• Nuclear Engineering and Technology
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• 제31권6호
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• pp.1-7
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• 1999

An innovative concept for a passive reactor shutdown system, so called self-actuated shutdown system(SASS), is inevitably required for the inherent safety in liquid metal reactor, which is designed with the totally different concept from the usual reactor shutdown system in LWR. SASS using Curie point electromagnet(CPEM) was selected as the passive reactor shutdown system for KALIMER (Korea Advanced Liquid MEtal Reactor). A mock-up of the SASS was designed, fabricated and tested. From the test it was confirmed that the mockup was self-actuated at the Curie point of the temperature sensing material used in the mockup. An articulated control rod was also fabricated and assembled with the CPEM to confirm that the control rod can be inserted into core even when the control rod guide tube is deformed due to earthquake. The operability of SASS in the actual sodium environment should be confirmed in the future. All the design and test data will be applied to the KALIMER design.

• Nuclear Engineering and Technology
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• 제34권2호
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• pp.117-131
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• 2002

Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of the work to demonstrate the inherent and ultimate safety of conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 MWe pool- type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method and associated computer program, SCHAMBETA, was developed using a modified Bethe-Tait method to simulate the kinetics and thermodynamic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of the energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the SCHAMBETA code for various reactivity insertion rates up to 100 S/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies were also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1802-1813
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• 2023

Conjugate heat transfer between liquid metal and solid is a common phenomenon in a liquid-metal-cooled fast reactor's fuel assembly and heat exchanger, dramatically affecting the reactor's safety and economy. Therefore, comprehensively studying the sophisticated conjugate heat transfer in a liquid-metal-cooled fast reactor is profound. However, it has been evidenced that the traditional Simple Gradient Diffusion Hypothesis (SGDH), assuming a constant turbulent Prandtl number (Pr_t,, usually 0.85 - 1.0), is inappropriate in the Computational Fluid Dynamics (CFD) simulations of liquid metal. In recent decades, numerous studies have been performed on the four-equation model, which is expected to improve the precision of liquid metal's CFD simulations but has not been introduced into the conjugate heat transfer calculation between liquid metal and solid. Consequently, a four-equation model, consisting of the Abe k - ε turbulence model and the Manservisi k_𝜃 - ε_𝜃 heat transfer model, is applied to study the conjugate heat transfer concerning liquid metal in the present work. To verify the numerical validity of the four-equation model used in the conjugate heat transfer simulations, we reproduce Johnson's experiments of the liquid lead-bismuth-cooled turbulent pipe flow using the four-equation model and the traditional SGDH model. The simulation results obtained with different models are compared with the available experimental data, revealing that the relative errors of the local Nusselt number and mean heat transfer coefficient obtained with the four-equation model are considerably reduced compared with the SGDH model. Then, the thermal-hydraulic characteristics of liquid metal turbulent pipe flow obtained with the four-equation model are analyzed. Moreover, the impact of the turbulence model used in the four-equation model on overall simulation performance is investigated. At last, the effectiveness of the four-equation model in the CFD simulations of liquid sodium conjugate heat transfer is assessed. This paper mainly proves that it is feasible to use the four-equation model in the study of liquid metal conjugate heat transfer and provides a reference for the research of conjugate heat transfer in a liquid-metal-cooled fast reactor.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2002년도 학술대회지
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• pp.47-50
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• 2002

The numerical simulation of a 271-rod fuel assembly of nuclear Liquid-Metal Fast Reactor (LMFR) with an infernal blockage has been carried out. Internal blockage within a subassembly is addressed in the safety assessment because it potentially has very serious consequences for the reactor as a whole. Three dimensional calculations were performed using the SABRE4 computer code for the range of blockage positions and sizes to investigate the seriousness and detectability of the internal blockage. The magnitude and location of the peak temperatures together with the temperature distribution at the subassembly exit were calculated in order to look at the potential for damage within the subassembly, and the possibility of blockage detection. The analysis result shows that the 6-subchannel blockage causes large temperature rise within a assembly with practically no change in mixed mean temperature at the assembly exit.

• Nuclear Engineering and Technology
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• 제39권1호
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• pp.43-50
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• 2007

A partial flow blockage in an assembly of a liquid metal reactor could result in a cooling deficiency of the core. To develop a partial blockage detection system, we have studied the changes of the temperature fluctuation characteristics in the upper plenum according to changes of the t10w blockage conditions in an assembly. We analyzed the temperature fluctuation in the upper plenum with the Large Eddy Simulation (LES) turbulence model in the CFX code and evaluated its statistical parameters. Based on the results of the statistical analyses, we developed a neural network model for detecting a partial flow blockage in an assembly. The neural network model can retrieve the size and the location of a flow blockage in an assembly from a change of the root mean square, the standard deviation, and the skewness in the temperature fluctuation data. The neural network model was found to be a possible alternative by which to identify a flow blockage in an assembly of a liquid metal reactor through learning and validating various flow blockage conditions.

• 에너지공학
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• 제10권4호
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• pp.342-348
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• 2001

A new design concept for a decay heat removal system in a liquid metal reactor is proposed. The new design utilizes a thermosyphon to enhance the heat removal capacity and its heat transfer characteristics are analyzed against the current PSDRS (Passive Safety Decay heat Removal System) in the KAL IMER (Korea Advanced LIquid MEtal Reactor) design. The preliminary analysis results show that the new design with a thermosyphon yields substantial increase of 20∼40% in the decay heat removal capacity compared to the current design that do not have the thermosyphon. The new design reduces the temperature rise in the cooling air of the system and helps the surrounding structure in maintaining its mechanical integrity for long term operation at an accident. Also the analysis revealed the characteristics of the interactions among various heat transfer modes in the new design.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2002년도 추계 학술대회논문집
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• pp.64-69
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• 2002

A numerical study has been peformed for evaluation of convection schemes for thermal hydraulic analysis in a liquid metal reactor Four convection schemes, HYBRID, QUICK, SMART and HLPA included in the CFX-4 code are considered. The performances of convection schemes are evaluated by applying them to the five test problems. The accuracy, stability and convergence are tested. It is shown that the HYBRID scheme is too diffusive, and the QUICK scheme exhibits overshoots and undershoots, and the SMART scheme shows convergence oscillations, and the HLPA scheme preserves the boundedness without causing convergence oscillations. The accuracies of SMART, QUICK and HLPA schemes are comparable. Thus, the use of HLPA scheme is highly recommended for thermal hydraulic analysis in a liquid metal reactor.

• 한국가시화정보학회지
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• 제9권4호
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• pp.63-68
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• 2011

The flow-field of a liquid-metal system is very important for the safety analysis and the design of the steam generator of liquid-metal fast breeder reactor. Dynamic neutron radiography (DNR) is suitable for a visualization and measurement of a liquid metal flow and a two-phase flow in a metallic duct. However, the three dimensional DNR techniques is not enough to obtain the velocity information in the wide channel up to now. In this research, a high speed DNR technique was applied to visualize the heavy liquid-metal flow field in the narrow channel with the HANARO-beam facility. The images were taken with a high frame-rate neutron radiography at 250 fps and analyzed with a Particle Image Velocimetry(PIV) method. The images were compared with the results of the commercial CFX code to study the feasibility of DNR technique for the measuring the heavy liquid-metal flow field. The PIV images could discern the turbulent vortex flow in the two-dimensional narrow channel.

• 한국전산유체공학회지
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• 제7권3호
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• pp.44-52
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• 2002

An analytic methodology was developed for free surface motions between liquid metal coolant and cover gas in order to calculate the phenomena of gas entrainment in hot pool surface through IHX EMP and reactor core. The methodology was setup by applying the first order VOF convection model to CFX4 general purpose fluid dynamics analysis code. The methodology was validated by applying it to an experimental apparatus designed for free surface motions of KALIMER reactor. The distributions of free surface calculated by the present methodology were almost coincident with the experimental data. The developed methodology was applied to the KALIMER reactor of full power operating condition. The shapes of the free surface were nearly uniform. From the results, it was found that the altitude of the free surface from the IHX inlet nozzle of KALIMER reactor is high enough not to affect to free surface motions of generating gas bubbles from the turbulent shear flows such as hydraulic jump and water falls.

• 비파괴검사학회지
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• 제24권2호
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• pp.171-179
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• 2004

가동중검사와 보수는 원자력발전소의 원자로계통설계에서 매우 중요한 설계개념이다. 액체금속로 KALIMER의 가동성 검증을 위해서 기계계통 설계에 가동중검사와 보수개념이 반영되어야 한다. 본 연구에서는 KALIMER 의 안전성과 신뢰성을 확보하기 위하여 KALIMER의 개념설계 단계에 필요한 가동중검사와 보수의 기본개념을 설정하였다. 액체금속로 가동중검사 규정인 ASME Section XI Division 3를 반영하고 KALIMER의 설계특성을 고려하여 원자로계통과 주요부품의 가동중검사와 보수에 대한 방법과 요건을 설계하고 기술하였다.