Modeling reactivity insertion experiments of TRISO particles in NSRR using BISON

Schappel, Danny; Brown, Nicholas R.; Terrani, Kurt A.

doi:10.1016/j.jnucmat.2019.151965

Cited by 18 publications

(3 citation statements)

References 32 publications

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“…BISON is maintained at INL and is under development at a wide variety of US institutions (Williamson et al, 2021). BISON is most often used to model UO 2 fuels in light-water reactors but includes capabilities to model a wide array of fuel forms and reactor types, including U-Zr-based fuels for sodium-cooled fast reactors (Greenquist et al, 2021a; and tristructural isotropic fuels (Schappel et al, 2020). BISON's open architecture enables it to model a wide array of materials, conditions, and geometries, including MiniFuel disks (Cheniour et al, 2022).…”

Section: Bison Descriptionmentioning

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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Section: Bison Descriptionmentioning

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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“…Temperature Gas-Cooled Reactor (HTGR) is one of the advanced nuclear reactors based on tri-structural isotropic (TRISO) particle [1] which has two types of fuel structure: prismatic block and pebble bed [2,3]. The progress of utilizing fuel coated technology have been reported as pebble fuel form are utilized for High Temperature Gas-cooled Reactor Pebblebed Module (HTR-PM) [4] and the 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) in China [5]. Meanwhile, prismatic fuel is used for GT-MHR in USA and the High-Temperature Engineering Test Reactor (HTTR) in Japan [5].…”

Section: Highmentioning

confidence: 99%

“…The progress of utilizing fuel coated technology have been reported as pebble fuel form are utilized for High Temperature Gas-cooled Reactor Pebblebed Module (HTR-PM) [4] and the 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) in China [5]. Meanwhile, prismatic fuel is used for GT-MHR in USA and the High-Temperature Engineering Test Reactor (HTTR) in Japan [5].…”

Section: Highmentioning

confidence: 99%

The Conceptual Design for Liquid Effluent Treatment of Uo2 Kernel Fabrication

Yusnitha

Kadarjono²,

Sartono³

et al. 2020

urania

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THE CONCEPTUAL DESIGN FOR LIQUID EFFLUENT TREATMENT OF UO2 KERNEL FABRICATION. The pebble fuel for HTGR is prepared through fabrication of UO2 kernel, coated particle and spherical element fuel. In the fabrication of UO2 kernel by external gelation method, a multicomponent of liquid effluent is generated. Therefore, the liquid effluent is required to be treated for safety reason before disposed to waste storage. In this paper, the conceptual design for the liquid effluent treatment of UO2 kernel fabrication is performed with the simulation process using CHEMCAD software. CHEMCAD is a software that can be utilized for chemical process design. The results showed that the proposed conceptual design is able to separate valuable components: isopropyl alcohol (IPA) and tetrahydrofurfuryl alcohol (THFA). The flowrate of IPA product is 5.28 kg/h with purity of 0.99 in mass fraction and the flowrate of THFA product is 1.01 kg/h with purity of 0.99 in mass fraction.Keywords: liquid effluent, UO2 kernel, CHEMCAD, HTGR.

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Preliminary numerical investigation of TRISO-matrix interface debonding characteristics in fully ceramic microencapsulated fuel

Zhang

Wang

et al. 2021

Annals of Nuclear Energy

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Modeling reactivity insertion experiments of TRISO particles in NSRR using BISON

Cited by 18 publications

References 32 publications

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

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

The Conceptual Design for Liquid Effluent Treatment of Uo2 Kernel Fabrication

Preliminary numerical investigation of TRISO-matrix interface debonding characteristics in fully ceramic microencapsulated fuel

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