• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.1027-1028
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• 2005

• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.49-57
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• 2018

The main objective of this paper is come up with methodology approach for FPGA-based system in verification and validation lifecycle regarding software reliability using system engineering approach. The steps of both reverse engineering and re-engineering are carried out to implement an FPGA-based of safety critical system in Nuclear Power Plant. The reverse engineering methodology is applied to elicit the requirements of the system as well as gain understanding of the current life cycle and V&V activities of FPGA based-system. The re-engineering method is carried out to get a new methodology approach of software reliability, particularly Software Reliability Growth Model. For measure the software reliability of a given FPGA-based system, the following steps are executed as; requirements definition and measurement, evaluation of candidate reliability model, and the validation of the selected system. As conclusion, a new methodology approach for software reliability measurement using software reliability growth model is developed.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1259-1265
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• 2021

In the present study, a large-area hybrid gamma imaging system was designed by adopting coded aperture imaging on the basis of a large-area Compton camera to achieve high imaging performance throughout a broad energy range (100-2000 keV). The system consisting of a tungsten coded aperture mask and monolithic NaI(Tl) scintillation detectors was designed through a series of Geant4 Monte Carlo radiation transport simulations, in consideration of both imaging sensitivity and imaging resolution. Then, the performance of the system was predicted by Geant4 Monte Carlo simulations for point sources under various conditions. Our simulation results show that the system provides very high imaging sensitivity (i.e., low values for minimum detectable activity, MDA), thus allowing for imaging of low-activity sources at distances impossible with coded aperture imaging or Compton imaging alone. In addition, the imaging resolution of the system was found to be high (i.e., around 6°) over the broad energy range of 59.5-1330 keV.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.805-806
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• 2005

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1431-1441
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• 2017

The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP) design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower), which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3591-3598
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• 2023

Photoneutron background spectroscopy observations at linac are essential for directing accelerator shielding and subtracting background signals. Therefore, we constructed a Bonner Sphere Spectrometer (BSS) system based on an array of BF₃ gas proportional counter tubes. Initially, the response of the BSS system was simulated using the MCNP5 code. Next, the response of the system was calibrated by using neutrons with energies of 2.86 MeV and 14.84 MeV. Then, the system was employed to measure the spectrum of the ²⁴¹Am-Be neutron source, and the results were unfolded by using the Gravel and EM algorithms. Using the validated system, the undesirable neutron spectrum of the 120 MeV electron linac was finally measured and acquired. In addition, it is demonstrated that the equivalent undesirable neutron dose at a distance of 3.2 m from the linac is 19.7 mSv/h. The results measured by the above methods could provide guidance for linac-related research.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.304-317
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• 2016

An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan), a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcrofte-Walton type accelerator, which generates the external neutron source by deuteriu-metritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.997-1005
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• 2018

The regulatory periodic safety inspection system is one of the most important methods for confirming the safety of nuclear power plants and the defense in depth in nuclear safety is the most important basic means for accident prevention and mitigation. Recently, a new regulatory technology based on risk-informed and safety performance has been developed and used in advanced countries. However, since the domestic periodic inspection system is being used in the same way over 30 years, it is necessary to know how the inspection contributes to the safety confirmation of the nuclear power plants. In this study, the domestic periodic inspection system currently in use was analyzed from the perspective of defense in depth in nuclear safety. In addition, the analysis results were compared to the U.S. NRC's safety inspection system to obtain consistency and lessons in this study. As a result of analysis, the NRC's safety inspections were distributed almost evenly at the all levels of defense in depth, while in the case of domestic inspection, they were heavily focused on the level 1 of defense in depth. Therefore, it appeared urgent to improve the inspection system to strengthen the other levels of defense in depth in nuclear safety.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.601-607
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• 2022

In recent years, monitoring of spent nuclear fuel inside dry cask storage has become an important area of national security. Muon tomography is a useful method for monitoring spent nuclear fuel because it uses high energy muons that penetrate deep into the target material and provides a 3-D structure of the inner materials. We designed a muon tomography system consisting of four 2-D position sensitive detector and characterized and optimized the system parameters. Each detector, measuring 200 × 200 cm², consists of a plastic scintillator, wavelength shifting (WLS) fibers and, SiPMs. The reconstructed image is obtained by extracting the intersection of the incoming and outgoing muon tracks using a Point-of-Closest-Approach (PoCA) algorithm. The Geant4 simulation was used to evaluate the performance of the muon tomography system and to optimize the design parameters including the pixel size of the muon detector, the field of view (FOV), and the distance between detectors. Based on the optimized design parameters, the spent fuel assemblies were modeled and the line profile was analyzed to conduct a feasibility study. Line profile analysis confirmed that muon tomography system can monitor nuclear spent fuel in dry storage container.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.19-29
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• 2022

The 2D/1D method has become the mainstream of the direct transport calculation considering the balance of accuracy and efficiency. However, the 2D/1D method still suffers from stability issues. Recently, a quasi-3D method has been proposed with axial Legendre expansion. Analysis and comparison of the 2D/1D and quasi-3D method is conducted in theory from the equation derivation. Besides, the C5G7 benchmark, the KUCA benchmark and the macro BEAVRS benchmark are calculated to verify the theory comparisons of these two methods with the direct transport code SHARK. All results show that the quasi-3D method has better stability and accuracy than the 2D/1D method with worse efficiency and memory cost. It provides a new option for direct transport calculation with the quasi-3D method.