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Taylor, Paul Allen

Publications and source records attributed to Taylor, Paul Allen.

Solidification of REDC Organics and Airborne Release Fraction Measurements

The Radiochemical Engineering Development Center (REDC) generates various liquid organic wastes from processing irradiated targets to recover heavy elements. In the past, these organics were discharged to the Oak Ridge National Laboratory (ORNL) liquid low-level waste (LLLW) system, along with the aqueous waste. Because of a reduction in aqueous LLLW from other ORNL generators and an increase in the radionuclide concentration, particularly 238 Pu, in the organic waste stream, the organics can no longer be discharged to the LLLW system. The plan is to solidify the liquid organic waste for disposal as solid debris waste. The documented safety analysis for REDC uses an airborne release fraction of 5 x 10 -4 with a respirable fraction of 0.2 for plutonium and other actinides from packaged solid debris waste during a fire scenario, for an overall airborne release fraction/respirable fraction of 10 -4 . Solidifying the organic liquids, either by using PM-199 Organoclay ® or by grouting, would produce solid waste forms with no free liquid, which should qualify for disposal at the Waste Isolation Pilot Plant. The use of Organoclay would be much simpler to implement inside a hot cell. Granular Organoclay can be added to the organic waste solutions until there is a dry layer on top of the liquid and then allowed to cure for a few hours. Grouting the liquids would require using multiple ingredients and mixing steps. Both methods produced waste forms that trapped cerium, as a surrogate for plutonium, inside the waste, with airborne release fractions during fire scenario tests of well below the 10 -4 limit.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Mu*STAR ADSR Fuel Conversion Facility Evaluation and Cost Analysis

The purpose of the reported effort is the development of a conceptual process for conversion of used nuclear fuel (UNF) from light water reactors (LWRs) to a fluoride salt form to facilitate its use for power generation in a Mu*STAR Accelerator-driven Subcritical Reactor (ADSR). This document evaluates potential processes for converting UNF into a fluoride salt, evaluates the costs/benefits of using enriched lithium in the carrier salt, and analyzes comparable costs of similar facilities to make a rough order of magnitude estimate for the cost to design, construct, and commission a Mu*STAR ADSR fuel conversion facility. This document also makes a high-level comparison of the Mu*STAR ADSR fuel cycle with other important fuel cycles. The results of the cost analysis show that the Mu*STAR ADSR fuel conversion facility will cost more than $100M, possibly up to around $1B, to design and construct. The cost will depend on the size of the hot cell required, which will depend on the processing options chosen for converting the UNF to a molten salt. Operations and maintenance of the facility will be on the order of $10M per year. The design, fabrication, and installation of the remote process equipment to support the fuel conversion will be an additional cost of more than $50M. Cost surveys of Department of Energy hot cell designs and previous Department of Energy hot cell construction suggest that the cost variance may be large, and the risk of escalating costs may be high in the regulation-mandated atmosphere of nuclear facility design and construction. The molten salt for the Mu*STAR ADSR contains all of the nonvolatile isotopes in the UNF, and there is never a separation of the plutonium from the fission products. This makes this conversion process much more proliferation resistant than the reprocessing options used for other closed fuel cycles. When normalized to the amount of electricity produced, the Mu*STAR ADSR fuel cycle’s expected cost might be less than any reprocessing fuel cycle and possibly less than the once-through cycle now used. Simulations show that once the conversion to fluoride is made, that up to seven times as much energy can be extracted from the UNF than was produced by the LWR while making it [Bowman, 2010]. Using once-only, on-site conversion to fluoride means no mining, no enrichment, no manufacturing of fuel rods, and no transportation costs. Typically, the UNF was produced over a 40-year period in an LWR; it can potentially be used for 280 years in Mu*STAR, producing the same power as the LWR did, ultimately requiring disposal of approximately the same volume of material as the LWR generated. Mu*STAR burns most of the higher actinides, which could reduce both the cost of disposal and significantly reduce the longevity of its radiotoxicity. Mu*STAR ADSR’s ability to demonstrate unique and valuable peripheral benefits are additional arguments for this technology, including (1) turning UNF into a valuable commodity (thus, its disposal will not be such an immediate problem, so continued operation of existing LWRs will be more acceptable); (2) burning fissile or fertile fuel including excess weapons-grade plutonium in subcritical mode; and (3) providing high-temperature process heat to convert natural gas to synthetic diesel fuel, water to hydrogen, or for other process-heat applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Solidified Organics Testing

The Radiochemical Engineering Development Center currently has plutonium-containing liquid organic waste that must be solidified before final disposal as solid debris waste. The documented safety analysis for the Radiochemical Engineering Development Center limits the amount of plutonium that can be released from solid debris waste forms during a fire scenario, which is less than 10 -4 airborne release fraction. Experimental results for plutonium release fractions from burning various materials has been compiled by the Department of Energy [DOE, 1994]. Combustion of organic liquids gave an average airborne release fraction of 6 × 10 -3 and combustion of surface-contaminated solids (paper, plastic, etc.) gave an average airborne release fraction of 5 × 10 -4 . Burning of 30% tri- n -butyl phosphate (TBP) in kerosene gave cerium release fractions of 5.6–7.7 × 10 -3 [Mishima and Schwendiman, 1973]. Earlier testing had evaluated many different grout-based formulations to solidify plutonium-containing organic waste liquids, and the ones used in these tests were shown to produce monolithic, fairly homogeneous solids. Each of the waste forms tested would meet the waste acceptance criteria for disposal at the Waste Isolation Pilot Plant as transuranic waste, but there were questions about whether they would meet the plutonium release criteria during a fire. The purpose of the work described here was to quantify the airborne release fraction for cerium, as a surrogate for plutonium, in these proposed waste forms. Cerium is often used as a simulant for plutonium in bench-scale testing, and cerium nitrate hexahydrate [Ce (NO 3 ) 3 ·6H 2 O] was added to the organic solutions that were solidified to prepare the surrogate waste forms for these tests. The waste forms were subjected to the designated fire scenario of 600ºC for 1 hour, and any cerium in the off-gas was trapped using a quartz filter followed by a scrubber containing 0.1 M NaOH, which is the same as used by Mishima and Schwendiman, [1973].

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