Continuous SiC fiber, CVI SiC matrix composites for nuclear applications: Properties and irradiation effects

Katoh, Yutai; Ozawa, Kazumi; Shih, Chunghao; Nozawa, Takashi; Shinavski, Robert J.; Hasegawa, Akira; Snead, Lance Lewis

doi:10.1016/j.jnucmat.2013.06.040

Cited by 416 publications

(235 citation statements)

References 116 publications

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“…85 While the properties of bulk SiC and its radiation behavior were extensively studied, the behavior of this ceramic matrix composite (CMC) is an active area of research. 86 Our specific focus here is to understand thermal transport in this composite material. TR techniques offer the capability to measure thermal properties of the matrix and fiber separately.…”

Section: Sic Fiber Compositesmentioning

confidence: 99%

Investigation of thermal transport in composites and ion beam irradiated materials for nuclear energy applications

et al. 2016

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Thermal transport in materials used for energy applications is a physical process directly tied to performance and reliability. As a result, a great deal of effort has been devoted to understanding thermal transport in materials whose ability to conduct heat is critical. Here, our objective is to discuss the utility of laser-based thermoreflectance (TR) approaches that provide microscale measurement of thermal transport. We provide several examples that implement the TR technique to investigate thermal transport in materials used in nuclear energy applications. First, we discuss utility of this technique to measure thermal conductivity in ion irradiated ceramic materials during investigations where the primary objective is to understand the impact of radiation induced crystalline structure defects on thermal transport. We also present the capability of TR approach to resolve thermal conductivity of each layer in tristructural isotropic fuel, silicon carbide fiber composites, and 2nd phase precipitates in uranium silicide. Finally, the ability to measure interface thermal resistance between adjacent layers in composites is demonstrated.

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Section: Sic Fiber Compositesmentioning

confidence: 99%

Investigation of thermal transport in composites and ion beam irradiated materials for nuclear energy applications

et al. 2016

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“…The silicon carbide layer has the ability to confine fission products and it can also withstand very high temperature. Barrier of fission products release of RDE-10 MWth consists of the pressure boundary, building, suppression pool and filters [1][2][3][4][5][6][7][8] .…”

Section: Introductionmentioning

confidence: 99%

“…Release mechanism includes fuel, core, building ventilation, helium purification, gaseous radioactive waste system, and neutron activation of circulating air through the Reactor Cavity of Cooling System (RCCS) [2][3][4][5][6][7] . Mechanistic approach for a source term development is required to setup technical basis and to take appropriate benefit of radionuclide retention capabilities of each of the multiple barriers in radionuclides transport.…”

Section: Introductionmentioning

confidence: 99%

Estimation of Routine Discharge of Radionuclides on Power Reactor Experimental Rde

Udiyani

Kuntjoro

2017

urania

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ESTIMATION OF ROUTINE DISCHARGE OF RADIONUCLIDES ON POWER REACTOR EXPERIMENTAL RDE.Experimental power reactor (RDE), which is planned to be constructed by BATAN, is a kind of High Temperature Gas Cooled Reactor (HTGR) with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products to be remained in the core. To ensure the safety for workers and public surroundings of the RDE site and to meet the regulatory body requirements for construction and operation, research is needed for environmental radiation in normal and abnormal condition. This study is focused on normal operating condition, while for the accident condition study will be carried out after the final design has been completed. Estimation of radiology in the environment involves the source term released into the environment under routine operation condition. The purpose of this study is to estimate the source terms released into the environment based on postulation of normal or routine operation of the RDE-10 MWth. The research approach starts with an assumption that there are defects and impurities in the TRISO fuel because of limitations during the fabrication. Mechanisms of fission products release from the fuel to the environment were created based on the safety features design of RDE. Radionuclides inventories in the reactor were calculated by the use of ORIGEN-2 whose library has been modified for HTGR type, while assumptions on TRISO fuel defects and release fraction for each compartment of the RDE safety systems are defined by a reference parameter. The results show that the important source terms of RDE are group of noble gases (Kr and Xe), halogen (I), Sr, Cs, H-3, and Ag. Activities of RDE source terms for routine operations have no significant difference with the HTR-10 source terms with the same power.Keywords: Routine discharge, radionuclide, source terms, RDE, HTGR.

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“…SiC is a ceramic that provides a higher melting temperature, reduced oxidation and heat of oxidation, reduced hydrogen generation under off-normal conditions, low chemical activity, and a lower neutron absorption cross section (Griffith, 2011;George et al, 2015). SiC cladding is a relatively new concept; manufacturing technologies are still under development, a proven solution to hermetically seal the fuel rod end plugs is yet to be found (Yueh and Terrani, 2014), neutron irradiation induces a reduction in thermal conductivity (Katoh et al, 2014), and recent hydrothermal corrosion experiments showed significant mass loss of SiC in BWR-like conditions . Ferritic stainless steel alloys, such as FeCrAl, have the ability to enhance oxidation resistance by creating a protective oxide layer; for instance, the aluminum found in FeCrAl combines with oxygen to produce an oxide layer (Al 2 O 3 ) capable of reducing oxidation kinetics in high temperature steam environments .…”

Section: Introductionmentioning

confidence: 99%

Neutronic evaluation of coating and cladding materials for accident tolerant fuels

Younker

Fratoni

2016

Progress in Nuclear Energy

111

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a b s t r a c tIn severe accident conditions with loss of active cooling in the core, zirconium alloys, used as fuel cladding materials for current light water reactors (LWR), undergo a rapid oxidation by high temperature steam with consequent hydrogen generation. Novel fuel technologies, named accident tolerant fuels (ATF), seek to improve the endurance of severe accident conditions in LWRs by eliminating or at least mitigating such detrimental steam-cladding interaction. Most ATF concepts are expected to work within the design framework of current and future light water reactors, and for that reason they must match or exceed the performance of conventional fuel in normal conditions. This study analyzed the neutronic performance of ATF when employed in both pressurized and boiling water reactors. Two concepts were evaluated: (1) coating the exterior of zirconium-alloy cladding with thin ceramics to limit the zirconium available for reaction with high-temperature steam; (2) replacing zirconium alloys with alternative materials possessing slower oxidation kinetics and reduced hydrogen production. Findings show that ceramic coatings should remain 10e30 mm thick to limit the neutronic penalty. Alternative cladding materials, with the exception of SiC, enhance neutron loss compared to zirconium-alloys. An extensive parametric analysis concluded that reference performance metrics can be met by employing 300-mm or less thick cladding or increasing fuel enrichment by up to 1.74% depending on material and geometry.

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Continuous SiC fiber, CVI SiC matrix composites for nuclear applications: Properties and irradiation effects

Cited by 416 publications

References 116 publications

Investigation of thermal transport in composites and ion beam irradiated materials for nuclear energy applications

Investigation of thermal transport in composites and ion beam irradiated materials for nuclear energy applications

Estimation of Routine Discharge of Radionuclides on Power Reactor Experimental Rde

Neutronic evaluation of coating and cladding materials for accident tolerant fuels

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