Design research for the porous hexagonal prism shaped fuel element adopted in nuclear thermal propulsion reactor based on neutronics and thermal-hydraulics coupled method

Zhang, Jun; Wang, Chenglong; Zhang, Jing; Zhang, Dalin; Tian, Wenxi; Qiu, Suizheng; Su, G.H.

doi:10.1016/j.anucene.2020.107847

Cited by 12 publications

(3 citation statements)

References 5 publications

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“…Uniaxial tension simulations were conducted for the fuel and the matrix, and the energy release rates (G c ) were adjusted iteratively until each material cracked at the correct fracture stress The phase-field fracture model was implemented using the finite element method with the Multiphysics Object-Oriented Simulation Environment (MOOSE), 38 as was done with the original model. 1 The J2 plasticity model in MOOSE was used to describe the constitutive behavior. Implicit time integration was used in the model.…”

Section: Materials Properties and Model Parametersmentioning

confidence: 99%

“…1. The gas also serves to cool the reactor, producing a temperature change in the axial direction of over 2000 K. 1 During operation, such a reactor would undergo multiple restart and short-burn propulsion cycles 2,3 for a combined burn time of under 2 h at very high temperature (2800 K to 3000 K). 2,4 In addition, the reactor may operate at low power and temperature between propulsion cycles to supply electrical power to the spacecraft 2,4 (hundreds of kW e to 1 MW e ).…”

Section: Introductionmentioning

confidence: 99%

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Review and Preliminary Investigation into Fuel Loss from Cermets Composed of Uranium Nitride and a Molybdenum-Tungsten Alloy for Nuclear Thermal Propulsion Using Mesoscale Simulations

Hirschhorn

Hilty

Tonks

et al. 2021

JOM

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A ceramic-metal composite (cermet) of uranium nitride (UN) particles embedded in a tungsten-molybdenum (W/Mo) alloy matrix is being considered as reactor fuel for nuclear thermal propulsion (NTP). One possible issue is the loss of fissile uranium atoms during reactor operation. We begin by reviewing historical data that suggest that a likely mechanism for fuel loss is transport of free uranium along metal grain boundaries and through cracks to free surfaces. We then employ simple two-dimensional (2D) mesoscale simulations to provide insights into crack formation and free uranium transport in a W/ Mo-UN cermet. Phase-field fracture simulations show that cracks form during cooling at the fuel-matrix interface and then within the fuel particles. Transport simulations show that cracks at the fuel-matrix interface and within the fuel accelerate fuel loss. Mechanical and diffusion data are needed for UN and W/Mo alloys to make these preliminary predictions more accurate.

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Section: Materials Properties and Model Parametersmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Review and Preliminary Investigation into Fuel Loss from Cermets Composed of Uranium Nitride and a Molybdenum-Tungsten Alloy for Nuclear Thermal Propulsion Using Mesoscale Simulations

Hirschhorn

Hilty

Tonks

et al. 2021

JOM

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“…The propellant flows through the subchannels of the fuel elements and absorbs thermal energy, producing an axial temperature difference exceeding 2000 K (Ref. 6). Then, the hightemperature hydrogen gas is exhausted through a nozzle and produces a reaction force that moves the rocket forward.…”

Section: Introductionmentioning

confidence: 99%

Multiscale Simulations of Thermal Transport in W-UO₂ CERMET Fuel for Nuclear Thermal Propulsion

Sessim

Tonks

2021

Nuclear Technology

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Nuclear thermal propulsion (NTP) provides a consistent source of thrust for long space missions. However, fuel development for NTP reactors is a major technological hurdle. Existing modeling and simulation tools developed by the U.S. Nuclear Engineering Advanced Modeling and Simulation (NEAMS) program for power reactors can be leveraged to help accelerate the fuel development. This work is a preliminary demonstration of the application of NEAMS tools to model NTP fuel. Specifically, the fuel performance tool BISON and the mesoscale reactor materials tool MARMOT are used to develop a multiscale model of thermal transport in a W-UO 2 CERMET fuel element for NTP reactors. Three-dimensional simulations in MARMOT are used to estimate the effective thermal conductivity (ETC) of fresh CERMET fuel at temperatures ranging from 1500 K to 3000 K. The ETC values from MARMOT are then used in BISON simulations that predict the steady-state temperature profile throughout a 61-subchannel hexagonal fuel element. The temperature varies by 83 K throughout the fuel element, with the highest temperature occurring near the outer edges of the element. BISON is also used to show that the temperature profile in prototype fuel elements with fewer subchannels does not vary significantly from that in the 61-subchannel element.

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Thermal hydraulic modeling of solid-fueled nuclear thermal propulsion reactors part II: Full-core coupled neutronic and thermal hydraulic analysis

Krecicki

Kotlyar

2022

Annals of Nuclear Energy

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Design research for the porous hexagonal prism shaped fuel element adopted in nuclear thermal propulsion reactor based on neutronics and thermal-hydraulics coupled method

Cited by 12 publications

References 5 publications

Review and Preliminary Investigation into Fuel Loss from Cermets Composed of Uranium Nitride and a Molybdenum-Tungsten Alloy for Nuclear Thermal Propulsion Using Mesoscale Simulations

Review and Preliminary Investigation into Fuel Loss from Cermets Composed of Uranium Nitride and a Molybdenum-Tungsten Alloy for Nuclear Thermal Propulsion Using Mesoscale Simulations

Multiscale Simulations of Thermal Transport in W-UO₂ CERMET Fuel for Nuclear Thermal Propulsion

Thermal hydraulic modeling of solid-fueled nuclear thermal propulsion reactors part II: Full-core coupled neutronic and thermal hydraulic analysis

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Design research for the porous hexagonal prism shaped fuel element adopted in nuclear thermal propulsion reactor based on neutronics and thermal-hydraulics coupled method

Cited by 12 publications

References 5 publications

Review and Preliminary Investigation into Fuel Loss from Cermets Composed of Uranium Nitride and a Molybdenum-Tungsten Alloy for Nuclear Thermal Propulsion Using Mesoscale Simulations

Review and Preliminary Investigation into Fuel Loss from Cermets Composed of Uranium Nitride and a Molybdenum-Tungsten Alloy for Nuclear Thermal Propulsion Using Mesoscale Simulations

Multiscale Simulations of Thermal Transport in W-UO2 CERMET Fuel for Nuclear Thermal Propulsion

Thermal hydraulic modeling of solid-fueled nuclear thermal propulsion reactors part II: Full-core coupled neutronic and thermal hydraulic analysis

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Multiscale Simulations of Thermal Transport in W-UO₂ CERMET Fuel for Nuclear Thermal Propulsion