• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.1
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• pp.43-63
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• 2022

The aim of this review is to communicate some essential knowledge of the underlying mechanism of the corrosion of structural containment alloys during molten salt reactor operation in the context of prospective online monitoring in future MSR installations. The formation of metal halide species and the progression of their concentration in the molten salt do reflect containment corrosion, tracing the depletion of alloying metals at the alloy salt interface will assure safe conditions during reactor operation. Even though the progress of alloying metal halides concentrations in the molten salt do strongly understate actual corrosion rates, their prospective 1^st order kinetics followed by near-linearly increase is attributed to homogeneous matrix corrosion. The service life of the structural containment alloy is derived from homogeneous matrix corrosion and near-surface void formation but less so from intergranular cracking (IGC) and pitting corrosion. Online monitoring of corrosion species is of particular interest for molten chloride systems since besides the expected formation of chromium chloride species CrCl₂ and CrCl₃, other metal chloride species such as FeCl₂, FeCl₃, MoCl₂, MnCl₂ and NiCl₂ will form, depending on the selected structural alloy. The metal chloride concentrations should follow, after an incubation period of about 10,000 hours, a linear projection with a positive slope and a steady increase of < 1 ppm per day. During the incubation period, metal concentration show 1^st order kinetics and increasing linearly with time^1/2. Ideally, a linear increase reflects homogeneous matrix corrosion, while a sharp increase in the metal chloride concentration could set a warning flag for potential material failure within the projected service life, e.g. as result of intergranular cracking or pitting corrosion. Continuous monitoring of metal chloride concentrations can therefore provide direct information about the mechanism of the ongoing corrosion scenario and offer valuable information for a timely warning of prospective material failure.

• Journal of Hydrogen and New Energy
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• v.33 no.5
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• pp.534-540
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• 2022

Liquid hydrogen has the best storage capacity per unit mass and is economical among storage methods for using hydrogen as fuel. As the demand for hydrogen increases, the need to develop a storage and supply system of liquid hydrogen is emphasizing. In order to liquefy hydrogen, it is necessary to pre-cool it to a maximum inversion temperature of -253℃. The Gifford-McMahon (GM) refrigerator is the most reliable and commercialized refrigerator among small-capacity cryogenic refrigerators, which can extract high-efficiency hydrogen through liquefied hydrogen production and boil of gas re-liquefaction. Therefore, in this study, the optimal conditions for liquefying gas hydrogen were sought using the GM cryocooler. The process was simulated by PRO/II under various cooling capacities of the GM refrigerator. In addition, the flow rate of hydrogen was calculated by comparing with specific refrigerator capacity depending on the pressure and flow rate of a refrigerant medium, helium. Simulations were performed to investigate the optimal values of the liquefaction flow rate and compression pressure, which aim for the peak refrigeration effect. Based on this, a liquefaction system can be selected in consideration of the cycle configuration and the performance of the refrigerator.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1676-1685
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• 2021

We develop a new high-fidelity multiphysics model to simulate boron chemistry in the porous Chalk River Unidentified Deposit (CRUD) deposits. Heat transfer, capillary flow, solute transport, and chemical reactions are fully coupled. The evaporation of coolant in the deposits is included in governing equations modified by the volume-averaged assumption of wick boiling. The axial offset anomaly (AOA) of the Seabrook nuclear power plant is simulated. The new model reasonably predicts the distributions of temperature, pressure, velocity, volumetric boiling heat density, and chemical concentrations. In the thicker CRUD regions, 60% of the total heat is removed by evaporative heat transfer, causing boron species accumulation. The new model successfully shows the quantitative effect of coolant evaporation on the local distributions of boron. The total amount of boron in the CRUD layer increases by a factor of 1.21 when an evaporation-driven increase of soluble and precipitated boron concentrations is reflected. In addition, the concentrations of B(OH)₃ and LiBO₂ are estimated according to various conditions such as different CRUD thickness and porosity. At the end of the cycle in the AOA case, the total mass of boron incorporated in CRUD deposits of a reference single fuel rod is estimated to be about 0.5 mg.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.307-319
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• 2022

60+ Years of nuclear power generation has led to a significant legacy of radioactively contaminated land at a number of nuclear licenced "mega sites" around the world. The safe management and remediation of these sites is key to ensuring there environmental stewardship in the long term. Bioremediation utilizes a variety of microbially mediated processes such as, enzymatically driven metal reduction or biominerialisation, to sequester radioactive contaminants from the subsurface limiting their migration through the geosphere. Additionally, some of these process can provide environmentally stable sinks for radioactive contaminants, through formation of highly insoluble mineral phases such as calcium phosphates and carbonates, which can incorporate a range of radionuclides into their structure. Bioremediation options have been considered and deployed in preference to conventional remediation techniques at a number of nuclear "mega" sites. Here, we review the applications of bioremediation technologies at three key nuclear licenced sites; Rifle and Hanford, USA and Sellafield, UK, in the remediation of radioactively contaminated land.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.4
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• pp.489-500
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• 2022

The decommissioning of nuclear facilities produces various types of radiologically contaminated waste. In addition, dismantlement activities, including cutting, packing, and clean-up at the facility site, result in secondary radioactive waste such as filters, resin, plastic, and clothing. Determining of the radionuclide content of this waste is an important step for the determination of a suitable management strategy including classification and disposal. In this work, we radiochemically characterized the radionuclide activities of filters used during the decommissioning of Korea Research Reactors (KRRs) 1 and 2. The results indicate that the filter samples contained mainly ³H (500-3,600 Bq·g^-1), ¹⁴C (7.5-29 Bq·g^-1), ⁵⁵Fe (1.1- 7.1 Bq·g^-1), ⁵⁹Ni (0.60-1.0 Bq·g^-1), ⁶⁰Co (0.74-70 Bq·g^-1), ⁶³Ni (0.60-94 Bq·g^-1), ⁹⁰Sr (0.25-5.0 Bq·g^-1), ¹³⁷Cs (0.64-8.7 Bq·g^-1), and ¹⁵²Eu (0.19-2.9) Bq·g^-1. In addition, the gross alpha radioactivity of the samples was measured to be between 0.32-1.1 Bq·g^-1. The radionuclide concentrations were below the concentration limit stated in the low- and intermediatelevel waste acceptance criteria of the Nuclear Safety and Security Commission, and used for the disposal of the KRRs waste drums to a repository site.

• Journal of Hydrogen and New Energy
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• v.33 no.6
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• pp.845-855
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• 2022

In response to climate change, the world is continuing efforts to reduce fossil fuels, expand renewable energy, and improve energy efficiency with the goal of achieving carbon neutrality. In particular, R&D is being made on the value chain covering the entire cycle of hydrogen production, storage, transportation, and utilization in order to shift the energy supply system to focus on hydrogen energy. Hydrogen-based energy sources can produce heat and electricity at the same time, so it is possible to utilize heat energy, which can increase overall efficiency. In this study, calculation of the optimal scale for hydrogen-based cogeneration and the composition of heat sources were reviewed. It refers to a method of the optimal heat source size according to the external heat supply and heat storage to be considered. The results of this study can be used as basic data for establishing a hydrogen-based energy supply model in the future.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2230-2245
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• 2023

KANT (KAIST Advanced Nuclear Tachygraphy) is a PWR core simulator recently developed at Korea Advance Institute of Science and Technology, which solves three-dimensional steady-state and transient multigroup neutron diffusion equations under Cartesian geometries alongside the incorporation of thermal-hydraulics feedback effect for multi-physics calculation. It utilizes the standard Nodal Expansion Method (NEM) accelerated with various Coarse Mesh Finite Difference (CMFD) methods for neutronics calculation. For thermal-hydraulics (TH) calculation, a single-phase flow model and a one-dimensional cylindrical fuel rod heat conduction model are employed. The time-dependent neutronics and TH calculations are numerically solved through an implicit Euler scheme, where a detailed coupling strategy is presented in this paper alongside a description of nodal equivalence, macroscopic depletion, and pin power reconstruction. For validation of the steady, transient, and depletion calculation with pin power reconstruction capacity of KANT, solutions for various benchmark problems are presented. The IAEA 3-D PWR and 4-group KOEBERG problems were considered for the steady-state reactor benchmark problem. For transient calculations, LMW (Lagenbuch, Maurer and Werner) LWR and NEACRP 3-D PWR benchmarks were solved, where the latter problem includes thermal-hydraulics feedback. For macroscopic depletion with pin power reconstruction, a small PWR problem modified with KAIST benchmark model was solved. For validation of the multi-physics analysis capability of KANT concerning large-sized PWRs, the BEAVRS Cycle1 benchmark has been considered. It was found that KANT solutions are accurate and consistent compared to other published works.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.371-381
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• 2023

Evaluating the effectiveness of the radiation protection measures deployed at the Centralized Radioactive Waste Management Facility in Ghana is pivotal to guaranteeing the safety of personnel, public and the environment, thus the need for this study. Radiagem^TM 2000 was used in measuring the dose rate of the facility whilst the personal radiation exposure of the personnel from 2011 to 2022 was measured from the thermoluminescent dosimeter badges using Harshaw 6600 Plus Automated TLD Reader. The decay store containing scrap metals from dismantled disused sealed radioactive sources (DSRS), and low-level wastes measured the highest dose rate of 1.06 ± 0.92 µSv·h^-1. The range of the mean annual average personnel dose equivalent is 0.41-2.07 mSv. The annual effective doses are below the ICRP limit of 20 mSv. From the multivariate principal component analysis biplot, all the personal dose equivalent formed a cluster, and the cluster is mostly influenced by the radiological data from the outer wall surface of the facility where no DSRS are stored. The personal dose equivalents are not primarily due to the radiation exposures of staff during operations with DSRS at the facility but can be attributed to environmental radiation, thus the current radiation protection measures at the Facility can be deemed as effective.

• Journal of Ceramic Processing Research
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• v.21 no.1
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• pp.57-63
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• 2020

Chemical looping hydrogen production (CLHP) is an attractive technology for H₂ production due to its ability to produce H₂ and capture CO₂ from fossil fuels simultaneously. In this paper, we present MgFe₂O₄ as an oxygen carrier material with high efficiency, low cost, and stable properties for CLHP. The redox reactions occurred reversibly in the fuel, steam, and air reactor as MgFe₂O₄→MgO/Fe, MgO/Fe→MgO/Fe₃O₄, and MgO/Fe₃O₄→MgFe₂O₄, respectively. The oxygen transfer capacities of MgFe₂O₄ for 5% H₂/N₂ and 5% CO/N₂ gases were about 23 wt% at 900 ℃. Both the oxygen transfer capacity and rate were well maintained during 10 redox cycles at 900 ℃. No phase changes or agglomeration occurred as the redox cycle number increased. Similarly, MgFe₂O₄ did not exhibit significant degradation in its total amount or maximum rate of H₂ production during four redox cycles. The average calculated amount of H₂ production for MgFe₂O₄ was 2,806 L/day per unit mass (kg).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.22 no.2
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• pp.227-235
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• 2024

After the permanent shut down of Kori Unit 1, various decommissioning activities will be implemented, including decontamination, segmentation, waste management, and site restoration. During the decommissioning period, waste management is among the most important activities to ensure that the process proceeds smoothly and within the expected timeframe. Furthermore, the radioactive waste generated during the operation should be sent to a disposal facility to complete the decommissioning project. Square and cylindrical concrete re-package drums were generated during the 1980s and 1990s. The square, containing boron concentrates, and cylindrical, containing spent resin, concrete re-package drums have been stored in a radioactive waste storage building. Homogeneous radioactive waste, including boron concentrates, spent resin, and sludge, should be solidified or packaged in high-integrity containers (HICs). This study investigates the sequential segmentation process for the separation of contaminated and non-contaminated regions, the re-packaging process of segmented or crushed cement-solidified boron concentrate, and re-packaging in HICs. The conceptual design evaluates the re-packaging plan for the segmented and crushed cement-solidified waste using HICs, which is acceptable in a disposal facility, and the quantity of generated HICs from the treatment process.