• Transactions of Materials Processing
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• v.14 no.6 s.78
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• pp.565-569
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• 2005

Reactor Vessel is one of the most important structural parts of nuclear power plant. It is manufactured by various steel forgings such as shell, head and transition ring. Head forgings have been made by open die forging process. After steel melting and ingot making, open die forging has been carried out to get a good quality which means high soundness and homogeniety of the steel forgings by using high capacity hydraulic press. This paper introduced the open die forging process and manufacturing experience of large head forgings which can be used for the reactor vessel of 1,000MW nuclear power plant.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2020.11a
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• pp.151-152
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• 2020

When concrete is exposed to fire, the thermal decomposition of hydrates of Portland cement paste results in critical damage to the concrete structure of a building. Recently, nanosilica arose as the effective nano-additive which can enhance the thermal resistance of the cementitious materials. However, the mechanism of the enhancement was not elucidated specifically. In this study, we investigated the properties of calcium silicate hydrates(C-S-H)of the nanosilica incorporated cement paste after heating to different heating temperatures (200℃, 500℃, and 800℃) by 29Si nuclear magnetic resonance. The results showed that the polymerization of C-S-H of nanosilica incorporated samples was larger than ordinary cement paste after heating to 200℃, and C-S-H formed during heating process to 500℃ due to the pozzolanic reaction during heating process.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2582-2590
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• 2021

The high-energy milling is one of the most extended techniques to produce Oxide dispersion strengthened (ODS) powder steels for nuclear applications. The consequences of the high energy mill process on the final powders can be measured by means of deformation level, size, morphology and alloying degree. In this work, an ODS ferritic steel, Fe-14Cr-5Al-3W-0.4Ti-0.25Y₂O₃-0.6Zr, was fabricated using two different mechanical alloying (MA) conditions (M_std and M_act) and subsequently consolidated by Spark Plasma Sintering (SPS). Milling conditions were set to evidence the effectivity of milling by changing the revolutions per minute (rpm) and dwell milling time. Differences on the particle size distribution as well as on the stored plastic deformation were observed, determining the consolidation ability of the material and the achieved microstructure. Since recrystallization depends on the plastic deformation degree, the composition of each particle and the promoted oxide dispersion, a dual grain size distribution was attained after SPS consolidation. M_act showed the highest areas of ultrafine regions when the material is consolidated at 1100 ℃. Microhardness and small punch tests were used to evaluate the material under room temperature and up to 500 ℃. The produced materials have attained remarkable mechanical properties under high temperature conditions.

• Journal of the Korean Society of Industry Convergence
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• v.24 no.3
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• pp.289-294
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• 2021

Tungsten (W) is used as a facing material for nuclear fusion reactors, and it is used in conjunction with structural materials such as copper alloy (CuCrZr), graphite, or stainless steel. On the other hand, since tungsten is a material with a high melting point, a method that can be manufactured at a lower temperature is important. Therefore, in this study, tungsten, which is a facing material, was attempted to be manufactured using a pressure sintering method. Material properties of sintered tungsten materials were analyzed for each solvent using two types of solvents, acetone and polyethylene glycol. The sintered tungsten material using acetone as a solvent exhibited a hardness value of about 255 Hv, and when polyethylene glycol was used, a hardness value of about 200 Hv was shown. The flexural strength of the sintered tungsten material was 870 MPa and 307 MPa, respectively, when acetone and polyethylene glycol were used as solvents. The sintered tungsten material using acetone as a solvent caused densification between particles, which served as a factor of increasing the strength.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.555-565
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• 2023

Despite many advantages as structural materials, austenitic stainless steels (SSs) have been avoided in many next generation nuclear systems due to poor void swelling resistance. In this paper, we report the results of heavy ion irradiation to the recently developed advanced radiation resistant austenitic SS (ARES-6P) with nanosized NbC precipitates. Heavy ion irradiation was performed at high temperatures (500 ℃ and 575 ℃) to the damage level of ~200 displacement per atom (dpa). The measured void swelling of ARES-6P was 2-3%, which was considerably less compared to commercial 316 SS and comparable to ferritic martensitic steels. In addition, increment of hardness measured by nano-indentation was much smaller for ARES-6P compared to 316 SS. Though some nanosized NbC precipitates were dissociated under relatively high dose rate (~5.0 × 10^-4 dpa/s), sufficient number of NbC precipitates remained to act as sink sites for the point defects, resulting in such superior radiation resistance.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3725-3731
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• 2023

Thorium dioxide (ThO₂)-silicon carbide (SiC) composite fuel pellets were fabricated via the spark plasma-sintering (SPS) method to investigate the role of the addition of SiC in enhancing the thermal conductivity of ThO₂ fuel. SiC particles with an average size of 1㎛ in 10 and 15 vol% were used to manufacture the composite pellets. The changes in the composites' densification, microstructure and thermal conductivity were explored by comparing them with pure ThO₂ pellets. The structural and microstructural characterization of the composite pellets has revealed that SPS could manufacture high-quality composite pellets without having any reaction products or intermetallic. The density measurement by the Archimedes principles and the grain size from the electron back-scattered diffraction (EBSD) analysis has indicated that the composites have higher densities and smaller grain sizes than the pellets without SiC addition. Finally, thermal conductivity as a function of temperature has revealed that sintered ThO₂-SiC composites showed an increase of up to 56% in thermal conductivity compared to pristine ThO₂ pellets.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1373-1380
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• 2019

Besides promising implications as fertile nuclear materials, thorium carbonitrides are of great interest owing to their peculiar physical and chemical properties, such as high density, high melting point, good thermal conductivity. This paper reports first-principles simulation results on the structural, electronic and magnetic properties of cubic thorium carbonitrides $ThC_xN_{(1-x)}$ (X = 0.03125, 0.0625, 0.09375, 0.125, 0.15625) employing formalism of density-functional-theory. For the simulation of physical properties, we incorporated full-potential linearized augmented plane-wave (FPLAPW) method while the exchange-correlation potential terms in Kohn-Sham Equation (KSE) are treated within Generalized-Gradient-Approximation (GGA) in conjunction with Perdew-Bruke-Ernzerhof (PBE) correction. The structural parameters were calculated by fitting total energy into the Murnaghan's equation of state. The lattice constants, bulk moduli, total energy, electronic band structure and spin magnetic moments of the compounds show dependence on the C/N concentration ratio. The electronic and magnetic properties have revealed non-magnetic but metallic character of the compounds. The main contribution to density of states at the Fermi level stems from the comparable spectral intensity of Th (6d+5f) and (C+N) 2p states. In comparison with spin magnetic moments of ThSb and ThBi calculated earlier with LDA+U approach, we observed an enhancement in the spin magnetic moments after carbon-doping into ThN monopnictide.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1423-1432
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• 2016

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

• Journal of the Korea institute for structural maintenance and inspection
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• v.18 no.1
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• pp.128-137
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• 2014

Sprayed fire resistive materials are mainly used at steel structures to satisfy fireproof construction standard. However, the regulations on bond strength have been not considered with the exception of structures in the nuclear power plants, although it is an important factor showing material properties. Therefore, this paper suggested guidelines for bond strength of sprayed fire resistive materials used in the substation, by measuring bond strength according to aging of structures and impact loading considering environment of substations. It is judged that the bond strength suggested in this paper is the minimum value because it was measured from specimens widely used.

• Journal of Powder Materials
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• v.31 no.3
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• pp.220-225
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• 2024

Molten salt reactors represent a promising advancement in nuclear technology due to their potential for enhanced safety, higher efficiency, and reduced nuclear waste. However, the development of structural materials that can survive under severe corrosion environments is crucial. In the present work, pure Ni was deposited on the surface of 316H stainless steel using a directed energy deposition (DED) process. This study aimed to fabricate pure Ni alloy layers on an STS316H alloy substrate. It was observed that low laser power during the deposition of pure Ni on the STS316H substrate could induce stacking defects such as surface irregularities and internal voids, which were confirmed through photographic and SEM analyses. Additionally, the diffusion of Fe and Cr elements from the STS316H substrate into the Ni layers was observed to decrease with increasing Ni deposition height. Analysis of the composition of Cr and Fe components within the Ni deposition structures allows for the prediction of properties such as the corrosion resistance of Ni.