Damage production and accumulation in SiC structures in inertial and magnetic fusion systems (original) (raw)
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Radiation Damage Parameters for SiC/SiC Composite Structure in Fusion Nuclear Environment
Fusion Science & Technology, 2003
The radiation effects in the fiber, matrix, and interface components of the SiC/SiC composite material are important input for lifetime assessment. Neutronics calculations were performed to determine the radiation damage parameters in the fiber/matrix and the candidate interface materials. The radiation damage parameters were calculated for both the carbon and silicon sublattices. The radiation damage parameters were evaluated for representative candidate breeding blankets. The breeder and/or coolant such as Pb 83 Li 17 , Flibe and Li 2 O affect the radiation damage parameters by impacting the neutron spectrum. The results provide an essential input for SiC/SiC composite lifetime assessment. The impact of the unique features of inertial fusion systems on damage parameters are identified.
Current status and critical issues for development of SiC composites for fusion applications
Silicon carbide (SiC)-based ceramic composites have been studied for fusion applications for more than a decade. The potential for these materials have been widely discussed and is now understood to be (1) the ability to operate in temperature regimes much higher than for metallic alloys, (2) an inherent low level of long-lived radioisotopes that reduces the radiological burden of the structure, and (3) perceived tolerance against neutron irradiation up to high temperatures. This paper reviews the recent progress in development, characterization, and irradiation effect studies for SiC composites for fusion energy applications. It also makes the case that SiC composites are progressing from the stage of potential viability and proof-of-principle to one where they are ready for system demonstration, i.e., for flow channel inserts in Pb–Li blankets. Finally, remaining general and specific technical issues for SiC composite development for fusion applications are identified.
Status and prospects for SiC SiC composite materials development for fusion applications
Silicon carbide (SIC) composites are very attractive for fusion applications because of their low afterheat and low activation characteristics coupled with excellent high temperature properties. These composites are relatively new materials that will require material development as well as evaluation of hermiticity, thermal conductivity, radiation stability, high temperature strength, fatigue, thermal shock, and joining techniques. The radiation stability of SIC-SIC composites is a critical aspect of their application as fusion components and recent results will be reported. Many of the non-fusion specific issues are under evaluation by other ceramic composite development programs, such as the US national continuous fiber ceramic composites.
Fusion Engineering and Design, 2005
A high fluence irradiation of SiC f /SiC composites has been performed to study the effect of neutron irradiation on the mechanical and physical behaviour of those composites. The fibre reinforced silicon carbide composites have been irradiated at two temperature levels of 600 • C and 900 • C up to a fluence of 3.5 × 10 25 n/m 2 (E > 0.1 MeV). The stiffness of the bending bars after irradiation changed with factors between 0.75 and 1.04 for the different composites while the bending strength after irradiation was reduced with a factor of 2 in some cases. The laser flash thermal diffusivity ratio's (α irr /α o) of the composites measured at 600 • C are from 0.1 to 0.5 and from 0.25 to 0.75 for an irradiation temperature of 600 • C and 900 • C, respectively. The dimensional changes observed are small.
Neutron Irradiation Swelling of SiC and SiCf/SiC for Advanced Nuclear Applications
Energy Procedia, 2015
Silicon carbide (SiC) is used as a layer in TRISO fuel of high temperature gas-cooled reactors because of its excellent thermal and mechanical properties. SiC fiber-reinforced SiC composites (SiC f /SiC) are also a candidate material for structural material for fusion reactor blanket and cladding materials of advance fission reactors. In this research, seven of monolithic SiC or SiC f /SiC composites materials with different fabrication processes were irradiated in the BR2 reactor up to a fluence of 2.0-2.5×10 24 (E>0.1 MeV) at 333-363 K. Changes in macroscopic lengths and lattice parameters before and after the neutron irradiation were measured. Furthermore, microstructure of SiC f /SiC composites was investigated using a scanning electron microscope, too. Results showed that after the neutron irradiation, the group of SiC and SiC f /SiC composites were swelled approximately 1.24~1.33% and 1.00~1.19% in length, respectively. Apparently the presence of fibers resulted in smaller swelling. It may be attributed for smaller swelling of SiC fibers. Further difference in swelling may be caused by the presence of different sintering additives.
A Monte Carlo Simulation of Radiation Damage of SiC and Nb Using JA-IPU Code
Journal of Energy and Power Engineering, 2015
MC (Monte Carlo) simulation code, JA-IPU is used to study radiation damage of SiC irradiated to spallation neutron and AmBe neutron spectra. The code is based on the major physical processes of radiation damage on incorporation of atomic collision cascade and limited to 10 MeV neutron energy. A phenomenological relation for radiation swelling is also derived. Based on the calculation of swelling, DPA (displacement per atom), defect production efficiency and effective threshold energy, E d eff from the data of MC simulation, SiC is inferred to be a highly radiation resistant material when compared with Nb and Ni metals which are used in composition of several reactor steels. Experimental results of hill-hock density measured using AFM (atomic force microscopy), also confirm radiation resistant behavior of SiC.
Mechanical properties of neutron irradiated SiC fibers
Journal of Nuclear Materials, 1986
SiC fibers were irradiated at neutron fluences of 8 x10 23 and 2 x 10 24 n/m 2 in JMTR (E > 1 MeV). The tensile strength of SiC fibers measured at gauge length of 5 mm essentially unchanged after the neutron irradiation, but the strength at a length of 25 mm and 50 mm increased by 40% and 25~ after the irradiation, respectively. The tensile strength variations of the irradiated SiC fibers with respect to gauge length were examined in terms of Weibull statistics. The Weibull modulus of the tensile strength in the SiC fibers increased slightly by the neutron irradiation. SiC fibers and SiC fiber reinforced aluminum preform wires (SiC/A1) were irradiated in JOYO (E > 0.1 MeV) at neutron fluences of 1 × 10 24, and 1 X10 24 and 1 X 1025 n/m 2, respectively. The mechanical properties of the irradiated SiC fibers and the SiC fibers extracted from the irradiated SiC/Al preforrff wires were essentially unchanged in comparison with the unirradiated SiC fibers.
Current status and prospects of SiCf/SiC for fusion structural applications
Journal of the European Ceramic Society, 2013
This article provides an overview of the main characteristics of SiC f /SiC that suggest the use of this SiC-based composite as a structural material for the blanket in future fusion reactors, a brief description of its structure and the role of its main constituents. The relevant fabrication processes and their ability to produce a material with the required properties are also summarised. The main part of the paper is devoted to an assessment of the state-of-the-art materials, and the basic requirements for the target material are discussed in terms of the achieved properties. The key issues and areas of uncertainty are described and suggestions for overcoming them are presented.