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Fusion Science and Technology
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Glass strategy: Hanford’s enhanced waste glass program
The mission of the Department of Energy’s Office of River Protection (ORP) is to complete the safe cleanup of waste resulting from decades of nuclear weapons development. One of the most technologically challenging responsibilities is the safe disposition of approximately 56 million gallons of radioactive waste historically stored in 177 tanks at the Hanford Site in Washington state.
ORP has a clear incentive to reduce the overall mission duration and cost. One pathway is to develop and deploy innovative technical solutions that can advance baseline flow sheets toward higher efficiency operations while reducing identified risks without compromising safety. Vitrification is the baseline process that will convert both high-level and low-level radioactive waste at Hanford into a stable glass waste form for long-term storage and disposal.
Although vitrification is a mature technology, there are key areas where technology can further reduce operational risks, advance baseline processes to maximize waste throughput, and provide the underpinning to enhance operational flexibility; all steps in reducing mission duration and cost.
R. Lässer, D. K. Murdoch, M. Glugla
Fusion Science and Technology | Volume 48 | Number 1 | July-August 2005 | Pages 337-342
Technical Paper | Tritium Science and Technology - Tritium Measurement, Monitoring, and Accountancy | doi.org/10.13182/FST05-A938
Articles are hosted by Taylor and Francis Online.
Unexpectedly large tritium amounts were trapped in the Plasma Facing Components of JET and TFTR during the respective tritium campaigns. Newly created co-deposited layers of carbon and hydrogen were identified as the main sinks. The first wall of ITER in contrast to JET and TFTR will be covered with beryllium, whereas the divertor tiles will be built of tungsten with the exception of a relatively small area of carbon fibre composites. Due to these three materials the composition of the newly created layers will change as a function of plasma operation. Their possible hydrogen content is not known yet and as a consequence the estimates of potentially trapped tritium differ strongly. To respect safety limits measurements of the mobilisable tritium inventories inside the vacuum vessel are required. The present strategy is to rely on the accountancy of the accessible tritium inside the fuel cycle and to derive the quantity of tritium trapped inside the vessel by difference. The tritium injected into the machine is only measured by mass flow meters and no effort is made to determine the tritium exhausted.Enhancements to determine the tritium and deuterium amounts injected into the torus and first proposals for enabling accountancy of the tritium and deuterium released from the torus cryo-pumps on a shot-by-shot basis are given. Only few additional buffer volumes and a micro gas chromatograph are required as the solutions are simple and inexpensive. These tools could be used already in the H-phase of ITER to obtain an integral value of the hydrogen trapped in the co-deposited layers by simple addition of small concentrations of deuterium to the protium and measuring the injected and released deuterium amounts.