Materials performance is central to the satisfactory operation of current and future nuclear energy systems. For example, the remarkable improvement in the operation and reliability of Generation-II light water reactors (LWRs) over the past 25 years has been largely associated with improvements in steam generator materials, fuel cladding technology (composition and fabrication), and improved understanding of water chemistry impacts on corrosion and deposition. Future fission and proposed fusion energy systems will be increasingly dependent on advanced structural materials to reliably deliver high performance with favorable safety attributes and acceptable economic cost. In many cases, the proposed operating temperatures are significantly higher than the experience base for light water reactors. This motivates development of structural materials with improved high temperature strength for prolonged operating periods and engineered corrosion resistance for the candidate coolants and other materials in the system. The high radiation fluxes in future nuclear energy systems will require the structural materials to have superior radiation resistance compared to currently available materials. The material performance demands are particularly challenging for the fuel cladding and wrapper of next-generation sodium-cooled fast reactors, where simultaneous resistance to high temperature thermal creep and radiation-induced property degradation up to doses of {approx}100-150 dpa is required. Several strategies can be utilized to develop structural materials with simultaneous high radiation resistance, high strength, good toughness and corrosion resistance, and moderate fabrication cost. There are three general approaches for designing radiation resistance: Nano-scale precipitates or interfaces to produce high point defect sink strength (e.g., oxide dispersion strengthened (ODS) and next-generation ferritic/martensitic steels with high particle densities); purposeful utilization of immobile vacancies (e.g., SiC/SiC ceramic composites); and utilization of radiation-resilient matrix phases (e.g., ferritic instead of austenitic steel matrix, etc.). High-performance steels designed using computational thermodynamics are demonstrating promising capability to produce a high density of highly stable nano-scale precipitates that could serve as efficient point defect recombination centers during irradiation, and also provide good thermal creep strength at high temperatures [1]. Figure 1 shows an example of improvements in high temperature thermal creep properties for a 9%Cr-1%Mo ferritic martensitic steel that was achieved simply by slightly altering the thermomechanical processing procedure. Figure 2 compares the fracture toughness behavior of an advanced ODS ferritic steel before and after low dose neutron irradiation at 300 C [2]. The ductile to brittle transition temperature (DBTT) in the LT orientation remained below -150 C with a shift in the DBTT of about 12 C; the corresponding shift in the DBTT for EUROFER97 (9Cr-2WVTa) ferritic/martensitic steel was about 39 C. Higher dose studies are in progress. It will be important for nuclear energy researchers to continue to closely interact with the broader materials science and engineering community in order to effectively leverage innovations that continue to occur in the broad field of materials science. For example, practical aspects used in the aerospace industry to reduce the time from invention of a new alloy system to code qualification and commercialization could be useful for development of new structural materials for nuclear energy systems. In the future, utilization of emerging advanced manufacturing processes such as additive manufacturing to produce near-net shape parts with precise microstructural control will be of increasing importance to control fabrication costs and to create high-performance fabrication architectures that could not be achieved using conventional fabrication methods. Following the accident at the Fukushima Daiichi site in Japan, there is increasing interest in exploring accident tolerant or enhanced safety margin fuel systems for existing and future reactors, which could potentially provide increased response time or reduced consequences via reduced enthalpy production, reduced hydrogen production, and delayed clad rupture or fission product release during a loss of coolant accident compared to conventional Zr alloy cladding/ monolithic UO{sub 2} fuel systems. Although all alternative fuel systems have technological or neutronic shortcomings, exploratory research would be useful to quantify their potential improved accident tolerance so that an informed decision on the best option(s) for future LWR fuel systems can be reached. If research results on a particular accident tolerant concept prove to be promising, it might be possible to initiate confirmatory tests in commercial reactors within about 10 years. (authors)
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Aug 7, 2013
E.G Moghaddam + 3
Open Access