John D. Stempien

Characteristics of Composite Silicon Carbide Fuel Cladding after Irradiation under Simulated PWR Conditions

Silicon carbide possesses a high melting point, low chemical activity, no appreciable creep at high temperatures, and a low neutron absorption cross section, making it an attractive material to investigate for use as fuel cladding in light water reactors. The cladding design investigated herein consists of three layers: an inner monolith of SiC, a central composite layer of SiC fibers infiltrated with SiC, and an outer SiC coating to protect against corrosion. The inner monolith provides strength and hermeticity for the tube, and the composite layer adds strength to the monolith while providing a pseudo-ductile failure mode in the hoop direction. The tube may be sealed by bonding SiC end caps to the SiC tube. A number of samples were irradiated in a test loop simulating pressurized water reactor coolant and neutronic conditions at the Massachusetts Institute of Technology research reactor. Postirradiation hoop stress testing via internal pressurization revealed 10% to 60% strength reduction due to physical properties mismatches between the three layers and corrosion. Weight loss measurements indicated that some irradiation-assisted corrosion occurred. Scanning electron microscope analysis allowed determination of the fracture mechanisms for specimens ruptured during hoop testing. The thermal diffusivities of the as-fabricated three-layer tube samples were measured to be roughly three times lower than those of the as-fabricated monolith layer. With irradiation, the thermal diffusivities decreased by factors of 14 and 8 for the monolith and three-layered samples, respectively. This change may be attributed to radiation damage and the formation of a silica layer on the sample surface. Anisotropic swelling of the bonded α-SiC blocks was sufficient to fail five of the six bond test specimens after a 1.5-month irradiation. Two of each of the calcium aluminate and Ti foil bonded samples failed. One of two TiC/SiC bond samples survived.

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors: Status, Challenges, and Path Forward

Three advanced nuclear power systems use liquid salt coolants that generate tritium and thus face the common challenges of containing and capturing tritium to prevent its release to the environment. The fluoride salt–cooled high-temperature reactor (FHR) uses clean fluoride salt coolants and the same graphite-matrix coated-particle fuel as high-temperature gas-cooled reactors. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt with release of fission product tritium into the salt. In most FHR and MSR systems, the baseline salts contain lithium where isotopically separated 7Li is proposed to minimize tritium production from neutron interactions with the salt. The Chinese Academy of Sciences plans to start operation of a 2-MW(thermal) molten salt test reactor by 2020. For high-magnetic-field fusion machines, the use of lithium enriched in 6Li is proposed to maximize tritium generation—the fuel for a fusion machine. Advances in superconductors that enable higher power densities may require the use of molten lithium salts for fusion blankets and as coolants. Recent technical advances in these three reactor classes have resulted in increased government and private interest and the beginning of a coordinated effort to address the tritium control challenges in 700°C liquid salt systems. We describe characteristics of salt-cooled fission and fusion machines, the basis for growing interest in these technologies, tritium generation in molten salts, the environment for tritium capture, models for high-temperature tritium transport in salt systems, alternative strategies for tritium control, and ongoing experimental work. Several methods to control tritium appear viable. Limited experimental data are the primary constraint for designing efficient cost-effective methods of tritium control.

Tritium Management and Control Using Carbon in a Fluoride-Salt-Cooled High-Temperature Reactor

Abstract Research characterizing hydrogen behavior on carbon has been primarily focused on collecting data at near-ambient temperatures and pressures for storage or for high volume applications such as fusion. Transport models of a pre-conceptual 236 MWt pebble-bed fluoride-salt-cooled, high-temperature reactor (PB-FHR) estimate that the production of tritium is relatively low resulting in partial pressures ranging between 0 and 20 Pa. Operating temperatures in an FHR range from 600 to 700°C. Under these operating conditions, the interaction between hydrogen and carbon is currently undefined. Since an FHR contains large quantities of carbon (reflectors, fuel, structures), the tritium behavior in carbon must be investigated in order to develop methods to control tritium release rates to the environment and material corrosion. Preliminary modeling and experiments demonstrate high performance is achieved in a carbon adsorption tower, which can reduce system release rates by greater than 99%. This research aims to (1) accurately measure hydrogen uptake and kinetics on different types of carbon at prototypic conditions and (2) use tritium transport modeling to demonstrate the potential of carbon materials for tritium capture and control.

Understanding and Pathways to Avoid Major Fuel Failures and Radionuclide Releases in Fluoride Salt–Cooled High-Temperature Reactor Severe Accidents

Abstract Fluoride salt–cooled High-temperature Reactors (FHRs) are a new type of power reactor that delivers heat to the power cycle between 600°C and 700°C. The FHR uses High-Temperature Gas-cooled Reactor (HTGR) graphite-matrix coated-particle fuel with failure temperatures of ~1650°C. The FHR coolants are clean fluoride salts that have melting points above 350°C and boiling points above 1400°C. This combination may enable the design of a large FHR that will not have significant fuel failure and thus radionuclide releases to the environment even in a beyond-design-basis accident (BDBA) that include failure of all cooling systems, the vessel, and containment systems. A first effort has been undertaken to understand FHR BDBAs and develop an FHR BDBA system to prevent major fuel failure if an accident occurs in a large FHR. Four design features limit BDBA fuel temperatures to lower than fuel failure temperatures. First, there is a large temperature drop to transfer decay heat from the fuel to the environment in a BDBA. Second, the large temperature difference between normal operating temperatures and fuel failure temperatures allows the use of increasing temperatures in an accident to degrade the insulation system and other barriers that prevent efficient transfer of decay heat from the reactor core to the environment in an accident. Third, the silo around the reactor vessel contains a BDBA salt that in an accident heats up, melts, and partly floods the silo to improve heat transfer from fuel to the environment. Fourth, the fuel and coolant retain fission products and actinides at high temperatures.