Emory D Collins

Chemical partitioning technologies for an ATW system

Abstract A roadmap for the development of the technology of an Accelerator Transmutation of Waste (ATW) system was recently submitted to the U.S. Congress by the U.S. Department of Energy. One element of this roadmap was a development plan for the separations technologies that would be required to support an ATW system operating with a sustained feed of 1,450 tonnes of commercial light water reactor spent fuel per year. A Technical Working Group was constituted to identify appropriate separations processes and prepare a plan for their development. The baseline process selected combines aqueous and pyrochemical processes to enable efficient separation of uranium, technetium, iodine, and the transuranic elements from LWR spent fuel in the head-end step. For the recycle of unburned transuranics and newly-generated technetium and iodine from irradiated ATW transmuter assemblies, which were given to be metallic in form, a second and quite different pyrochemical process was identified. The diversity of processing methods was chosen for both technical and economic factors; aqueous methods are deemed to be better suited to large tonnages of commercial oxide spent fuel, while it is considered that pyrochemical processes can be exploited effectively in smaller-scale operations, particularly when the application is to metallic fuels or targets. A six-year technology evaluation and development program is foreseen, by the end of which an informed decision can be made on proceeding with demonstration of the ATW system.

Advanced reprocessing for fission product separation and extraction

Abstract: The United States inventory of used nuclear fuels contains approximately 2 to 5 wt % fission products, depending on the extent of fuel burnup during irradiation, with the greater amounts produced in higher burnup fuels. For reprocessing of used nuclear fuels, fission products are more often divided into categories according to their chemical and radiological properties. Advanced reprocessing includes further separations processes to enable capture and disposal of the volatile fission product elements in improved solid waste forms, as well as additional separations processes being developed (1) to enable recovery and recycle of the remaining minor transuranium element actinides, neptunium, americium, and curium, and (2) to segregate the lanthanide fission products and the intermediate-lived heat-generating radionuclides, 137 Cs/ 137m Ba and 90 Sr/ 90 Y.

The Influence of Lewis Acid/Base Chemistry on the Removal of Gallium by Volatility from Weapons-Grade Plutonium Dissolved in Molten Chlorides

Abstract It has been proposed that GaCl3 can be removed by direct volatilization from a Pu-Ga alloy that is dissolved in a molten chloride salt. Although pure GaCl3 is quite volatile (boiling point: 201°C), the behavior of GaCl3 dissolved in chloride salts is quite different because of solution effects and is critically dependent upon the composition of the solvent salt (i.e., its Lewis acid/base character). In this technical note, the behavior of gallium in prototypical Lewis acid and Lewis base salts is contrasted. It is found that gallium volatility is suppressed in basic melts and promoted in acidic melts. These results have an important influence on the potential for simple gallium removal in molten salt systems.

Analysis of Data from Leaching Concrete Samples Taken from the Three Mile Island Unit 2 Reactor Building Basement

Samples of contaminated concrete from the basement of the reactor building at Three Mile Island Unit 2 were tested and analyzed at Oak Ridge National Laboratory to determine the potential for decon...

Development and Operation of a Unique Conversion/Solidification Process for Highly Radioactive and Fissile Uranium

A unique evaporation/thermal denitration process was developed, operated, and maintained successfully to permit the solidification and safe storage of --1000 kg of highly radioactive and fissile uranium, containing --75% /sup 235/U, --10% /sup 233/U, and --140 ppm /sup 232/U. The project, called the Consolidated Edison Uranium Solidification Program, was carried out to prepare a stable uranium form for long-term, safe storage. During the project, the uranium nitrate solution was divided into --400 batches, which were successfully processed. Details of the process development, equipment maintenance, and operating expertise are described.

Reestablishing the Supply of Plutonium-238

The US Department of Energy has presented a plan to use existing reactors at Oak Ridge National Laboratory (ORNL) and Idaho National Laboratory (INL) and processing facilities at ORNL, modified as needed, to produce Pu. The basic capabilities that need to be put into place to produce new Pu are (1) neptunium storage and transport, (2) target fabrication, (3) target irradiation, and (4) chemical processing of irradiated targets to recover Pu. Neptunium currently in storage at INL will be shipped to ORNL during CY 2015. The target design has progressed to a prototypic target design that is expected to be used for production. Initial chemical processing experiments have shown successful recovery of neptunium and plutonium, but overall product purity has not been as high as desired.

ORNL experience and perspectives related to processing of thorium and 233U for nuclear fuel

Abstract Thorium-based nuclear fuel cycles have received renewed attention in both research and public circles since about the year 2000. Much of the attention has been focused on nuclear fission energy production that utilizes thorium as a fertile element for producing fissionable 233U for recycle in thermal reactors, fast reactors, or externally driven systems. Lesser attention has been paid to other fuel cycle operations that are necessary for implementation of a sustainable thorium-based fuel cycle such as reprocessing and fabrication of recycle fuels containing 233U. This paper first identifies recent literature that has resulted from the renewed interest in thorium-based fuel cycles. Next, differences in the radiation characteristics of nuclear materials associated with thorium-based and uranium-based fuels are discussed, and the generic implications of the differences to nuclear material processing are identified. Then, experience at Oak Ridge National Laboratory concerning processing of thorium and 233U is described in terms of the processing projects and campaigns undertaken and the facilities in which the processing was implemented. This experience then provides the basis for a generalized discussion of processing nuclear materials associated with thorium-based fuel cycles as compared to uranium-based fuel cycles. This comparative discussion focuses on key out-of-reactor fuel cycle operations: reprocessing of metal-clad oxide and graphite-matrix oxide used nuclear fuels (UNFs) including head-end processing (shearing and dissolution), solvent extraction, product conversion, fuel fabrication, and waste management. It is concluded that the recycle of thorium-based UNF constituents (233U and thorium) is more technically challenging than the recycle of uranium-based UNF constituents (plutonium and uranium) based on the radiation, chemical, and physical characteristics of nuclear materials in thorium-based fuel cycles as compared to uranium-based fuel cycles.

Complete Non-Radioactive Operability Tests for Cladding Hull Chlorination

Non-radioactive operability tests were made to test the metal chlorination reactor and condenser and their accessories using batch chlorinations of non-radioactive cladding samples and to identify optimum operating practices and components that need further modifications prior to installation of the equipment into the hot cell for tests on actual used nuclear fuel (UNF) cladding. The operability tests included (1) modifications to provide the desired heating and reactor temperature profile; and (2) three batch chlorination tests using, respectively, 100, 250, and 500 g of cladding. During the batch chlorinations, metal corrosion of the equipment was assessed, pressurization of the gas inlet was examined and the best method for maintaining solid salt product transfer through the condenser was determined. Also, additional accessing equipment for collection of residual ash and positioning of the unit within the hot cell were identified, designed, and are being fabricated.

Advanced thermal denitration conversion processes for aqueous-based reprocessing and recycling of spent nuclear fuels

This chapter focuses on advanced thermal denitration methods for conversion of nitrate solutions of uranium and plutonium to oxide particles with good ceramic properties for fabrication of recycle fuel. Specifically, the Japanese microwave heating process and the U.S. modified direct denitration process are described and compared, with emphasis on co-conversion of mixed uranium-plutonium nitrate solutions to produce mixed oxide. Process chemistry, processing equipment, oxide product characteristics, and co-conversion process comparisons are described.

Zirconium Recycle Test Equipment for Hot Cell Operations

The equipment components and assembly support work were modified for optimized, remote hot cell operations to complete this milestone. The modifications include installation of a charging door, Swagelok connector for the off-gas line between the reactor and condenser, and slide valve installation to permit attachment/replacement of the product salt collector bottle.