• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.34-39
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• 1996

A consistent general order nodal method for solving the 3-D neutron diffusion equation in (x-y-z) geometry has ben derived by using a weighted integral technique and expanding the spatial variables by the Legendre orthogonal series function. The equation set derived can be converted into any order nodal schemes. It forms a compact system for general order of nodal moments. The method utilizes the analytic solutions of the transverse-integrated quasi -one dimensional equations and a consistent expansion for the spatial variables so that it renders the use of an approximation for the transverse leakages no necessary. Thus, we can expect extremely accurate solutions and the solution would converge exactly when the mesh width is decreased or the approximation order is increased since the equation set is consistent mathematically.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4238-4245
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• 2023

In order to predict the lifetime of InP Heterojunction Bipolar Transistor (HBT) devices and related circuits in the space radiation environment, a novel model including gamma radiation effects is proposed in this paper. Based on the analysis of radiation-induced device degradation effects including both DC and AC characteristics, a set of empirical expressions describing the device degradation trend are presented and incorporated into the Keysight model. To validate the effective of the proposed model, a series of radiation experiments are performed. The correctness of the novel model is validated by comparing experimental and simulated results before and after radiation.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.678-699
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• 2015

Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for nearterm human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of $100MW_{th}$ and an electricity generation mode of $100MW_{th}$, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and thermohydraulics was carried out. The result indicates that the innovative design has great potential for high propellant efficiency and thrust-to-weight of engine ratio, compared with the existing NTR designs. However, the build-up of fission products in fuel has a significant impact on the bimodal operation of the moderated reactor such as xenon-induced dead time. This issue can be overcome by building in excess reactivity and control margin for the reactor design.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.13-30
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• 2019

This paper presents the recent improvements in the DeCART code for HTGR analysis. A new 190-group DeCART cross-section library based on ENDF/B-VII.0 was generated using the KAERI library processing system for HTGR. Two methods for the eigen-mode adjoint flux calculation were implemented. An azimuthal angle discretization method based on the Gaussian quadrature was implemented to reduce the error from the azimuthal angle discretization. A two-level parallelization using MPI and OpenMP was adopted for massive parallel computations. A quadratic depletion solver was implemented to reduce the error involved in the Gd depletion. A module to generate equivalent group constants was implemented for the nodal codes. The capabilities of the DeCART code were improved for geometry handling including an approximate treatment of a cylindrical outer boundary, an explicit border model, the R-G-B checker-board model, and a super-cell model for a hexagonal geometry. The newly improved and implemented functionalities were verified against various numerical benchmarks such as OECD/MHTGR-350 benchmark phase III problems, two-dimensional high temperature gas cooled reactor benchmark problems derived from the MHTGR-350 reference design, and numerical benchmark problems based on the compact nuclear power source experiment by comparing the DeCART solutions with the Monte-Carlo reference solutions obtained using the McCARD code.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.262-268
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• 2022

This paper introduces a novel concept of the pulsed neutron facility (PNF) for maximizing the production of the thermal neutrons and its application to medical use based on prompt gamma neutron activation analysis (PGNAA) using Monte Carlo simulations. The PNF consists of a compact D-T neutron generator, a graphite pile, and a detection system using Cadmium telluride (CdTe) detector arrays. The configuration of fuel pins in the graphite monolith and the design and materials for the moderating layer were studied to optimize the thermal neutron yields. Biological samples - normal and cancerous breast tissues - including chlorine, a trace element, were used to investigate the sensitivity of the characteristic γ-rays by neutron-trace material interactions and the detector responses of multiple particles. Around 90 % of neutrons emitted from a deuterium-tritium (D-T) neutron generator thermalized as they passed through the graphite stockpile. The thermal neutrons captured the chlorines in the samples, then the characteristic γ-rays with specific energy levels of 6.12, 7.80 and 8.58 MeV were emitted. Since the concentration of chlorine in the cancerous tissue is twice that in the normal tissue, the count ratio of the characteristic g-rays of the cancerous tissue over the normal tissue is approximately 2.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.922-932
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• 2024

This paper presents the design features and preliminary design analysis results of the Wielenga Innovation Static Salt Reactor (WISSR). The WISSR incorporates features that make it both flexible and inherently safe. It is based on innovative technology that controls a nuclear reactor by moving molten salt fuel into or out of the core. The reactor is a low-pressure, fast spectrum transuranic (TRU) burner reactor. Inherent shutdown is achieved by a large negative reactivity feedback of the liquid fuel and by the expansion of fuel out of the core. The core is made of concentric, thin annular fuel chambers containing molten fuel salt. A molten salt coolant passes between the concentric fuel chambers to cool the core. The core has both fixed and variable volume fuel chambers. Pressure, applied by helium gas to fuel reservoirs below the core, pushes fuel out of a reservoir and up into a set of variable volume chambers. A control system monitors the density and temperature of the fuel throughout the core. Using NaCl-(TRU,U)Cl₃ fuel and NaCl-KCl-MgCl₂ coolant, a road-transportable compact WISSR core design was developed at a power level of 1250 MWt. Preliminary neutronics and thermal-hydraulics analyses demonstrate the technical feasibility of WISSR.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.699-714
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• 2021

Supercritical CO₂ (S-CO₂) Brayton cycle has been applied to various heat sources in recent decades, owing to the characteristics of compact structure and high efficiency. Understanding the research development in this emerging research field is crucial for future study. Thus, a bibliometric approach is employed to analyze the scientific publications of S-CO₂ cycle field from 2000 to 2019. In Scopus database, there were totally 724 publications from 1378 authors and 543 institutes, which were distributed over 55 countries. Based on the software-BibExcel, these publications were analyzed from various aspects, such as major research areas, affiliations and keyword occurrence frequency. Furthermore, parameters such as citations, hot articles were also employed to evaluate the research output of productive countries, institutes and authors. The analysis showed that each paper has been cited 13.39 times averagely. United States was identified as the leading country in S-CO₂ research followed by China and South Korea. Based on the contents of publications, existing researches on S-CO₂ are briefly reviewed from the five aspects, namely application, cycle configurations and modeling, CO₂-based mixtures, system components, and experiments. Future development is suggested to accelerate the commercialization of S-CO₂ power system.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1626-1633
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• 2021

In an industrial field, non-destructive testing (NDT) is commonly used to inspect industrial products. Among NDT methods using radiation sources, X-ray laminography has several advantages, such as high depth resolution and low computational costs. Moreover, an X-ray laminography system with stationary source array and compact detector is able to reduce mechanical motion artifacts and improve inspection efficiency. However, this system, called stationary inverse-geometry X-ray laminography (s-IGXL), causes truncation artifacts in reconstructed images due to limited fields-of-view (FOVs). In this study, we proposed a projection data correction (PDC) method to reduce the truncation artifacts arisen in s-IGXL images, and the performance of the proposed method was evaluated with the different number of focal spots in terms of quantitative accuracy. Comparing with conventional techniques, the PDC method showed superior performance in reducing truncation artifacts and improved the quantitative accuracy of s-IGXL images for all the number of focal spots. In conclusion, the PDC method can improve the accuracy of s-IGXL images and allow precise NDT measurements.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.164-169
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• 2020

Use of SiPM has been considered as an alternative to PMT, because of its compact size, low-operating voltage, non-sensitive to electromagnetic, low costs and so on. The main limitation for the use of SiPM is due to its small sensitive area compared to PMT that limits the light collection, and therefore the sensor energy resolution. In this article we studied the effect of increasing the number of SiPM by connecting them in parallel to increase the active detection area. This allowed us to compare the different energy resolution measurements. ¹³⁷Cs has been selected as reference to study the energy resolution for 662 keV gamma-rays. Another investigation was to compare the minimum detectable gamma energy under various SiPM configurations. It has been found that the use of 4 SiPM arrays can greatly improve the energy resolution up to 4% than only one SiPM array, meanwhile use of more than 2 SiPM arrays does not increase the energy resolution significantly. Thus we can conclude that for a large area of cylindrical scintillator (3 × 3 inches), the use of SiPMs are limited to a certain number or certai active area depending on the commercial SiPMs, and its cost should be less than traditional PMT for the cost-effective and compact size considerations. It is well known that the gain of SiPM varies with temperature. In this article, we also calibrated gain to guarantee the same position of photoelectric peak in response of different temperatures.

• Nuclear Engineering and Technology
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• v.47 no.4
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• pp.479-487
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• 2015

The thallium-doped sodium iodide [NaI(Tl)] scintillation detector is preferred as a gamma spectrometer in many fields because of its general advantages. A silicon photomultiplier (SiPM) has recently been developed and its application area has been expanded as an alternative to photomultiplier tubes (PMTs). It has merits such as a low operating voltage, compact size, cheap production cost, and magnetic resonance compatibility. In this study, an array of SiPMs is used to develop an NaI(Tl) gamma spectrometer. To maintain detection efficiency, a commercial NaI(Tl) $2^{\prime}{\times}2^{\prime}$ scintillator is used, and a light guide is used for the transport and collection of generated photons from the scintillator to the SiPMs without loss. The test light guides were fabricated with polymethyl methacrylate and reflective materials. The gamma spectrometer systems were set up and included light guides. Through a series of measurements, the characteristics of the light guides and the proposed gamma spectrometer were evaluated. Simulation of the light collection was accomplished using the DETECT 97 code (A. Levin, E. Hoskinson, and C. Moison, University of Michigan, USA) to analyze the measurement results. The system, which included SiPMs and the light guide, achieved 14.11% full width at half maximum energy resolution at 662 keV.