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McDuffee, Joel

Publications and source records attributed to McDuffee, Joel.

9 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Simulation of natural circulation cartridge loop experiments and application to molten salt reactors

This work uses the TRAC/RELAP Advanced Computational Engine (TRACE) thermal hydraulics (TH) code to model natural circulation cartridge loop experiments previously conducted at Oak Ridge National Laboratory (ORNL) using water and compares the simulated and experimental results. TRACE is also used to characterize natural circulation in the cartridge loop vehicle using FLiNaK as the working fluid. The experimental vehicle is a buoyancy-aided, annular cartridge loop, referred to as a thermosyphon, and is designed to aid in qualifying liquid–fueled and/or liquid–cooled irradiation experiments for the Versatile Test Reactor (VTR), which is currently being designed in the United States. Out-of-pile water experiments have been conducted using the cartridge and the Thermosyphon Test Loop facility at ORNL, and future experiments are anticipated that would use other molten salt surrogates as the working fluid, followed by eventual insertion of a similar cartridge into VTR. Additionally, this work aims to determine how well TRACE can replicate the natural convection conditions that were observed experimentally; this serves as an initial step for validating the modeling tool for design and safety calculations to support future irradiation experiments in VTR. Initial predictions of potential experiments were made using FLiNaK as the natural circulation fluid to demonstrate the relevance of the cartridge design to molten salt reactors (MSRs). Results from this study indicate that TRACE can accurately capture natural convection phenomena in the thermosyphon and that several design changes to the current cartridge vehicle are necessary to achieve hydraulic conditions similar to those expected in MSRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Plutonium-238 Production Program Results, Implications, and Projections from Irradiation and Examination of Initial NpO 2 Test Targets for Improved Production

An alternative target design with potential improvements, including a major increase in 238 Pu production rate and annual capacity; fewer targets to be fabricated, irradiated, and processed; and a significant replacement of a large volume of caustic-nitrate, aluminum-bearing radioactive liquid waste with a smaller volume of solid metal waste, has been conceived and evaluated using reactor physics and thermal-hydraulic analyses. The alternative target design uses pressed pellets of 237 NpO 2 , sintered to 92% to 93% of theoretical density, and stacked inside a Zircaloy-4 cladding tube. Additionally, four test targets were fabricated, irradiated, and examined. No melting or other potential problems were indicated. Projections from measured constituents indicated annual production could be increased by a factor of ~2, and the number of targets required to be fabricated, irradiated, and processed could be reduced by a factor of ~5.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Conceptual design of HFIR irradiation experiment for material compatibility study on liquid Sn divertor

Liquid Sn is one of the promising coolants for liquid surface divertor concept of fusion reactors. However, the compatibility between liquid Sn and structural materials is an important issue that has to be addressed, because liquid Sn is extremely corrosive to steels at high temperatures. The corrosion may be mitigated when a protective Al2O 3 layer is formed on the surface of alumina forming steels. However, the effect of neutron irradiation on the integrity of protective layer is not made clear so far. Japan and US joint research project “FRONTIER” started in 2019 to investigate the material compatibility under neutron irradiation. The purpose of the present study is to develop the conceptual design of the irradiation test capsule which enables material compatibility tests for the alumina forming steels - liquid metal systems under neutron irradiation in the High Flux Isotope Reactor at Oak Ridge National Laboratory, TN, USA. The three dimensional drawing of capsule structure was then developed. The validity of the material selections for the capsule design was investigated by means of corrosion tests of SiC, Si3N4, Ti, and Mo in liquid Sn at 773 K for 262 hr.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Wireless Instrumented RB Experiment Preliminary Design and Analysis

The ability to deploy new nuclear fuels for current or future reactor concepts requires carefully designed experiments to generate data to support fuel qualification. Ideally these experiments would include state of-the-art sensing to maximize the amount of in situ data that can be collected during operation. Furthermore, advanced reactor systems can take advantage of integrated in-core sensing technologies to maximize fuel utilization, reduce unnecessary conservativism in design margins, and improve operator’s understanding of limiting peaking factors. Before any novel sensing technologies can be readily adopted for nuclear applications, they must first demonstrate acceptable performance in test reactors. This report summarizes the preliminary design and analysis of the most highly instrumented irradiation experiment ever performed in the removable beryllium (RB) positions of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The Wireless Instrumented RB Experiment 2021 (WIRE-21) will test a wide range of sensors including wireless sensors being developed by Westinghouse Electric Company (WEC) that could provide in situ measurements of peak fuel temperatures and fuel rod pressurization due to fission gas release. The ability to wirelessly transmit a signal through the fuel rod’s cladding is critical to improving fuel monitoring capabilities without requiring signal penetrations through the cladding pressure boundary, which would significantly impact fuel fabrication, handling, and operation. Other sensors that will be tested in WIRE-21 include an array of thermocouples, self-powered neutron detectors (SPNDs), and spatially distributed fiber-optic temperature sensors. More generally, WIRE-21 will establish a flexible irradiation vehicle design to allow accelerated, economical testing of advanced sensor technologies while leveraging the extremely high neutron flux that is available in HFIR. This report summarizes the mechanical design for WIRE-21, the experimental test matrix, initial neutronic and thermal design analyses, and the active monitoring and control system enhancements necessary to support testing of advanced sensor technologies. The containment for WIRE-21 is similar to previous RB irradiation vehicles but includes a few modifications, most notably the use of integrated compression seals to pass a larger number of sensor leads through the experiment’s pressure boundary. In addition to the sensor leads, inert gas lines are passed into the experiment to enable active temperature control and the ability to pneumatically actuate a bellows-driven pressure sensor. WIRE-21 is targeting temperatures (300–350°C) and neutron fluence levels (~10 22 n/cm 2 ) relevant to light water reactors (LWRs), but the flexible design of the experiment vehicle allows much higher operating temperatures (>1,100°C). Neutronic calculations determine the neutron flux conditions as well as the nuclear heating within the experiments. These results are used as inputs to detailed thermal finite element calculations, which are required to evaluate the complex, three-dimensional heat transfer that occurs within WEC’s wireless sensor enclosures. Initial results show that the temperatures of the sensors’ enclosures and the metal bellows can be operated near the temperature range of LWR coolants and cladding while simultaneously increasing the temperature of a surrogate fuel material to values in the range of 800–1200°C to simulate centerline fuel temperatures during LWR operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗