• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.579-587
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• 2015

To support the use of nuclear power as a sustainable electric energy generating technology, long-term supply of uranium is very important. The objective of this research is to investigate the use of new adsorbent material for cost effective uranium extraction from seawater. An activated carbon-based adsorbent material is developed and tested through an electrosorption technique in this research. Adsorption of uranium from seawater by activated carbon electrodes was investigated through electrosorption experiments up to 300 minutes by changing positive potentials from +0.2V to +0.8V (vs. Ag/AgCl). Uranium adsorption by the activated carbon electrode developed in this research reached up to 3.4 g-U/kg-adsorbent material, which is comparable with the performance of amidoxime-based adsorbent materials. Electrosorption of uranium ions from seawater was found to be most favorable at +0.4V (vs. Ag/AgCl). The cost of chemicals and materials in the present research was compared with that of the amidoxime-based approach as part of the engineering feasibility examination.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1567-1576
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• 2023

The reactor containment building (RCB) in nuclear power plants (NPPs) plays an important role in protecting the reactor systems from external loads as well as preventing radioactive leaking. As we witnessed the nuclear disaster at Fukushima Daiichi (Japan) in 2011, the earthquake is one of the major threats to NPPs. The purpose of this study is to evaluate effects of concrete material models and presstressing forces on the seismic performance evaluation of RCB in NPPs. A typical RCB designed in Korea is employed for a case study. Detailed three-dimensional nonlinear finite element models of RCB are developed in ANSYS. A series of pushover analyses are then performed to obtain the pushover curves of RCB. Different capacity curves are compared to recognize the influence of different material models on the nonlinear behavior of RCB. Additionally, the effects of prestressing forces on the seismic performances of the structure are also investigated. Moreover, a set of damage states corresponding to damage evolutions of the structures is proposed in this study.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.416-417
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.05a
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• pp.23-24
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• 2016

• Journal of Radiation Industry
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• v.17 no.4
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• pp.345-357
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• 2023

Nuclear forensics has been understood as a mendatory component in the international society for nuclear material control and non-proliferation verification. Radiochronometry of nuclear activities for nuclear forensics are decay series characteristics of nuclear materials and the Bateman equation to estimate when nuclear materials were purified and produced. Radiochronometry values have uncertainty of measurement due to the uncertainty factors in the estimation process. These uncertainties should be calculated using appropriate evaluation methods that are representative of the accuracy and reliability. The IAEA, US, and EU have been researched on radiochronometry and uncertainty of measurement, although the uncertainty calculation method using the Bateman equation is limited by the underestimation of the decay constant and the impossibility of estimating the age of more than one generation, so it is necessary to conduct uncertainty calculation research using computer simulation such as Monte Carlo method. This highlights the need for research using computational simulations, such as the Monte Carlo method, to overcome these limitations. In this study, we have analyzed mathematical models and the LHS (Latin Hypercube Sampling) methods to enhance the reliability of radiochronometry which is to develop an uncertainty algorithm for nuclear material radiochronometry using Bateman Equation. We analyzed the LHS method, which can obtain effective statistical results with a small number of samples, and applied it to algorithms that are Monte Carlo methods for uncertainty calculation by computer simulation. This was implemented through the MATLAB computational software. The uncertainty calculation model using mathematical models demonstrated characteristics based on the relationship between sensitivity coefficients and radiative equilibrium. Computational simulation random sampling showed characteristics dependent on random sampling methods, sampling iteration counts, and the probability distribution of uncertainty factors. For validation, we compared models from various international organizations, mathematical models, and the Monte Carlo method. The developed algorithm was found to perform calculations at an equivalent level of accuracy compared to overseas institutions and mathematical model-based methods. To enhance usability, future research and comparisons·validations need to incorporate more complex decay chains and non-homogeneous conditions. The results of this study can serve as foundational technology in the nuclear forensics field, providing tools for the identification of signature nuclides and aiding in the research, development, comparison, and validation of related technologies.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1711-1716
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• 2017

Uranium tetrafluoride ($UF_4$) is the most used nuclear material for producing metallic uranium by reduction with Ca or Mg. Metallic uranium is a raw material for the manufacture of uranium silicide, $U_3Si_2$, which is the most suitable uranium compound for use as nuclear fuel for research reactors. By contrast, ammonium uranyl carbonate is a traditional uranium compound used for manufacturing uranium dioxide $UO_2$ fuel for nuclear power reactors or $U_3O_8-Al$ dispersion fuel for nuclear research reactors. This work describes a procedure for recovering uranium and ammonium fluoride ($NH_4F$) from a liquid residue generated during the production routine of ammonium uranyl carbonate, ending with $UF_4$ as a final product. The residue, consisting of a solution containing high concentrations of ammonium ($NH_4^+$), fluoride ($F^-$), and carbonate ($CO_3^{2-}$), has significant concentrations of uranium as $UO_2^{2+}$. From this residue, the proposed procedure consists of precipitating ammonium peroxide fluorouranate (APOFU) and $NH_4F$, while recovering the major part of uranium. Further, the remaining solution is concentrated by heating, and ammonium bifluoride ($NH_4HF_2$) is precipitated. As a final step, $NH_4HF_2$ is added to $UO_2$, inducing fluoridation and decomposition, resulting in $UF_4$ with adequate properties for metallic uranium manufacture.

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

• Journal of Radiation Protection and Research
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• v.34 no.3
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• pp.144-150
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• 2009

This study examines how much the radiation dose rate around it varies if a crack occurs on the spent nuclear fuel rod. The spent nuclear fuel rod to be examined is that of Kori unit 3&4. The source terms are evaluated using the ORIGEN-ARP that is part of the version 5.1 of the SCALE package. The radiation dose rate is assessed using the TORT. To check if the structure of a fuel rod is appropriately modeled in the TORT calculation, the calculation results by the TORT are compared with those by the ANISN for the same case. From the code simulation, it is known that if a crack occurs on the spent nuclear fuel rod, the neutron dose rate varies depending on what material is the crack filled with, but the gamma dose rate varies irrespective of type of the material that the crack is filled with.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1160-1167
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• 2018

Detectability of the loose parts (LPs) in steam generator (SG) was studied with eddy current testing technique such as X-probe, bobbin and rotating coils ($MRPC^{(R)}$) as a function of LP size and spacing between LP and tube or between LP and support structures. SG mockup simulating SG tube and support structures with LP was fabricated. The X-probe showed slightly better detectability than $MRPC^{(R)}$ for LP of ferrous (F-LP) material and vice versa for LP of nonferrous (NF-LP) material. In terms of feasibility, inspection rate and other predictable features of the SG tubing inspections, X-probe can be used reliably for monitoring the LPs and the flaws formed by LPs on SG tubes.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2489-2497
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• 2023

Co and Eu impurities in the SSCs are nuclides that dominantly influence the neutron-induced radioactive inventory in metal and concrete radwastes (radioactive wastes) during NPP decommission. The impurity concentrations provided by NUREG/CR-3474 were used for the practical range of Co and Eu impurity concentrations to be applied to the code calculations. Metal structures near the core were evaluated to be ILW (intermediate-level waste) for the whole range of Co impurity concentration, so the boundary line between ILW and LLW (low-level waste) has no change for the whole concentration range provided by NUREG/CR-3474. Also, the boundary line between VLLW (very low-level waste) and CW (clearance waste) in the concrete shield could alter a little depending on the Eu impurity concentration within the range provided by NUREG/CR-3474. From this work, it is found that the concentration of material impurities of SSCs gives no critical impact on determining radwaste levels.