• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1142-1153
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• 2019

This paper presents the effect of fusion procedure on the fusion performance of the thermal butt fusion in the safety class III buried HDPE piping per various tests performed, including high speed tensile impact, free bend, blunt notched tensile, notched creep, and PENT tests. The suitability of fusion joints and qualification procedures was evaluated by comparing test results from the base material and buttfusion joints. From the notched tensile test result, it was found that the fused joints have much lower toughness than the base material. It was also identified that the notched tensile test is more desirable than the high speed tensile impact and free bend tests presented in the ASME Code Case N-755-3 as a fusion qualification test method. In addition, with regard to the single low-pressure fusion joint performances, the procedure given by the ISO 21307 was determined to be better that the one specified in the Code Case N-755-3.

• Corrosion Science and Technology
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• v.22 no.3
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• pp.214-219
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• 2023

Considering the case in the United States where most nuclear power plants with an initial design life of 40 years continue to operate until 60 or 80 years after undergoing material soundness evaluation, it is time to plan a more robust long-term operation strategy for nuclear power plants in Korea. There are some reports that SRO/LRO might be formed when Alloy 690 is heat treated for 10,000 hours to 100,000 hours at 360 to 450 ℃. The possibility of LRO formation in Alloy 690 steam generator tubings of Kori nuclear power plant unit 1 (Kori-1) was investigated using existing research papers. The mechanism in which SRO/LRO occurred was also surveyed. Alloy 690 was found to be more likely to cause ordering than Alloy 600 in terms of alloy composition. The ordering could be evaluated through changes in material properties. However, it is difficult to evaluate it from a microstructural point of view. The likelihood of LRO in Alloy 690 of the Kori-1 plant operated at 320 ℃ for 19 years seemed to be low in terms of time and exposure temperature.

• Journal of Ocean Engineering and Technology
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• v.37 no.5
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• pp.205-214
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• 2023

As the regulations on greenhouse gas emissions at sea become strict, efforts are being made to minimize environmental pollutants emitted from fossil fuels used by ships. Considering the large sizes of ships in conjunction with securing stable supplies of environment-friendly energy, interest in nuclear energy to power ships has been increasing. In this study, the neutron irradiation that occurs during the nuclear reactor operation and its effect on the structural responses of the stiffened-plate structures are investigated. This is done by changing the material properties of DH36 steel according to the research findings on the neutron-irradiated steels and then performing the structural response analyses of the structures using analytical and finite-element numerical solutions. Results reveal the influence of neutron irradiation on the structural responses of the structures. It is shown that both the strength and stiffness of the structures are affected by the neutron-irradiation phenomenon as their maximum flexural stress and deflection are increased with the increase in the amount of neutron irradiation. This implies that strength and stiffness need to be considered in the design of ships equipped with marine nuclear reactors.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2570-2575
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• 2024

The study aims to develop a novel, lead-free, flexible and lightweight composite shielding material against ionizing radiation. For this, it was used bismuth oxide (Bi₂O₃) in RTV-2 silicon matrix. The shielding tests were carried out in both diagnostic X-ray energies and intermediate gamma-ray energy range of up to 662 keV to determine the radiation attenuation properties of this material in terms of attenuation ratio, half value layer, tenth value layer, mean free path and lead equivalency of samples in weight of 30%, 40%, 50% in Bi₂O₃. In the diagnostic X-ray energy range, half value layer, tenth value layer and lead equivalency (in mm Pb) of the produced samples were measured at 80 and 100 kVp narrow beam conditions according to the requirements of EN IEC 61331-1 standard. The results show that lead equivalent values of the produced novel sheets was measured to be 0.16 mm Pb, corresponding to a 6 mm thickness of the flexible sample when it contains 30% wt. Bi₂O₃ in RTV matrix. The experimental findings for durability and flexibility also indicated that this new RTV-based flexible, lead -free shielding composite can be used safely for especially for manufacturing aprons, garments and thyroid guards used in mammography, radiology, nuclear medicine and dental applications in practice.

• Journal of Energy Engineering
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• v.23 no.3
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• pp.21-26
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• 2014

EU-APR1400, the Korean nuclear reactor design for European market adopts a so-called core catcher for ex-vessel molten corium retention and cooling as a severe-accident mitigation system. Sacrificial material, which controls melt properties and modifies melt conditions favorable for corium cooling and retention, is usually employed to protect core catcher body from molten corium. Since molten corium can be ejected through a breach of a reactor pressure vessel and impinged on the sacrificial material with enhanced heat transfer at a severe accident, it is very important to predict ablation rate of sacrificial material due to corium jet impingement accurately for core catcher design. In this paper, sacrificial-material ablation model based on boundary layer theory is suggested and compared with the experimental results by KAERI.

• Korean Journal of Materials Research
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• v.12 no.12
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• pp.941-946
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• 2002

Both Cu and Cu-oxide nanopowders have great potential as conductive paste, solid lubricant, effective catalysts and super conducting materials because of their unique properties compared with those of commercial micro-sized ones. In this study, Cu and Cu-oxide nanopowders were prepared by Pulsed Wire Evaporation (PWE) method which has been very useful for producing nanometer-sized metal, alloy and ceramic powders. In this process, the metal wire is explosively converted into ultrafine particles under high electric pulse current (between $10^4$ and $10^{ 6}$ $A/mm^2$) within a micro second time. To prevent full oxidations of Cu powder, the surface of powder has been slightly passivated with thin CuO layer. X-ray diffraction analysis has shown that pure Cu nanopowders were obtained at $N_2$ atmosphere. As the oxygen partial pressure increased in $N_2$ atmosphere, the gradual phase transformation occurred from Cu to $Cu_2$O and finally CuO nanopowders. The spherical Cu nanopowders had a uniform size distribution of about 100nm in diameter. The Cu-oxide nanopowders were less than 70nm with sphere-like shape and their mean particle size was 54nm. Smaller size of Cu-oxide nanopowders compared with that of the Cu nanopowders results from the secondary explosion of Cu nanopowders at oxygen atmosphere. Thin passivated oxygen layer on the Cu surface has been proved by XPS and HRPD.

• Nuclear Engineering and Technology
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• v.19 no.3
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• pp.169-179
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• 1987

The change of material J-R (J-T) curve with crack extension and J-calculation method was investigated to give experimental and analytical method for reliable J-R (J-T) curve, which was adapted recently as a tool for instability analysis of Nuclear Pressure Vessel. Experiments were carried out by Single Specimen Unloading Compliance Method using 1/2"T, Compact-Tension Type fracture mechanic specimens which were the same size and material as domestic nuclear pressure vessel material surveillance specimens. The results revealed that crack extension up to 25~30% of initial uncracked ligament and JD (Deformation theory J) calculation method, currently being used in NUREG-0744, could give rather reliable material J-R (J-T) curve than the small crack extension and JM (Modified J) calculation method. But as JM results more or less higher J at instability, the application of JM should be considered regarding to the problem of power plant availability.lity.

• Korean Journal of Metals and Materials
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• v.46 no.3
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• pp.118-124
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• 2008

Abstract: A minimum commitment method(MCM) was applied to predict the long-term creep rupture life for type 316LN stainless steel(SS). Lots of the creep-rupture data for the type 316LN SS were collected through world-wide literature surveys and the experimental data of KAERI. Using these data, the long-term creep rupture life above ${10}^5$ hour was predicted by means of the MCM. In order to obtain the most appropriate value for the constant A being used in the MCM equation, trial and error method was used for the wide ranges from -0.12 to 0.12, and the best value was determined by using the coefficient of determination, $R^2$ which is a statistical parameter. A suitable value for the A in type 316LN stainless steel was found to be at -0.02 ~ -0.05 ranges. It is considered that the MCM will be superior in creep-life prediction to commonly-used timetemperature parametric method, because the P(T) and G($\sigma$) functions are determined from the regression method based on experimental data.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.21-26
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• 1997

Recently, thorium is receiving increasing attention as an important fertile material for the expanding nuclear power programs around the world. The superior nuclear and physical properties of thorium-based fuels could lead to very low fuel cycle cost and make thorium reactors economically attractive. In addition, the use of thorium in reactors would permit more efficient utilization of low cost uranium reserves and reduction nuclear wastes. In this work, the nuclear characteristics of a new type thorium fueled reactor (Radkowsky Thorium Reactor) consisting seed-blanket assemblies are addressed and compared with those typical assemblies of a PWR (CE type). Also, an assessment on several advantages of thorium fueled reactors is provided. All these results are based on the HELIOS code calculation.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.501-512
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• 2014

HANARO is a multipurpose research reactor located at the Korea Atomic Energy Research Institute (KAERI). Since the commencement of its operation in 1995, various neutron irradiation facilities, such as rabbit irradiation facilities, fuel test loop (FTL) facilities, capsule irradiation facilities, and neutron transmutation doping (NTD) facilities, have been developed and actively utilized for various nuclear material irradiation tests requested by users from research institutes, universities, and industries. Most irradiation tests have been related to national R&D relevant to present nuclear power reactors such as the ageing management and safety evaluation of the components. Based on the accumulated experience as well as the sophisticated requirements of users, HANARO has recently supported national R&D projects relevant to new nuclear systems including the System-integrated Modular Advanced Reactor (SMART), research reactors, and future nuclear systems. This paper documents the current state and utilization of irradiation facilities in HANARO, and summarizes ongoing research efforts to deploy advanced irradiation technology.