Recent advances in the treatment of irradiated graphite: A review

The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrainmore » all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim storage, packaging, transportation, waste forms, waste treatment, decontamination and decommissioning issues; and low-level waste (LLW) and high-level waste (HLW) disposal.« less

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Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

The overall purpose of this study is to provide information and guidance to the Office of Environmental Management of the US Department of Energy (DOE) about the level of characterization necessary to dispose of DOE-owned spent nuclear fuel (SNF) The disposal option modeled was codisposal of DOE SNF with defense high-level waste (DHLW) A specific goal was to demonstrate the influence of DOE SNF, expected to be minor, in a predominately commercial repository using modeling conditions similar to those currently assumed by the Yucca Mountain Project (YMP) A performance assessment (PA) was chosen as the method of analysis The performance metric for this analysis (referred to as the 1997 PA) was dose to an individual; the time period of interest was 100,000 yr Results indicated that cumulative releases of 99Tc and 237Np (primary contributors to human dose) from commercial SNF exceed those of DOE SNF both on a per MTHM and per package basis Thus, if commercial SNF can meet regulatory performance criteria for dose to an individual, then the DOE SNF can also meet the criteria This result is due in large part to lower burnup of the DOE SNF (less time for irradiation) and to the DOE SNF's small percentage of the total activity (15%) and mass (38%) of waste in the potential repository Consistent with the analyses performed for the YMP, the 1997 PA assumed all cladding as failed, which also contributed to the relatively poor performance of commercial SNF compared to DOE SNF

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Update to Assessment of Direct Disposal in Unsaturated Tuff of Spent Nuclear Fuel and High-Level Waste Owned by U.S. Department of Energy

The fluoride salt–cooled high-temperature reactor (FHR) is a new reactor type that combines the graphite-matrix coated-particle fuel and graphite moderator from high-temperature gas-cooled reactors...

https://www.tandfonline.com/doi/pdf/10.13182/NT14-88

Spent Nuclear Fuel and Graphite Management for Salt-Cooled Reactors: Storage, Safeguards, and Repository Disposal

This paper presents a deterministic performance assessment for spent fuel from deep-burn modular high-temperature reactors (DBMHRs) in the proposed Yucca Mountain repository. Typical DBMHR designs ...

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Performance Assessment for Geological Disposal of Graphite Waste Containing TRISO Particles

This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options Fort St Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel Future US HTGRs will be based on the Fort St Vrain (FSV) fuel form The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel From the standpoint of process cost and schedule (not considering repositorymore » cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing« less

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A.L. Lotts

Papers

Options for treating high-temperature gas-cooled reactor fuel for repository disposal