• Nuclear Engineering and Technology
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• v.38 no.7
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• pp.591-618
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• 2006

An integrated R&D program is being conducted to study, qualify, and in some cases, develop materials with required properties for the reactor systems being developed as part the U.S. Department of Energy's Generation IV Reactor Program. The goal of the program is to ensure that the materials research and development (R&D) needed to support Gen IV applications will comprise a comprehensive and integrated effort to identify and provide the materials data and its interpretation needed for the design and construction of the selected advanced reactor concepts. The major materials issues for the five primary systems that have been considered within the U.S. Gen IV Reactor Program-very high temperature gas-cooled, supercritical water-cooled, gas-cooled fast spectrum, lead-cooled fast spectrum, and sodium-cooled fast spectrum reactors-are described along with the R&D that has been identified to address them.

• Nuclear Engineering and Technology
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• v.39 no.1
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• pp.21-30
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• 2007

Current high-temperature gas-cooled reactors (HTGRs) are based on a closed Brayton cycle with helium gas as the working fluid. Thermodynamic performance of the axial-flow helium gas turbines is of critical concern as it considerably affects the overall cycle efficiency. Helium gas turbines pose some design challenges compared to steam or air turbomachinery because of the physical properties of helium and the uniqueness of the operating conditions at high pressure with low pressure ratio. This report present a review of the helium Brayton cycle experiences in Germany and in Japan. The design and availability of helium gas turbines for HTGR are also presented in this study. We have developed a new throughflow calculation code to calculate the design-point performance of helium gas turbines. Use of the method has been illustrated by applying it to the GTHTR300 reference.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.115.1-115.1
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• 2003

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3685-3693
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• 2021

This paper presents a disturbance observer-based robust backstepping load-following control (DO-RBLFC) scheme for modular high-temperature gas-cooled reactors (MHTGRs) in the presence of actuator saturation and disturbances. Based on reactor kinetics and temperature reactivity feedback, the mathematical model of the MHTGR is first established. After that, a DO is constructed to estimate the unknown compound disturbances including model uncertainties, external disturbances, and unmeasured states. Besides, the actuator saturation is compensated by employing an auxiliary function in this paper. With the help of the DO, a robust load-following controller is developed via the backstepping technique to improve the load-following performance of the MHTGR subject to disturbances. At last, simulation and comparison results verify that the proposed DO-RBLFC scheme offers higher load-following accuracy, better disturbances rejection capability, and lower control rod speed than a PID controller, a conventional backstepping controller, and a disturbance observer-based adaptive sliding mode controller.

• Nuclear Engineering and Technology
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• v.56 no.11
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• pp.4851-4858
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• 2024

Matrix graphite is used as a structural material, thermal conductor, moderator, and secondary fission product barrier for fuel elements in high-temperature gas-cooled reactors (HTRs). Due to its high graphitization degree and compressibility, natural flake graphite (NFG) is used as the main filler in traditional A3-3 matrix graphite, whereas artificial graphite (AG), with a lower graphitization degree than NFG, serves as an additive for toughness and gas permeability. Matrix graphite could be improved in terms of thermal conductivity, oxidation resistance, and irradiation performance by increasing the degree of graphitization. However, reports on the development of new matrix graphite formulations are scarce. In this study, MG-20 matrix graphite was prepared by mixing 60 wt % NFG, 20 wt% natural microcrystalline graphite (MG), and 20 wt% phenolic resin. Due to the high graphitization degree (higher than AG) and low coefficient of thermal expansion (CTE) of MG, MG-20 exhibited higher thermal conductivity (~6%) and lower CTE (~2.4%) than A3-3. Thus, MG-20 with higher graphitization degree and better thermal properties than A3-3 could improve the performance of HTR fuel elements in the future.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.39-40
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• 2005

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2989-2998
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• 2022

Line contact structures, such as the contact between graphite brick and graphite tenon, widely exist in high-temperature gas-cooled reactors. Due to the stress concentration effect, the line contact area is one of the dangerous positions prone to failure in the nuclear reactor core. In this paper, the failure mechanism of line contact structures composed of IG11 nuclear graphite column and brick were investigated by means of experiment and finite element simulation. It was found that the failure process mainly includes three stages: firstly, the damage accumulation in nuclear graphite material led to the characteristic yielding of the line contact structure, but no macroscopic failure can be observed at this stage; secondly, the stresses near the contact area met Mohr failure criterion, and a crack initiated and propagated laterally in the contact zone, that is, local macroscopic failure occurred at this stage; finally, a second crack initiated in the contact area and developed in to a Y-shape, resulting in the final failure of the structure. This study lays a foundation for the structural design and safety assessment of high-temperature gas-cooled reactors.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1679-1686
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• 2024

Graphite material plays an important role in nuclear reactors especially the high-temperature gas-cooled reactors (HTGRs) by its outstanding comprehensive nuclear properties. The structural integrity of graphite pebble fuel elements is the first barrier to core safety under any circumstances. The correct knowledge of the stiffness coefficient of the graphite pebble fuel element inside the reactor's core is significant to ensure the valid design and inherent safety. In this research, a vertical extrusion device was set up to measure the stiffness coefficient of the graphite pebble fuel element by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. The stiffness coefficient equations of graphite pebble fuel elements at different temperatures are given (in a helium atmosphere). The result first provides the data on the high-temperature stiffness coefficient of pebbles in helium gas. The result will be helpful for the engineering safety analysis of pebble-bed nuclear reactors.

• Korean Journal of Materials Research
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• v.16 no.5
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• pp.318-322
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• 2006

Short-term high temperature degradation test was conducted on Hastelloy X, a candidate tube material for high temperature gas-cooled reactors (HTGR), to evaluate the variation of microstructure and mechanical property in air at $1050^{\circ}C$ during 2000 h. The dominant oxide layer was Cr-oxide and a very shallow Cr-depleted region was observed below the oxide layer. At the beginning of degradation, the island shape $M_6C$ precipitate (M=Mo-rich, Fe, Ni, Cr) was observed in matrix region. After 2000 h degradation, precipitate shape was changed to the chain shape and increased amount of precipitate. These results influenced mechanical property of the specimen which exposed in high temperature. Yield strength was decreased from 115MPa to 89 MPa after 24 h and 2000 h exposure, respectively.

• Journal of the Korean Ceramic Society
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• v.44 no.3 s.298
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• pp.169-174
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• 2007

TRISO coated fuel particle is one of the most important materials for hydrogen production using HTGR (high temperature gas cooled reactors). It is composed of three isotropic layers: inner pyrolytic carbon (IPyC), silicon carbide (SiC), outer pyrolytic carbon (OPyC) layers. In this study, TRISO coated fuel particle layers were deposited through CVD process in a horizontal hot wall deposition system. Also the computational simulations of input gas velocity, temperature profile and pressure in the reaction chamber were conducted with varying process variable (i.e temperature and input gas ratios). As deposition temperature increased, microstructure, chemical composition and growth behavior changed and deposition rate increased. The simulation showed that the change of reactant states affected growth rate at each position of the susceptor. The experimental results showed a close correlation with the simulation results.