• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2823-2834
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• 2023

Austenitic stainless steel welds (ASSWs) of nuclear components undergo aging-related degradations caused by high temperature and neutron radiation. Since irradiation leads to the change of material characteristics, relevant quantification is important for long-term operation, but limitations exist. Although ion irradiation is utilized to emulate neutron irradiation, its penetration depth is too shallow to measure bulk properties. In this study, a systematic approach was suggested to estimate mechanical properties of ion irradiated 308 ASSW. First of all, weld specimens were irradiated by 2 MeV proton to 1 and 10 dpa. Microstructure evolutions due to irradiation in δ-ferrite and austenite phases were characterized and micropillar compression tests were performed. In succession, dislocation density based stress-strain (S-S) relationships and quantification models of irradiation defects were adopted to define phases in finite element analyses. Resultant microscopic S-S curves were compared to verify material parameters. Finally, macroscopic behaviors were calculated by multiscale simulations using real microstructure based representative volume element (RVE). Validity of the approach was verified for the unirradiated specimens such that the estimated S-S curves and 0.2% offset yield strengths (YSs) which was 363.14 MPa were in 10% agreement with test. For irradiated specimens, the estimated YS were 917.41 MPa in 9% agreement.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2288-2297
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• 2023

Current designs of both large reactor units and small modular reactors utilize a nuclear fuel with increasing enrichment. This increasing demand for better nuclear fuel utilization is a challenge for nuclear fuel handling facilities. The operation with higher enriched fuels leads to reduced reserves to legislative and safety criticality limits of spent fuel transport, storage and final disposal facilities. Design changes in these facilities are restricted due to a boron content in steel and aluminum alloys that are limited by rolling, extrusion, welding and other manufacturing processes. One possible solution for spent fuel pools and casks is the burnup credit method that allows decreasing very high safety margins associated with the fresh fuel assumption in spent fuel facilities. This solution can be supplemented or replaced by an alternative solution based on placing the neutron absorber material directly into the fuel assembly, where its efficiency is higher than between fuel assemblies. A neutron absorber permanently fixed in guide tubes decreases system reactivity more efficiently than absorber sheets between the fuel assemblies. The paper summarizes possibilities of fixed neutron absorbers for various nuclear fuel and fuel handling facilities. Moreover, an absorber material was optimized to propose alternative options to boron. Multiple effective absorbers that do not require steel or aluminum alloy compatibility are discussed because fixed absorbers are placed inside zirconium or steel cladding.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4664-4670
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• 2023

Radiation is the main safety issue for almost all nuclear applications, which must be controlled to protect living organisms and the surrounding materials. In this context, radiation shielding materials have been investigated and used in nuclear technologies. The choice of materials depends on the radiation usage area, type, and energy. Polymer materials are preferred in radiation shielding applications due to their superior characteristics such as chemical inertness, resistivity, low weight, flexibility, strength, and low cost. In the presented work, ABS polymer material, which is possibly the most commonly used material in 3D printers, is mixed with Gd₂O₃ and Er₂O₃ nanoparticles. ABS filaments containing these rare-earth elements are then produced using a filament extruder. These produced filaments are used in a 3D printer to create shielding samples. Following the production of shielding samples, SEM, EDS, and gamma-ray shielding analyses (including experiments, WinXCOM, GEANT4, and FLUKA) are performed. The results show that 3D printing technology offers significant enhancements in creating homogeneous and well-structured materials that can be effectively used in gamma-ray shielding applications.

• Proceedings of the Korean Institute of IIIuminating and Electrical Installation Engineers Conference
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• 2005.11a
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• pp.151-155
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• 2005

Though the nuclear fusion system has been fused into hydro-nuclear based on thermodynamics by tokamak system, there has been no success story. Because it's impossible to confine high temperatured plasma long Time actually. New nuclear-fusion-system using this nuclear-fusion-method will gather toroidal-magnetic-field by putting Core Block(C shaped torus iron) and toroidal-aluminium coil into toroidal magnetic-field-aluminium. That will arrange the nuclear-fusion-route on a gap fallen out by a part of cut torus-core and mkee the toroidal-an electric-current flow and electrolyze the fusioned-material (an electrolyte) into troidal-electrocity. That consists of troidal-magnetic-fild coil, toroidal-coial fusioned- material, series circuit. So toroidal-electocity will be changed into filament-electrocity and be introjected into fusioned-material. In a sapce on filament-electrocity, the magnetic inhaling-powr will enlarge to input-electrocity outside. This will exceed the Coulomb force and reache the nuclear-fusion. By this phenomenon there be quantity-loss. By this process I could confirmed that Einstein euation$(E=mC^2)$ releases into thermal energy.

• Progress in Medical Physics
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• v.28 no.4
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• pp.207-217
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• 2017

For effective patient treatment in proton therapy, it is therefore important to accurately measure the beam range. For measuring beam range, various researchers determine the beam range by measuring the prompt gammas generated during nuclear reactions of protons with materials. However, the accuracy of the beam range determination can be lowered in heterogeneous phantoms, because of the differences with respect to the prompt gamma production depending on the properties of the material. In this research, to improve the beam range determination in a heterogeneous phantom, we derived a formula to correct the prompt-gamma distribution using the ratio of the prompt gamma production, stopping power, and density obtained for each material. Then, the prompt-gamma distributions were acquired by a multi-slit prompt-gamma camera on various kinds of heterogeneous phantoms using a Geant4 Monte Carlo simulation, and the deduced formula was applied to the prompt-gamma distributions. For the case involving the phantom having bone-equivalent material in the soft tissue-equivalent material, it was confirmed that compared to the actual range, the determined ranges were relatively accurate both before and after correction. In the case of a phantom having the lung-equivalent material in the soft tissue-equivalent material, although the maximum error before correction was 18.7 mm, the difference was very large. However, when the correction method was applied, the accuracy was significantly improved by a maximum error of 4.1 mm. Moreover, for a phantom that was constructed based on CT data, after applying the calibration method, the beam range could be generally determined within an error of 2.5 mm. Simulation results confirmed the potential to determine the beam range with high accuracy in heterogeneous phantoms by applying the proposed correction method. In future, these methods will be verified by performing experiments using a therapeutic proton beam.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.29 no.1
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• pp.17-22
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• 2016

The activation energy of a material is an important factor that significantly affects the lifetime and can be used to develop a degradation model. In this study, a thermal analysis was carried out to evaluate and collect quantitative data on the degradation of insulation materials like EPR and CSP used for nuclear power plant cables. The activation energy was determined from the relationship between log ${\beta}$ and 1/T based on the Flynn.Wall.Ozawa method, by a TGA test. The activation energy was also derived from the relationship between ln(t) and 1/T based on isothermal analysis, by an OIT test. The activation energy of EPR derived from thermal analysis was used to calculate the accelerated aging time corresponding to the number of years of use, employing the Arrhenius equation, and determine the elongation corresponding to the accelerated aging time.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.186-192
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• 2022

Flow accelerated corrosion (FAC) is a type of pipe corrosion in which the pipe thickness decreases depending on the fluid flow conditions. In nuclear power plants, FAC mainly occurs in the carbon steel pipes of a secondary system. However, because the temperature of a secondary system pipe is over 150 ℃, in situ monitoring using a conventional ultrasonic non-destructive testing method is difficult. In our previous study, we developed a waveguide ultrasonic thickness measurement system. In this study, we applied a waveguide ultrasonic thickness measurement system to monitor the thinning of the pipe according to the change in pH. The Korea Atomic Energy Research Institute installed FAC-proof facilities, enabling the monitoring of internal fluid flow conditions, which were fixed for ~1000 h to analyze the effect of the pH. The measurement system operated without failure for ~3000 h and the pipe thickness was found to be reduced by ~10% at pH 9 compared to that at pH 7. The thickness of the pipe was measured using a microscope after the experiment, and the reliability of the system was confirmed with less than 1% error. This technology is expected to also be applicable to the thickness-reduction monitoring of other high-temperature materials.

• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.417-422
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• 2006

The present paper describes the 45 MW High Flux Reactor (HFR) which is located in Petten, The Netherlands. This paper focuses on selected technical aspects of this reactor and on nuclear fuel irradiation experiments. These fuel experiments are mainly experiments on Light Water Reactor (LWR) and Very/High Temperature Reactor (V/HTR) fuels, but also on Fast Reactor (FR) fuels, transmutation fuels and Material Test Reactor (MTR) fuels.

• Transactions of the Korean Society of Mechanical Engineers
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• v.18 no.5
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• pp.1106-1116
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• 1994

Leak Before Break(LBB) design concept is applied to piping systems of newly-built Yonggwang 3, 4 nuclear generating stations as a design alternative to the provision of pipe whip restraints, in recognition of the questionable benefits of providing such restraints. The objective of this paper is to evaluate the material properties (tensile and fracture toughness) of SA312 TP316 stainless steel and their associated welds manufactured for shutdown cooling system of Yonggwang 3, 4 nuclear generating stations. Effect of various parameters such as specimen orientation, test temperature, welding on material properties were examined.

• Proceedings of the Korea Committee for Ocean Resources and Engineering Conference
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• 2000.10a
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• pp.42-46
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• 2000

In this paper, joints of Cu-1Cr-0.1Zr alloy to STS316L were performed by friction welding method. Cu-1Cr-0.1Zr alloy is attractive candidate as nuclear power plant material and exibit the best combination of high sts good electrical and thermal conductivity of any copper alloy examined. The stainless steel is a structural material who alloy acts as a heat sink material for the surface heat flux in the first wall. So, in this paper, not only the develop optimizing of friction welding with more reliability and more applicabililty but also the development of in-process rear quility(such as strength and toughness) evaluation technique by acoustic emission for friction welding of such nuclear component of Cu-1Cr-0.1Zr alloy to STS316L steel were performed.