Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for a nuclear hydrogen generation. The VHTR can produce hydrogen from heat and water by using a thermo-chemical process or from heat, water, and natural gas by steam reformer technology. Korea Atomic Energy Research Institute (KAERI) is in the process of carrying out a nuclear hydrogen system by considering the indirect cycle gas cooled reactors that produce heat at temperatures in the order of 950°. The nuclear hydrogen system is planning to produce hydrogen by using nuclear energy and a thermo-chemical process. Helium gas is the choice for the coolant of the nuclear hydrogen system since it is an inert gas, with no affinity to a chemical or nuclear activity; therefore a radioactivity transport in the primary circuit of the nuclear hydrogen system is minimal under a normal operation. Moreover, its gaseous nature avoids problems related to a phase change and water-metal reactions and therefore improves its safety. A coaxial double-tube hot gas duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger (IHX) for the nuclear hydrogen system. In this study, a preliminary design analysis for the primary and secondary HGDs of the nuclear hydrogen system was carried out. These preliminary design activities include a preliminary decision on the geometric dimensions, a preliminary strength evaluation and an appropriate material selection. A preliminary decision on the geometric dimensions of the HGDs was undertaken based on three engineering concepts, such as a constant flow velocity model (CFV model), a constant flow rate model (CFR model), a constant hydraulic head model (CHH model), and also based on a heat balanced model (HB model). We compared the geometric dimensions and their preliminary strength evaluation results from the various models.
Preliminary Design Analysis of Hot Gas Ducts for a Nuclear Hydrogen System of 200 MWt
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Song, K, & Kim, Y. "Preliminary Design Analysis of Hot Gas Ducts for a Nuclear Hydrogen System of 200 MWt." Proceedings of the Fourth International Topical Meeting on High Temperature Reactor Technology. Fourth International Topical Meeting on High Temperature Reactor Technology, Volume 2. Washington, DC, USA. September 28–October 1, 2008. pp. 699-703. ASME. https://doi.org/10.1115/HTR2008-58056
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