• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집B
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• pp.386-391
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• 2001

In this paper, the vibration characteristics of IHTS(Intermediate Heat Transfer System) piping system of LMR(Liquid Metal Reactor) conveying hot liquid sodium are investigated to eliminate the pipe supports for economic reasons. To do this, a 3-dimensional straight pipe element and a curved pipe element conveying fluid are formulated using the dynamic stiffness method of the wave approach and coded to be applied to any complex piping system. Using this method, the dynamic characteristics including the natural frequency, the frequency response functions, and the dynamic instability due to the pipe internal flow velocity are analyzed. As one of the design parameters, the vibration energy flow is also analyzed to investigate the disturbance transmission paths for the resonant excitation and the non-resonant excitations.

• 에너지공학
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• 제12권4호
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• pp.281-288
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• 2003

The MATRA-LMR code has been developed based on a subchannel analysis method for LMR (Liquid Metal Reactor) core subassembly thermal hydraulic design and analysis. The code was improved to allow a seven assembly calculation and can account for inter-assembly heat transfer based on a lumped parameter model. This paper describes the main modifications and improvements of the code and shows reference calculation results which compared single assembly calculation with seven assembly calculation cased for driver and blanket subassemblies of the KALIMER 150 MWe breakeven conceptual design core. KAL- IMER is a pool-type sodium cooled reactor with a thermal output of 392.0 MWth, which have inherently safe, environmentally friendly, proliferation-resistant and economically viable reactor concepts.

• Nuclear Engineering and Technology
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• 제36권2호
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• pp.165-174
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• 2004

The distributed resistance model has been recently implemented into the MATRA-LMR code in order to improve its prediction capability over the wire-wrap model for a flow blockage analysis in the LMR. The code capability has been investigated using experimental data observed in the FFM (Fuel Failure Mock-up)-2A and 5B for two typical flow conditions in a blocked channel. The predicted results by the MATRA-LMR with a distributed resistance model agreed well with the experimental data for wire-wrapped subchannels. However, it is suggested that the parameter n in the distributed resistance model needs to be calibrated accurately for a reasonable prediction of the temperature field under a low flow condition. Finally, the analyses of a blockage for the assembly of the KALIMER design are performed. Satisfactory results by the MATRA-LMR code were obtained through and rerified a comparison with results of the SABRE code.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.489-494
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• 1997

A comparative transient analyses were performed for oxide and nitride cores or a large (3000 MWt), pool-type, liquid-metal-cooled reactor (LMR). The study was focused on three representative accident initiators with failure to scram : the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected fast transient overpower (UFTOP). The margins to fuel melting and sodium boiling have been evaluated for these representative transients. The results show that there is an increase in safety margin with nitride core which maintains the physical dimensions of the oxide core.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 1998년도 춘계 학술발표회 논문집 Proceedings of EESK Conference-Spring 1998
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• pp.147-154
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• 1998

The purpose of this paper is to propose the draft guidelines of seismic isolation design of Liquid Metal Reactor (LMR) using high damping laminated rubber bearings. The scopes of guidelines include design requirements of a seismically isolated system and components, seismic isolator, isolation system, interface system between seismic isolation and non-seismic isolation part, qualification and acceptance tests of seismic isolator, seismic isolation reliability, and seismic safety and monitoring system. Proposed guidelines shall be revised to extend to general design guideline for nuclear facilities by further research and discussions.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.752-762
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• 2021

The Smoothed Particle Hydrodynamics is one of the most widely used mesh-free numerical method for thermo-fluid dynamics. Due to its Lagrangian nature and simplicity, it is recently gaining popularity in simulating complex physics with large deformations. In this study, the 3D single/two-phase numerical simulations are performed on the Liquid Metal Reactor (LMR) centralized sloshing benchmark experiment using the SPH parallelized using a GPU. In order to capture multi-phase flows with a large density ratio more effectively, the original SPH density and continuity equations are re-formulated in terms of the normalized-density. Based upon this approach, maximum sloshing height and arrival time in various experimental cases are calculated by using both single-phase and multi-phase SPH framework and the results are compared with the benchmark results. Overall, the results of SPH simulations show excellent agreement with all the benchmark experiments both in qualitative and quantitative manners. According to the sensitivity study of the particle-size, the prediction accuracy is gradually increasing with decreasing the particle-size leading to a higher resolution. In addition, it is found that the multi-phase SPH model considering both liquid and air provides a better prediction on the experimental results and the reality.

• Nuclear Engineering and Technology
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• 제29권3호
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• pp.240-250
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• 1997

Liquid sodium is widely used as a coolant of LMR(Liquid Metal Reactor) because of its physical and nuclear properties. However, the liquid sodium is very chemically reactive with oxygen and water so that the study on the sodium fire plays an important role in the LMR safety analysis. In this study, a sodium fire model is suggested to analyze the sodium pool fire where both the flame and the reaction products are considered. And also, sodium pool fire analysis computer code, SOPA, is developed. The sensitivity study on the experimental parameters such as the thermal radiation from flame to atmospheric gas, the vessel cooling and the duration of sodium spill was performed. The results showed good agreements with experimental data in the literature.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.797-806
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• 2009

The Multichannel Analyzer for Transient and steady-state in Rod Array - Liquid Metal Reactor for Flow Blockage analysis (MATRA-LMR-FB) code for the analysis of a subchannel blockage has been developed and evaluated through several experiments. The current version of the code is improved here by the implementation of a distributed resistance model which accurately considers the effect of flow resistance on wire spacers, by the addition of a turbulent mixing model, and by the application of a hybrid scheme for low flow regions. Validation calculations for the MATRA-LMR-FB code were performed for Oak Ridge National Laboratory (ORNL) 19-pin tests with wire spacers and Karlsruhe 169-pin tests with grid spacers. The analysis of the ORNL 19-pin tests conducted using the code reveals that the code has sufficient predictive accuracy, within a range of 5 $^{\circ}C$, for the experimental data with a blockage. As for the results of the analyses, the standard deviation for the Karlsruhe 169-pin tests, 0.316, was larger than the standard deviation for the ORNL 19-pin tests, 0.047.

• 전기의세계
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• 제42권7호
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• pp.3-8
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• 1993

무한 동력을 얻고자 하는 생각은 "기술"이나 "과학"이라는 단어가 사용된 이래 수많은 두뇌들에 의해 얼핏 한번은 떠올려진 이상이었을 것이다. 지금과 같이 자원의 고갈과 위기를 쉴새없이 부르짖는 시기에 "무한동력"이란 존재는 인류의 모든 어려움을 해결할만한 방법이고, 특히 과소비성 현대의 생활에는 절대적인 수단일 것이다. 무한 동력과는 비교조차도 할 수 없는 한참아래쪽에서 서성거리고 있는 에너지 발생원인 액체금속 냉각고속증식로의 존재에 대하여, 언제, 어떻게, 접근해야 할까\ulcorner 하는 고민을 우리는 계속해야 할 필요가 있겠는가의 관한 의문에 대한 대답은 항상 긍정적이어야 하겠다. 이제는 국가적 규모의 연구개발사업이 요구되며 자원빈국이면서 공업선진국을 지향하고 있는 우리나라의 현 상황으로는 아직 실용화되고 있지 않은 LMR 개발기술에 접근할 수 있는 시간적 여지가 남아있다. 따라서 우리특유의 기술소화내지는 토착화의 의지를 결집하여 독작적인 개발계획을 수립하여 적극추진한다면 선두주자들과 어깨를 나란히 하여 우리도 꿈의 원자로인 LMR의 혜택을 누릴 수 있을 것으로 기대한다.자로인 LMR의 혜택을 누릴 수 있을 것으로 기대한다.

• 한국전산구조공학회논문집
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• 제12권4호
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• pp.583-589
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• 1999

본 논문에서는 MDO기법에 의한 핵연료교환장치의 구조해석 단계 중 핵연료교환장치의 휨 변형을 구하는 재료역학해석을 수행하였다. 이는 액체 금속로(LMR) 핵연료교환장치의 기본설계를 위하여 매우 중요하다. 해석대상 핵연료교환장치의 정적구조는 기 수행한 핵연료교환장치의 기구 동역 학 해석 결과를 활용하였다. 네 가지 핵연료교환동작에 대하여 핵연료 봉의 무게를 100㎏에서 500㎏까지 100㎏씩 증가시켜 휨 변형의 크기를 구하였다. 그 결과 회전 중심 축에서 가장 멀리 있는 핵연료 봉을 교환하는 핵연료교환동작에서 최대 휨 변형이 발생함이 밝혀졌다. 또한 이 최대 휨 변형이 발생하는 핵연료교환장치구조에 대하여 부재의 단면두께를 축소하면서, 또 단면형상을 여러 가지로 바꾸면서 휨 변형크기를 구하여 비교하였다. 비교결과 비교대상 단면형상 중에서 중공직사각형 단면이 최소 휨 변형이 발생하는 최적단면형상임이 밝혀졌다.