, 40 CFR Part 191, set quantities of permitted release, by isotope, assessed at 10,000 years. These permitted release quantities are associated with health effects imputed to collective population doses. By the Energy Policy Act of 1992 (P.L. 102-486) the Congress set aside that EPA standard for Yucca Mountain, the first candidate high-level waste repository under investigation. Instead of that standard, Section 801 of the Act directed EPA to promulgate standards to ensure protection of public health from high-level radioactive wastes in a deep geologic repository that might be built under Yucca Mountain in Nevada. By this provision, EPA must set

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    the standards to ensure protection of the health of individual members of the public. The standards will apply only to the Yucca Mountain site. To assist EPA in this endeavor, Congress asked the National Academy of Sciences to advise EPA on the technical bases for such standards.

    The National Research Council Board on Radioactive Waste Management formed a Committee on Technical Bases for Yucca Mountain Standards. That Committee has issued its detailed report, Technical Bases for Yucca Mountain Standards (National Research Council, National Academy Press, 1995). The EPA standard 40 CFR Part 191 has since been revised and reissued applicable to the Waste Isolation Pilot Plant (WIPP) and high-level waste repositories other than Yucca Mountain, but EPA has not yet issued a revised Yucca Mountain standard.

    Regulatory Approach for Yucca Mountain

    Yucca Mountain is being investigated as a possible repository for high-level radioactive waste under the terms of the Nuclear Waste Policy Act of 1982 and the Nuclear Waste Policy Amendments Act of 1987. The repository, if approved, would be used principally for the disposal of commercial reactor spent fuel, with a statutory limit of 70,000 tons (from the 1987 Act). A decision by President Reagan permitted some of this capacity to be used for defense high-level wastes. It is assumed here that disposal of the aluminum-based spent fuel would be in Yucca Mountain, co-disposed with glass logs from defense high-level wastes. Therefore, the regulatory approach and waste acceptance basis for aluminum-based spent nuclear fuel are the same as those for Yucca Mountain.

    The USNRC regulation, 10 CFR Part 60, promulgates the phased regulatory approach required by the Nuclear Waste Policy Act of 1982. That approach consists of several phases: site characterization, license application (after Presidential notice to and approval by the Congress), construction authorization, license to receive waste, and finally license amendment for closure. At the present time, Yucca Mountain is still in the site characterization phase, with DOE acting on a Site Characterization Plan first submitted to USNRC in late 1988. That Site Characterization Plan was reviewed by USNRC following the procedures specified in 10 CFR Part 60.18, Review of Site Characterization Activities. By those procedures the USNRC does not approve the activities or thereby commit

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    to issuance of a license. Rather, the USNRC reviews the activities, raising questions, providing comments, or even raising objections in matters that appear to be so seriously deficient as to raise doubt about the possibility of future licensing. Thus, the first Yucca Mountain Site Characterization Plan was reviewed and received many comments with two objections, that the quality assurance program for data gathering was inadequate and that the repository design control process was not coherent (letter R. M. Bernero, USNRC, to S. M. Rousso, DOE-RW, dated July 31, 1989). Those objections and major comments have long since been resolved and the site characterization phase has continued with a clear public record of the characterization, continuing analysis, and review. Following the procedures of characterization phase review the USNRC does not specifically approve any data or analysis but does document its review thereof to reveal any objections, comments, or questions. The DOE is then able to address and resolve them with USNRC review and acknowledgment, but without explicit USNRC approval.

    Until the EPA has acted to promulgate a new performance standard for Yucca Mountain disposal, the DOE approach for performance assessment of Yucca Mountain is to conduct total system performance assessment, using state of the art methods, assessing performance against the standards likely to result, considering the advice of the Committee on Technical Bases for Yucca Mountain Standards. For example, performance assessments are not stopped at 10,000 years but are carried out over the Committee-recommended 1 million years (which indicate that peak projections appear beyond 10,000 years, at several tens of thousands of years).

    In another area of significance here, DOE has established a process for determining the acceptability of candidate forms of high-level radioactive waste. That process is laid out in DOE/RW-0351P Waste Acceptance System Requirements Document, December 1996. An important element of determining waste acceptability is the conduct of a performance assessment to determine that disposal of this particular candidate waste will not significantly affect the overall performance of the repository. In addition, if the candidate waste, like spent nuclear fuel, contains fissile material, the performance assessment must include a criticality safety analysis. The December workshop on this project was

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    told how DOE is now systematically analyzing the various forms of spent fuel in DOE possession, including the aluminum-based spent fuel, in this way. Consistent with the site characterization phase of Yucca Mountain activities, DOE is conducting this work with review and comment by USNRC, not seeking approval.

    Evaluation of DOE Aluminum-Based Spent Fuel Alternatives

    The evaluation of the alternatives being considered by DOE for management of aluminum-based spent nuclear fuel began with four questions:

    • Has DOE-Savannah River identified the appropriate criteria for aluminum-based spent fuel from the draft waste acceptance criteria document that has been prepared by the Yucca Mountain program?
    • Which of the waste forms is likely to be the most acceptable for disposal in a repository relative to commercial spent fuel and vitrified logs? For those waste forms that are unlikely to be acceptable, has DOE considered alternate processing options?
    • Are the waste acceptance criteria that have been identified by DOE suitable for selecting among the alternative processing options?
    • Is DOE-Savannah River making an adequate effort to stay current with changes in waste acceptance criteria?

    The evaluation here addresses each of these questions in order, beginning with the first. DOE-Savannah River has not attempted to identify the appropriate criteria for aluminum-based spent fuel acceptance unilaterally. Rather, it has agreed with and supported the expert team in the service of DOE's Office of Civilian Radioactive Waste Management (DOE-RW) to conduct the performance assessments needed for the waste acceptance process described in DOE/RW-0351P. That analysis is also available for review by the USNRC. Thus, DOE-Savannah River has identified the appropriate process for determining waste form acceptance and joined in that process.

    In considering the second question, it should be noted that there is some ambiguity in the DOE approach to simply reprocessing the aluminum-based spent fuel using the existing facilities at the Savannah River Site. The recovered fissile material could join other high-enriched

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    uranium streams in the DOE system, and the waste liquid could pass to the vitrification process to become ordinary glass logs, already established as a waste form. Instead of following this option, the DOE-RW and DOE-Savannah River waste acceptance process team has conducted detailed analysis of only one alternative, direct co-disposal. Direct co-disposal is the option where the intact spent fuel is loaded in a suitable canister, small enough in size to fit in the center of the proposed disposal array of five glass logs in a ring within a large canister. The direct co-disposal is evidently more attractive than the base case of direct disposal, because it uses the otherwise unused space in the center of the glass log canister and thereby obtains a substantial degree of self-protection from the surrounding waste logs. Direct co-disposal is also the evident limiting case for waste acceptance since it brings the unmodified corrosion characteristics and the higher nuclear reactivity of this spent fuel directly into the repository. The approach seems to be that if the direct co-disposal option yields an acceptable waste form, there is no need for further consideration of waste acceptance in weighing alternatives. That approach is reasonable as revealed by the consideration of the next question.

    The third question now simplifies to whether direct co-disposal is an acceptable waste form. The results of analysis so far obtained by the DOE-RW/DOE-SR team were presented at the December workshop. The team appears to be using appropriately conservative models to explore the alternative scenarios of corrosion and material slumping in order to discern the possibility of nuclear criticality. It should be noted that the team analyzed two types of this spent fuel as bounding, the fuel from the MIT Reactor and that from the Oak Ridge Reactor, each with its own tailored canister. Careful choice of and arrangement of poison in the canister for each specific fuel is important to obtain satisfactory results. The analytical results indicate that in situ criticality is highly unlikely, and we were told that if one did occur it would be a low yield, Oklo-type event, not a prompt criticality as suggested by some recent authors. The acceptability of these nuclear reactivity conclusions was linked to the expectation that the relevant USNRC regulation, 10 CFR Part 60.131(h), would be changed by USNRC rulemaking before repository licensing. This is a reasonable assumption since that regulation is a standard deterministic, double contingency requirement typically applied to fissile

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    material handling. It is my understanding that USNRC recognizes and agrees with the need to change this regulation to a probabilistic, performance assessment model for the repository.

    With the apparently acceptable nuclear reactivity conclusions, the team presented its current analysis results for all waste releases from the repository over a million year time period and the aluminum-based spent fuel contribution to those releases. The releases from the aluminum-based spent fuel are not significant; they are several orders of magnitude lower than the repository overall. Therefore, the results so far indicate that direct co-disposal is an acceptable waste form and any other option, if chosen, could yield an acceptable waste form.

    The fourth question really addresses where DOE is going from here. It is my impression, based on the presentations and discussions at the workshop, that reprocessing, direct co-disposal, or a combination of the two will be the options of choice. If direct co-disposal is pursued, the analyses should be completed and documented, especially the parts related to specific canister design and nuclear criticality analysis. The USNRC should be asked for detailed review of the analysis and for a commitment to the rulemaking to change 10 CFR Part 60.131(h).

    Summary

    DOE-Savannah River is following a reasonable regulatory approach for establishing the acceptability of waste disposal forms for aluminum-based spent nuclear fuel. It has nearly completed substantial analysis to demonstrate the waste acceptability of the bounding and most promising option, direct co-disposal. The work is not yet complete but the path to completion is clear.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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    Topic: Processing and Remote Handling Considerations

    Consultant: Joseph Byrd

    Introduction and Background

    Background and details of the aluminum-Based Spent Fuel Program were presented in a plenary session on 2 December. Materials that were presented are summarized below.

    In consideration of a 1992 DOE decision to phase out reprocessing activities throughout the DOE complex, a program was initiated to develop non-reprocessing technologies to be ready for implementation by 2000. DOE assembled a Task Team in November 1995 to develop a technical strategy for safe and economical handling, treatment, and disposal of aluminum-based SNF in hand and to be received. Eight non-reprocessing options along with the reference processing case were selected and studied. The options included direct disposal, co-disposal, dilution, and advanced treatments. The findings and conclusions of this Task Team were consensus-based using four weighted evaluation criteria: confidence in success, cost, technical suitability, and timeliness. Direct/co-disposal and melt/dilute options were recommended as alternatives to reprocessing.

    An Alternate Technology Program was initiated to move forward with development of the options recommended by the Task Team. Objectives and deliverables have been set through FY-98. All eight of the non-reprocessing options were discussed including reviews of waste forms, waste containers, and handling/processing issues. Results of the studies for metallurgy and corrosion issues and disposal criticality analysis were presented and discussed.

    Processing/Remote Handling Issues

    On December 3, small group breakout discussions were held to discuss details and remaining issues on specific topics. The groups concentrated on the topics specified in the technical expert assignments for the workshop.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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    The NRC project's principal investigator provided the following questions to be addressed for the processing/remote handling considerations. These specific questions and other issues were discussed in the breakout session with knowledgeable meeting participants including representatives from Savannah River Site, Argonne National Laboratory, and Idaho National Laboratory. Each of the eight technology options was discussed relative to the following questions. The participants were able to provide in-depth discussions relative to the following questions. The following answers to the questions are based on those discussions.

    1.  

    For each of the processing options evaluated by DOE, are the processing steps used as the basis for assessment and comparison (other than direct co-disposal) technically credible? That is, are they likely to work as described and produce the products and results assumed?

    Yes. The Task Team did a good job in its evaluation of the options. It considered a reasonable set of criteria on which to base its decisions and recommendations. Some options were eliminated and some added during the evaluation process. The Task Team was diverse.

    2.  

    Are there other processing options that should be considered by DOE for disposition of aluminum-based spent fuel?

    No other options that should have been considered are apparent to this reviewer.

    3.  

    Do the inner container designs appear adequate to contain the waste forms resulting from the various processing options? Are they over-designed for the intended application?

    The inner containers appear to be reasonable approaches for containment of the waste from various processing options.

    4.  

    Are DOE's basic material handling plans, pool use, and other facility needs reasonable, and are remote handling technologies available to meet these needs?

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

      Material handling for pools and repositories is not an issue. There are some remote handling issues related to some of the process options.

      Remote handling systems for all options would require some systems development (i.e. systems are made up of standard components but are not ''off the shelf' items). Remote handling equipment would require more development for the handling laminates for the Press/Dilute option, but proven technologies are available to meet this need. Significant remote handling problems are anticipated for the Plasma Arc Treatment option. More design information would need to be available before remote handling assessments could be made. No major problems are anticipated for the remaining options. Technologies are available to meet the remote handling needs. However, more design information would have to be available for all options before an assessment could be made on exactly how much remote handling system development would be required.

      A preliminary design has been done for a Transportation and Storage Facility that would accommodate any of the process options. The material handling and remote handling are assumed to be the same for getting materials into and out of whatever process is selected. This is a reasonable assumption when considering those processes most likely to be selected. Some chopping and material conditioning will be unique to the individual process and must be addressed in the development of those options (Plasma Arc Treatment, Chop/Dilute, and GMODS). The preliminary facility design uses proven methods for manipulators and overhead cranes. An electric cart system is used for transportation within the Facility. For the final design more attention should be given to remote handling. State-of-the-art Automated Guided Vehicles (AGVs) should be evaluated. Also, enhancements to remote handling are available to provide more effective and efficient operations: for example, graphical operator interfaces, simulator operation modes, telerobotic operations (combined teleoperator and semi-autonomous modes), and swing-free cranes. These technology enhancements should be

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

        

    considered while keeping in mind the overall objectives of efficiency and safety. Databases and database management should be incorporated to inventory and track materials through the entire process.

    5.  

    Are the technical requirements of the various alternative processing options sufficiently well defined so that reasonable judgments can be made about the likelihood of success of implementing them?

    Yes. The Task Team in its analyses adequately considered technical requirements.

    6.  

    Are there large differences in likelihood of success of implementing the various processing options?

    In general, some uncertainties such as the percentage of uranium that will be allowed will have an impact on most options. Lower percentage requirements will result in higher temperatures, off-gas problems, etc.

    The GMODS (Glass Material Oxidation and Dissolution System) and Plasma Arc Treatment options have more unknowns and would require significantly more development in order to confirm their likelihood of success. The Plasma Arc Treatment has anticipated high temperature problems and complex issues related to the shredder mechanism and process. The GMODS has been proved on bench scale mockup. However, lead in the off-gas and the shear mechanism/process into the melter will present problems.

    Handling and characterization will be problems with the Press/Dilute Option. More design work would be required.

    General Issues

    The following are general concerns about issues related to the project.

    • Political decisions appear to be the project drivers that override any technical considerations beginning with the 1992 decision to phase out all reprocessing operations throughout the weapons complex. This may make sense for consideration of new major facilities. However, this
    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×
    • decision has even affected the consideration of using an already existing facility, H-canyon at SRS, an option that would be most cost-effective.
    • Nonproliferation issues appear to be overemphasized for this problem. Unknown international "perceptions" appear to be serious concerns, although the research reactors' SNF handling, processing, and disposal are only the "tip of the iceberg" of the national waste problems. This appears to be a very small nonproliferation issue compared with the overall nonproliferation issues in total waste management, waste disposal, and waste storage. Emphases on past examples do not seem appropriate or necessary for this problem.
    • Remote handling systems development costs have not been factored into the costs associated with the technology options. This probably does not substantially alter the total costs for those options under consideration but may change the ranking since many of the cost estimates were very close together.

    Conclusions and Recommendations

    1.  

    Advanced technologies in remote handling. No major problems associated with remote handling for the options under consideration are anticipated. However, there were no indications that the latest technologies in remote handling were being considered in the facilities and processes designs. Basic manipulators, cranes, and electric carts can be greatly enhanced using proven advanced technologies such as swing-free cranes, telerobotic manipulators, graphical operator interfaces, system simulators, and advanced AGVs (automated guided vehicles). These technologies should be considered and evaluated for all appropriate phases of the process in order to assure the most effective and efficient operations.

    2.  

    Computer databases. There were no indications that advanced computerized databases were being considered for inventory and tracking of materials through the processes. It is recognized that the workshop did not specifically address this issue. However, the use of these systems is recommended to enhance security, safety, record keeping, reporting, and efficiency of the entire operation.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    3.  

    Uranium content in final product. There is still an undecided issue concerning allowable percentage of uranium in the final products. This issue should be settled as soon as possible so that technical uncertainties associated with this factor can be resolved.

    4.  

    Nonproliferation and criticality issues. Many nonproliferation and criticality issues are still unresolved. Since these major issues are option dependent, they should be resolved as soon as possible so that the process can be selected and designs can proceed in order to meet the desired schedule.

    5.  

    H-canyon option. Due to earlier political decisions, the use of H-canyon at SRS was not considered as a viable solution to the research reactor spent nuclear fuel problem. The use of H-canyon at SRS appears to be the most cost-effective and practical solution. Since budget issues are major concerns in the overall national environmental waste problem, the decision not to use the existing H-canyon should be reevaluated and reconsidered.

    References

    1. 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. Volumes I and II, U. S. Department of Energy, Office of Spent Fuel Management. June 1996.

    2. Alternative Aluminum Spent Nuclear Fuel Treatment Technology Development Status Report . Savannah River Technology Center, Strategic Materials Technology Department, Materials Technology Section. WSRC-TR-97-00345. October 1997.

    3. Savannah River Site: FY97 Spent Nuclear Fuel Interim Management Plan. Westinghouse Savannah River Company. October 1996.

    4. Technical Strategy for the Management of INEEL Spent Nuclear Fuel: A Report of the INEEL Spent Nuclear Fuel Task Team. Idaho National Engineering and Environmental Laboratory. March 1997.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Topic: Metallurgy and Corrosion

    Consultant: R.L. Dillon

    Introduction

    The emphasis in this report is on questions concerning options for the disposition of aluminum spent nuclear fuel (Al-SNF) raised by Milt Levenson, Principal Investigator, and Kevin Crowley, Study Director, in their "Instructions to the Technical Experts" (1). The instructions indicate that focus of the study shall be (a) dry storage of fuel and achievement of a road-ready status for the fuel package, (b) processing and preparation of optional waste forms and corresponding waste form properties, (c) interactions of waste form with the canister, and (d) review of the status of Co-Disposal and Melt-Dilute/Press-Dilute disposal options.

    The reference material made available to the review panel was researched and written up by a small team of scientists and engineers who appear as authors in various combinations. A given piece of work may be covered several times in various status and topical reports. In view of the interlocking authorship and repetitive reporting I have not felt it necessary to cite all the documents in which a given piece of information appears.

    Question 1a: Are the DOE plans for fuel handling, drying, and interim storage technically credible? Are these steps adequate to prevent significant fuel corrosion?

    Response: (1) Dry storage criteria for Al-SNF require the fuel remain in a condition mechanically suitable for "full safe retrievability" (2) for manipulations such as fuel rearrangement within a container, movement of fuel between canisters, adjustment of the gas environment, or withdrawal of fuel assemblies for examination or testing. Corrosion limits must assure that residual cladding has the requisite strength. In addition, processes that distort the clad-fuel material interface by blistering must also be avoided.

    These requirements have led to procedures that limit the reactant inventory within the container, after drying, to amounts that do not

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    significantly reduce effective cladding wall thickness. Limiting maximum cladding-fuel temperature to less than 200 °C in dry storage controls fuel blistering and creep (3, Sec 3.2).

    The effectiveness of the outgassing during the drying process has been shown, but there may still something to learn from tests on irradiated MTR fuel (3). Such tests are dependent on the availability of the Instrumented, Shielded Test Canister System (3). The response to the first part of questions 1 is that observance of corrosion limits during storage and drying and access to the Shielded Test Canister System for validation purposes will permit handling even with pitted fuel (2).

    Question 1b: Are these steps adequate to prevent significant fuel corrosion?

    Response: Laboratory studies have measured the effects of drying on residual moisture in a test chamber. Scaling to a fuel canister, quantitative consumption of the after-drying water vapor inventory does not approach corrosion or hydrogen limits. These extrapolations need validation that the Shielded Test Canister System will make possible. Studies performed at INEEL and Hanford have shown that mechanical pumps can outgas fuel and large nuclear components to water levels acceptable for dry storage of Al-SNF (4).

    Question 2: For each of the processing options evaluated by DOE, are the processing steps used as the basis for the assessment and comparison technically credible? Are they likely to work as described?

    Response: It is proposed here to consider only four of the eight options: Direct/Co-Disposal (3, Sec 1.1); Press and Dilute (5, App F), or Melt-Dilute (3, Sec 5.1), and the Electrometallurgical Option (5, App F). I have taken the cue from the Research Reactor Spent Nuclear Fuel Task Team who indicate in reference 3, Sec 6.2, that Direct/Co-Disposal may be regarded as the primary approach with Press and Dilute or Melt and Dilute as backup. It also recommend electrometallurgy as a secondary and diverse option. This report considers only these technologies.

    It is not the object in this discussion of processing options to undertake a general review. Comments are confined to four of the eight

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    options and to the corrosion implications of the process itself and of the waste products that emerge from the processing.

    • Direct/Co-Disposal Treatment (3, Sec 4.0): This process proposes to take fuel that has met criteria relating to its condition (cladding integrity, pit depth and diameter and, oxide thickness as it exists in a basin) and place it with other waste forms in an efficient array within a container. After outgassing to 4.6 torr to remove free water, the container is sealed for placement with other waste forms (e.g., glass logs in a canister) for eventual shipment to the repository. What is expected of the Direct/Co-Disposal Option is a method of treating Al-SNF using current technology, that will provide a waste form for safe interim storage, and will be road ready when regulatory approvals are obtained for shipment to the repository. The methodology depends on proven techniques or those sufficiently developed to be confidently applied to the waste form package (e.g., fuel drying), all at minimum cost. There are questions regarding the admissibility of the Direct/Co-Disposal product to the repository that require resolution.

      Regulatory problems regarding waste form admissibility to the repository are matters outside the scope of our charge. Beyond that, the technology is in place, and I see no corrosion reason not to put the process in operation.

    • The Direct/Co-Disposal waste treatment process does not stand alone. The Melt-Dilute (3, Sec 5.0) option offers some superior features, but it involves technologies proven only on a reduced scale. In this option, fuel assemblies are melted with additions of depleted uranium for isotopic dilution and additions of aluminum—and, where needed, depleted uranium—for a desired aluminum-uranium (Al-U) alloy composition. One advantage of the M-D approach is that it permits isotopic dilution if that is required. There are process options that remain to be considered with respect to melt pouring temperatures and methods and melt composition, which can improve the cast product for specific applications. Sampling during processing of the molten, homogenized pool of several fuel assemblies at once may provide superior and less expensive analytical data. The cast product is compact relative to intact fuel assemblies, with the potential to reduce the number of canisters that
    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×
    • must be provided for in the repository. In principle, M-D could deal with uranium oxides and silicides as well as Al-U alloys. However, difficulties are encountered in alloying UO2, which is a more stable than A12O3 at higher melt temperatures.

      Three test methods have been selected for assessment of SNF degradation and the release of radionuclides from M-D product (3, Sec 6.1). However, the greatest value for these techniques is likely to come in studying reactions between canister materials and repository waters, or fuel materials and repository waters. This is not an environment we are to consider. For evaluation of these test procedures the test matrix proposed by Savannah River is needed (3, Sec 6.0-6.25).

    • Press-Dilute Treatment (5, App F): This is a waste volume compaction process with isotopic dilution. Dilution is accomplished by layering the fuel material with depleted uranium and pressing to form a block. Six pressed blocks are placed in a stainless steel can and welded shut. This is a simple technology that could be put into operation with relative ease. No special corrosion problems are foreseen.
    • The Electrometallurgical treatment of Al-SNF (5, App F) is a two stage process plus a head-end facility. The Al-SNF is melted and cast into anodes for the electrorefiner. The first electrorefining step is the transport of aluminum from the Al-SNF anode in a LiF-KF electrolyte to the electrorefiner cathode, leaving uranium and metallic fission products behind in the basket of the anode. The aluminum is disposable as low-level waste. The contents of the anode basket are transferred to a second electrorefiner where high purity uranium is recovered. Fission products are converted to oxides that provide a feed stock for a glass melter. Fission products xenon and krypton are released to the inerted enclosure; other volatile fission products are trapped.

      High throughput of electrorefined uranium has been demonstrated at ANL. High throughput electrorefining of aluminum has yet to be shown, along with scrubbing alkaline earth products from the aluminum electrorefiner salt. Design of an engineering-scale electrorefiner is underway. Nothing is said about corrosion during electrorefining. Until more is known about the feasibility of the process it is probably too early to give corrosion much priority.

      The benefits of a functional electrometallurical process are the recovery of enriched uranium for potential reuse and the consequent reduction of nuclear waste storage canisters requiring disposal. If reuse of

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×
    • the recovered enriched uranium is not an option, then the EM product has few apparent advantages over the Melt-Dilute Process.

    Question 3: Since the amount of water in the repository is likely to be somewhat limited, would filling the canister void space with aluminum or some other sacrificial material make any difference in the long-term corrosion of the waste form?

    Response: The drying operation is designed to remove free water in the container prior to its sealing off and isolation. Through the drying operation, residual water is limited to amounts consistent with an aluminum corrosion allowance of 0.3 mil. and a hydrogen gas limit of less than 4 percent by volume, the lower limit of hydrogen gas flammability (3). Void-space pressurization and the limits on corrosion product hydrogen generation are more restrictive than the corrosion limit (2,3). The water that will remain in the void space after water vapor levels are reduced to a few torr in the drying operation will be insufficient to endanger either the corrosion limit or the void space gas pressure or composition limits. In "dry storage," without some inexplicable entry of water, the addition of aluminum to the void space is largely without consequence except to the extent that the reduced free volume would increase the gas space pressure for a given amount of corrosion.

    This question is more meaningful in the context of repository behavior, which is outside our charge. It is most plausible to me that water for entry into the waste container fuel material will first have penetrated the massive carbon steel overpack and then the inner canister liner. The possibility of crevice corrosion in the annulus between the inner barrier (alloy 625, Ref. 6) and the outer barrier (A516, Ref. 6) of the canister will require assessment. Access of considerable water to the repository is implied. If this scenario is rational, the addition of sacrificial aluminum to the canister void space would be inconsequential.

    Question 4: Will any waste forms resulting from any of the alternative processing options be likely to increase internal corrosion of a standard repository container compared to spent commercial fuel or vitrified glass logs? That is, are there likely to be interactions between the waste form

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    and the inner container in excess of what would be expected for commercial fuel or vitrified glass logs.

    Response: Direct/Co-disposal. Aluminum clad fuel removed from pool storage for direct disposal is expected to bring with it a small amount of water along with any deposits from the reactor coolant or pool storage. Treatment in a tumbling washer will remove the bulk of the crud and loose deposits (4). After vacuum-drying, residual water levels for encapsulated fuel assemblies are too low to challenge limits set for incremental corrosion of Al-clad in the fuel storage container. Gas pressure (water vapor or H2 or a combination of those) in the vacuum-dried waste container will be well under acceptable pressure limits. There is no reason to suspect nonvolatile corrosive substances to be introduced to the container by way of the spent fuel charge.

    Press and Dilute. Vacuum drying is applicable to press and dilute Al-SNF processing. Similar assurances regarding water vapor and nonvolatile corrosive substances are justified.

    Melt and Dilute. This Al-SNF treatment process generates a melt of uranium, aluminum, transuranics, and fission products cast compactly into a mold for disposal in a waste container. Given the high temperature of processing, the possibility of release of water or corrosive impurities to the container from the waste form is further reduced over the processes for direct/co-disposal or press and dilute treatment.

    Electrometallurgical Processing. This process also involves melting of the waste form and negligible likelihood of introducing water or corrosive species to the container environment.

    Question 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 and are they likely to lead to useful outcomes?

    Response: An instrumented canister was designed and fabricated to validate the drying and storage criteria for a road-ready container. The instrumented test chamber will accept an Materials and Test Reactor (MTR) fuel assembly in a chamber instrumented to measure and record the temperature of the fuel cladding, the ambient gas temperature, the gas species present, relative humidity, and windows to determine visually the condition of fuel material surfaces. The instrumented canister is suitable

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    for corrosion measurements as determined by water consumption and hydrogen generation. Alternatively, drying of a fuel assembly can be followed as a function of temperature, relative humidity, and time. The instrumented canister is a necessary capability and validates more oblique laboratory studies to the same purpose for the corrosion and drying studies of irradiated fuel (3, Sec 4.2).

    In the corrosion area, work done and the proposed future efforts are generally viable and support proposed handling plans. The corrosion models and the range over which they apply are examined later at the conclusion of question responses.

    There are several items which have not been treated explicitly in the literature handouts that may deserve some consideration:

    • Question 4 above points up the lack of information concerning the choice of material for the waste container and its interactions with the waste form. I refer here to the can into which the waste is placed, whether fuel assemblies, or castings or molten metal from the Melt-Dilute process. Certainly the can could serve better as a barrier to water intrusion if the material and its metallurgical condition were wisely chosen. There is casual reference to alloys XM19 and 316L SS (6) but without comment or any indication of how these materials were selected. Melt-can interactions may be understood, but they are not discussed.
    • Radiation reacts with water vapor to generate nitric acid. Both radiation and dilute nitric acid have been shown to accelerate aluminum corrosion rates in vapor phase reactions. Possibly the radiation effect on aluminum corrosion can be accounted for by nitric acid generation. Such is the implication of the parallel corrosion studies: on one hand, using a solution of one part concentrated nitric acid and 6 parts of water and, on the other, water and a gamma source of about 1.8 x 106 rads/hr. It is not clear otherwise why that particular acid solution was chosen for comparison with the gamma irradiated solution. The equilibrium of nitric acid, air and water vapor, and gamma radiation is not discussed. A better understanding of radiation effects requires such a parametric analysis.
    • Here are some comments about the corrosion model that come out of my involvement with aluminum corrosion some years ago. Corrosion is strongly influenced by the aluminum surface temperature.
    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×
    • The storage temperature, like all storage parameters, is based on the requirement that Al-SNF must be safely retrievable during treatment and subsequent storage. The interim storage temperature limit has been identified as 200 °C. To keep the concentration of corrosion product hydrogen below the combustion limit it must constitute less than 4 percent of the gas volume. The maximum acceptable gas pressure within a waste container is 60 psi. With these limits set, the tolerable amount of reaction between stored fuel and the gas environment can be determined.

      The rates and temperature dependence of the interactions between the container environment and the clad and fuel surfaces not only are important in themselves but can lead to a mathematical model helpful in understanding the corrosion mechanism. The principal virtue of a model is the ability to extrapolate to temperatures where data are not at hand. It is very much to the point to know when accelerated corrosion is likely to occur and its approximate rate.

      In liquid water environments, cladding alloys also corrode by a parabolic rate process. There are differences among alloys, but for relatively brief exposures at 200 °C or less, corrosion is proportional to the square root of time. For longer periods of time or higher water temperatures the corrosion process ''breaks away." At breakaway the rate becomes linear with time and is much accelerated. Hanford work related time to breakaway and breakaway corrosion rates to temperature, both by Arrhenius type expressions. Savannah River has not looked for this particular type of relationship. However I found it very interesting that specimens in high temperature vapor, corroding in the rapid linear breakaway corrosion mode, continued to corrode by the same breakaway mechanism when transferred to a lower temperature vapor environment. In this present work, recognizing that accelerating rates are probable at long exposure or high temperature is probably sufficient information to avoid underestimating corrosion effects whether rates are predicated by the Savannah River or Hanford approach.

    • Inclusion of other corrosion variables in the mathematical model may be possible with information already at hand. Corrosion rates have been determined in isothermal vapor environments at several initial relative humidities, which depending on the experiment may or may not decrease with time. Such an inquiry should start at the beginning and establish the order of the water vapor-aluminum reaction. There could be
    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×
    • some use for a corrosion model in which water vapor pressure or relative humidity could be directly cranked into the rate expression.

    Conclusions

    The body of corrosion work presented to the expert committee is quite impressive. The storage basin-related corrosion studies not only are complete but have resulted in pool cleanup procedures that have eliminated fuel pitting in the basin. It is difficult to find much of a corrosion threat during dry storage, but the work that should be done on container drying requirements and the corrosion consequences of the small amount of residual water has been done. Of particular interest to this observer was the characterization of Al-U alloys and the rationalization of the corrosion behavior of these materials. While similar work has been done on the solubility of radionuclides from glass, the work on radioisotope release from Al-SNF material is new information on an entirely different class of materials and of general technical interest. This is in part because of the projected use of the three test protocols newly applied to Al-SNF characterization. Along with these activities, which are presently laboratory tests, are engineering-scale validation studies in which irradiated MTR fuel assemblies can be monitored under dry storage environmental conditions for fuel and gas temperature, gas species present, pressure, relative humidity, and the visual condition of the fuel material.

    I found no significant deficiencies in the Savannah River corrosion program. The water basin corrosion effects have been thoroughly studied. There is little reason to expect new phenomena to show up. Corrosion in dry storage environments will be limited by the small and controlled availability of water.

    References

    1. Levenson, M., and K. Crowley, Study Director, "Instructions to the Technical Experts," November 14, 1997.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    2. Sindelar, R.L., H.B. Peacock Jr, P.S. Lam, N.C. Iyer, and M.R. Louthan, "Acceptance Criteria for Interim Dry Storage of Aluminum-Alloy Clad Spent Nuclear Fuels," WSRC-TR-95-0347 (U), March 1966.

    3. "Alternative Aluminum Spent Nuclear Fuel Treatment Technology Development Report (U)," WSRC-TR-97-00345, October 1997.

    4. Large, W.S., R.L., Sindelar, "Review of Drying Methods for Spent Fuel," WSCR-TR-97-0075, April 1997.

    5. "Technical Strategy for the Treatment, Packaging and Disposal of Aluminum-Based Spent Nuclear Fuel," Vol. 11.

    6. Benton, H.A., "Waste Form and Co-Disposal Waste Package for Aluminum-Based Research Reactor Fuel," NAS Review of Al-Based SNF Alternative Technology Selection, December 2, 1997.

    7. Abashian, M.S., "Mined Geological Disposal System Waste Acceptance Criteria," B00000000-01717040600-00095 REVOO, September 1997.

    8. Dillon, R.L. and V.H. Troutner, "Observations on the Mechanisms and Kinetics of Aqueous Aluminum Corrosion," HW-51849, September 30, 1957.

    9. Dillon, R.L., "Observations on the Mechanisms and Kinetics of Aqueous Aluminum Corrosion II," HW-71756, November 1971.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Topic: Cost and Schedule

    Consultant: G. Brian Estes

    A list of questions relating to cost and schedule issues follows together with my responses. Information was gathered from a review of references and, a series of briefings and interactive meetings held in Augusta, Georgia on December 2 and 3, 1997, and documents that were provided subsequent to the meetings.

    Question 1. Are the cost data provided by DOE reasonably complete and transparent?

    Response. The cost data contained in the report of the Research Reactor Spent Nuclear Fuel Task Team, Technical Strategy for the Treatment, Packaging, and Disposal of Aluminum-Based Spent Nuclear Fuel, Vols. I and II, appear to have been reasonably complete and transparent as of the time they were prepared and appropriate for the level of development of the alternatives considered.

    Cost estimates were built in a logical manner, and data from approved studies were used with a reasonable application of scaling factors, uncertainty factors, and adjustments for inflation. In addition, experience from similar or related projects underway has been factored into the estimating factors.1

    While some elements of cost appear to be excessive (for example, the combination of construction inspection, project support, project management, and construction management totals 23 percent),2 they are consistently applied and do not affect the cost comparison.

    The question of whether the use of existing facilities rather than constructing new was considered was not addressed in Vol. I of the Task Team report. However, Vol. II3 and briefings in Augusta, Georgia4 disclosed that preliminary cost estimates showed higher costs for modifying existing facilities. The primary drivers of the higher costs in this case are operation and maintenance costs of the wet basins, and the fact that other existing facilities would need significant upgrades to meet

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    current Nuclear Regulatory Commission standards, which may be required for facilities completed after 2002.5

    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 appear to be adequate to limit the field of candidate technologies to two or three for further refinement prior to final decision. The schedules contained in the report are unrealistic, but this does not appear to affect the final choice of technologies. Accordingly, there appears to be a reasonable basis to proceed with the refinement of three technologies currently being pursued in detail.

    The Transfer Facility Project was not included in the FY98 program as listed in the Savannah River Site FY97 Spent Nuclear Fuel Interim Management Plan6 on which the Task Force Report schedule was based. The original schedule was probably unachievable even with perfect DOE focus and universal support (personal opinion) but was directed by DOE in the Task Force ground rules.7 A schedule summary presented at the Augusta meeting indicates a three to five year slip in initial operating capability and substantial increases in costs.8 This affects all technologies, and a review of the time-sensitive cost elements, primarily wet basin operating costs, indicates there would be no change in relative standings among the alternatives. A detailed cost report issued on December 12, 1997, to support an ongoing independent report on non-proliferation issues shows reasonable agreement in proportional costs. For example, life cycle costs for electrometallurgy were 33 percent higher than direct co-disposal and are now 40 percent higher. Likewise the same costs for dissolve and vitrify were 67 percent higher than direct co-disposal and are now 66 percent higher.9 This latest report further confirms the original Task Force report conclusions on proceeding with a limited number of options.

    Credit for sale of commercial grade uranium resulting from processing/co-disposal and electrometallurgy options is listed in the comparisons.10 During the Augusta meeting it was determined that while a sale has not yet taken place, negotiations are underway with the Tennessee Valley Authority to sell commercial type uranium and a signed agreement is expected to be in hand by July 1998.11 While the total

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    supply and demand picture is not known and credit may be somewhat overstated, failure to realize these savings would not in itself affect a decision on alternatives.

    The question of whether the processing/co-disposal baseline cost estimates included handling the process waste stream as well as the product was raised during the Augusta meeting. The costs of handling, converting the waste to glass logs, and repository disposal are included in the estimates.12

    Question 3. Are the cost and schedule estimates developed by DOE for the alternative processing options suitable for budget planning purposes?

    Response. The estimates and schedules contained in the Task Force Report are not of budget quality. As discussed under question 2 above, the schedules appear to be unachievable even with universal DOE support and absence of controversy. That condition does not exist, because during the week of November 24 the budget disappeared and then was restored during the week of December 1.13 Further, the cost estimates, while sufficient to pick among alternatives for further development, are not refined to the point to support budget expectations for line item projects.

    As discussed under question 2, the Transfer Facility project was not submitted as a part of the FY98 DOE budget, and there has been no exposure of the project to the Congress to determine support there.14 The program has been discussed with the Nuclear Regulatory Commission, but work on licensing of facilities has not yet begun. The December 1997 cost study indicates facilities will be constructed to USNRC standards but will not be licensed.15

    In the time since the Task Force Report was published a preconceptual design of the Transfer Facility has been performed by Bechtel.16 The roughly $240 million cost is about 10 percent under the like facility estimate in a non-proliferation cost study prepared in July 1997.17 This bottoms-up estimate has the type of detail required to support a line item project and is now available for that purpose. In addition, studies on the modification and use of existing facilities at L Basin for the receipt portion of the Transfer Facility are being revisited.18

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    The cost estimates prepared to support the non-proliferation study show that while there have been adjustments as a result of better information and more detail available than when the original Task Force report was prepared, by far the greatest increase in life cycle costs among all alternatives is in operational costs. These costs increased by a factor greater than 2:1 and are responsible for the bulk of the cost growth.19 While the unit costs for processing and other radioactive materials handling are well known, the budget questions will need to focus on the drivers of these costs (i.e., manpower and time). Again, the relative positions of alternatives remain unchanged.

    Question 4. Has DOE considered the costs of program delays in its budget development or budget planning for this program?

    Response. The costs contained in the Task Force report do not reflect program delays since they were prepared for a schedule with admittedly forced dates. At the Augusta meeting it was reported that the DOE decision on whether to proceed and with what alternatives is expected by October 1999,20 and a project will be submitted either as an FY00 privatization project or an FY01 line item project.21

    The cost estimates in the July 1997 and December 1997 nonproliferation cost studies have been adjusted for programming as currently foreseen. Inflation factors used appear to be reasonable.

    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. Cost estimate development and schedules contained in the Research Reactor Task Force Report are not consistent with DOE procedures because the forced schedule mentioned in questions 2, 3, and 4 did not permit development of budget quality estimates. The revised schedules provided at the Augusta meeting now support adherence to DOE procedures.

    Estimates have been prepared for submission of the project under the privatization program.22 There is no evidence of significant waivers of environmental, safety, and health procedures, DOE site work

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    complications, and risk assumption, which would seem to be necessary to support the type of cost savings projected. The project was included in the FY99 privatization program but was dropped from the list in an FY99 pass-back. Five of seven projects submitted were reprogrammed as line item projects. DOE has yet to sign a privatization contract requiring debt financing, and experience to date with funded privatization contracts has been poor.23 The December 1997 estimates include costs for privatization, including debt financing.24 While factors are scaled to account for varying levels of relative risk, the track record to date doesn't answer the question of whether, in fact, investors will be willing to support such a venture. Pursuing this means of executing the project appears to guarantee additional delays (personal opinion).

    Detailed scheduling of work for the processing canyons has been accomplished since publication of the Research Reactor Task Force Report. It has been confirmed that options involving reprocessing addressed in the report can be accommodated in the overall workload consistent with DOE procedures.25 The December 1997 cost estimates also show a $240 million life cycle cost saving for the reprocess/co-disposal option.26 This appears to be an attractive option, but the policy issue on whether to exercise it must be decided by DOE.

    The Research Reactor Task Force Report recommends a project approach to the program.27 If the line item project route is chosen, DOE procedures provide that critical milestone decisions for projects under $500 million are made by the local DOE site office.28 Both a project approach and local decision making authority are essential to timely execution of the program (personal opinion).

    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. The schedule milestones laid out in the Research Reactor Task Force Report are not being met for reasons discussed under questions 2 through 5 above.29 Since the project has slipped and cost estimates have been revised, cost performance cannot be evaluated yet.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Revised schedules presented at the Augusta meeting appear to be achievable with strong DOE support and commitment (personal opinion).

    A non-proliferation study currently under preparation is not expected to make go/no-go recommendations on any alternatives.30 It should, however, assist in answering the political question on whether and how much reprocessing to do in order to support a timely decision on which alternatives to pursue and meet budget programming windows.

    References

    1. Devine, J., et al., Technical Strategy for the Treatment, Packaging, and Disposal of Aluminum-Based Spent Nuclear Fuel, a report to the U.S. Department of Energy (DOE) by the Research Reactor Spent Nuclear Fuel Task Team, June 1996, Vol. II, pp. C-38,54,57. Also breakout session at Augusta, Georgia meeting, December 3, 1997 (John Hurd, WSRC/Bechtel).

    2. Ibid., Table C7.2-2f.

    3. Ibid., pp. C-124-127.

    4. Breakout session at Augusta, Georgia meeting, December 3, 1997 (John Hurd, WSRC/Bechtel).

    5. Breakout session at Augusta, Georgia meeting, December 3, 1997 (Randy Polnick, DOE-SR).

    6. Dupont, M.E., et al., Savannah River Site FY97 Spent Nuclear Fuel Interim Management Plan (U), October 1996, p. 15.

    7. Briefing at Augusta, Georgia meeting, December 2, 1997 ( Jack Devine, Polestar) .

    8. Briefing at Augusta, Georgia meeting, December 2, 1997 ( Joe Krupa, WSRC).

    9. Krupa, J.F., Savannah River Site Aluminum-Clad Spent Nuclear Fuel Alternative Cost Study Rev 1 (U), December 12, 1997, p. 25.

    10. Devine, J., et al., Technical Strategy for the Treatment, Packaging, and Disposal of Aluminum-Based Spent Nuclear Fuel, a report to the U.S. Department of Energy (DOE) by the Research Reactor Spent Nuclear Fuel Task Team, June 1996, Vol. I, p. 58.

    11. Breakout session at Augusta, Georgia meeting, December 3, 1997 (John Dickinson, WSRC).

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    12. Breakout session at Augusta, Georgia meeting, December 3, 1997 (Joe Krupa, WSRC).

    13. Briefing at Augusta, Georgia meeting, December 2, 1997 (Jon Wolfstal, DOE-HQ).

    14. Breakout session at Augusta, Georgia meeting, December 3, 1997 (John Hurd, WSRC/Bechtel).

    15. Krupa, J.F., Savannah River Site Aluminum-Clad Spent Nuclear Fuel Alternative Cost Study Rev 1 (U), December 12, 1997, p. 25.

    16. Breakout session at Augusta, Georgia meeting, December 3, 1997 (Jane Carter, WSRC/Bechtel).

    17. Briefing at Augusta, Georgia meeting, December 2, 1997 (Joe Krupa, WSRC) .

    18. Breakout session at Augusta, Georgia meeting, December 3, 1997 (Mark Barlow, WSRC).

    19. Devine, J., et al., Technical Strategy for the Treatment, Packaging, and Disposal of Aluminum-Based Spent Nuclear Fuel, a report to the U.S. Department of Energy (DOE) by the Research Reactor Spent Nuclear Fuel Task Team, June 1996, Vol. II, Table C8.2f. Also Krupa, J.F., Savannah River Site Aluminum-Clad Spent Nuclear Fuel Alternative Cost Study Rev I (U), December 12, 1997, Table D-lf.

    20. Briefing at Augusta, Georgia meeting, December 2, 1997 (Karl Waltzer, DOE-SR).

    21. Breakout session at Augusta, Georgia meeting, December 3, 1997 (Randy Polnick, DOE-SR).

    22. Dupont, M.E., et al., Savannah River Site FY97 Spent Nuclear Fuel Interim Management Plan (U), October 1996, pp. 4, 27f.

    23. Breakout session at Augusta, Georgia meeting, December 3, 1997 (Randy Polnick, DOE-SR).

    24. Krupa, J.F., Savannah River Site Aluminum-Clad Spent Nuclear Fuel Alternative Cost Study Rev 1 (U), December 12, 1997, p. 21f.

    25. Breakout session at Augusta, Georgia meeting, December 3, 1997 (John Dickinson, WSRC).

    26. Krupa, J.F., Savannah River Site Aluminum-Clad Spent Nuclear Fuel Alternative Cost Study Rev 1 (U), December 12, 1997, p. 25.

    27. Devine, J., et al., Technical Strategy for the Treatment, Packaging, and Disposal of Aluminum-Based Spent Nuclear Fuel, a report to the U.S.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Department of Energy (DOE) by the Research Reactor Spent Nuclear Fuel Task Team, June 1996, Vol. I, p. 58.

    28. Breakout session at Augusta, Georgia meeting, December 3, 1997 (Randy Polnick, DOE-SR).

    29. Briefing at Augusta, Georgia meeting, December 2, 1997 (Karl Waltzer, DOE-SR).

    30. Briefing at Augusta, Georgia meeting, December 2, 1997 (Jon Wolfstal, DOE-HQ).

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Topic: Processing and Remote Systems

    Consultant: Harry Harmon

    Introduction

    We were asked to participate in the presentations on December 23, 1997, and study the materials provided in order to address a set of specific questions that were given to us. Joe Byrd and I served as a subteam on Processing/Remote Handling. In addition to the reports and handouts received on the first day of the meeting, we met with a number of technical personnel involved in the program on the second day. These personnel included program participants from Savannah River Site, Argonne, Idaho National Engineering and Environmental Laboratory, and interested members of the public. My response to the questions is provided below. Mr. Byrd will provide more detail than I have offered in the area of remote handling.

    Responses

    Q: For each processing option evaluated by DOE, are the processing steps used as the basis for assessment and comparison (other than direct co-disposal) technically credible? That is, are they likely to work as described and produce the products and results assumed?

    A: I believe that enough is known about each process option to conclude that they are all technically credible; i.e., the physical and chemical processes employed should work in principle. However, whether they will work as described in a remote plant environment is more difficult to state with certainty. The latter part of this question depends more on the level of development and prior experience with the process steps. Unit operations such as packaging, mechanical size reduction, melting, dissolution, vitrification, and electrorefining of uranium and aluminum are well known and demonstrated.1,2 The electrometallurgical process was depicted as having little or no secondary waste, but similar processes at Rocky Flats have generated significant quantities of salt waste and other residues.3 A key development need for

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    the electrometallurgical process is actual demonstration of the ability to recycle all potential waste streams. I believe that the Glass Material Oxidation and Dissolution System (GMODS) and Plasma Arc Treatment will require extensive development programs to support a reliable plantscale process. For GMODS, I believe that feeding pieces of fuel elements to the melter, pouring the glass product, and off-gas processing will be challenging steps. In the case of plasma arc treatment, feeding fuel elements, remote maintenance of the rotating furnace, control of the ceramic waste form composition, and off-gas processing will be significant development concerns.1

    Q: Are there other processing options that should be considered by DOE for disposition of aluminum-based spent fuel?

    A: Reprocessing of portions of the spent fuel in 221-H canyon at the Savannah River site should be given more consideration in this study. While the timing of fuel receipts may preclude processing all the fuel in 221-H (even if policy considerations allowed it), the program could be simplified by reprocessing portions of the fuel. First, part of the fuel could be reprocessed to alleviate basin capacity concerns. The resulting purified HEU uranium solution could be diluted to the desired level (less than 20 percent for proliferation concerns or less than 5 percent for LWR fuel use) and fission products would go to the Defense Waste Processing Facility (DWPF) and, eventually, to the repository. How much is processed would depend on shipping schedules, basin space, and demand for LEU uranium for LWR fuel. Secondly, the program would be simplified by eliminating small quantities of U-Al fuels that are significantly different in size from conventional Materials and Test Reactor (MTR) fuel elements.1 For example, long rods (like NRU/NRX fuel) will complicate design of fuel handling and feed preparation steps and will require some size reduction even for direct co-disposal. (Some believe that non-standard fuel dimensions are a greater problem for direct co-disposal than for the processing options.) All these U-Al fuels, uranium metal in aluminum cans, and UO2 in aluminum cans are chemically compatible with SRS canyon processes (although the UO2 powder materials will require some special nuclear safety controls during dissolution). Thus, they could be eliminated from design considerations by reprocessing them.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Two variations briefly considered in the Technical Strategy Report1 should, in fact, be given further evaluation: (1) Small casks designed for interim dry storage and transportation could be used by the reactor sites for shipment of MTR fuel to SRS. Casks, such as the GNB CASTOR MTR cask, are commercially available and should be considered in cost analyses of storage options. (2) Not shipping INEEL SNF to SRS reduces SRS storage needs and minimizes unnecessary transportation. If direct co-disposal is selected, INEEL will have the packaging facilities required, based on its role with stainless steel and zircalloy-clad fuels.4

    Q: Do inner container designs appear adequate to contain the waste forms resulting from the various processing options?

    A: Although I did not examine this in great depth, I see no reason why the designs would not be adequate. Clearly, the processing options must size their product containers appropriately, but this is straightforward.

    Q: Are they overdesigned for the intended application?

    A: No. The canisters were described as being fabricated of steel with neutron poison inserts as required. DWPF canisters are stainless steel, so it seems appropriate that the SNF canister should be of similar durability.

    Q: Are DOE's basic material handling plans, pool use, and other facility needs reasonable, and are remote handling technologies available to meet these needs?

    A: The transfer and storage facility concepts1 employed standard equipment and conventional techniques for remote material handling. For direct co-disposal, remote size reduction equipment will be required to accommodate the dimensional restrictions of the inner containers for other than standard MTR fuel elements. All required remote handling

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    technologies are readily available, and advancements in this field continue to be made.

    Q: Are the technical requirements for the alternative processing options sufficiently well defined so that reasonable judgments can be made about the likelihood of success of implementing them?

    A: The processing options can be divided into three families: (1) HEU dilution technologies; (2) advanced treatment technologies; and (3) options involving canyon processing. Technical requirements for each family of options have been defined, at least in general terms. The Technical Strategy1 identified four criteria that all process options must meet to be considered:

    • Development work must be completed by 2000 (this is no longer viewed as required by DOE).
    • Funding required during the development period (e.g., during the first five years) must be within that reasonably expected to be available in that time frame.
    • The waste form must be compatible with anticipated repository requirements.
    • The treatment technology cannot present any environmental, safety, and health operational concerns.

    Also, requirements for repository disposal of the Al-SNF form are listed in Reference 2, but most of those criteria relate to the canister and waste package.

    However, I was not able to find a complete set of process requirements or product requirements that all options must meet to be successful. (Such a document may exist, but I have not seen it.) For example, if all SNF HEU must be diluted to less than 20 percent before going to the repository, then direct disposal and direct co-disposal are eliminated. Also, if separation of fission products from fissile material is forbidden, then all canyon processing options and electrometallurgical treatment are eliminated. Without firm requirements identified that all processes and products must meet, the options can only be evaluated as possible approaches, each with potentially different end products.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Also, since product requirements are not specifically identified, evaluations of characterization needs must be very broad and comprehensive at this juncture.5 Extensive dialogue and teamwork between the spent fuel program and the repository will be required to develop an achievable and acceptable characterization plan. I would recommend use of commercially available instrumentation for burnup and fissile content measurements on fuel elements. With these data, fission product content can be calculated with sufficient accuracy to provide characterization via process knowledge.

    Q: Are there large differences in likelihood of success of implementing the various process options?

    A: Yes. To be considered successfully implemented, the process option must be capable of being implemented in an operating facility within the budget and schedule constraints. (Given sufficient time and funding, all options could be implemented in my opinion.) Thus, the likelihood of success in this context is inversely related to the extent of technology development needed. Based on my judgment and the information in the technical strategy document1, I would rank them as follows (highest likelihood of success at top of list):

    1. Processing/co-disposal
    2. Direct disposal and direct co-disposal
    3. Press/dilute and melt/dilute
    4. Dissolve and vitrify
    5. Electrometallurgical
    6. Plasma arc
    7. GMODS
    8. There are not large differences in likelihood of success between the first three groups, but the last four will require significantly more effort to be successful.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    References

    1. "Technical Strategy for the Treatment, Packaging, and Disposal of Aluminum-Based Spent Nuclear Fuel," Volumes I and II, June 1996.

    2. L. Sindelar et al. "Alternative Aluminum Spent Nuclear Fuel Treatment Technology Development Status Report (U)," WSRC-TR97-00345, October 1997.

    3. Implementation Plan for Defense Nuclear Facility Safety Board Recommendation 94-1. May 1994.

    4. P. Hoskins et al. "Technical Strategy for the Management of INEEL Spent Nuclear Fuel," March 1997.

    5. E. Skidmore et al. "Task Plan for Characterization of DOE Aluminum Spent Nuclear Fuel (U)," SRT-MTS-97-2004, January 31, 1997.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Topic: Nuclear Criticality Safety

    Consultant: Valerie L. Putman

    Introduction

    Aluminum-based Spent Nuclear Fuel (Al-SNF) is to be collected at the Savannah River Site (SRS) for treatment and interim storage, then shipped to a permanent repository for final storage. Currently stored at SRS, various research reactors and U.S. Department of Energy (DOE) sites throughout the United States, and various foreign research reactors, Al-SNF is in a wide variety of configurations. Most Al-SNF is highly enriched in 235U at beginning and end of life. Al-SNF is therefore considered to be a greater criticality safety concern throughout treatment, interim storage, transport, and final storage than Spent Nuclear Fuel (SNF) from commercial power plants.

    SRS management must adequately identify, characterize, and weigh options for DOE to select a path forward for disposing this fuel. SRS staff must work closely with organizations that currently have the Al-SNF and with repository personnel to ensure that options adequately address all issues for the Al-SNF itself and for the Al-SNF in the repository in an efficient, cost-effective manner.

    SRS adequacy in addressing criticality safety issues of the options is reviewed here. Two criticality safety questions were specified for this review (Levenson and Crowley, 1997):

    Question 1: What are the significant criticality issues that must be considered during processing, interim storage of the waste form after processing, and shipment of the waste form to a repository? Has DOE adequately addressed these issues in its technology planning?

    Question 2: Do any of the waste forms produced by the alternative processing options pose significant internal or external criticality hazards in a repository—either from material degradation in the waste container or in the near field of the repository after the container is breached—relative to commercial spent fuel or vitrified high-level waste? NOTE: comments on the use of poisons or isotopic dilution are

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    appropriate as are comments on filling the void space in the canister so as to limit the volume of water that could be present in case of canister leakage.

    Conclusion

    Adequate criticality safety can be assured for configurations and activities to treat, move, and store DOE aluminum-based spent nuclear fuel (Al-SNF). Criticality safety analysis methodologies are well developed, and existing computer codes and neutron cross section libraries appear sufficient. Although few, if any, specific criticality safety evaluations are complete, scoping work indicates sufficient safety can be provided by standard means such as limiting fissile quantity, including neutron absorbers, and for the near-term, limiting configurations. In addition, information from in-progress and planned tests will be used to determine if additional criticality safety resources are needed.

    SRS presenters identified a co-disposal option as currently most favored. Strategies are identified to prevent a critical excursion for this option during all stages of treatment, interim storage, and final disposal. Controls include fixed neutron absorbers in canister baskets for some Al-SNFs. However, continued mixing of fissile material and neutron absorber is less certain after many millennia, when fuel and canisters are fully degraded and material might migrate outside the repository. Although information to date indicates the consequences of a critical excursion with Al-SNF at this point would be negligible, accident prevention apparently is still considered very important.

    Therefore, this reviewer believes, if critical excursion prevention continues to be a very high priority after fuel and canisters are fully degraded and material might migrate outside the repository, Al-SNF treatment(s) that significantly dilute 235U and/or introduce significant neutron absorbers in the fuel matrix should be selected.

    Discussion

    Except for scoping calculations, little fuel-specific or activity-specific criticality safety work is complete for proposed activities with these Al-SNFs. More specific work is premature until fuel-treatment and storage-configuration options are narrowed further. It is sufficient to

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    ensure that adequate methodology (codes and cross-section libraries; tool selection methods; modeling, calculation, validation, and documentation requirements and methods; criticality accident scenario identification; and criticality safety contingency analysis) and information (fuel, facility, and treatment descriptions) will be available to perform specific criticality safety evaluation(s) when needed. Current availability is not required if measures are taken to identify and obtain necessary items in a timely manner to support option selection and specific evaluations.

    Criticality Safety at Savannah River Site (SRS)

    Many proposed activities with Al-SNF at SRS are similar to past activities, there and elsewhere, with aluminum-based nuclear fuels. These activities include underwater interim SNF storage for some Al-SNF types, packaging or repackaging into fuel-specific baskets in approximately 17-inch-diameter Al-SNF canisters, drying Al-SNF and/or loaded canisters as needed, normally dry Al-SNF storage, and associated fissile material handling and transportation. For criticality safety purposes, most of these activities should represent minor perturbations from previously evaluated conditions. Although, these activities must be evaluated to develop specific criticality safety limits and critically safe designs, or to show applicability of limits and designs established for Al-SNF already at SRS, SRS criticality safety methodology should be adequate for these evaluations.

    Additional treatment options discussed are dissolve and vitrify (glass or ceramic waste), electrometallurgical uranium separation, glass material oxidation and dissolution, melt and dilute (metal waste), plasma arc vitrification (ceramic waste), press and dilute (metal waste), and, as a baseline, continued wet-chemical uranium separation. With the exception of the press-and-dilute option, these treatments present opportunities to dilute 235U and/or introduce neutron absorbers in the fuel matrix itself, which would better assure criticality safety in the final repository after many millennia when the fuel matrix and canisters are completely degraded. Options involving uranium separation also include uranium reuse, if appropriate, based on economics and non-proliferation policies.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    With the exception of wet-chemical processing and, possibly, dissolve and vitrify options, some steps of these treatments could represent major perturbations from previously evaluated conditions. However, U.S.-wide experience in developing new fissile material processes indicate that adequate criticality safety control should be possible through vessel and/or container geometry, fissile mass or dilution limits, fixed or soluble neutron absorbers, and/or moderation limits for each step of these further treatment processes. The problem is usually one of adequately balancing criticality safety with process efficiency during process development.

    Criticality safety methodology at SRS is well developed due to its many years of nuclear experience with reactor operations, fuel storage, fuel processing via chemical reactions, and associated fissile material handling. Its past missions include significant experience with aluminum-based fuels. Therefore, at least some of criticality safety staff have considerable experience with software, cross-section data, and fuel-characterization data available for evaluating such fuels (see Gough et al., 1997, as an example).

    Criticality-safety-evaluation methodology at SRS includes a well-developed validation program which should identify and adequately compensate for any problems that might exist in codes and/or cross-section data. Staff experience and the validation program itself appear sufficient to ensure calculation validation effectiveness (Kimball and Trumble, 1997; Chandler and Trumble, 1997).

    None of the discussed Al-SNF activities at SRS are anticipated to involve conditions under which basic nuclear data and/or data processing are currently questioned (for example, 235U very highly diluted with aluminum, 235U with massive aluminum reflectors, or uranium in the resonance neutron energy range). Current SRS validation practices should be sufficient to identify if a less-than-desirable code and cross-section combination is used for particular conditions (for example, 235U in a fast energy system with SCALE 4.3 and ENDF/B-VI.3 cross-sections). Therefore, available information and data should be sufficient to perform and validate the necessary criticality safety evaluations. Additionally, ongoing tests should provide sufficient information to identify conditions that significantly deviate from expected conditions, enabling staff to determine if criticality safety tools are adequate in a timely manner.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Criticality Safety on Public Roads, Transport to Final Repository

    Criticality safety requirements for transportation of fissile material over public roads are well established and are updated as needed. Transportation requirements specifically identify many conditions that must be evaluated, relying less upon the evaluating or operating organization to identify all important conditions.

    Neither SRS nor Yucca Mountain criticality safety personnel typically perform many transportation evaluations. However, cask vendors, license holders, and several DOE contractor sites have well established criticality safety methodology to satisfy transportation criticality safety evaluation requirements. The methodologies are similar in many respects to the SRS criticality safety methodology but are specifically tailored to the types of fissile materials with which these other organization are concerned. General principles apply but some specifics (for example, detailed modeling methods) might not apply.

    Whichever organization performs transportation analyses for treated Al-SNF, its methodology will need to be reviewed for applicability to the treated fuel. Few if any problems are expected in developing a methodology because most options would produce a treated AI-SNF less reactive than previously shipped configurations of respective beginning of life aluminum-based fuel and/or untreated Al-SNF. Although there are larger-than-typical uncertainties in reflection cross sections of lead and iron (materials in transportation casks that would not necessarily be addressed elsewhere), established transportation criticality safety methodologies must already have resolved any problems caused by these uncertainties.

    A major difference between SRS and typical transportation criticality safety methodology is procedures for determining calculation-method bias and uncertainty. Although SRS's procedure is less conservative than typically used for transportation evaluations, the SRS procedure is defensible and valid.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Criticality Safety at Final Repository

    Al-SNF treatment and packaging options to be implemented at SRS greatly affect strategies to assure criticality safety for specific fuel and storage configurations in the final repository, and vice-versa. Candidate options result in fuel matrices and configurations ranging from packages enveloped by commercial-SNF packages to pockets of highly enriched uranium surrounded by highly radioactive waste, depleted uranium, and/or commercial SNF. In the latter case, SRS prefers a co-disposal option because it limits pocket size, providing a more assured fissile material dilution than direct disposal as canisters degrade over millennia.

    Repository criticality safety methodology is developed for final repository at the Yucca Mountain site. Initially developed for commercial SNF, it was recently revised to address highly enriched DOE SNF, including Al-SNF. Revisions included expanding the validation database and developing strategies for handling conditions not already addressed by methodology for commercial SNF (Thomas et al., 1997).

    Repository criticality safety work to date for Al-SNF primarily evaluates a co-disposal option. Of SRS favored options, this one is judged to be most vulnerable to a criticality accident because the waste form is repackaged fuel assemblies, still qualifying as highly enriched uranium in which each assembly, if flooded, is nearly optimally moderated. If adequate criticality safety can be assured for this option, adequate safety can be assured, perhaps with different strategies, for other favored but less vulnerable options.

    These Al-SNF criticality safety studies are based on two SRS-identified representative Al-SNFs, Massachusetts Institute of Technology (MIT) and Oak Ridge Research (ORR) reactor fuel assemblies (Doering and Gottlieb, 1997). Some Al-SNFs are more reactive than the representative fuels (for example, University of Missouri Research Reactor (MURR) fuel) but might not be adequately representative of other Al-SNF characteristics (Sentieri, 1996, p. 69). Unless additional evaluations are performed for more reactive fuels, it will probably be necessary to limit canister loadings to ensure each fuel array in a canister is no more reactive than the most reactive loading of representative Al-SNF. Such a limitation is not necessarily inefficient depending on relative individual and cumulative fuel assembly volumes.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    With the exception of specifically evaluating the most reactive Al-SNF fuel assemblies, evaluated conditions were selected to envelope a wide variety of Al-SNF assemblies and repository conditions. This strategy is adequate and economical for DOE to determine a path forward and might be adequate for final evaluations if the co-disposal option is selected.

    A comparison of evaluated critical-experiments and repository-condition characteristics, with the representative fuels, indicates there are sufficient experiment data to validate most enveloping repository conditions (Anderson, 1977; Doering and Gottlieb, 1997; Gottlieb, 1997; Gottlieb et al., 1997). In some cases, calculations might be more conservative than absolutely necessary. Additional experiment data are not essential but such data might allow minor storage efficiency improvements.

    There is a general lack of critical experimental data for extremely dilute 235U systems and for fissile systems in the intermediate neutronenergy range. This is not currently a concern because calculations indicate repository conditions involving these characteristics have extremely low subcritical keffs (Gottlieb, 1997; Gottlieb et al., 1997). In these cases, calculation validation is much less of a concern. For example, if a condition's calculated keff is 0.3, it is more important to demonstrate that the condition's actual keff satisfies the required margin, in this case does not exceed 0.95, than to demonstrate that the calculated keff is within a few percent of the actual keff.

    Intact Canisters and Al-SNF, Initial Repository Conditions

    Traditional criticality safety requirements, concerns, and issues apply to the final repository initially because workers could be at risk if a criticality accident were to occur and because evaluations and corrective actions could be undertaken in a reasonable manner if problems develop. Initial co-disposal conditions involve handling and storing intact or nearly intact (non-leaking, very close to initial configuration) Al-SNF canisters, each surrounded by five co-disposed waste canisters.

    Specific canister-array configurations probably differ but methodologies and data used in addressing criticality safety for treated

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Al-SNF in interim storage and for transportation should apply to evaluations for initial repository conditions. Although possibly less well known, neutron reflection and moderation properties of materials that could credibly be between and/or around Al-SNF canisters in the repository should be no worse than materials (water, concrete, transportation cask, and adjacent fuels) that must be considered for these earlier activities.

    Initial (Phase 1) criticality safety analysis of intact or nearly intact AI-SNF canisters and fuel is nearly complete for the co-disposal option with the two representative fuels (Doering and Gottlieb, 1997; Gottlieb, 1997). One representative fuel, ORR, is sufficiently reactive to require fixed neutron absorbers for a critically safe, efficient Al-SNF canister loading under flooded conditions (Doering and Gottlieb, 1997). If ORR fuel with fixed neutron absorbers is acceptable, it should be possible to qualify canisters of the more reactive Al-SNFs, possibly with fewer fuel assemblies and the same neutron absorber or with the same number of fuel assemblies and more effective neutron absorbers (for example, thicker metal plates or higher concentrations in metal).

    Degraded Canisters and/or Al-SNF, Long Term Repository Conditions

    Long term repository criticality safety is handled in a nontraditional manner because there would eventually be no mechanism to detect developing problems, and because the location is very well shielded if an accident were to occur. In this case, evaluations examine environmental consequences more closely than human radiological exposure because humans are not at direct risk from a criticality accident. An inadvertent critical excursion is still undesirable and preventative measures are required. However, an extremely low-probability event might be acceptable if environmental consequences are negligible.

    Initial criticality safety analysis of degraded Al-SNF canisters and/or Al-SNF fuel (Phase 2) is nearing completion for the co-disposal option with two representative fuels (Gottlieb et al., 1997). Controls to minimize critical excursions in or near a canister are considered in this phase. In some cases it is important for criticality safety to ensure that, where neutron absorbers were required for intact fuel, absorbers continue to be ''mixed in" with degrading fuel. Gadolinium in carbon steel baskets

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    is the currently preferred absorber because it is expected to degrade in a manner that should encourage mixing with degraded fuel.

    Criticality safety analysis is initiated for storage conditions after millennia; canisters, baskets, and fuel are theorized to be fully degraded and completely uncontained (Phase 3). It is again preferred that, where neutron absorbers were required for intact fuel, absorbers continue to be mixed with fissile material. However, mixing is less assured at this stage. This reviewer believes that, if critical excursion prevention continues to be a very high priority for this phase, Al-SNF treatment(s) that significantly dilute 235U and/or which introduce significant neutron absorbers in the fuel matrix should be selected to better assure continued mixing.

    Criticality accident prevention is desirable but prevention methods are less assured. Critical excursion consequences therefore are more important for ensuring acceptably low risk during Phase 3. Information to date indicates that, if a critical excursion were to occur only in the highly enriched SNF, humans would still be very well shielded from the excursion. Additionally, resultant increases in fission products would be negligible compared to the already large inventory from commercial SNF, even when considering decay of the commercial SNF fission product inventory before this hypothetical excursion.

    Summary

    Adequate criticality safety can be assured for configurations and activities to treat, move, and store DOE Al-SNF. Criticality safety analyses methodologies are well developed and existing computer codes and neutron cross section libraries appear sufficient. Scoping work indicates sufficient safety can be provided by standard means such as limiting fissile quantity, including neutron absorbers, and, for the near-term, limiting configurations. Information from in-progress and planned tests will be used to determine if additional criticality safety resources are needed.

    Most criticality safety issues for proposed activities to treat, store in interim facilities, and ship Al-SNF will be the same as, or very similar to, criticality safety issues already addressed for existing SRS fuel

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    processing, storage, transfer, and transport activities. Methods for addressing criticality safety issues under such circumstances and for such processes are well established and apparently adequate (specifically, acceptable to the operating contractor and to DOE). Although analyses to define specific criticality safety limits are not yet initiated, proposed activities should not pose a technical challenge to this analysis methodology. Also, although there is a difference between SRS and typical transportation criticality safety procedures for determining bias, the SRS procedure is defensible and valid.

    Activity similarity exceptions involve advanced treatment options because some aspects of these treatments might challenge adequacy of existing SRS data and/or expertise. However, SRS criticality safety analysis methodology would still apply, but updated, upgraded, or different specific tools (modeling conventions, codes, cross-section libraries) might be needed. Planned and in-progress testing should soon identify Al-SNF characteristics for process stages of concern, at which point any need for updated, upgraded, or different criticality safety tools can be determined. Information to date indicates current tools are sufficient.

    In most cases, treatment and packaging options result in an Al-SNF waste that is significantly more reactive and has a significantly different fuel composition than commercial SNF waste. At this time, criticality safety analyses of Al-SNF in a permanent repository focus on a co-disposal option. Of SRS favored options, this one will result in the most reactive permanent storage configuration. If adequate criticality safety can be assured for this configuration and waste form, then staff should be able to assure adequate criticality safety for less reactive configurations and waste forms.

    Appropriate criticality safety methodology including validation is developed for the Yucca Mountain final repository. Preliminary evaluations are nearly complete, based on two representative Al-SNFs, the selected co-disposal option, and a few conditions selected to envelope all credible conditions identified for Al-SNF in the repository. A comparison of characteristics for these enveloping conditions and for critical experiments in the repository's validation database indicates the methodology is adequate.

    Although representative Al-SNFs do not include the most reactive Al-SNF assemblies, preliminary evaluations are sufficient to indicate

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    adequate criticality safety can be established for this option. Depending on specific baskets and storage configurations for the more reactive fuels, these preliminary evaluations might qualify as enveloping evaluations if the co-disposal option is selected.

    Final repository critical excursion prevention relies in some cases on neutron absorbers, which should be mixed with the Al-SNF, either between assemblies, between assembly components, or within the fuel matrix itself. After many millennia, fuel and baskets could be completely degraded and materials would not be contained. Under such conditions, continued mixing of neutron absorbers and 235U is less assured. Information to date indicates that the consequences of a critical excursion would be negligible and should be acceptable. However, if critical excursion prevention is paramount even at this fully degraded stage, this reviewer recommends treatment option(s) that dilute 235U and/or introduce neutron absorbers in the fuel matrix be selected to better assure continued fuel and absorber mixing.

    References

    Anderson, M. J. 1977. "Summary Report of Laboratory Critical Experiment Analyses Performed for the Disposal Criticality Analysis Methodology," B00000000-01717-5705-00076, Rev. 0. Civilian Radioactive Waste Management System Management & Operating Contractor: Las Vegas, Nevada (September 4, 1997).


    Chandler, John R., and E. Fitz Trumble. 1997. "Use of Bias, Uncertainty, and Subcritical Margins at the Savannah River Site." In Proceedings, Criticality Safety Challenges in the Next Decade, Chelan, Washington, September 7-11, 1997. American Nuclear Society: La Grange, Illinois (September 1997), pp. 262-267.


    Doering, Thomas W., and Peter Gottlieb. 1997. "Evaluation of co-disposal Viability for Aluminum-Clad DOE-Owned Spent Fuel: Phase 1, Intact co-disposal Canister," BBA000000-01717-5705-00011, Rev. 1. Civilian Radioactive Waste Management System Management & Operating Contractor: Las Vegas, Nevada (August 15, 1997).

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Gottlieb, P[eter]. 1997. "Degraded Waste Package Criticality: Summary Report of Evaluations through 1996," BBA000000-017175705-00012, Rev. 0. Civilian Radioactive Waste Management System Management & Operating Contractor: Las Vegas, Nevada (August 28, 1997).

    Gottlieb, P[eter], et al. 1997. "Waste Package Probabilistic Criticality Analysis Summary Report of Evaluations in 1997," BBA000000-01717-5705-00015, Rev. 0 . Civilian Radioactive Waste Management System Management & Operating Contractor: Las Vegas, Nevada (September 16, 1997).

    Gough, Sean T., et al., 1997 "Process Error-Induced Criticality Accident Analysis for Savannah River Site Receiving Basin for Offsite Fuel." In Proceedings, Criticality Safety Challenges in the Next Decade, Chelan, Washington, September 7-11, 1997. American Nuclear Society: La Grange, Illinois (September 1997), pp. 306309. (Primarily used as an example of SRS criticality safety methodology.)


    Kimball, Kevin D., and E. F[itz] Trumble. 1997. "Statistical Methods for Accurately Determining Criticality Code Bias." In Proceedings, Criticality Safety Challenges in the Next Decade, Chelan, Washington, September 7-11, 1997. American Nuclear Society: La Grange, Illinois (September 1997), pp. 247-254.


    Levenson, Milt, and Kevin Crowley. 1997. "Instruction to the Technical Experts, Draft." Memoranda for Technical Options for Disposition of Al-Based Spent Nuclear Fuel. Board on Radioactive Waste Management, National Research Council: Washington D.C. (November 1997).


    Sentieri, P[aul] J. 1996. "Criticality Safety Evaluation for Various Fuels Proposed for Dry Canning," LITCO Internal Report INEL-95/306. Lockheed Idaho Technologies Company: Idaho Falls, Idaho (April 1996).


    Thomas, D. A., et al. 1997. "Disposal Criticality Analysis Methodology Technical Report," B00000000-01717-5705-00020, Rev. 1. Civilian Radioactive Waste Management System Management & Operating Contractor: Las Vegas, Nevada (September 4, 1997).

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Other Documents Reviewed

    ANS-8.1 Working Group. 1988. Nuclear Criticality Safety in Operations with Fissile Materials Outside Reactors, ANSI/ANS-8.1-1983(R1988). American Nuclear Society: La Grange Park, Illinois (approved October 7, 1983).

    Bradley, Terry L., et al. 1996. "Direct and Co-Disposal Treatment Technologies," Appendix D. In Technical Strategy for the Treatment, Packaging, and Disposal of Aluminum-Based Spent Nuclear Fuel, Vol. 2. Research Reactor Spent Nuclear Fuel Task Team (June 1996).

    Fisher, L. E., et al. 1988. Package Review Guide for Reviewing Safety Analysis Reports for Packaging, UCID-21218, Rev. 1. Lawrence Livermore National Laboratory: Livermore, California (October 1988), pp. 8-6-8-10.

    Huria, H., and M. Ouisloumen. 1997. "ENDF/B-VI: 238U Resonance Integral Reduction—A Closer Look." In Transactions of the American Nuclear Society, Vol. 76, American Nuclear Society: La Grange, Illinois (June 1997), pp. 329-330.

    Lovett, Phyllis A., et al. 1996. "Criticality Control Bases for Repository Licensed Under 10 CFR Part 60," Appendix B. In Technical Strategy for the Treatment, Packaging, and Disposal of Aluminum-Based Spent Nuclear Fuel, Vol. 2. Research Reactor Spent Nuclear Fuel Task Team (June 1996). Prepared for U.S. Department of Energy, Office of Spent Fuel Management: Washington, D.C.

    Savannah River Technology Center. 1997. "Alternative Aluminum Spent Nuclear Fuel Treatment Technology Development Status Report (U)," WSRC-TR-97-00345 (U). Westinghouse Savannah River Company: Aiken, South Carolina (October 1997).

    Spencer, Robert R., et al. 1997. "Neutron Total and Capture Cross-Section Measurements on Aluminum at ORELA." In Transactions of the American Nuclear Society, Vol. 77. American Nuclear Society: La Grange, Illinois (November 1997), pp. 237-238.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    TRW. 1995. "Preliminary Requirements for the Disposition of DOE Spent Nuclear Fuel in a Deep Geologic Repository," A00000000-00811-1708-00006, Rev 0. TRW Environmental Safety Systems Inc.: Vienna, Virginia (December 15, 1995).

    TRW. 1997. "OCRWM Data Needs for DOE Spent Nuclear Fuel," A00000000-01717-2200-00090, Rev 2 Draft. TRW Environmental Safety Systems Inc.: Vienna, Virginia (September 1997).

    U.S. Department of Energy. 1995. Facility Safety, DOE O 420.1 with Changes 1 and 2. U.S. Department of Energy: Washington D.C. (October 13, 1995; November 16, 1995; October 24, 1996), part 4.3.

    U.S. Nuclear Regulatory Commission. Disposal of High-Level Radioactive Waste in Geologic Repositories, Part 60 in Title 10 of Code of Federal Regulations (10 CFR 60).

    U.S. Nuclear Regulatory Commission. Packaging and Transportation of Radioactive Material, Part 71 in Title 10 of Code of Federal Regulations (10 CFR 71).

    Weinman, J. P. 1997. "Monte Carlo Cross-Section Testing for Thermal and Intermediate 235U/238U Critical Assemblies-ENDF/B-V Versus ENDF/B-VI." In Transactions of the American Nuclear Society, Vol. 76. American Nuclear Society: La Grange, Illinois (June 1997), pp. 325-327.

    Wright, R. Q. and L[uiz] C. Leal. 1997. "Benchmark Testing and Status of ENDF/B-VI Release 3 Evaluations." In Transactions of the American Nuclear Society, Vol. 77. American Nuclear Society: La Grange, Illinois (November 1997), pp. 232-234.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Topic: Proliferation Aspects of the Treatment Options

    Consultant: David Rossin

    Several points are made below based on my personal knowledge and experience over 43 years in nuclear reactor technology, materials and fuel cycle work. Much of that experience has involved licensed commercial nuclear power plants and fuel cycle facilities. Where the text reflects my personal opinions, it is so noted.

    The discussion is divided into two main topics: (1) proliferation resistance and (2) assessment of alternative technologies under the Environmental Impact Statement process.

    Proliferation Assessment

    United States is a Weapons State—Safeguards Capability in the United States is internationally accepted. The Task Team Report included proliferation resistance for 5 percent of its Kepner-Tregoe evaluation. No significant differences between alternatives were identified. The report shows that further consideration of proliferation resistance is not needed for a meaningful comparison of alternatives. In my opinion, this is a reasonable approach. Because proliferation resistance is definitely a topic of concern to DOE, it would not be fitting to give the topic zero weight or to omit it, since that would not provide transparency for the analysis, and since the topic must be covered in the Programmatic EIS.

    Proliferation Potential for Disposition Options. Obviously, all operations at SRL are conducted under DOE safeguards. Thus there is no proliferation risk associated with any of the actual operations that are under consideration.

    All future operations including storage and handling of waste packages prior to insertion in the repository will also be conducted at safeguarded DOE sites. All transportation will also have to have appropriate safeguards. It is certainly reasonable at this time to assume that appropriate safeguards will be required regardless of the extent to which specific regulations or commitments exist at this time. In my

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    opinion, there is confidence that appropriate regulations will be developed and applied when needed in the future, and therefore there should be no delays in the decision process because of transportation or handling safeguards.

    Dilution of HEU may be needed for criticality considerations, but not for proliferation resistance at SRL. In my opinion, dilution of unique and costly HEU should not be done unless a valid and cost-effective case can be made for it. Physical security, accounting and safeguards protocols for handling HEU have been applied for decades, so a demonstration of diluting a quantity of HEU has no significance in international understanding of nonproliferation policy.

    Identification of Separated Fissile Material as a Proliferation Risk

    If DOE were to designate any separated fissile material as a proliferation risk, this could have extensive and costly ramifications. It could rule out separation of uranium rich in U-235 into a waste form suitable for disposal or require dilution at a point in the process that might not be necessary or cost-effective.

    Such a designation could even preclude selection of co-disposal as an alternative. This does not appear to be a desirable result, based on the Task Team analysis. It might even force the choice to the melt-dilute alternative, even if other considerations point elsewhere.

    In theory, it would be fine if all fissionable material were in forms or storage that meet either the "Spent Fuel Standard" or the "Stored Weapons Standard" described by the NAS report Management and Disposition of Excess Weapons Material. I believe there are commonsense levels on either side of each of these concepts that provide adequate safeguards for certain categories of fissionable materials. Neither of these concepts are really standards, in that they are not promulgated by any international or even national standards organization. They are valuable concepts since they represent known states, but are not to be applied blindly.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Container

    The proposed co-disposal container has a canister surrounded by vitrified HLW. This container design has enough radioactivity to meet a self-protection criterion. However, as with the spent fuel standard, this concept is only one of many factors in a safeguard solution. I do not believe that these "standards" were conceived of as absolute design standards to be universally applied. They play appropriate roles in establishing an effective and cost-effective safeguards approach.

    Repository

    Until sealed, the repository is a facility under safeguards. Material once placed in the repository is no longer accessible. The regulations governing the repository require that it will not be sealed for 50 years after loading to provide assurance that nothing dangerous is happening. DOE has said that it will be conservative and wait 100 years.

    These DOE containers will be surrounded by thousands of containers of spent fuel or HLW in the repository. Therefore radiation and geometry provide proliferation resistance. For assessment of proliferation risk, the canisters will be surrounded by material with the same or higher proliferation risk potential.

    Self-Protection Criterion

    The IAEA's Self-Protection Guideline is a definition, not a necessity for proliferation resistance. Meeting nonproliferation acceptability involves a set of safeguards technologies. Materials that are not regarded as "self-protecting" can successfully be safeguarded by many other means. Therefore, meeting a Self-Protection Guideline is not in itself a figure of merit for nonproliferation

    Study requested is An Assessment of Alternative Technologies without Processing—Needs Assessment of Processing Case for Comparison. The EIS must include assessment of alternatives not restricted by the proposing organization. Alternatives currently in operation need to be assessed, even if different from national policy.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    Current operations provide actual cost data for comparison. The discussion covered experience, problems and solutions. The base case data have less uncertainty than for the alternatives included in the assessment.

    An Independent Study by DOE. The final environmental impact statement summary on policy concerning foreign research reactor spent fuel (DOE/EIS-0218F Feb. 1996, referred to as FEIS-96S) explains use of the EIS process, indicates that some spent fuel will be processed at SRL, and discusses (Summary p. 23) considerations that would affect decisions to process spent fuel.

    The ROD on FRRFM and FEIS-96S call for an ''independent study on the nonproliferation and other implications of reprocessing of spent fuel from foreign research reactors" to be initiated in mid-1996. The panel heard a presentation on DOE's plans for this effort and its progress to date. The study, however, is being done by DOE, and the presentation stated that the Office of Arms Control and Nonproliferation is conducting it. A mid-1998 target date was indicated.

    Questions were raised about whether this approach could be justified as an independent study. Jon Wolfstal of DOE responded that one outside expert had prepared a preliminary draft and a second had reviewed and edited it. Despite statements of Mr. Wolfstal about following standard procedures for public comment and response, "process" does not produce credibility.

    Publication of a draft for comment is used to assist the DOE in preparing the final report. The comments that are received are addressed in a section of the final report, and generally some brief discussion is included about how comments were handled and why. This particular approach to "process" does not afford opportunity for discussion or feedback between commentors and the authors, nor is there any independent arbiter. It is recognized that the authors are the ones responsible for the report, but it is seldom made clear when Department or White House policy sets the tone or the conclusions, and makes open discussion of certain comments impossible.

    Since it is this same limited process that DOE intends to use in this "independent study," I am concerned that there will be no opportunity for open discussion or debate on the critical issue of evaluation of all alternatives including processing options and the implications of them. I

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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    have personally experienced frustration with the comment process on DOE decision documents. Therefore, I would be very dubious that the current approach would permit DOE to obtain the benefit of a truly independent study. A different approach may be advisable in resolving this issue.

    Can an Adequate Record to Support this Contention be Provided for Assessment? As noted above, the ROD on FRRFM and the FEIS-96S call for an "independent study on the nonproliferation and other implications of reprocessing of spent fuel from foreign research reactors." This statement makes it obvious that a study should include processing options among the alternatives. However, this was specifically not done in the Task Team Report.

    In my opinion, whether or not use of processing encourages plutonium processing in other nations is a matter of conjecture. I have observed that this opinion is widely shared both internationally and within the United States. Therefore, a discussion of this key point should be included for any technology option for which it is offered as a criterion for exclusion.

    Countless precedents for this point can be found in commercial nuclear power plant licensing cases. EISs were rejected and returned to their sponsoring organizations for failure to adequately document their assessment of alternatives, even if there were special reasons why those alternatives might not be chosen. Many of the Environmental Impact Statements that were prepared for nuclear reactor construction permits were found to be "incomplete" by the Council on Environmental Quality in the early 1970s because they did not deal exhaustively with alternatives to the proposed nuclear power plants.

    I made a presentation, along with Ruble Thomas of Southern Services and Charley Wylie of Duke Power Company, to Nuclear Regulatory Commission (USNRC) staff on the system planning concepts used by utilities in selecting among alternatives (about 1974; no reference available to me). We explained siting alternatives and fuel alternatives and the reasoning that was used to make successive decisions. The staff response was that these concepts were acceptable and logical, but that they should be described in environmental reports submitted by utilities

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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    and would be included in the environmental impact statements the USNRC would prepare. We pointed out that sometimes certain alternatives were impractical or unreasonable on their face, and argued that analyzing such alternatives should not be necessary in the EIS process, but we were advised to include all "reasonable" alternatives in our assessment.

    Use of Processing for Waste Management

    Processing has been used and will continue to be used as long as it is needed and facilities to do it are available for damaged or corroded spent fuel that might leak or cause contamination. However, DOE has taken the position that unless it is a health and safety matter, processing should not be used as a step in radioactive waste management.

    A point was made that other nations may be considering use of reprocessing as a step in their waste management program. Obviously, this was once the case in the United States. In response to questions, Mr. Wolfstal mentioned France and Taiwan as nations that wish to do this. France does it commercially. Taiwan needs our consent to do it with United States origin fuel. These nations are nonproliferation treaty members and have stated that they will not use commercial reprocessing for weapons purposes.

    It is certainly not clear to me that whether or not the United States uses processing to manage its assortment of DOE spent fuel and foreign research reactor spent fuel would be of any influence on these nations or any others.

    Mr. Wolfstal pointed out that the French appeared to use the U.S. acceptance of MOX for disposition of excess weapons plutonium as an argument for why the rest of the world should continue to use civil plutonium. (A copy of the cited Newsletter from France on the French nuclear program is attached.)

    Summary of Findings

    Proliferation Resistance. There is no actual difference in proliferation resistance between the several alternatives considered in the Task Team report. The same would be true if the base case involving

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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    further use of processing in canyon facilities at Savannah River were chosen.

    Following U.S. Policy. Where selection decisions are based on interpretation of U.S. policy rather than technical, cost-effectiveness and environmental merit, this must be explicitly stated and discussed in an EIS.

    Follow Basic EIS Requirements. All realistic alternatives must be treated in an EIS.

    International Implications. It is not credible to say that the choice of technology the United States makes for managing this spent fuel will have any effect on nonproliferation decisions other nations will make.

    Independence. It is difficult to accept a report done by the nuclear nonproliferation of DOE as independent in such a controversial area.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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    Topic: Metallurgy and Corrosion

    Consultant: Paul Shewmon

    Introduction

    The Westinghouse Savannah River (WSR) people are working only with Direct Disposal (D-D) and Melt-Dilute (M-D) processes for disposing of aluminum clad fuel so these are the only processes my comments will cover. The electrolytic process would use the M-D ingots as a uniform feed/electrode product. The WSR people agree that the aluminum clad fuel could also be reprocessed in the canyons at SR as long as these canyons are available. However, the return of fuel from research reactors could stretch on into the future until long after it would be economical to maintain the canyons in operation and then some other process like D-D or M-D would need to be used.

    Processing

    Direct Disposal. In this process the fuel element is dried; non-fuel-bearing material at the ends of the assemblies is removed; and the fuel-bearing material is sealed in a canister with dry air for shipment to interim storage and ultimately to a repository. The exact composition of the canister material has not yet been specified, but it will probably be an austenitic stainless steel. The compatibility of the fuel with the canister would not be a concern under repository conditions (temperatures) due to the inertness of the aluminum cladding and the stainless steel canister in dry air. It is expected that each fuel type can be put in a form acceptable for the repository. Requirements for the fuel form needed for a road-ready package and interim dry storage of aluminum-clad fuels received from basin storage have been recommended [1,2]. Also, the requirements for repository disposal of such fuel have been established and can be met [2].

    Each fuel type will require some adjustment of the process to put in an acceptable form. The packing density of the fuel in the canister will be low due to the irregular shape of the fuel, and free space built into the subassembly to allow cooling water flow.

    Melt-Dilute. Here the fuel assemblies will be dried and then melted in an induction furnace. The volatile fission product will be

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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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

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    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.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    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.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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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.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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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

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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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

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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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.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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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.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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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.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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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

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
    ×

    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.

    Suggested Citation:"Appendix D. Consultant Reports." National Research Council. 1998. Research Reactor Aluminum Spent Fuel: Treatment Options for Disposal. Washington, DC: The National Academies Press. doi: 10.17226/6099.
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    The U.S. Department of Energy (DOE) is preparing an environmental impact statement (EIS) for management of aluminum spent fuel from foreign and domestic research reactors, much of which is highly enriched in uranium-235. This EIS will assess the need for additional treatment and storage facilities at the Savannah River Site to accommodate the receipt of this fuel, and it also will assess and select a treatment technology to prepare this fuel for interim storage and eventual shipment to a repository for disposal.

    This National Research Council book, which was prepared at the request of DOE's Savannah River Office, provides a technical assessment of the technologies, costs, and schedules developed by DOE for eight alternative treatment options and the baseline reprocessing option. It also provides comments on DOE's aluminum spent fuel disposal program, a program that is slated to last for about 40 years and cost in excess of $2 billion.

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