Conceptual Design of the Sodium-Cooled Fast Reactor Kalimer-600

Hahn, Dohee; Kim, Yeong-Il; Lee, Chan Bock; Kim, Seong-O; Lee, Jae-Han; Lee, Yongbum; Kim, Byung-Ho; Jeong, Hae-Yong

doi:10.5516/net.2007.39.3.193

Cited by 44 publications

(18 citation statements)

References 6 publications

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“…The KALIMER-600 is a pool-type sodium-cooled fast reactor and has two heat transport systems: a primary pool, and an intermediate loop [1]. Figure 1 shows the heat balance of the KALIMER-600 at the rated full power condition while detailed data for KALIMER-600 are described in Reference 1.…”

Section: Features Of Kalimer-600mentioning

confidence: 99%

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Development of Reactor Power Control Logic for the Power Maneuvering of Kalimer-600

Seong¹,

Kang²,

Kim³

2010

Nuclear Engineering and Technology

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Section: Features Of Kalimer-600mentioning

confidence: 99%

“…It is a pooltype fast reactor that uses sodium as the coolant in the primary pool and the intermediate loop [1]. In this study, we developed a reactor power control logic for the power maneuvering event of the plant.…”

Section: Introductionmentioning

confidence: 99%

Development of Reactor Power Control Logic for the Power Maneuvering of Kalimer-600

Seong¹,

Kang²,

Kim³

2010

Nuclear Engineering and Technology

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“…Thermal-hydraulic analysis of crossflow PCHE was carried out based on the values in Tables 3 and 4. [38] Na/Na 545/390 320/526 0.1/0.1 Molten Salt-cooled [39] FLiBe/FLiNaK 700/650 600/690 0.1/0.1…”

Section: Thermal-hydraulic Analysis Of Crossflow Pchementioning

confidence: 99%

Analytical Study on Thermal and Mechanical Design of Printed Circuit Heat Exchanger

Yoon

Sabharwall

Kim

2013

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The analytical methodologies for the thermal design, mechanical design, and cost estimation of printed circuit heat exchanger are presented in this study. Three flow arrangements for PCHE of parallel flow, countercurrent flow and crossflow are taken into account. For each flow arrangement, the analytical solution of temperature profile of heat exchanger is introduced. The size and cost of printed circuit heat exchangers for advanced small modular reactors are also presented using various coolants such as sodium, molten salts, helium, and water. viii ix SUMMARYPrinted Circuit Heat Exchanger (PCHE) is one of the candidate designs of Very High Temperature Reactor (VHTR) or Advanced High Temperature Reactor (AHTR) heat exchanger (HX). Fine grooves in the plate of PCHE are made by using the technique that is employed for making printed circuit board. This heat exchanger is formed by the diffusion bonding of stacked plates whose grooved surfaces are the flow paths. Thermal design of heat exchanger is required to determine the size and effectiveness of heat exchanger. To evaluate the structural integrity, mechanical design of heat exchanger must be investigated. In a previous study, the parallel/counter flow PCHE analysis code was developed. The methodologies for the thermal and mechanical design of parallel/counter flow PCHE used in previous study are also summarized in this report.The objective of this work is to develop the analysis code for the crossflow PCHE to determine the size and cost of crossflow PCHE for AHTR. The size of the crossflow PCHE is determined through thermal design process of heat exchanger. Two dimensional temperature profiles in primary and secondary sides of crossflow PCHE are obtained from the solution of analytical model of crossflow PCHE assuming a single pass, both with unmixed fluid, and no contribution from longitudinal heat conduction. The mechanical design of crossflow PCHE is to determine the criteria of geometric parameters of structure for maintaining the integrity of the heat exchanger. The method for mechanical design of crossflow PCHE is developed based on that of parallel/counter flow PCHE.To verify the developed code, the grid sensitivity test and effectiveness-number of transfer units (NTU) analysis were carried out. The grid sensitivity test for the developed code was performed to evaluate the effect of grid number on the result. The result of grid sensitivity test shows that the effect of grid number is negligible. The effectiveness-NTU analysis was carried out to verify the developed code. The calculated effectiveness by the code and -NTU correlation for crossflow heat exchanger shows a good agreement with each other.As a parametric study, uncertainty analyses for fluid properties and heat transfer correlation were performed. The uncertainty of fluid properties was investigated by assuming ±30% of uncertainty in fluid material properties. The result of uncertainty analysis shows that the uncertainty of fluid property was negligible in the thermal design of crossflow...

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“…Burnup 100-150 MW·days/kg has been achieved in the FBTR fast reactor with uranium-plutonium carbide fuel, irradiated fuel with burnup 25, 50, and 100 MW·days/kg has been reprocessed using a novel technology based on the Purex process [3]. Further improvements will require using metallic (alloyed) fuel in a closed fuel cycle: PFBR-500 (startup by 2020, India) [4] and KALIMER-600 (South Korea) [5].Different promising fuel compositions including carbide fuel (UC, PuC, UPuC) have been investigated in our country. A series of experiments with carbide fuel has been performed at the Research Institute for Nuclear Reactors and burnup 10.4% h.a.…”

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confidence: 99%

Neutron-physical characteristics and fuel utilization in a sodium-cooled fast reactor with high-density fuel at the initial stage of the transition to a closed cycle

Mitenkov¹,

Peskov²,

Mitenkova³

2009

At Energy

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The results of neutron-physical investigations of a fast reactor using high-density fuel (UC) at the initial stage of the transition to a closed fuel cycle are presented. Validation is given for the possibility of making the transition to a closed fuel cycle with self-supply of fissile nuclei starting with the first recycle.The high density of the fuel used in fast reactors makes it possible to improve the neutron-physical characteristics and to use the fuel more efficiently.In some western countries, India, China, and South Korea intensive research is being conducted to develop fast reactors using different kinds of fuel. Experiments performed on the Phoenix reactor (France) studying nitride and carbide fuels showed the advantage of the latter type. Investigations of carbide fuel in a closed fuel cycle are continuing as part of the Generation IV international program as well as in a program with France, USA, and Japan collaborating to build FBR by 2020-2025 [2].India has made substantial advances in perfecting carbide fuel. Burnup 100-150 MW·days/kg has been achieved in the FBTR fast reactor with uranium-plutonium carbide fuel, irradiated fuel with burnup 25, 50, and 100 MW·days/kg has been reprocessed using a novel technology based on the Purex process [3]. Further improvements will require using metallic (alloyed) fuel in a closed fuel cycle: PFBR-500 (startup by 2020, India) [4] and KALIMER-600 (South Korea) [5].Different promising fuel compositions including carbide fuel (UC, PuC, UPuC) have been investigated in our country. A series of experiments with carbide fuel has been performed at the Research Institute for Nuclear Reactors and burnup 10.4% h.a. has been achieved [6]. Reassuring results have also been obtained for metallic alloyed uranium-plutonium-zirconium fuel, and a technology for fabricating fuel elements with such fuel has been perfected [7]. A federal targeted program has the BN-800 reactor starting up in 2012. The implementation of a closed fuel cycle based on this reactor is still not completely clear. Consequently, in order to transition to a closed fuel cycle taking account of the operation of this reactor in an open cycle it is helpful to compare the neutron-physical characteristics of the core with several types of fuel in order to determine its dimensions and the optimal parameters of the fuel assemblies.The present article together with [8] presents a conceptual investigation of this question for oxide and carbide fuels. In [8], optimization of fuel-assembly composition and core characteristics was performed for oxide duel and BN-800 oper-

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Conceptual Design of the Sodium-Cooled Fast Reactor Kalimer-600

Cited by 44 publications

References 6 publications

Development of Reactor Power Control Logic for the Power Maneuvering of Kalimer-600

Development of Reactor Power Control Logic for the Power Maneuvering of Kalimer-600

Analytical Study on Thermal and Mechanical Design of Printed Circuit Heat Exchanger

Neutron-physical characteristics and fuel utilization in a sodium-cooled fast reactor with high-density fuel at the initial stage of the transition to a closed cycle

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