Mitchell T. Farmer

The objective of the Versatile Test Reactor (VTR) is to enable testing of advanced reactor fuels and materials in a fast spectrum neutron environment. Internal cartridge loop testing capabilities are being developed that will allow the cartridge coolant to be isolated from the reactor coolant. This approach will allow various cartridge coolants to be investigated, thereby maximizing testing capability. A sodium cartridge loop testing capability is being developed by a team that includes Argonne National Laboratory (Argonne) as the laboratory partner, Framatome Inc. as the industrial partner, and Purdue University along with the University of Wisconsin–Madison as university partners. Specific elements of the current work include overall cartridge loop design development that is being led by Framatome, Inc. Coolant chemistry monitoring and control are key elements of any high-pedigree irradiation testing capability; the University of Wisconsin is leading this effort by developing and experimentally verifying methods for achieving this capability in pile. Purdue University is developing a scaling methodology, and on that basis, a thermal-hydraulic testing capability to validate fluid flow and heat transfer models for the cartridge that will be used to support design and safety analysis activities. Argonne has focused on developing and testing technologies specifically targeted at simplifying VTR operations, as well as developing modeling tools to support cartridge loop design and safety analysis. The purpose of this paper is to summarize the current status of the sodium fast reactor cartridge development, including details on the cartridge functional requirements, physical design, chemistry control, operations, and safety.

The Three Mile Island Unit 2 (TMI-2) accident, which occurred on March 28, 1979, led industry and regulators to enhance strategies to protect against severe accidents in commercial nuclear power plants. Investigations in the years after the accident concluded that at least 45% of the core had melted and that nearly 19 tonnes of the core material had relocated to the lower head. Postaccident examinations indicate that about half of that material formed a solid layer near the lower head and above it was a layer of fragmented rubble. As discussed in this paper, numerous insights related to pressurized water reactor accident progression were gained from postaccident evaluations of debris, reactor pressure vessel (RPV) specimens, and nozzles taken from the RPV. In addition, information gleaned from TMI-2 specimen evaluations and available data from plant instrumentation were used to improve severe accident simulation models that form the technical basis for reactor safety evaluations. Finally, the TMI-...