David Carpenter

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.

Irradiation Testing of High-Power-Density Vibropacked Annular Fuel

This paper describes an irradiation test of high-power-density internally and externally cooled annular fuel samples in the 5-MW Massachusetts Institute of Technology (MIT) research reactor MITR-II. The design of the irradiation facility is briefly reviewed, with an emphasis on the thermal-hydraulic behavior of the irradiation capsules. The irradiation test is described, including the thermal history of the two irradiated samples. A discussion of the observed asymmetrical temperature profiles is provided. Results of preliminary postirradiation examination consisting of collimated gamma scans of the irradiation capsules to confirm burnup estimates and estimate fission gas release (FGR) are also presented. It is concluded that the vibropacked fuel samples’ FGR is below 1%, and that is within the predictable range by a specially equipped FRAPCON model.

Progress towards developing neutron tolerant magnetostrictive and piezoelectric transducers

Current generation light water reactors (LWRs), sodium cooled fast reactors (SFRs), small modular reactors (SMRs), and next generation nuclear plants (NGNPs) produce harsh environments in and near the reactor core that can severely tax material performance and limit component operational life. To address this issue, several Department of Energy Office of Nuclear Energy (DOE-NE) research programs are evaluating the long duration irradiation performance of fuel and structural materials used in existing and new reactors. In order to maximize the amount of information obtained from Material Testing Reactor (MTR) irradiations, DOE is also funding development of enhanced instrumentation that will be able to obtain in-situ, real-time data on key material characteristics and properties, with unprecedented accuracy and resolution. Such data are required to validate new multi-scale, multi-physics modeling tools under development as part of a science-based, engineering driven approach to reactor development. It is n...

Radiation resilient fiber Bragg grating sensors for sensing applications in nuclear reactor cores

This paper reports testing results of radiation resilient fiber Bragg grating (FBG) in radiation resistant fibers in the nuclear reactor core at MIT Research Reactor Lab. FBGs were fabricated by 140-fs ultrafast laser pulse using a phase mask approach. In-core test of fiber Bragg gratings was carried out in the core region of a 6-MW research reactor at temperature > 600°C and an average fast neutron (>1 MeV) flux >1×1014 n/s/cm2. First 100-day tests of FBG sensors shows less than 5 dB reduction in FBG peak strength after over 1×1020 n/cm2 of accumulated fast neutron dosage. To test temporal responses of FBG sensors, a number of reactor anomaly events were artificially created to abruptly change reactor power, temperature, and neutron flux over short periods of time. The thermal optical coefficients and temporal responses of FBG sensors are determined at different accumulated dosages of neutron flux. Results presented in this paper reveals that temperature-stable Type-II FBGs fabricated in radiation-hardened fibers could be used as sensors to perform in-pile measurements to improve safety and efficiency of existing and next generation nuclear reactors.

Tritium Control and Capture in Salt-Cooled Fission and Fusion Reactors

Abstract Three advanced power systems use liquid salt coolants that generate tritium and thus face common challenges to prevent release of the tritium to the environment. The Fluoride-salt-cooled High-temperature Reactor (FHR) uses the same graphite-matrix coated-particle fuel as High-Temperature Gas-cooled Reactors (HTGRs) and clean fluoride salt coolants. Molten salt reactors (MSRs) dissolve the fuel in a fluoride or chloride salt and release the fission product tritium to the salt. High-magnetic-field fusion machines may use liquid salt cooling and blankets because of the very high power densities of this new class of fusion machine. The three technologies can be coupled to a Nuclear Air-Brayton Combined Cycle (NACC) enabling variable electricity with base-load reactor operation. Converging requirements for tritium control in 700°C liquid salts are leading to cooperative programs across technologies; tritium models that combined generation, chemistry, metal corrosion and transport; and new tritium control technologies using advanced carbon forms, metals produced by additive manufacturing and other technologies.

Tritium Production and Partitioning from the Irradiation of Lithium-Beryllium Fluoride Salt

Abstract The MIT Nuclear Reactor Laboratory (NRL) has irradiated lithium-beryllium fluoride (flibe) salt as part of an on-going U.S. Department of Energy-funded Integrated Research Project to develop a Fluoride Salt High-Temperature Reactor (FHR). As part of this project, the NRL has carried out two irradiations of FHR materials in static flibe at 700°C in the MIT Research Reactor. These irradiations marked the start of a program evaluating the tritium production and release from the fluoride salt system at high temperature; in particular, there is interest in the evolution of tritium from the salt into solid materials and cover gasses. This paper describes the experience gained from the irradiation of flibe with respect to the detection of tritium. It covers the development of techniques for monitoring the evolution of tritium from the salt during irradiation and the factors particular to the FHR system that influence this process, including the radiolytic production and release of volatile fluorine and fluoride products as a function of temperature. In addition, it discusses the measurement of tritium partitioning between the different materials in the experiment due to the confluence of diffusion, adsorption, and chemical and radiolytic reactions.

Fuel Management of PWR Cores with Silicon Carbide Cladding

Abstract The primary motivation for using silicon carbide rather than zirconium alloy cladding is its putative improvement in accident resistance, due to slow reactions with water, even at high temperatures. But, fuel management performance will also be an important consideration in its commercial acceptance. Whether backfittable 18- and 24-month cycles can be designed for existing light water reactors, their enrichments, operating characteristics, and fuel costs are questions that the present study undertakes to answer. Also evaluated is the possibility of leveraging silicon carbide’s ability to sustain higher fuel duty for increasing power levels and discharge burnups in pressurized water reactors. A preliminary design using fuel rods with the same dimensions as in typical Westinghouse fuel, but with fuel pellets having a 10 vol % central void, has been adopted to mitigate the higher fuel temperatures when silicon carbide is used. This allows design of 18- and 24-month cycles that meet present-day operating constraints on peaking factor, boron concentration, reactivity coefficients, and shutdown margin, while achieving batch average discharge burnups up to 80 MWd/kg U, as well as power uprates of 10% and possibly 20%. Control rod configuration modifications may be required to meet the shutdown margin criterion for the 20% uprate. For nonuprated cores, silicon carbide–clad fuel may have a fuel cost advantage, especially with increasing discharge burnup, provided the fuel manufacturing cost is close to that of Zircaloy tubes. The economics of the fuel cycle also improve with power uprates, as the value of the additional energy generated may substantially exceed the advantage from fuel cost alone.