• Journal of the Korean Society of Mechanical Technology
• /
• v.13 no.2
• /
• pp.145-153
• /
• 2011

The irradiation tests of materials in HANARO have been performed usually at temperatures below $300^{\circ}C$ at which the RPV(Reactor Pressure Vessel) materials of the commercial reactors such as the light water reactor and CANDU are operated. As VHTR(Very High Temperature Reactor) and SFR(Sodium-cooled Fast Reactor) projects are being carried as a part of the present Gen-IV program in Korea, the requirements for irradiation of materials at temperatures higher than $500^{\circ}C$ are recently being gradually increased. To overcome the restriction in the use at high temperature of the existing Al thermal media, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated more advanced than the single thermal media capsule. At the irradiation test of the capsule, the temperature of the specimens successfully reached $700^{\circ}C$ and the integrity of Al, Ti and graphite material was maintained.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.21 no.4
• /
• pp.571-576
• /
• 2023

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.28 no.5
• /
• pp.436-442
• /
• 2008

Graphite material has been recognized as a very competitive candidate for reflector, moderator, and structural material for very high temperature reactor (VHTR). Since VHTR is operated up to $900-950^{\circ}C$, small amount of impurity may accelerate the oxidation and degradation of carbon graphite, which results in increased porosity and lowered fracture toughness. In this study, ultrasonic wave propagation properties were investigated for both as-received and degradated material, and the feasibility of ultrasonic testing (UT) was estimated based on the result of ultrasonic property measurements. The ultrasonic properties of carbon graphite were half, more than 5 times, and 1/3 for velocity, attenuation, and signal-to-noise (S/N) ratio respectively. Degradation reduces the ultrasonic velocity slightly by 100 m/s, however the attenuation is about 2 times of as-receive state. The results of probability of detection (POD) estimation based on S/N ratio for side-drilled-hole (SDHs) of which depths were less than 100 mm were merely affected by oxidation and degradation. This result suggests that UT would be reliable method for nondestructive testing of carbon graphite material of which thickness is not over 100 mm. In accordance with the result produced by commercial automated ultrasonic testing (AUT) system, human error of ultrasonic testing is barely expected for the material of which thickness is not over 80 mm.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.34 no.9
• /
• pp.1241-1248
• /
• 2010

In large-scale production of hydrogen, a PHE (Process Heat Exchanger) is a key component because the heat required to carry out the Sulfur-Iodine chemical reaction that yields hydrogen is transferred from a VHTR (Very High Temperature Reactor) by the PHE. Korea Atomic Energy Research Institute established a helium gas loop for conducting performance test of components that are used in the VHTR. In this study, as a part of high-temperature structural-integrity evaluation of a designed PHE prototype that is scheduled to be tested in the helium gas loop, we carried out high-temperature structural-analysis modeling, thermal analysis, and thermal-expansion analysis for the designed PHE prototype. An appropriate constraint condition is proposed at the end of the in-flow and out-flow pipelines of the primary and secondary coolants and the proposed constraint condition will be applied to the design of the performance-test loop setup for the designed PHE prototype.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.35 no.11
• /
• pp.1499-1506
• /
• 2011

The IHX (intermediate heat exchanger) is a key component of nuclear hydrogen systems for the production of massive amounts hydrogen. The IHX transfers the $950^{\circ}C$ heat generated by the VHTR (very high temperature reactor) to a hydrogen production plant. The Korea Atomic Energy Research Institute established a small-scale gas loop to test the performance of key VHTR components and manufactured a small-scale PCHE (printed circuit heat exchanger) prototype, which is being considered as a candidate for the IHX, for testing in the small-scale gas loop. In this study, as a part of the high-temperature structural integrity evaluation of the small-scale PCHE prototype, we carried out high-temperature structural analysis modeling and macroscopic thermal and structural analysis for the small-scale PCHE prototype under the small-scale gas loop test conditions. This analysis serves as a precedent study to scheduled PCHE performance test in the small-scale gas loop. The results obtained in this study will be compared with the test results for the small-scale PCHE and then used to design the medium-scale PCHE prototype.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.34 no.5
• /
• pp.553-561
• /
• 2010

A medium-scale helium loop that can simulate a VHTR (very-high-temperature reactor) is now under construction at the Korea Atomic Energy Research Institute. The heaters of the test helium loop electrically heat helium fluid up to $950^{\circ}C$ at pressures of 1 to 9 MPa. To optimize the design specifications of the experimental helium loop, the conjugate heat transfer in the high-temperature helium heater was analyzed by performing a CFD simulation. The analysis results indicate that the maximum temperature does not exceed the allowable limit. It is confirmed that the thermal characteristics of the loop with the given geometry satisfy the design requirements.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.41 no.3
• /
• pp.193-198
• /
• 2017

Alloy 617 is one of the primary candidate materials to be used in a very high temperature reactor (VHTR) system as an intermediate heat exchanger (IHX). To investigate the low cycle fatigue behavior of Alloy 617 weldments at a high temperature of $850^{\circ}C$, fully reversed strain-controlled fatigue tests were conducted with the total strain values ranging from 0.6~1.5％. The weldment specimens were machined using the weld pads fabricated with a single V-grove configuration by gas tungsten arc welding (GTAW) process. The fatigue life is reduced as the total strain range increases. For all testing conditions, the cyclic stress response behavior of the Alloy 617 weldments exhibited the initial cyclic strain hardening phenomenon during the initial small number of cycles. Furthermore, the overall fatigue cracking and the propagation or cracks showed a transgranular failure mode.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.36 no.10
• /
• pp.1283-1288
• /
• 2012

A process heat exchanger (PHE) in a nuclear hydrogen system is a key component for transferring the considerable heat generated in a very high temperature reactor (VHTR) to a chemical reaction that yields a large quantity of hydrogen. A performance test on a medium-scale PHE prototype made of $Hastelloy^{(R)}$-X is scheduled in a small-scale gas loop at the Korea Atomic Energy Research Institute. In this study, as a preliminary study before carrying out the performance test in the gas loop, high-temperature structural analysis modeling and macroscopic thermal and structural analysis of the medium-scale PHE prototype by imposing the established displacement boundary constraints were carried out under the gas loop test condition. The results obtained in this study will be compared with the performance test results of the medium-scale PHE prototype in the gas loop.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.35 no.10
• /
• pp.1249-1255
• /
• 2011

A PHE (Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature Reactor) to a chemical reaction that yields a large quantity of hydrogen. A small-scale PHE prototype made of Hastelloy-X was scheduled for testing in a small-scale gas loop at the Korea Atomic Energy Research Institute. In this study, as a part of the evaluation of the high-temperature structural integrity of the PHE prototype, high-temperature structural analysis modeling, and macroscopic thermal and elastic-plastic structural analysis of the PHE prototype were carried out under the gas-loop test conditions as a preliminary qwer123$study before carrying out the performance test in the gas loop. The results obtained in this study will be used to design the performance test setup for the modified PHE prototype.

• Proceedings of the Materials Research Society of Korea Conference
• /
• 2012.05a
• /
• pp.95.1-95.1
• /
• 2012

Very High Temperature gas cooler Reactor (VHTR) has been considered as one of the most promising nuclear reactor because of many advantages including high inherent safety to avoid environmental pollution, high thermal efficiency and the role of secondary energy source. The TRISO coated fuel particles used in VHTR are composed of 4 layers as OPyC, SiC, IPyC and buffer PyC. The significance of CVD-SiC coatings used in tri-isotropic(TRISO) nuclear coated fuel particles is to maintain the strength of the whole particle. Various methods have been proposed to evaluate the mechanical properties of CVD-SiC film at room temperature. However, few works have been attempted to characterize properties of CVD-SiC film at high temperature. In this study, micro tensile system was newly developed for mechanical characterization of SiC thin film at elevated temperature. Two kinds of CVD-SiC films were prepared for micro tensile test. SiC-A had [111]-preferred orientation, while SiC-B had [220]-preferred orientation. The free silicon was co-deposited in SiC-B coating layer. The fracture strength of two different CVD-SiC films was characterized up to $1000^{\circ}C$.The strength of SiC-B film decreased with temperature. This result can be explained by free silicon, observed in SiC-B along the columnar boundaries by TEM. The presence of free silicon causes strength degradation. Also, larger Weibull-modulus was measured. The new method can be used for thin film material at high temperature.