• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.571-576
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• 2023

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2753-2760
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• 2021

This paper proposes using sodium-cooled fast reactor technologies for use in hydrogen vapor methane (SMR) modification. Using three independent energy rings in the Russian BN-600 fast reactor, steam is generated in one of the steam-generating cycles with a pressure of 13.1 MPa and a temperature of 505 ℃. The reactor's second energy cycles can increase the gas-steam mixture's temperature to the required amount for efficient correction. The 620 ton/hr 540 ℃ steam generated in this cycle is sufficient to supply a high-temperature synthesis current source (700 ℃), which raises the steam-gas mixture's temperature in the reactor. The proposed technology provides a high rate of hydrogen production (approximately 144.5 ton/hr of standard H₂), also up to 25% of the original natural gas, in line with existing SMR technology for preparing and heating steam and gas mixtures will be saved. Also, exergy analysis results show that the plant's efficiency reaches 78.5% using HTR heat for combined hydrogen and power generation.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.31-44
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• 2019

This study simulated a control rod assembly (CRA), which is a part of reactor shutdown systems, in immersed and fluid flow conditions. The CRA was inserted into the reactor core within a predetermined time limit under normal and abnormal operating conditions, and the CRA (which consists of complex geometric shapes) drop behavior is numerically modeled for simulation. A full-scale prototype CRA drop test is established under room temperature and water-fluid conditions for verification and validation. This paper describes the details of the numerical modeling and analysis results of the several conditions. Results from the developed numerical simulation code are compared with the test results to verify the numerical model and developed computer code. The developed code is in very good agreement with the test results and this numerical analysis model and method may replace the experimental and CFD method to predict the drop behavior of CRA.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.107-114
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• 2016

Four cylindrically shaped IHXs(Intermediate Heat Exchangers) are installed in the PHTS(Primary Heat Transfer System) of the PGSFR(Prototype Gen IV Sodium cooled Fast Reactor). As for the IHX, the temperature difference of structure is inevitable result caused by heat transfer between primary coolant sodium and IHTS(Intermediate Heat Transport System) sodium. It is necessary to evaluate the high temperature structural integrity of IHXs which operate at the elevated temperature condition over the creep temperature. In this paper, the high temperature structural integrity of IHX under assumed loading conditions has been reviewed according to ASME code.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1025-1044
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• 2009

Systematic research has been conducted by KAERI to develop a supercritical carbon dioxide Brayton cycle energy conversion system coupled with a sodium cooled fast reactor. For the development of the supercritical $CO_2$ Brayton cycle ECS, KAERI researched four major fields, separately. For the system development, computer codes were developed to design and analyze the supercritical $CO_2$ Brayton cycle ECS coupled with the KALIMER-600. Computer codes were developed to design and analyze the performance of the major components such as the turbomachinery and the high compactness PCHE heat exchanger. Three dimensional flow analysis was conducted to evaluate their performance. A new configuration for a PCHE heat exchanger was developed by using flow analysis, which showed a very small pressure loss compared with a previous PCHE while maintaining its heat transfer rate. Transient characteristics for the supercritical $CO_2$ Brayton cycle coupled with KALIMER-600 were also analyzed using the developed computer codes. A Na-$CO_2$ pressure boundary failure accident was analyzed with a computer code that included a developed model for the Na-$CO_2$ chemical reaction phenomena. The MMS-LMR code was developed to analyze the system transient and control logic. On the basis of the code, the system behavior was analyzed when a turbine load was changed. This paper contains the current research overview of the supercritical $CO_2$ Brayton cycle coupled to the KALIMER-600 as an alternative energy conversion system.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.41 no.3
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• pp.233-239
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• 2017

The CA (Control Assembly) of an SFR has a CRA(Control Rod Assembly) with an inner duct and control rod. During an emergency situation, the CRA falls into the duct of the CA for a rapid shut-down. The drop time and impact velocity of the CRA are important parameters with respect to the reactivity insertion time and the structural integrity of the CRA. The objective of this study was to investigate the dynamic behavior and integrity of the CRA owing to a drop impact. The impact analysis of the CRA under normal/abnormal drop conditions was carried out using the commercial FEM code LS-DYNA. Results of the drop impact analysis demonstrated that the CRA maintained structural integrity, and could be safely inserted into the flow hole of the damper under abnormal conditions.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4125-4133
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• 2022

KAERI (Korea Atomic Energy Research Institute) developed the conceptual design of PGSFR (Prototype Gen-IV Sodium Cooled Fast Reactor) and Burner Reactor. Since the reactor characteristics of the PGSFR and Burner Reactor are different, the shape, size and the arrangement of the main components in the reactors must be different. Therefore, the conceptual design for the fuel handling and transfer systems needs to be performed coinciding with the structure of the reactor. Especially, because a redan structure dividing hot and cold pool is installed in the reactor vessel, the conceptual design of the fuel handling and transfer system largely changes depending on the location of the redan structure. Various elements of the conceptual design and an integral arrangement for the fuel handling and transfer system were arranged according to the characteristics, sizes and shapes of the reactors. In this paper, the conceptual designs of the fuel handling and transfer system for PGSFR and Burner Reactor are described. Especially, an A-frame method is selected as the fuel handling and transfer system for the Burner Reactor, considering the layout of the internal structure. The tilt angle, diameter and length of A-frame is determined and the strength evaluation of the A-frame is performed.

• Korean Journal of Metals and Materials
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• v.48 no.5
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• pp.410-416
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• 2010

Studies were carried out to establish the technology for sodium-clad compatibility and to analyze the compatibility behavior of the Sodium-cooled Fast Reactor (SFR) cladding material under a flowing sodium environment. The natural circulation facility caused by the thermal convection of the liquid sodium was constructed and the 316-series stainless steels were exposed at $650{^{\circ}C}$ liquid sodium for 1458 hours. The weight change and related microstructural change were analyzed. The results showed that the quasi-dynamic facility represented by the natural convection exhibited similar results compared to the conventional dynamic facility. Selective leaching and local depletion of the chromium, re-distribution of the carbide, and the decarburization process took place in the 316-series stainless steel under a flowing sodium environment. This process decreased as the sodium flowed along the channel, which was caused by the change in the dissolved oxygen and carbon activity in the liquid sodium.

• Journal of the Korean Society of Mechanical Technology
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• v.13 no.2
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• pp.145-153
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• 2011

The irradiation tests of materials in HANARO have been performed usually at temperatures below $300^{\circ}C$ at which the RPV(Reactor Pressure Vessel) materials of the commercial reactors such as the light water reactor and CANDU are operated. As VHTR(Very High Temperature Reactor) and SFR(Sodium-cooled Fast Reactor) projects are being carried as a part of the present Gen-IV program in Korea, the requirements for irradiation of materials at temperatures higher than $500^{\circ}C$ are recently being gradually increased. To overcome the restriction in the use at high temperature of the existing Al thermal media, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated more advanced than the single thermal media capsule. At the irradiation test of the capsule, the temperature of the specimens successfully reached $700^{\circ}C$ and the integrity of Al, Ti and graphite material was maintained.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1429-1435
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• 2021

The temperature distribution of an electromagnetic pump was analyzed with a flow rate of 1380 L/min and a pressure of 4 bar designed for the sodium thermo-hydraulic test in the Sodium Test Loop for Safety Simulation and Assessment-Phase 1 (STELLA-1). The electromagnetic pump was used for the circulation of the liquid sodium coolant in the Intermediate Heat Transport System (IHTS) of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) with an electric power of 150 MWe. The temperature distribution of the components of the electromagnetic pump was numerically analyzed to prevent functional degradation in the high temperature environment during pump operation. The heat transfer was numerically calculated using ANSYS Fluent for prediction of the temperature distribution in the excited coils, the electromagnet core, and the liquid sodium flow channel of the electromagnetic pump. The temperature distribution of operating electromagnetic pump was compared with cooling of natural and forced air circulation. The temperature in the coil, the core and the flow gap in the two conditions, natural circulation and forced circulation, were compared. The electromagnetic pump with cooling of forced circulation had better efficiency than natural circulation even considering consumption of the input power for the air blower. Accordingly, this study judged that forced cooling is good for both maintenance and efficiency of the electromagnetic pump.