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High-Temperature Reactors
Published in William J. Nuttall, Nuclear Renaissance, 2022
HTGR systems make use of fuel/moderator composites usually in the form of Triso fuel. Lyman focused on the risk of a graphite fire, but in so doing he may have over-stated the problem. Nuclear graphite cannot ignite or maintain self-sustaining combustion, even at very high temperatures [113]. In pebble bed reactors, the Triso fuel is assembled in spherical fuel elements of a few centimetres diameter coated in pyrolytic graphite. Perhaps the concern of the critics should be regarded as the possibility of graphite oxidation, but if that is the case the risks and consequences are not so severe. Strictly graphite does not burn, as there is no flame, but one might say that: at its worst graphite does not burn, it smoulders. The PBMR is designed to operate at approximately 900°C with the graphite-clad fuel elements in an inert helium atmosphere. Despite the lack of conventional fire risk, it would be a serious problem if air or water were to encounter a bare graphite surface at such temperatures. Nevertheless, it is important to note that oxidation of the outer graphite cladding of the fuel pebbles probably would not in itself lead to a breakdown of the SiC-clad Triso fuel or the release of radioactive materials. The designers are confident that the ingress of oxidisers (air or water) would always be inhibited by the nature of their design and that the barriers to release could cope with the gaseous and sooty products of oxidation. Experience at the Jülich AVR points to concerns about hot spots in the reactor core and the physical integrity and quality of the fuel pebbles as constructed and handled at that time. Defects in Triso fuel construction, especially poor integrity of the SiC shell, might lead to the possibility of the release of radioactive fission products and minor actinides in the event of a serious accident. Defects would also relate to the cleanliness of the internal reactor components. The structural integrity of pebble bed Triso fuel is of particular importance, given the high temperature anticipated in some HTGR accident scenarios.
Investigation on the oxidation behavior and multi-step reaction mechanism of nuclear graphite SNG742
Published in Journal of Nuclear Science and Technology, 2020
Wei Lu, Xiaowei Li, Xinxin Wu, Libin Sun, Zhengcao Li
Nuclear graphite has been used as a structural material and moderator in high-temperature gas-cooled reactors (HTGR) due to its excellent neutron moderation performance, radiation resistance, and stable mechanical properties at high temperatures. Taking the commercial demonstration HTGR power plant HTR-PM in China, for example, the amount of graphite in fuel elements is about 80 tons and the amount of structure graphite is about 1200 tons [1]. HTR-PM uses helium as coolant, and the average helium temperature at the outlet of the reactor core is 750°C. During normal operation, a trace amount of oxygen, water vapor, and other impurities will exist in the primary circuit, causing oxidative corrosion of structural graphite and graphite in the fuel elements [2]. In severe loss of coolant accidents (LOCAs), some amount of air may enter the core via natural circulation, causing oxidation (or corrosion) of the structural graphite and the base graphite of the fuel spheres. Corrosion may increase the porosity of the graphite, which will decrease its strength [3–5]. Therefore, the study of oxidation performance of graphite and development of new nuclear graphite materials with excellent anti-oxidation performance are important for increasing the safety of HTGR.
Adsorption and Desorption of Tritium in Nuclear Graphite at 700°C: A Gas Chromatographic Study Using Hydrogen
Published in Nuclear Technology, 2019
Ke Deng, Mingjun Zhang, Xijun Wu, Qin Zhang, Guo Yang, Zhaowei Ma, Fei Wei, Guanghua Wang, Wei Liu
Nuclear graphite has been widely used for the moderator, reflector, and other structural materials in different types of gas-cooled reactors. As a result, a significant amount of irradiated nuclear graphite is now waiting to be treated before final disposal. According to data collected by the International Atomic Energy Agency, more than 230 000 tons of irradiated nuclear graphite has been temporarily retained.1 In addition to the accumulation of nuclear graphite resulting from the operation of existing or decommissioned reactors, graphite will also be used in Generation IV nuclear reactors, high-temperature gas-cooled reactors (HTGRs), and molten salt reactors (MSRs) (Refs. 2 and 3). Therefore, a great deal of attention should be paid to disposing of decommissioned nuclear graphite.
The evolution of He+ irradiation-induced point defects and helium retention in nuclear graphite
Published in Journal of Nuclear Science and Technology, 2019
Mingyang Li, Chuanqing Shi, Henk Schut, Zhengjun Zhang, Zhengcao Li
Nuclear graphite refers to the graphite intended for application in nuclear reactors. Graphite has been used as the neutron moderator and structural material in several kinds of graphite-moderated nuclear fission reactors, including Magnox, Advanced Gas-cooled Reactor (AGR) and High-Temperature Gas-cooled Reactors (HTGR). The advantages of nuclear graphite include low neutron absorption cross section, high chemical purity, good thermal and mechanical properties at very high temperatures (>1000°C), good compatible with the other materials used in the reactor core, high chemical stability, and inexpensiveness, which enable nuclear graphite to be an important component in nuclear power plants [1–3]. However, Very High-Temperature Reactor (VHTR), which is a Generation IV reactor concept that comprises HTGR but with a potential core outlet temperature of at least 1273 K [4], has set higher demands to nuclear graphite [5–9]. In the moderation process, neutrons collide with carbon nuclei and thereby transfer kinetic energy to carbon matrix, and then the carbon atoms are knocked and will be displaced from their equilibrium positions and thus distort the surrounding lattices [10,11]. As a result of long-term neutron irradiation, the structure and properties of graphite would be seriously harmed, and the safety of nuclear power plants would be threatened by the failure of graphite components. Therefore, it is of great importance to study the response of graphite to radiation damage and to understand the defect behavior in thermal environment.