Thermal Aspects of Using Uranium Nitride in SuperCritical Water-Cooled Nuclear Reactors

Grande, Lisa; Peiman, Wargha; Mikhael, Sally; Villamere, Bryan; Rodriguez-Prado, Adrianexy; Allison, Leyland; Pioro, Igor

doi:10.1115/icone18-29790

Cited by 3 publications

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“…Using UO 2 fuel has many advantages, however it is limited due to its low thermal conductivity [9]. Therefore, alternative fuels such as uranium nitride and carbide were considered as applicable options.…”

Section: Introductionmentioning

confidence: 99%

Neutronic analysis of alternative ceramic fuels in the next generation of VVER-1000/V446 reactor

Rahimi,

Heidari,

Kheradmand Saadi

et al. 2023

Phys. Scr.

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One of the most important issues in nuclear fuel management is burnup increase, which impacts significantly on the reactor cycle length. Burnup enlargement specifically leads to reducing the cost of electricity generation and increasing efficiency. In the present study, the neutronic analysis of some alternative ceramic fuels includingUN, 15UN, and UChas been performedthe VVER-1000/V446 reactor, comparing with the conventional UO2. Being single-ligand, the alternative fuels’ density of heavy metals such as uranium is greater than UO2’s one. Therefore, the absorption and fissionamounts in thesefuels will be higher than theones in UO2 fuel. By modellingVVER-1000/V446 reactor with MCNPX2.6 Monte Carlo code, we showed that the excess reactivity value for UC fuel (169.173 ∆K/K) increased compared to UO2 fuel (165.526). Due to the ultra-high burnup, the reactor cycle length is 398 days for 15UN fuel, while it is 270, 308, and 130 days for UO2, UC, and UN fuels, respectively, and has increased by 47%. Therefore, the use of alternative fuels has advantages like plant construction cost reduction.excess reactivity, discharge burn up, making them suitable substitution of conventional UO2 for the next generation of LWRs.

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Section: Introductionmentioning

confidence: 99%

Neutronic analysis of alternative ceramic fuels in the next generation of VVER-1000/V446 reactor

Rahimi,

Heidari,

Kheradmand Saadi

et al. 2023

Phys. Scr.

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Thermal-Hydraulic and Neutronic Analysis of a Reentrant Fuel-Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors

Peiman

Pioro

Gabriel

2015

Journal of Nuclear Engineering and Radiation Science

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To address the need to develop new nuclear reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of nuclear reactors called Generation IV. The Generation IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts, one of which is supercritical water-cooled reactor (SCWR). Among the Generation IV nuclear-reactor concepts, only SCWRs use water as a coolant. The SCWR concept is considered to be an evolution of water-cooled reactors (pressurized water reactors (PWRs), boiling water reactors (BWRs), pressurized heavy water reactors (PHWRs), and light-water, graphite-moderated reactors (LGRs)), which comprise 96% of the current fleet of operating nuclear power reactors and are categorized under Generation II, III, and III+ nuclear reactors. The latter water-cooled reactors have thermal efficiencies of 30–36%, whereas the evolutionary SCWR will have a thermal efficiency of approximately 45–50%. In terms of a pressure boundary, SCWRs are classified into two categories, namely, pressure-vessel (PV) SCWRs and pressure-channel (PCh) SCWRs. A generic pressure-channel SCWR, which is the focus of this paper, operates at a pressure of 25 MPa with inlet and outlet coolant temperatures of 350°C and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the reactor core should be designed, which depends on a fuel-channel design. Second, a nuclear fuel and fuel cycle should be selected. Several fuel-channel designs have been proposed for SCWRs. These fuel-channel designs can be classified into two categories: direct-flow and reentrant channel concepts. The objective of this paper is to study thermal-hydraulic and neutronic aspects of a reentrant fuel-channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB, which calculates fuel-centerline-temperature, sheath-temperature, coolant-temperature, and heat-transfer-coefficient profiles. A lattice code and diffusion code were used to determine a power distribution inside the core. Then, heat flux in a channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents a fuel centerline temperature of a newly designed fuel bundle with UO2 as a reference fuel. The results show that the maximum fuel centerline temperature exceeds the design temperature limits of 1850°C for fuel.

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Thermal-Hydraulic and Neutronic Analysis of a Re-Entrant Fuel Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors

Peiman

Pioro

Gabriel

2014

Volume 5: Innovative Nuclear Power Plant Design and New Technology Application; Student Paper Competition

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To address the need to develop new nuclear reactors with higher thermal efficiency, a group of countries, including Canada, have initiated an international collaboration to develop the next generation of nuclear reactors called Generation IV. The Generation IV International Forum (GIF) Program has narrowed design options of the nuclear reactors to six concepts one of which is the SuperCritical Water-cooled Reactor (SCWR). Among the Generation IV nuclear-reactor concepts, only SCWRs use water as the coolant. The SCWR concept is considered to be an evolution of Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs), which comprise 81% of the current fleet of operating nuclear reactors and are categorized under Generation II nuclear reactors. The latter water-cooled reactors have thermal efficiencies in the range of 30–35% while the evolutionary SCWR will have a thermal efficiency of about 40–45%. In terms of a pressure boundary SCWRs are classified into two categories, namely, Pressure Vessel (PV) SCWRs and Pressure Channel (PCh) SCWRs. A generic pressure channel SCWR, which is the focus of this paper, operates at a pressure of 25 MPa with inlet and outlet coolant temperatures of 350 and 625°C, respectively. The high outlet temperature and pressure of the coolant make it possible to improve the thermal efficiency. On the other hand, high operating temperature and pressure of the coolant introduce a challenge for material selection and core design. In this view, there are two major issues that need to be addressed for further development of SCWR. First, the reactor core should be designed, which depends on a fuel channel design (for PCh SCWR). Second, a nuclear fuel and fuel cycle should be selected. Third, materials for core components and other key components should be selected based on material testing and experimental results. Several fuel-channel designs have been proposed for SCWRs. These fuel-channel designs can be classified into two categories: direct-flow and re-entrant channel concepts. The objective of this paper is to study thermal-hydraulic and Neutronic aspects of a re-entrant fuel channel design. With this objective, a thermal-hydraulic code has been developed in MATLAB which calculates the fuel centerline temperature, sheath temperature, coolant temperature and heat transfer coefficient profiles. A lattice code and a diffusion code were used in order to determine the power distribution inside the core. Then, the heat flux in a channel with the maximum thermal power was used as an input into the thermal-hydraulic code. This paper presents the fuel centerline temperature of a newly designed fuel bundle with UO2 as a reference fuel. The results show that the maximum fuel centerline temperature and the sheath temperature exceed the temperature limits of 1850°C and 850°C for fuel and sheath, respectively.

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Thermal Aspects of Using Uranium Nitride in SuperCritical Water-Cooled Nuclear Reactors

Cited by 3 publications

References 0 publications

Neutronic analysis of alternative ceramic fuels in the next generation of VVER-1000/V446 reactor

Neutronic analysis of alternative ceramic fuels in the next generation of VVER-1000/V446 reactor

Thermal-Hydraulic and Neutronic Analysis of a Reentrant Fuel-Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors

Thermal-Hydraulic and Neutronic Analysis of a Re-Entrant Fuel Channel Design for Pressure-Channel Supercritical Water-Cooled Reactors

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