• Proceedings of the KSME Conference
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• 대한기계학회 2004년도 춘계학술대회
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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.

• KSTLE International Journal
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• 제4권2호
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• pp.43-46
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• 2003

Water-lubricated frictional characteristics of a stainless steel ball bearings are not well known compared to the oil-lubricated frictional characteristics. Furthermore a study on friction at a high temperature is rare because the bearing maintenance strategy for water-lubricated or chemicals-lubricated bearings of equipment is generally based on the replacement of the failed bearings-and parts. Ball bearings and ball screw are installed in the power transmission for the newly developing integral reactor and these are lubricated with chemically-controlled pure water at a high temperature and a high pressure. Bearings and power transmitting mechanical elements for an atomic reactor requires high reliability and high performance during the estimated lifetime, and it should be verified. In this paper, experimental research results of the frictional characteristics for water-lubricated ball bearings are presented as a preliminary investigation.

• Proceedings of the KIEE Conference
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• 대한전기학회 1999년도 추계학술대회 논문집 학회본부 A
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• pp.98-100
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• 1999

A thermal analysis was performed for the electromagnet which is installed in the control element drive mechanism(CEDM) of the integral reactor SMART. A model for the thermal analysis of the electromagnet was developed and theoretical bases for the model were established. It is important that the temperature of the electromagnet windings be maintained within the allowable limit of the insulation, since the electromagnet of CEDM is always supplied with current during the reactor operation. So the thermal analysis of the winding insulation which is composed of polyimide and air were performed by finite element method. The thermal properties obtained here will be used as input for the optimization analysis of the electromagnet.

• Nuclear Engineering and Technology
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• 제44권6호
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• pp.689-696
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• 2012

A small integral reactor SMART (System Integrated Modular Advanced ReacTor), being developed in Korea since late 1990s and targeted to obtaining a standard design approval by the end of 2011, is introduced. The design scope and level for design certification (DC) is well described in the U.S. NRC SECY documents published the early 1990s. However, the documents are valid for a one-step licensing process called a combined operating license (COL) by the U.S. NRC, while Korea still uses a two-step licensing process. Thus, referencing the concept of the SECY documents, we have established the design scope and level for the SMART DC using the contexts of the standard review plan (SRP). Some examples of the results and issues raised during our review are briefly presented in this paper. The same methodology will be applied to other types of reactor under development in Korea, such as future VHTR reactors.

• Nuclear Engineering and Technology
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• 제14권4호
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• pp.163-171
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• 1982

An ABBN-type 26-group constant set, KAERI-26G, which can be reliably applicable to fast reactor calculations has been generated using the nuclear data of ENDF/B-IV or ENDL-78 and a processing code ETOX-K4. The KAERI-26G set was evaluated by analysing measured integral quantities such as effective multiplication factor, central reaction-rate ratio, and central reactivity coefficient for a variety of critical assemblies. All these calculated quantities were compared with results from other workers using similar-type sets.

• Proceedings of the Korean Nuclear Society Conference
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.623-628
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• 1998

A new advanced integral reactor of 330 MWt thermal capacity named SMART (System-Integrated Modular Advanced Reactor) is currently under development in Korea Atomic Energy Research Institute (KAERI) for multi-purpose applications. Modular once-through steam generator (SG) and self-pressurizing pressurizer equipped with wet thermal insulator and cooler are essential components of the SMART. The SMART Provides safety systems such as Passive Residual Heat Removal System (PRHRS). In this study, a computer code for performance analysis of the PRHRS is developed by modeling relevant components and systems of the SMART. Using this computer code, a performance analysis of the PRHRS is performed in order to check whether the passive cooling concept using the PRHRS is feasible. The results of the analysis show that PRHRDS of the SMART has excellent passive heat removal characteristics.

• Tribology and Lubricants
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• 제20권3호
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• pp.140-144
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• 2004

Water-lubrication ball bearings are required to install in aqueous medium where water is used as coolant or working fluid. However water-lubricated frictional characteristics of stainless steel ball bearing is not will known compared to oil-lubricated frictional characteristics. Furthermore study on friction at high temperature is rare because bearing maintenance strategy for water-lubricated or chemicals-lubricated bearings of equipment is mostly based on change of failed bearings and parts. Ball bearings and ball screws are used to transmit power in the control rod drive mechanism for an integral reactor and are lubricated with high temperature and high pressure chemically-controlled water. Bearings and power transmitting mechanical elements for a nuclear reactor require high reliability and high performance during estimated lifetime, and their performance should be verified. In this paper, experimental research results of frictional characteristics of water-lubricated ball bearing are reported.

• Nuclear Engineering and Technology
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• 제8권4호
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• pp.209-217
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• 1976

A finite element method is formulated for one-speed integral equation it or the neutron transport in a slab reactor. The formulation incorporates both the linear and the cubic Hermite interpolating polynomials and is used to compute the approximate solutions for the slab criticality and Milne problem. The results are compared with the exact solutions available and then the effectiveness of the method is extensively discussed.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 한국윤활학회 2002년도 proceedings of the second asia international conference on tribology
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• pp.289-290
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• 2002

Stainless steel ball bearings are used in the control element drive mechanism and driving mechanisms such as step motor and gear boxes for the integral nuclear reactor, SMART. The bearings operate in pressurized pure water (primary coolant) at high temperature and should be lubricated with only this water because it is impossible to supply greases or any additional lubricant since the whole nuclear rector system should be perfectly sealed and the coolant cannot contain ingredients for bearing lubrication. Temperature of water changes from room temperature to about 120 degree Celsius and pressure rises up to 15MPa in the nuclear reactor. It can be anticipated that the frictional characteristics of the ball bearings changes according to the operating conditions, however little data are available in the literature. It is found that friction coefficient of 440C stainless steel itself does not change sharply according to temperature variation from the former research, and the friction coefficient is about 0.45 at low speed range. In this research frictional characteristics of the assembled ball bearings are investigated. A special tribometer is used to simulate the axial loading and the bearing operating conditions, temperature and pressure in the driving mechanism in the nuclear reactor. Highly purified water is used as lubricant ‘ and the water is heated up to 120 degree Celsius and pressurized to 15MPa. Friction force is monitored by the torque transducer.

• Nuclear Engineering and Technology
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• 제41권7호
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• pp.953-968
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• 2009

It is important to accurately predict the temperature and density distributions in large stratified enclosures both for design optimization and accident analysis. Current reactor system analysis codes only provide lumped-volume based models that can give very approximate results. Previous scaling analysis has shown that stratified mixing processes in large stably stratified enclosures can be described using one-dimensional differential equations, with the vertical transport by jets modeled using integral techniques. This allows very large reductions in computational effort compared to three-dimensional CFD simulation. The BMIX++ (Berkeley mechanistic MIXing code in C++) code was developed to implement such ideas. This paper summarizes major models for the BMIX++ code, presents the two-plume mixing experiment simulation as one validation example, and describes the codes' application to the liquid salt buffer pool system in the AHTR (Advanced High Temperature Reactor) design. Three design options have been simulated and they exhibit significantly different stratification patterns. One of design options shows the mildest thermal stratification and is identified as the best design option. This application shows that the BMIX++ code has capability to provide the reactor designers with insights to understand complex mixing behavior with mechanistic methods. Similar analysis is possible for liquid-metal cooled reactors.