• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.521-528
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• 2014

The surface modification of engineering materials by laser beam scanning (LBS) allows the improvement of properties in terms of reduced wear, increased corrosion resistance, and better strength. In this study, the laser beam scan method was applied to produce an oxide dispersion strengthened (ODS) structure on a zirconium metal surface. A recrystallized Zircaloy-4 alloy sheet with a thickness of 2 mm, and $Y_2O_3$ particles of $10{\mu}m$ were selected for ODS treatment using LBS. Through the LBS method, the $Y_2O_3$ particles were dispersed in the Zircaloy-4 sheet surface at a thickness of 0.4 mm, which was about 20% when compared to the initial sheet thickness. The mean size of the dispersive particles was 20 nm, and the yield strength of the ODS treated plate at $500^{\circ}C$ was increased more than 65 % when compared to the initial state. This strength increase was caused by dispersive $Y_2O_3$ particles in the matrix and the martensite transformation of Zircaloy-4 matrix by the LBS.

• Korean Journal of Materials Research
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• v.14 no.7
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• pp.462-469
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• 2004

The effect of Cu addition on the corrosion characteristics of Zr alloys that developed for nuclear fuel cladding in KAERI (Korea Atomic Energy Research Institute) was evaluated. The alloys having different element of Nb, Sn, Fe, Cr and Cu were manufactured and the corrosion tests of the alloys were performed in static autoclave at $360^{\circ}C$, distilled water condition. The alloys were also examined for their microstructures using the optical microscope and the TEM equipped with EDS and the oxide property was characterized by using X-ray diffraction. From the result of corrosion test more than 450 days, the corrosion rate of the Zr-based alloys was changed with alloying element such as Nb, Sn, Fe, Cr and especially affected by Cu addition. The corrosion resistance was increased with increasing the Cu content and the tetragonal $ZrO_2$ layer was more stabilized on the Cu-containing alloys.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1330-1337
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• 2016

A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.

• 제어로봇시스템학회:학술대회논문집
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• 2003.10a
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• pp.1220-1224
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• 2003

In this paper, we measured the radiation dose of a fuel handling machine of the CANDU type Wolsong nuclear reactor directly during operation, in spite of the high radiation level. In this paper we will describe the sensor development, measurement techniques, and results of our study. For this study, we used specially developed semiconductor sensors and matching dosimetry techniques for the mixed radiation field. MOSFET dosimeters with a thin oxide, that are tuned to a high dose, were used to measure the ionizing radiation dose. Silicon diode dosimeters with an optimum area to thickness ratio were used for the radiation damage measurements. The sensors are able to distinguish neutrons from gamma/X-rays. To measure the radiation dose, electronic sensor modules were installed on two locations of the fuel handling machine. The measurements were performed throughout one reactor maintenance cycle. The resultant annual cumulative dose of gamma/X-rays on the two spots of the fuel handling machine were 18.47 Mrad and 76.50 Mrad, and those of the neutrons were 17.51 krad and 60.67 krad. The measured radiation level is high enough to degrade certain cable insulation materials that may result in electrical insulation failure.

• Nuclear industry
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• v.36 no.1
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• pp.75-76
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• 2016

일본은 사용후연료의 재처리 정책을 진행하고 있으나 고속증식로 계획이 지연됨에 따라 그 처리의 목표가 확실하지 못하다. MOX(Mixed-Oxide Fuel : 혼합 산화물 연료)에 의한 처리가 계획되어 있으나 상당히 어려운 상황이다. 또 2018년에 미 일 원자력협력협정의 갱신이 예정되어 있는데 이 플루토늄의 취급이 문제가 될 예정이다. 미국은 핵무기의 원료나 테러에 사용할지도 모르는 플루토늄의 확산을 경계하고 있기 때문이다.

• Nuclear Engineering and Technology
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• v.41 no.2
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• pp.149-154
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• 2009

The microstructure, the corrosion behavior and the oxide properties were examined for Zr-1.5Nb-0.4Sn-0.2Fe-0.1Cr (HANA-4) alloys which were subjected to two different final annealing temperatures: $470^{\circ}C$ and $570^{\circ}C$. HANA-4 was shown to have $\ss$-enriched phase with a bcc crystal structure and Zr(Nb,Fe,Cr)$_2$ with a hcp crystal structure with $\ss$-enriched phase being more frequently observed compared with Zr(Nb,Fe,Cr)$_2$. The corrosion rate of HANA-4 was increased with an increase of the final annealing temperature in the PWR-simulating loop, $360^{\circ}C$ pure water and $400^{\circ}C$ steam conditions, which was correlated well with a reduction in the size of the columnar grains in the oxide/metal interface region. The oxide growth rate of HANA-4 was considerably affected by the alloy microstructure determined by the final annealing temperature.

• Korean Journal of Materials Research
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• v.33 no.10
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• pp.385-392
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• 2023

A solution combustion process for the synthesis of hollandite (BaAl₂Ti₆O₁₆) powders is described. SYNROC (synthetic rock) consists of four main titanate phases: perovskite, zirconolite, hollandite and rutile. Hollandite is one of the crystalline host matrices used for the disposal of high-level radioactive wastes because it immobilizes Sr and Lns elements by forming solid solutions. The solution combustion synthesis, which is a self-sustaining oxi-reduction reaction between a nitrate and organic fuel, generates an exothermic reaction and that heat converts the precursors into their corresponding oxide products in air. The process has high energy efficiency, fast heating rates, short reaction times, and high compositional homogeneity. To confirm the combustion synthesis reaction, FT-IR analysis was conducted using glycine with a carboxyl group and an amine as fuel to observe its bonding with metal element in the nitrate. TG-DTA, X-ray diffraction analysis, SEM and EDS were performed to confirm the formed phases and morphology. Powders with an uncontrolled shape were obtained through a general oxide-route process, confirming hollandite powders with micro-sized soft agglomerates consisting of nano-sized primary particles can be prepared using these methods.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.657-665
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• 2021

Nondestructive evaluation methods play an important role in ensuring component integrity and safety in many industries. Operator fatigue can play a critical role in the reliability of such methods. This is important for inspecting high value assets or assets with a high consequence of failure, such as aerospace and nuclear components. Recent advances in convolution neural networks can support and automate these inspection efforts. This paper proposes using residual neural networks (ResNets) for real-time detection of corrosion, including iron oxide discoloration, pitting and stress corrosion cracking, in dry storage stainless steel canisters housing used nuclear fuel. The proposed approach crops nuclear canister images into smaller tiles, trains a ResNet on these tiles, and classifies images as corroded or intact using the per-image count of tiles predicted as corroded by the ResNet. The results demonstrate that such a deep learning approach allows to detect the locus of corrosion via smaller tiles, and at the same time to infer with high accuracy whether an image comes from a corroded canister. Thereby, the proposed approach holds promise to automate and speed up nuclear fuel canister inspections, to minimize inspection costs, and to partially replace human-conducted onsite inspections, thus reducing radiation doses to personnel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.19-32
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• 2010

Electrolytic reduction technology is essential for the purpose of adopting pyroprocessing into spent oxide fuel as an alternative option in a back-end fuel cycle. Spent fuel consists of various metal oxides, and each metal oxide releases an oxygen element depending on its chemical characteristic during the electrolytic reduction process. In the present work, an electrolytic reduction behavior was estimated for voloxidized spent fuel based on the assumption that each metal-oxygen system is independent and behaves as an ideal solid solution. The electrolytic reduction was considered as a combination of a Li recovery and chemical reactions between the metal oxides such as uranium oxide and the produced Li metal. The calculated result revealed that most of the metal oxides were reduced by the process. It was evaluated that a reduced fraction of lanthanide oxides increased with a decreasing $Li_2O$ concentration. However, most of the lanthanides were expected to be stable in their oxide forms. In addition, a semi-empirical model for describing $U_3O_8$ electrolytic reduction behavior was proposed by considering Li diffusion and a chemical reaction between $U_3O_8$ and Li. Experimental data was used to determine model parameters and, then, the model was applied to calculate the reduction yield with time and to estimate the required time for a 99.9% reduction.

• Nuclear Engineering and Technology
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• v.38 no.6
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• pp.535-550
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• 2006

The objectives of this paper are twofold to summarize the new findings and confirmed results from the Phebus FPT-1 experimental data and to report useful information to MELCOR users regarding the better use of MELCOR. For the core damage behavior, the early stage of a melt progression was predicted well; however, the late phase models, concerned with fuel dissolution, oxide cladding failure, fuel slumping, rubble debris heat up, effects of burn-up fuel, and so on, still showed limitations in MELCOR. For the fission product behavior, the comparison showed unexpected phenomena, various limitations, unresolved issues, and even absence of models. The issues summarized in this study have revealed the main areas where our endeavors need to be intensified in order to improve our understanding of severe accident phenomena. From the analysis of the Phebus FPT-1 test results, not only new core damage features, such as foaming or core expansion, but also possible new fission product release patterns due to effects from a high burn-up fuel have raised alternative challenging phenomena that should be solved in the next severe accident research phase.