• Nuclear Engineering and Technology
• /
• v.33 no.5
• /
• pp.498-513
• /
• 2001

Performed here is a comparative assessment study for the generation of the pressure- temperature (P/T) limit curve of the reactor vessel. Using the cooling or heating rate and vessel material properties, the stress distribution is obtained to calculate stress intensity factors, which are compared with the material fracture toughness to determine the relations between operating pressure and temperature during cool-down and heat-up. P/T limit curves are generated with respect to crack direction, clad thickness, toughness curve, cooling or heating rate and neutron fluence, and their results are compared.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.27 no.2
• /
• pp.231-238
• /
• 2003

The fuel of light water reactor is used for several years under high temperature and pressure, so it needs to be clad with high corrosion resistance material. The cladding materials must have the characteristics of low absorption of a neutron and high corrosion resistance. Zircaloy-2 in Boiling Water Reactor, Zircaloy-4 in Pressurized Water Reactor have been used as cladding materials and Zirlo has been developed as the material for preventing the corrosion. If the fracture of the cladding tube occurs during operation, it will cause the economic loss to shut down and replace the system. So it is needed to evaluate the integrity of the cladding materials. In this paper, the tensile characteristics of the cladding materials were investigated for the basic research of fracture characteristics. Also the residual stress was analyzed to compare the tube type(original type) specimen and the flattened type specimen.

• Journal of Ocean Engineering and Technology
• /
• v.14 no.1
• /
• pp.44-51
• /
• 2000

Safety and integrity are required for reactor pressure vessels because they are operated in high temperature. There are single specimen method multiple specimen method and load ratio analysis method which used as evaluation of safety and integrity for reactor pressure vessels. In this study the fracture resistance curve(J-R curve) elastic-plastic fracture toughness($J_{IC}$) and material tearing modulus ($T_{mat}$) of SA 508 class 3 alloy steel used as reactor pressure vessel steel are measured and evaluated at room temperature 20$0^{\circ}C$ and 30$0^{\circ}C$ according to unloading compliance method and load ration analysis method. And then the comparison with experimental $J_{IC}$ and theoretical$J_{IC}$ by local fracture strain is managed.

• Corrosion Science and Technology
• /
• v.5 no.3
• /
• pp.94-98
• /
• 2006

The next generation nuclear power reactor will use Pb-Bi as the cooling material. The steel structure materials such as HT-9 used in the reactor suffer from corrosion when they are exposed to high temperature Pb-Bi. Therefore corrosion should be prevented to use Pb-Bi as the coolant material without any safety problem. One method is to control the oxygen content in Pb-Bi. An appropriate amount of oxygen in Pb-Bi can produce a thin oxide layer on steel, and this layer protects the steel from corrosion attack. Since the required oxygen content in Pb-Bi is in the range of $10^{-5}$ to $10^{-7}$ wt%, this small oxygen content can be controlled by flowing a mixture of hydrogen gas and water vapor. The stagnant corrosion test of HT-9 samples was performed by controlling the oxygen content up to 2,000 hours. The corrosion behavior of HT-9 was analyzed at the temperatures of $500^{\circ}C$ and $650^{\circ}C$ with a reduced condition and a oxygen content of $10^{-6}$ wt%.

• 한국연소학회:학술대회논문집
• /
• 2013.06a
• /
• pp.51-54
• /
• 2013

Rotary kiln reactors are frequently equipped with an axial burner with which solid burden material is directly heated. Lifters are commonly used along the length of the system to lift particulate solids and increase the heat transfer between the solid bed and the combustion gas. The material cascading from the lifters undergoes drying and reacting through direct contact with the gas stream. In this study, volume distribution of materials held within lifters was modeled according to the different lifter configuration and appropriate configuration was used for the design purpose. This was applied to the one-dimensional heat balance model of a counter-current flow reactor, which contributes to the increase of the effective contact surface, and thereby enhances the heat transfer.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2120-2134
• /
• 2022

A hypothetical core destructive accident (HCDA) has received widespread attention as one of the most serious accidents in sodium-cooled fast reactors. This study combined recent advantages in numerical methods to realize realistic modeling of the complex fluid-structure interactions during HCDAs in a full-scale sodium-cooled fast reactor. The multi-material arbitrary Lagrangian-Eulerian method is used to describe the fluid-structure interactions inside the container. Both the structural deformations and plug rises occurring during HCDAs are evaluated. Two levels of expansion energy are considered with two different reactor models. The simulation results show that the container remains intact during an accident with small deformations. The plug on the top of the container rises to an acceptable level after the sealing between the it and its support is destroyed. The methodology established in this study provides a reliable approach for evaluating the safety feature of a container design.

• Journal of the Korean Crystal Growth and Crystal Technology
• /
• v.18 no.3
• /
• pp.101-104
• /
• 2008

GaN films were grown on the vertical and horizontal reactors by the hydride vapour phase epitaxy (HVPE). The structural and optical characteristics of the GaN films were investigated depending on the reactor-type. GaN epilayers were characterized by double crystal X-ray diffraction (DC-XRD), transmission electron microscopy (TEM) and photoluminescence (PL). Surface defects of two kinds of the GaN films were revealed by the wet chemical etching method, using $H_3PO_4$ acid at $200^{\circ}C$ for 8 minutes. Hexagonal etch pits were analyzed by optical microscopy and SEM. Etch pit densities were calculated to be approximately $1.4{\times}10^7$ and $1.2{\times}10^6\;cm^{-2}$ for GaN layers grown on horizontal and vertical reactors, respectively. Those results show GaN grown in the vertical reactor having a better quality of optical properties and crystallinity than that in the horizontal reactor.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
• /
• 2003.07b
• /
• pp.1115-1118
• /
• 2003

A great part of flint fuel is Hydrogen and carbon's combination. they should be generated steam and carbon gas when combine with oxygen for perfect combustion in the inside of engine, but the pollutive material especially NOx and Soot are exhausted, even though fuel is combine perfectly. In this paper the characteristics of exhaust material from diesel engine is investigated for the corona discharge-EGR hybrid plasma reactor, which combined corona discharge with EGR system. NOx is decreased in all load and applied voltage to plasma reactor. However Soot is increased as increasing EGR rate and it is decreased as increasing applied voltage.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2094-2106
• /
• 2022

The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K with thermal and irradiation-induced expansion during burnup. The expansion changes the gap thickness between the solid components and the material properties, and may even cause the gap closure, which then significantly influences the thermal and mechanical characteristics of the reactor core. This study developed an irradiation behavior model for HPRTRAN, a heat pipe reactor system analysis code, to introduce the irradiation effects such as swelling and creep. The megawatt heat pipe reactor MegaPower was chosen as an application case. The coupled irradiation-thermal-mechanical model was developed to simulate the irradiation effects on the heat transfer and stresses of the whole reactor core. The results show that the irradiation deformation effect is significant, with the irradiation-induced strains up to 2.82% for fuel and 0.30% for monolith at the end of the reactor lifetime. The peak temperatures during the lifetime are 1027:3 K for the fuel and 956:2 K for monolith. The gap closure enhances the heat transfer but caused high stresses exceeding the yield strength in the monolith.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.14 no.2
• /
• pp.91-100
• /
• 2016

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.