Updated Generation IV Reactors Integrated Materials Technology Program Plan, Revision 2

The Department of Energy's (DOE's) Generation IV Nuclear Energy Systems Program will address the research and development (R&D) necessary to support next-generation nuclear energy systems. Such R&D will be guided by the technology roadmap developed for the Generation IV International Forum (GIF) over two years with the participation of over 100 experts from the GIF countries. The roadmap evaluated over 100 future systems proposed by researchers around the world. The scope of the R&D described in the roadmap covers the six most promising Generation IV systems. The effort ended in December 2002 with the issue of the final Generation IV Technology Roadmap [1.1]. The six most promising systems identified for next generation nuclear energy are described within the roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor - SCWR and the Very High Temperature Reactor - VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor - GFR, the Lead-cooled Fast Reactor - LFR, and the Sodium-cooled Fast Reactor - SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides, and may provide an alternative to accelerator-driven systems. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. Accordingly, DOE has identified materials as one of the focus areas for Gen IV technology development

Analysis of grain boundary sinks and interstitial diffusion in neutron-irradiated SiC

The widths of the interstitial loop denuded zone (DZ) along grain boundaries were examined for 3C-SiC irradiated at 1010–1380 °C by transmission electron microscopy (TEM) in an effort to obtain the activation energy of interstitial migration. Denuded-zone widths as small as 17 nm were observed below 1130 °C, indicating that a substantial population of “TEM invisible” voids of diameter <0.7 significantly contribute to interstitial annihilation. By using the obtained loop DZ width and the matrix sink strength (including the invisible voids), the activation energy of interstitial diffusion was determined to be 1.5 eV for the slower moving Si interstitial of SiC by application of simple reaction-diffusion equations

Design and strategy for next-generation silicon carbide composites for nuclear energy

Silicon carbide (SiC) ceramic-based composites continue to be attractive material options for fusion in-vessel components and fission reactor core structures because of their exceptional high-temperature capability and favorable neutronic properties. As performance data accumulates, the limitations of the current generation of nuclear-grade SiC composites are becoming more apparent. These limitations mainly involve strength degradation during high-dose neutron irradiation. This paper discusses several options for improving the performance of the next generation of SiC composites to enhance the radiation resistance, along with new experimental results on neutron irradiation resistance. The main emphasis is on the selection of the fibers and the design and development of alternative interphase layers for advanced composites

Mechanical properties of neutron irradiated F82H using micro-tensile testing

Room temperature micro-tensile tests were successfully performed on F82H specimens neutron-irradiated at573 K up to 5 dpa and unirradiated using a focused ion beam (FIB) device. Dimensions of the gauge section formicro-tensile specimens were about 10 μm (length) × 1 μm2(area). These specimens were fabricated in a FIBdevice. The tensile properties obtained on specimens of micrometer size qualitatively agreed with results frommillimeter size specimens. Using this micro-tensile testing technique is considered to become very useful data incombining the micro-tensile data on ion irradiated F82H with data on neutron irradiated material. This methodovercomes the limitation on obtaining mechanical property on ion irradiated specimens to only micro-in-dentation testing

High-dose, intermediate-temperature neutron irradiation effects on silicon carbide composites with varied fiber/matrix interfaces

SiC/SiC composites are promising structural candidate materials for various nuclear applications over the wide temperature range of 300–1000 °C. Accordingly, irradiation tolerance over this wide temperature range needs to be understood to ensure the performance of these composites. In this study, neutron irradiation effects on dimensional stability and mechanical properties to high doses (11–44 dpa) at intermediate irradiation temperatures (˜600 °C) were evaluated for Hi-Nicalon Type-S or Tyranno-SA3 fiber–reinforced SiC matrix composites produced by chemical vapor infiltration. The influence of various fiber/matrix interfaces, such as a 50–120 nm thick pyrolytic carbon (PyC) monolayer interphase and 70–130 nm thick PyC with a subsequent PyC (˜20 nm)/SiC (˜100 nm) multilayer, was evaluated and compared with the previous results for a thin-layer PyC (˜20 nm)/SiC (˜100 nm) multilayer interphase. Four-point flexural tests were conducted to evaluate post-irradiation strength, and SEM and TEM were used to investigate microstructure. Regardless of the fiber type, monolayer composites showed considerable reduction of flexural properties after irradiation to 11–12 dpa at 450–500 °C; and neither type showed the deterioration identified at the same dose level at higher temperatures (> 750 °C) in a previous study. After further irradiation to 44 dpa at 590–640 °C, the degradation was enhanced compared with conventional multilayer composites with a PyC thickness of ˜20 nm. Multilayer composites have shown comparatively good strength retention for irradiation to ˜40 dpa, with moderate mechanical property degradation beginning at 70–100 dpa. Irradiation-induced debonding at the F/M interface was found to be the major cause of deterioration of various composites