• 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.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.572-594
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• 2016

As one of the Gen-IV nuclear energy systems, a sodium-cooled fast reactor (SFR) is being developed at the Korea Atomic Energy Research Institute. As a long-term national research project, advanced radiation resistant oxide dispersion strengthened steel (ARROS) is being developed as an in-core fuel cladding tube material for a SFR in the future. In this paper, the current status of ARROS development is reviewed and its future prospective is discussed.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.968-976
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• 2019

This paper presents a number of verification case studies for a recently developed sensitivity/uncertainty code package. The code package, ROMUSE (Reduced Order Modeling based Uncertainty/Sensitivity Estimator) is an effort to provide an analysis tool to be used in conjunction with reactor core simulators, in particular the Virtual Environment for Reactor Applications (VERA) core simulator. ROMUSE has been written in C++ and is currently capable of performing various types of parameter perturbations and associated sensitivity analysis, uncertainty quantification, surrogate model construction and subspace analysis. The current version 2.0 has the capability to interface with the Design Analysis Kit for Optimization and Terascale Applications (DAKOTA) code, which gives ROMUSE access to the various algorithms implemented within DAKOTA, most importantly model calibration. The verification study is performed via two basic problems and two reactor physics models. The first problem is used to verify the ROMUSE single physics gradient-based range finding algorithm capability using an abstract quadratic model. The second problem is the Brusselator problem, which is a coupled problem representative of multi-physics problems. This problem is used to test the capability of constructing surrogates via ROMUSE-DAKOTA. Finally, light water reactor pin cell and sodium-cooled fast reactor fuel assembly problems are simulated via SCALE 6.1 to test ROMUSE capability for uncertainty quantification and sensitivity analysis purposes.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.283-291
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• 2024

Sodium cooled Fast Reactors (SFR) are built with several engineered safety features and hence a severe accident such as a core melt accident is hypothetical with a probability of <10^-6/ry. However, in case of such accidents, the mixture of the molten fuel and structural materials interacts with sodium. This phenomenon is known as Molten Fuel Coolant Interaction (MFCI) and results in fragmentation of the melt due to various instabilities. The fragmented particles settle as a debris bed on the core catcher at the bottom of the reactor vessel, and continue to generate decay heat. Characteristics of the debris particles play a vital role in heat transfer from the bed and need thorough investigation. The size, shape, and physical state of the debris depend on the associated fragmentation mechanism, superheating of the melt, and sodium temperature. Experiments have been conducted by releasing simulated corium, a molten mixture of alumina and iron generated by the aluminothermy process at ~2400 ℃ into liquid sodium, to study the fragmentation phenomena. After the experiment, the fragmented debris was retrieved and the particle size distribution was determined by sieve analysis. The debris was subjected to microscopic investigation for obtaining morphological characteristics. Based on the characteristics of debris, an attempt has been made to assess of fragmentation mechanism of simulated corium in sodium.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2077-2083
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• 2022

In the unlikely event of core disruptive accident in sodium cooled fast reactors, the reactor containment building would be bottled up with sodium and fission product aerosols. The behavior of these aerosols is crucial to estimate the in-containment source term as a part of nuclear reactor safety analysis. In this work, the evolution of sodium aerosol characteristics (mass concentration and size) is simulated using HAARM-S code. The code is based on the method of moments to solve the integro-differential equation. The code is updated to FORTRAN-77 and run in Microsoft FORTRAN PowerStation 4.0 (on Desktop). The sodium aerosol characteristics simulated by HAARM-S code are compared with the measured values at Aerosol Test Facility. The maximum deviation between measured and simulated mass concentrations is 30% at initial period (up to 60 min) and around 50% in the later period. In addition, the influence of humidity on aerosol size growth for two different aerosol mass concentrations is studied. The measured and simulated growth factors of aerosol size (ratio of saturated size to initial size) are found to be matched at reasonable extent. Since sodium is highly reactive with atmospheric constituents, the aerosol growth factor depends on the hygroscopic growth, chemical transformation and density variations besides coagulation. Further, there is a scope for the improvement of the code to estimate the aerosol dynamics in confined environment.