• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.785-796
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• 2009

A KALIMER-600 concept which is a type of sodium-cooled fast reactor, has been developed at KAERI. It uses sodium as a primary coolant and is a pool-type reactor to enhance safety. Also, a supercritical carbon dioxide ($CO_2$) Brayton cycle is considered as an alternative to an energy conversion system to eliminate the sodium water reaction and to improve efficiency. In this study, a simplified model for analyzing the thermodynamic performance of the KALIMER-600 coupled with a supercritical $CO_2$ Brayton cycle was developed. To develop the analysis model, a commercial modular modeling system (MMS) was adopted as a base engine, which was developed by nHance Technology in USA. It has a convenient graphical user interface and many component modules to model the plant. A new user library for thermodynamic properties of sodium and supercritical $CO_2$ was developed and attached to the MMS. In addition, some component modules in the MMS were modified to be appropriate for analysis of the KALIMER-600 coupled with the supercritical $CO_2$ cycle. Then, a simplified performance analysis code was developed by modeling the KALIMER-600 plant with the modified MMS. After evaluating the developed code with each component data and a steady state of the plant, a simple power reduction and recovery event was evaluated. The results showed an achievable capability for a performance analysis code. The developed code will be used to develop the operational strategy and some control logics for the operation of the KALIMER-600 with a supercritical $CO_2$ Brayton cycle after further studies of analyzing various operational events.

• KIPS Transactions on Computer and Communication Systems
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• v.2 no.12
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• pp.561-568
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• 2013

Instrumentation and control system in nuclear power plant performs protecting, controling and monitoring safety operation of Nuclear Power Plant. As cyber attack to the control equipment of instrumentation and control system can cause reactor shutdown and radiation release, it is required to design the instrumentation and control system considering cyber security in accordance with regulatory guides and industrial standards. In this paper, we proposed a design method of uni-directional communication structure which is required in the design of defense-in-depth model according to regulatory guides and industrial standards and we implemented a communication board with the proposed method. This communication board was tested in various test environments and test items and we concluded it can provide uni-directional communication structure required to design of defense-in-depth model against cyber attack by analyzing the results. The proposed method and implemented communication board were applied in the design of SMART (system-integrated modular advanced reactor) I&C (instrumentation and control) systems.

• Nuclear Engineering and Technology
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• v.39 no.2
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• pp.103-110
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• 2007

The High Temperature Gas-cooled Reactor (HTGR) possesses inherent safety features and is recognized as a representative advanced nuclear system for the future. Based on the success of the HTR-10, the long-time operation test and safety demonstration tests were carried out. The long-time operation test verifies that the operation procedure and control method are appropriate for the HTR-10 and the safety demonstration test shows that the HTR-10 possesses inherent safety features with a great margin. Meanwhile, two new projects have been recently launched to further develop HTGR technology. One is a prototype modular plant, denoted as HTR-PM, to demonstrate the commercial capability of the HTGR power plant. The HTR-PM is designed as $2{\times}250$ MWt, pebble bed core with a steam turbine generator that serves as an energy conversion system. The other is a gas turbine generator system coupled with the HTR-10, denoted as HTR-10GT, built to demonstrate the feasibility of the HTGR gas turbine technology. The gas turbine generator system is designed in a single shaft configuration supported by active magnetic bearings (AMB). The HTR-10GT project is now in the stage of engineering design and component fabrication. R&D on the helium turbocompressor, a key component, and the key technology of AMB are in progress.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.344-363
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• 1999

A multi-dimensional thermal-hydraulic system analysis code, MARS 1.3.1, has been developed in order to have the realistic analysis capability of two-phase thermal-hydraulic transients for pressurized water reactor (PWR) plants. As the backbones for the MARS code, the RELAP5/MOD3.2.1.2 and COBRA-TF codes were adopted in order to take advantages of the very general, versatile features of RELAP5 and the realistic three-dimensional hydrodynamic module of COBRA-TF. In the MARS code, all the functional modules of the two codes were unified into a single code first. Then, the source codes were converted into the standard Fortran 90, and then they were restructured using a modular data structure based on "derived type variables" and a new "dynamic memory allocation" scheme. In addition, the Windows features were implemented to improve user friendliness. This paper presents the developmental work of the MARS version 1.3.1 including the hydrodynamic model unification, the heat structure coupling, the code restructuring and modernization, and their verifications.their verifications.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.151-156
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• 2001

Experimental frictional and wear characteristics of silicon graphite materials is studied in this paper. Those specimens are lubricated with high temperature and highly pressurized water to simulate the same operating condition for the journal bearing and the thrust bearing on the main coolant pump bearing in the newly developing nuclear reactor named SMART(System-integrated Modular Advanced ReacTor). Operating condition of the bearings is realized by the tribometer and the autoclave. Friction coefficient and wear loss are analyzed to choose the best silicon graphite material. Pin on plate test specimens are used and coned disk springs are used to control the applied force on the specimens. Wear loss ana wear width are measured by a precision balance and a micrometer. The friction force is measured by the strain gauge which can be used under high temperature and high pressure. Three kinds of silicon graphite materials are examined and compared with each other, and each material shows similar but different results on frictional and wear characteristics.

• The Transactions of the Korean Institute of Electrical Engineers B
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• v.50 no.10
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• pp.483-493
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• 2001

The 3-phase induction is used for the MCP(main coolant pump) and the pure water is used for lubrication of bearing because of the difficulty of repair. Therefore the type of motor is the canned-motor that is welded by sealed can to prevent the stator and rotor from the lubricating water. A lot of Eddy currents are produced in the can because of the conductivity of material. And these eddy currents in the can are the most important cause that decrease the efficiency of induction motor. Therefore we have to find the method to decrease these eddy currents in the can for the improvement of efficiency of motor. In this paper, we proposed the method of design and analysis of canned-motor using equivalent circuit with consideration of can loss for the improvement of efficiency of motor.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2053-2058
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• 2004

The pressurized light water cooled, medium power (330 MWt) SMART (System-integrated Modular Advanced ReacTor) has been under development at KAERI for a dual purpose : seawater desalination and electricity generation. The SMART design verification phase was followed to conduct various separate effects tests and comprehensive integral effect tests. The high temperature / high pressure thermal-hydraulic test facility, VISTA(Experimental Verification by Integral Simulation of Transient and Accidents) has been constructed to simulate the SMART-P (the one fifth scaled pilot plant) by KAERI. Experimental tests have been performed to investigate the thermal-hydraulic dynamic characteristics of the primary and the secondary systems. Heat transfer characteristics and natural circulation performance of the PRHRS (Passive Residual Heat Removal System) of SMART-P were also investigated using the VISTA facility. The coolant flows steadily in the natural circulation loop which is composed of the steam generator (SG) primary side, the secondary system, and the PRHRS. The heat transfers through the PRHRS heat exchanger and ECT are sufficient enough to enable the natural circulation of the coolant.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.4
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• pp.481-496
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• 2020

After nuclear power plants are permanently shut down and decommissioned, the remaining irradiated metal components such as stainless steel, carbon steel, and Inconel can be used as neutron absorber. This study investigates the possibility of reusing these metal components as neutron absorber materials, that is burnable poison. The absorption cross section of the irradiated metals did not lose their chemical properties and performance even if they were irradiated over 40-50 years in the NPPs. To examine the absorption capability of the waste metals, the lattice calculations of WH 17×17 fuel assembly were analyzed. From the results, Inconel-718 significantly hold-down fuel assembly excess reactivity compared to stainless steel 304 and carbon steel because Inconel-718 contains a small amount of boron nuclide. From the results, a 20wt% impurity of boron in irradiated Inconel-718 enhances the excess reactivity suppression. The application of irradiated Inconel-718 as a burnable absorber for SMR core was investigated. The irradiated Inconel-718 impurity with 20wt% of boron content can maintain and suppress the whole core reactivity. We emphasize that the irradiated metal components can be used as burnable absorber materials to control the reactivity of commercial reactor power and small modular reactors.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1464-1470
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• 2022

This paper presents a high performance burnable absorber named as CIMBA (Cylindrically Inserted and Mechanically Separated Burnable Absorber) for the soluble-boron-free SMR. The CIMBA is the cylindrical gadolinia inserted into the annular fuel pellets. Although the CIMBA utilizes the spatial self-shielding effect of the fuel material, a large reactivity upswing occurs when the gadolinia is depleted. To minimize the reactivity swing of the CIMBA-loaded FA, two approaches were investigated. One is controlling the spatial self-shielding effect of the CIMBA as burnup proceeds by a multi-layered structure of the CIMBA (ML-CIMBA) and the other is the mixed-loading of two different types of CIMBA (MIX-CIMBA). Both approaches show promising performances to minimize the reactivity swing, where the MIX-CIMBA is more preferable due to its simpler fabrication process. In conclusion, the MIX-CIMBA is expected to accelerate the commercialization of the CIMBA and can be used to achieve an optimal soluble-boron-free SMR core design.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1890-1901
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• 2022

We have assessed the applicability of the thermal-hydraulic system analysis code, SPACE, to a small modular reactor called SMART. For the assessment, the experimental data from a scale-down integral-test facility, SMART-ITL, were used. It was conformed that the SPACE code unrealistically calculates the safety injection flow rate through the CMT and SIT during a small-break loss-of-coolant experiment. This unrealistic behavior was due to the overprediction of interfacial heat transfer at the steam-water interface in a vertically stratified flow in the tanks. In this study, a special thermal-hydraulic component model has been developed to realistically calculate the interfacial heat transfer when a strong non-equilibrium two-phase flow is formed in the CMT or SIT. Additionally, we developed a special heat structure model, which analytically calculates the heat transfer from the hot steam to the cold tank wall. The combination of two models for the tank are called the special component model. We assessed it using the SMART-ITL passive safety injection system (PSIS) test data. The results showed that the special component model well predicts the transient behaviors of the CMT and SIT.