• Proceedings of the KSME Conference
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• 2008.11b
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• pp.3181-3188
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• 2008

The supercritical $CO_2$ Brayton cycle energy conversion system is presented as a promising alternative to the present Rankine cycle. The principal advantage of the S-$CO_2$ gas is a good efficiency at a modest temperature and a compact size of its components. The S-$CO_2$ Brayton cycle coupled to a SFR also excludes the possibilities of a SWR (Sodium-Water Reaction) which is a major safety-related event, so that the safety of a SFR can be improved. KAERI is conducting a feasibility study for the supercritical carbon dioxide (S-$CO_2$) Brayton cycle power conversion system coupled to KALIMER(Korea Advanced LIquid MEtal Reactor). The purpose of this research is to develop S-$CO_2$ Brayton cycle energy conversion systems and evaluate their performance when they are coupled to advanced nuclear reactor concepts of the type under investigation in the Generation IV Nuclear Energy Systems. This paper contains the research overview of the S-$CO_2$ Brayton cycle coupled to KALIMER-600 as an alternative energy conversion system.

• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.115-125
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• 2016

The prototype SFR has currently been under design by KAERI. The size of its major components is much larger than that of APR1400 and high temperature materials are applied for it. The increased size of components and those specific materials effect on material procurement, manufacturing process and fabrication facilities. The manufacturing methods are studied for Reactor Vessel/Guard Vessel, Control Rod Drive Mechanism, Heat Exchanger, Primary Pump, Reactor Vessel Internals, Steam Generator and In-Vessel Transfer Machine. The proper manufacturing methods are suggested for each component including side forging technology for ultra large forgings of Reactor Vessel to minimize the weld seams on which In-service Inspection should be conducted.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.70-77
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• 2016

In this paper, the structural design and integrity evaluations for a reactor vessel of PGSFR sodium-cooled fast reactor(150MWe) are carried out in compliance with ASME BPV III, Division 5 Subsection HB. The reactor vessel is designed with a direct contact of primary sodium coolant to its inner surface and has a double vessel concept enclosing by containment vessel. To assure the structural integrity for 60 years design lifetime and elevated operating temperature of $545^{\circ}C$, which can invoke creep and creep-fatigue damage, the structural integrity evaluations are carried out in compliance with the ASME code rules. The design loads considered in this evaluations are primary loads and operation thermal cycling loads of normal heat-up and cool-down. From the evaluations, the PGSFR reactor vessel satisfies the ASME code limits but it was found that there is a little design margin of creep damage for inner surface at the region of cold pool free surface.

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.89-94
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• 2014

The 4th generation SFR is being designed with a milestone of construction by 2028. It is important to understand the subchannel flow characteristics in fuel assembly through the experimental investigations and to estimate the calculation uncertainties for insuring the confidence of the design code calculation results. The friction coefficient and the mixing coefficient are selected as primary parameters. The two parameters are related to the flow distribution and diffusion. To identify the flow distribution, an iso-kinetic method was developed based on the previous study. For the mixing parameters, a wire mesh system and a laser induced fluorescence methods were developed in parallel. The measuring systems were adopted on 37 rod bundle test geometry, which was developed based on the Euler number scaling. A scaling method for a design of experimental facility and the experimental identification techniques for the flow distribution and mixing parameters were developed based on the measurement requirement.

• The KSFM Journal of Fluid Machinery
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• v.18 no.6
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• pp.5-11
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• 2015

In a pressurized water reactor (PWR), control rod assembly (CRA) falls into the guide tubes of a fuel assembly due to gravity for scram. Various theoretical approaches and numerical analyses have been performed because its shape is simple and its design was completely developed several decades ago. A control rod assembly for a sodium-cooled faster reactor (SFR) which is geometrically more complicated is being actively developed in Korea nowadays. Drop time and impact velocity of a CRA are important parameters with respect to reactivity insertion time and the mechanical robustness of a CRA and a guide duct. In this paper, computational method considering simultaneously the equation of motion for rigid body and the Navier-Stokes equations for fluid is suggested and verified by comparison with theoretical analysis results. Through this valuable CFD analysis method, drop time and impact velocity of initially designed SFR CRA are evaluated before performing scram tests with it.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.25 no.1
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• pp.48-57
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• 2015

In a sodium-cooled fast reactor, which is a Generation-IV reactor, refueling is conducted by rotating, but not opening, the reactor head to prevent a reaction between the sodium, water and air. Therefore, an inspection technique that checks for the presence of any obstacles between the reactor core and the upper internal structure, which could disturb the rotation of the reactor head, is essential prior to the refueling of a sodium-cooled fast reactor. To this end, an ultrasound-based inspection technique should be employed because the opacity of the sodium prevents conventional optical inspection techniques from being applied to the monitoring of obstacles. In this study, a ranging inspection technique using a plate-type ultrasonic waveguide sensor was developed to monitor the presence of any obstacles between the reactor core and the upper internal structure in the opaque sodium. Because the waveguide sensor installs an ultrasonic transducer in a relatively cold region and transmits the ultrasonic waves into the hot radioactive liquid sodium through a long waveguide, it offers better reliability and is less susceptible to thermal or radiation damage. A 10 m horizontal beam waveguide sensor capable of radiating an ultrasonic wave horizontally was developed, and beam profile measurements and basic experiments were carried out to investigate the characteristics of the developed sensor. The beam width and propagation distance of the ultrasonic wave radiated from the sensor were assessed based on the experimental results. Finally, a feasibility test using cylindrical targets (corresponding to the shape of possible obstacles) was also conducted to evaluate the applicability of the developed ranging inspection technique to actual applications.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.8
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• pp.831-836
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• 2014

In this study, high temperature integrity evaluation on a pressure vessel of the expansion tank operating at elevated temperature of $510^{\circ}C$ in the sodium test facility of the SEFLA(Sodium Thermal-hydraulic Experiment Loop for Finned-tube Sodium-to-Air heat exchanger) to be constructed at KAERI has been performed. Evaluations of creep-fatigue damage based on a full 3D finite element analyses were conducted for the expansion tank according to the recent elevated temperature design codes of ASME Section III Subsection NH and French RCC-MRx. It was shown that the expansion tank maintains its integrity under the intended creep-fatigue loads. Quantitative code comparisons were conducted for the pressure vessel of austenitic stainless steel 316L.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.41 no.5
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• pp.375-380
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• 2017

This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the $B_{0.004}$ life would be guaranteed for the SFR fuel rods with 99％ confidence if no failure was observed at any of the contact surfaces of the five specimens.

• Proceedings of the Korean Institute of Surface Engineering Conference
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• 2018.06a
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• pp.98.1-98.1
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• 2018

소듐냉각 고속로 (SFR)는 원자력 발전의 가장 시급한 문제점으로 부각되고 있는 사용 후 핵연료를 재활용 하여 가동하는 원자로 이다. Generation IV로 명명되는 차세대 원자로 중에 하나로 국제 공동연구와 자체 연구를 통해 우리나라 고유의 기술이 축적되고 개발되고 있다. 현재 소듐냉각 고속로의 가장 큰 문제점 중의 하나는 금속핵연료와 피복관의 상호반응이다. 상호반응이 일어나면 공융현상을 일으켜 피복관의 녹는점이 낮아지고 피복관의 두께가 얇아져 원자로의 안전에 치명적인 위협이 된다. 이러한 문제를 해결하기 위해 전해도금 (electro-plating)을 활용하여 HT9 피복관 내면에 크롬을 도금하여 금속핵연료와 피복관의 상호반응을 억제하는 연구가 본 연구팀에서 진행되고 있다. 크롬과 전해도금을 코팅 물질과 코팅 방법으로 선정한 이유는 튜브 내면에 적용하기 용이하고 경제적인 코팅 방법이기 때문이다. 본 연구에서는 여러 가지 전해도금 인자 중 온도와 pulse 전류의 파형이 상호반응 방지 효과에 미치는 영향에 대하여 고찰하였다. 도금액의 온도를 $50{\sim}80^{\circ}C$, 전류 파형 중 on/off time을 1:1, 10:1, 1:10으로 하여 여러 HT9 시편을 도금하였고 모의 금속 핵연료 합금인 Ce-Nd와 확산 반응 실험을 수행하여 상호반응 방지 효과를 분석하였다. 광학현미경과 전자현미경을 이용한 미세구조 분석 결과 도금액의 온도가 $65^{\circ}C$ 이하인 시편에서는 미세균열이 심하게 발생하였고 그 균열을 통해서 물질이 확산하고 상호반응을 한다는 것이 관찰되었다. $65^{\circ}C$보다 높은 도금액의 온도에서 형성된 크롬막은 균열이 없고 상호반응 방지 효과가 좋은 것이 확인되었다. 특히 전류 파형의 on/off time이 1:1일 때 상호반응 방지 효과가 가장 좋은 것을 확인하였다. 이러한 결과는 크롬 전해도금의 코팅 조건이 상호반응 방지 효과에 매우 중요한 요인으로 작용한다는 것을 말해주고 있다.