• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1705-1716
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• 2021

Bangladesh, as a newcomer country, is expecting to start her nuclear power journey by 2022. Due to evident reasons, newcomer nuclear countries face several key challenges concerning the development of national nuclear safety infrastructure. The paper investigates the status of the 7 key safety infrastructure issues out of the 19 and readiness of the supportive organizations, laboratories, and workforces following the International Atomic energy Agency's status evaluation guide at milestone 3 and foreign countries' practice. Much progress has been achieved at phase 3 regarding the establishments of a few Acts, a regulator, and an operator. However, comprehensive regulatory frameworks, skilled workforces, establishments of a few supportive organizations, and laboratories for managing environmental radioactivity, radiological accidents, and radioactive wastes are yet to ready. Several suggestions are made for establishing and expediting radiation monitoring laboratories, a radiological emergency management center, a radioactive waste management company, and technical support organizations for the safety infrastructure. To avoid perceived risks, policymakers and competent authorities need to emphasize creating an optimized safety infrastructure before commissioning and operating the 1^st nuclear power plant safely, securely, and cost-sustainably.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1097-1104
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• 2023

A preliminary criticality analysis for novel pyrochemical apparatuses for the reprocessing of mixed uranium-plutonium nitride spent nuclear fuel from the BREST-OD-300 reactor was performed. High-temperature processing apparatuses, "metallization" electrolyzer, refinery remelting apparatus, refining electrolyzer, and "soft" chlorination apparatus are considered in this work. Computational models of apparatuses for two neutron radiation transport codes (MCU-FR and MCNP) were developed and calculations for criticality were completed using the Monte Carlo method. The criticality analysis was performed for different loads of fissile material into the apparatuses including overloading conditions. Various emergency situations were considered, in particular, those associated with water ingress into the chamber of the refinery remelting apparatus. It was revealed that for all the considered computational models nuclear safety rules are satisfied.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2723-2733
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• 2023

Transuranic nuclides (such as ²³⁸Pu, ²⁵²Cf, ²⁴⁹Bk, etc.) have a wide range of application in industry, medicine, agriculture, and other fields. However, due to the complex conversion chain and remarkable fission losses in the process of transuranic nuclides production, the generation amounts are extremely low. High flux reactor with high neutron flux and flexible irradiation channels, is regarded as the promising candidate for producing transuranic nuclides. It is of great significance to increase the conversion ratio of transuranic nuclides, resulting in higher efficiency and better economy. In this paper, we perform an optimization design evaluation of producing transuranic nuclides in high flux reactor, which includes optimization design of irradiation target and influence study of reactor core loading. It is demonstrated that the production rate increases with appropriately determined target material and target structure. The target loading scheme in the irradiation channel also has a significant influence on the production of transuranic nuclides.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.29 no.10
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• pp.538-550
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• 2017

In a PWR (Pressurized Water Reactor), the appropriate heat removal from the surface of fuel rod bundle is important for ensuring thermal margins and safety. Although many CFD (Computational Fluid Dynamics) software have been used to predict complex flows inside fuel assemblies with mixing vanes, there is no domestic regulatory guideline for the comprehensive evaluation of CFD software. Therefore, from the nuclear regulatory perspective, it is necessary to perform the systematic assessment and prepare the domestic regulatory guideline for checking whether valid CFD software is used for nuclear safety problems. In this study, to provide systematic evaluation and guidance on the applicability of CFD software to the domestic nuclear safety area, the results of the sensitivity analysis for the effect of the discretization scheme accuracy for the convection terms and turbulence models, which are main factors that contribute to the uncertainty in the calculation of the nuclear safety problems, on the prediction performance for the turbulent flow distribution inside the fuel assembly with split-type mixing vanes were explained.

• Journal of the Korean Society of Safety
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• v.30 no.4
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• pp.174-180
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• 2015

This study developed a Safety Culture Impact Assessment Model (SCIAM) which consists of a safety culture assessment methodology and a safety culture impact quantification methodology. The SCIAM uses safety culture impact index (SCII) to monitor the status of safety culture of the NPPs periodically and it uses relative core damage frequency (RCDF) to present the impact of safety culture on the safety of the NPPs. As a result of applying SCIAM to the reference plant (Kori 3), the standard for the healthy safety culture of the reference plant is suggested. SCIAM might contribute to improve the safety of the NPPs (Nuclear Power Plants) by monitoring the status of safety culture periodically and presenting the standard of healthy safety culture.

• Proceedings of the Korea Institute of Fire Science and Engineering Conference
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• 2008.11a
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• pp.444-449
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• 2008

Fire model shall be verified and validated to reliably predict the consequences of fires within its limitations. This study aims to predict pool fire phenomena with two rooms using CFAST and to compare CFAST simulation results with PRISME experimental data which can be applicable to the fire of nuclear power plant facility. Five different mass loss rate(MLR) are used in the simulation and the simulated results of specific quantities such as temperature, chemical composition are compared to the experimental data. From this study, the CFAST simulation results with the proper MLR show better similarity and trend with pool fire experimental data.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.489-490
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.365-366
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.275-276
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1333-1334
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• 2004