Nuclear characterization of a general-purpose instrumentation and materials testing location in TREAT

Bess, John D.; Woolstenhulme, Nicolas E.; Jensen, Colby; Parry, James R.; Hill, Connie M.

doi:10.1016/j.anucene.2018.10.011

Cited by 19 publications

(4 citation statements)

References 19 publications

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“…The capsule is placed within the Minimal Activation Retrievable Capsule Holder (MARCH) and the capsule holder is placed within the Broad Use Specimen Transient Experiment Rig (BUSTER). Details for both MARCH and BUSTER are given in reference [6].…”

Section: Sirius-3 Experiments Descriptionmentioning

confidence: 99%

Thermal Analysis of the SIRIUS3 Nuclear Propulsion Fuel Calibration Experiment

Gleicher,

Jaradat,

Schunert

et al. 2023

Nuclear and Emerging Technologies for Space (NETS 2023)

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To support long space missions, NASA is designing and testing different fuels for nuclear thermal propulsion. These designs encompass ceramic metallic fuels which are to be irradiated in the TREAT reactor to test the fuel's performance. This type of fuel is a ceramic fuel type dispersed in a metallic matrix. A set of experiments named the SIRIUS experiments are designed to test the performance of ceramic metallic fuels under conditions that are intended to be prototypical of mission conditions. These conditions include going from cold conditions ( 300K) to very hot conditions (2700K-3000K) in a very short amount of time (tens of seconds). The SIRIUS-3 experiment tests a column of 16 fuel specimens under irradiation conditions. For this paper, a Bison model is constructed and thermal analysis of the SIRIUS-3 experiment is performed. These analyses reveal the importance of different local experimental components to characterize the heat transport phenomena of the ceramic metallic fuel. This modeling and analysis also informs predictive modeling for future experiments such as the SIRIUS-2C and SIRIUS-5 tests, which will be a ceramic fuel type dispersed in a ceramic matrix.

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Section: Sirius-3 Experiments Descriptionmentioning

confidence: 99%

Thermal Analysis of the SIRIUS3 Nuclear Propulsion Fuel Calibration Experiment

Gleicher,

Jaradat,

Schunert

et al. 2023

Nuclear and Emerging Technologies for Space (NETS 2023)

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“…Hence, by combining this capsule-in-pipe mechanical layout with capsule materials that do not transmute into significant radioisotopes (principally titanium alloys), the MARCH system enabled fresh fuel capsules to be irradiated, removed from BUSTER on the TREAT working floor by using the storage holes, and shipped for post-transient exams using glovebox facilities, all in a matter of weeks. A detailed characterization of the BUSTER nuclear environment was performed via Monte Carlo neutronics modeling and can be found in [21]. This approach enables BUSTER to function as a reusable device manufactured in accordance with exacting pressure vessel code and quality assurance requirements, whereas capsules are typically treated as consumable hardware with function-specific engineering requirements.…”

Section: Current Efforts and Future Outlookmentioning

confidence: 99%

The Transient Reactor Test Facility (TREAT)

Woolstenhulme¹

2022

Nuclear Reactors - Spacecraft Propulsion, Research Reactors, and Reactor Analysis Topics

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Constructed in the late 1950s, the Transient Reactor Test facility (TREAT) provided numerous transient irradiations until operation was suspended in 1994. It was later refurbished, and resumed operations in 2017 to meet the data needs of a new era of nuclear fuel safety research. TREAT uses uranium oxide dispersed in graphite blocks to yield a core that affords strong negative temperature feedback. Automatically controlled, fast-acting transient control rods enable TREAT to safely perform extreme power maneuvers—ranging from prompt bursts to longer power ramps—to broadly support research on postulated accidents for many reactor types. TREAT’s experiment devices work in concert with the reactor to contain specimens, support in situ diagnostics, and provide desired test environments, thus yielding a uniquely versatile facility. This chapter summarizes TREAT’s design, history, current efforts, and future endeavors in the field of nuclear-heated fuel safety research.

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“…Simulation tools, such as BISON, maximize the value of separate effects tests by allowing iterative model improvements and pre-test predictions of behavior. Similar tools to MiniFuel, such as FAST, BUSTER, and SETH, have been developed at Idaho National Laboratory over the last few years (Bess et al, 2019;Beausoleil et al, 2020;Terrani et al, 2020). In the present work, the MiniFuel irradiation capability was leveraged to study the effects of temperature on the swelling behavior of two U-Mo alloys with increasing burnup.…”

Section: Introductionmentioning

confidence: 99%

Accelerated fission rate irradiation design, pre-irradiation characterization, and adaptation of conventional PIE methods for U-10Mo and U-17Mo

Doyle¹,

Massey²,

Richardson³

et al. 2023

Front. Nucl. Eng.

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Metallic U alloys have high U density and thermal conductivity and thus have been explored since the beginning of nuclear power research. Alloys of U with modest amounts of Mo, such as U-10 wt % Mo (U-10Mo), are of particular interest because the γ-U crystal structure in this alloying addition shows prolonged stability in reactor service. Historically, radiation data on U-10Mo fuels were collected in Na fast reactors or lower temperature research reactor conditions, but little is known about irradiation behavior, particularly swelling and creep, at irradiation temperatures between 250 and 500°C. This work discusses the methodology and pre-irradiation characterization results from a U-Mo irradiation campaign performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. U-10Mo and U-17Mo samples irradiations are being completed at temperatures ranging from 250 to 500°C to three targeted fission densities between 2 × 1020 and 1.5 × 1021 fissions per cubic centimeter. Swelling measurement of the specimen sizes studied here required development and assessment of new methods for volume determination before and after irradiation. Laser profilometry and X-ray computation tomography (XCT) were used to provide preirradiation characterization of samples to determine the error and applicability of each to determine swelling following irradiation. These outcomes are contextualized through use of BISON simulations performed to assess the predicted expansion of U-Mo fuels subjected to the irradiation conditions of this work. Use of existing BISON fuel performance models predicted a maximum of 7% swelling under the irradiation conditions of this study. Pre-irradiation characterization revealed the as-cast U-Mo fuel samples were uniformly large-grained fully cubic U crystals with small U-C/N bearing precipitates and pores distributed throughout. Samples were found to contain a bulk porosity between .4 and 3% because of the casting process. Local porosity in areas far from large, interconnected pores was found by Slice-and-View to be under .2%. Nanometer-sized precipitates rich in C and N were identified in all samples, likely because of impurities during the fabrication process. Dendritic bands were also observed throughout the samples. These bands were characterized by variable Mo content that deviated from the overall Mo content by 2–3 wt %. No other microstructural features were correlated to these bands. Mechanical properties were found to be slightly strengthened compared to literature reports of bulk U-Mo fuels due to the nano-scale precipitates throughout the sample.

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Nuclear characterization of a general-purpose instrumentation and materials testing location in TREAT

Cited by 19 publications

References 19 publications

Thermal Analysis of the SIRIUS3 Nuclear Propulsion Fuel Calibration Experiment

Thermal Analysis of the SIRIUS3 Nuclear Propulsion Fuel Calibration Experiment

The Transient Reactor Test Facility (TREAT)

Accelerated fission rate irradiation design, pre-irradiation characterization, and adaptation of conventional PIE methods for U-10Mo and U-17Mo

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