• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1964-1969
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• 2021

We developed a Dual-PPACs detector for fast neutron measurements that consists of two sets of PPAC: conventional PPAC and fission PPAC. A²³⁸U(U₃O₈) coating is placed in the fission PPAC's anode, which is used as the neutrons conversion layer. An experiment was performed to measure neutron time-of-flight (TOF) in which ²⁵²Cf spontaneous fission source was used. An excellent time resolution of 164ps has been observed at 6 mbar in isobutene gas. With the excellent time resolution of Dual-PPACs detector, exact neutron energy can be extracted from the timing measurement. The experimental detection efficiency was 1.9 × 10^-7, consistent with the efficiency of 2.5 × 10^-7 given by a Geant4 simulation. Ultimately, the results show that the Dual-PPACs detector is a suitable candidate for measuring fast neutrons in the future CiADS system.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1099-1109
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• 2020

The performance of gadolinium burnable absorber (GdBA) for reactivity control in UO₂ and (U, Th)O₂ fuels and its impact on spent fuel characteristics was performed. Five fuel assemblies: one without GdBA fuel rod and four each containing 16, 24, 34 and 44 GdBA fuel rods in both fuels were investigated. Reactivity swing in all the FAs with GdBA rods in UO₂ fuel was higher than their counterparts with similar GdBA fuel rods in (U, Th)O₂ fuel. The excess reactivity in all FAs with (U, Th)O₂ fuel was higher than UO₂ fuel. At the end of single discharge burn-up (~ 49.64 GWd/tHM), the excess reactivity of (U, Th) O₂ fuel remained positive (16,000 pcm) while UO₂ fuel shows a negative value (-6,000 pcm), which suggest a longer discharge burn-up in (U, Th)O₂ fuel. The concentration of plutonium isotopes and minor actinides were significantly higher in UO₂ fuel than in (U, Th)O₂ fuel except for ²³⁶Np. However, the concentration of non-actinides (gadolinium and iodine isotopes) except for ¹³⁵Xe were respectively smaller in (U, Th)O₂ fuel than in UO₂ fuel but may be two times higher in (U, Th)O₂ fuel due to its potential longer discharge burn-up.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.709-714
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• 2009

An ion implanter, which can serve as a metal-ion supply, has been constructed and performance tested. Copper ions are generated and extracted from a Bernas ion source with a heating crucible that provides feed gases to sustain the plasma. Sable arc plasmas can be sustained in the ion source for a crucible temperature in excess of $350^{\circ}C$. Stable extraction of the ions is possible for arc Currents less than 0.3 A. Arc currents increase with the induced power of a block cathode and the transverse field in the ion source. $Cu^+$ ions in the extracted beam are separated using a dipole magnet. A $20{\mu}A$ $Cu^+$ ion current can be extracted with a 0.2 A arc current. The ion current can support a dose of $10^{16}ions/cm^2$ over an area of $15\;cm^2$ within a few hours.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1736-1746
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• 2021

The results of full-scale numerical experiments of a hybrid thorium-containing fuel cell facility operating in a close-to-critical state due to a controlled source of fusion neutrons are discussed in this work. The facility under study was a complex consisting of two blocks. The first block was based on the concept of a high-temperature gas-cooled thorium reactor core. The second block was an axially symmetrical extended plasma generator of additional neutrons that was placed in the near-axial zone of the facility blanket. The calculated models of the blanket and the plasma generator of D-T neutrons created within the work allowed for research of the neutronic parameters of the facility in stationary and pulse-periodic operation modes. This research will make it possible to construct a safe facility and investigate the properties of thorium fuel, which can be continuously used in the epithermal spectrum of the considered hybrid fusion-fission reactor.

• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.1-15
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• 2020

Nuclear power plants in South Korea are operated to cover the baseload demand. Hence they are operated at 100% rated power and do not deploy power tracking control except for startup, shutdown, or during transients. However, as the contribution of renewable energy in the energy mix increases, load follow operation may be needed to cover the imbalance between consumption and production due to the intermittent nature of electricity produced from the conversion of wind or solar energy. Load follow operation may be quite challenging since the operators need to control the axial power distribution and core reactivity while simultaneously conducting the power maneuvering. In this paper, a systems engineering approach for multi-physics load follow simulation of APR1400 is performed. RELAP5/SCDAPSIM/MOD3.4/3DKIN multi-physics package is selected to simulate the Korean Advanced Power Reactor, APR1400, under load follow operation to reflect the impact of feedback signals on the system safety parameters. Furthermore, the systems engineering approach is adopted to identify the requirements, functions, and physical architecture to provide a set of verification and validation activities that guide this project development by linking each requirement to a validation or verification test with predefined success criteria.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3388-3400
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• 2023

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.

• Journal of the Korean Society of Systems Engineering
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• v.19 no.2
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• pp.46-58
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• 2023

Deterministic safety analysis is a crucial part of safety assessment, particularly when it comes to demonstrating the safety of nuclear power plant designs. The traditional approach to deterministic safety analysis models is to model the nuclear core using point kinetics. However, this simplified approach does not fully reflect the real core behavior with proper moderator and fuel reactivity feedbacks during the transient. The use of Multi-Physics approach allows more precise simulation reflecting the inherent three-dimensionality (3D) of the problem by representing the detailed 3D core, with instantaneous updates of feedback mechanisms due to changes of important reactivity parameters like fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC). This paper addresses a CEA ejection accident at hot full power (HFP), in which the underlying strong and un-symmetric feedback between thermal-hydraulics and reactor kinetics exist. For this purpose, a multi-physics analysis tool has been selected with the nodal kinetics code, 3DKIN, implicitly coupled to the thermal-hydraulic code, RELAP5, for real-time communication and data exchange. This coupled approach enables high fidelity three-dimensional simulation and is therefore especially relevant to reactivity initiated accident (RIA) scenarios and power distribution anomalies with strong feedback mechanisms and/or un-symmetrical characteristics as in the CEA ejection accident. The Systems Engineering approach is employed to provide guidance in developing the work in a systematic and efficient fashion.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.11a
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• pp.253-253
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• 2004

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.355-356
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• 2018

• Bulletin of the Korean Space Science Society
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• 2009.04a
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• pp.64.3-64.3
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• 2009