• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.286-300
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• 2002

An algorithm for the sodium boiling model has been developed for calculation of the void reactivity feedback as well as the fuel and cladding temperatures in the KALIMER core after onset of sodium boiling. Modeling of sodium boiling in liquid metal reactors using sodium as a coolant is necessary because of phenomenon difference comparing with that observed generally in light water reactor systems. The applied model to the algorithm is the multiple-bubble slug ejection model. It allows a finite number of bubbles in a channel at any time. Voiding is assumed to result from formation of bubbies that (ill the whole cross section of the coolant channel except for the liquid film left on the cladding surface. The vapor pressure, currently, is assumed to be uniform within a bubble The present study is focused on not only demonstration of the vapor bubble behavior predicted by the developed model, but also confirmation of a qualitative acceptance for the model. As a result, the model can represent important phenomena in the sodium boiling, but it is found that further effort is also needed for its completition.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3517-3527
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• 2021

New 80-MW (electric) ultra-long-life sodium cooled fast reactor core having inherent safety characteristics is designed with heterogeneous fuel assemblies comprised of driver and blanket fuel rods. Several options using upper sodium plenum and SSFZ (Special Sodium Flowing Zone) for reducing sodium void reactivity are neutronically analyzed in this core concept in order to improve the inherent safety of the core. The SSFZ allowing the coolant flow from the peripheral fuel assemblies increases the neutron leakage under coolant expansion or voiding. The Monte Carlo calculations were used to design the cores and analyze their physics characteristics with heterogeneous models. The results of the design and analyses show that the final core design option has a small burnup reactivity swing of 618 pcm over ~54 EFPYs cycle length and a very small sodium void worth of ~35pcm at EOC (End of Cycle), which leads to the satisfaction of all the conditions for inherent safety with large margin based on the quasi-static reactivity balance analysis under ATWS (Anticipated Transient Without Scram).

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1747-1755
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• 2021

This paper presents floating absorber for safety at transient (FAST) which is a passive safety device for sodium-cooled fast reactors with a positive coolant temperature coefficient. Working principle of the FAST makes it possible to insert negative reactivity passively in case of temperature rise or voiding of coolant. Behaviors of the FAST in conventional oxide fuel-loaded and metallic fuel-loaded SFRs are investigated assuming anticipated transients without scram (ATWS) scenarios. Unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected transient overpower (UTOP) and unprotected chilled inlet temperature (UCIT) scenarios are simulated at end of life (EOL) conditions of the oxide and the metallic SFR cores, and performance of the FAST to improve the reactor safety is analyzed in terms of reactivity feedback components, reactor power and maximum temperatures of fuel and coolant. It is shown that FAST is able to improve the safety margin of conventional burner-type SFRs during ULOF, ULOHS, UTOP and UCIT.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.660-665
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• 1998

The feasibility study on conceptual design methodology for accelerator-driven sodium-cooled sub-critical transmutation reactors has been conducted to optimize the design parameters from the scale laws and validates reactor performance with the integrated code system. A 1000 MWth sodium-cooled sub-critical transmutation reactor has been scale and verified through the methodology in this paper, which is referred to advanced Liquid Metal Reactor (ALMR). a Pb-Bi target material and a partitioned fuel are the liquid phases, and they are cooled by the circulation of secondary Pb-Bi coolant and by primary sodium coolant, respectively. Overall key design parameters are generated from the scale laws and they are improved and validated by the intergrated code system. Intergrated Code System (ICS) consist of LAHET, HMCNP, ORIGEN2, and COMMIX codes and some files. Through ICS the target region, the core region, and thermal-hydraulic related are analyzed once-through. Results of conceptual design are attached in this paper.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.227-239
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• 2015

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, doublefault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel-coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

• Journal of Energy Engineering
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• v.13 no.2
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• pp.101-111
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• 2004

The main purpose of a liquid metal reactor core thermal-hydraulic design is to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power distribution in the core. The thermal-hydraulic design procedure consists of the coolant flow distribution to the sub-assemblies, the coolant/fuel temperature calculations and detailed subchannel analysis. This paper describes the LMR core thermal-hydraulic design methodology and summarizes the major design and analysis results of KALIMER breeder and breakeven cores and subassemblies. KALIMER is a 150 MWe rated (392 MWth) heterogeneous core with U-TRU-Zr ternary alloy fuel and sodium coolant.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.283-291
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• 2024

Sodium cooled Fast Reactors (SFR) are built with several engineered safety features and hence a severe accident such as a core melt accident is hypothetical with a probability of <10^-6/ry. However, in case of such accidents, the mixture of the molten fuel and structural materials interacts with sodium. This phenomenon is known as Molten Fuel Coolant Interaction (MFCI) and results in fragmentation of the melt due to various instabilities. The fragmented particles settle as a debris bed on the core catcher at the bottom of the reactor vessel, and continue to generate decay heat. Characteristics of the debris particles play a vital role in heat transfer from the bed and need thorough investigation. The size, shape, and physical state of the debris depend on the associated fragmentation mechanism, superheating of the melt, and sodium temperature. Experiments have been conducted by releasing simulated corium, a molten mixture of alumina and iron generated by the aluminothermy process at ~2400 ℃ into liquid sodium, to study the fragmentation phenomena. After the experiment, the fragmented debris was retrieved and the particle size distribution was determined by sieve analysis. The debris was subjected to microscopic investigation for obtaining morphological characteristics. Based on the characteristics of debris, an attempt has been made to assess of fragmentation mechanism of simulated corium in sodium.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.170-170
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• 2001

• Journal of Mechanical Science and Technology
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• v.19 no.4
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• pp.1044-1051
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• 2005

The isothermal characteristics of a rectangular parallelepiped sodium heat pipe were inves­tigated for high-temperature applications. The heat pipes was made of stainless steel of which the dimension was $140\;m\;(L)\;{\times}\;95m\;(W)\;{\times}\;46 m\;(H)$ and the thickness of the container was 5 mm. Both inner surfaces of evaporator and condenser were covered with screen meshes to help spread the liquid state working fluid. To provide additional path for the working fluid, a lattice structure covered with screen mesh wick was inserted in the heat pipe. The bottom surface of the heat pipe was heated by an electric heater and the top surface was cooled by circulating coolant. The concern in this study was to enhance the temperature uniformity at the bottom surface of the heat pipe while an uneven heat source up to 900 W was in contact. The temperature distribution over the bottom surface was monitored at more than twenty six locations. It was found that the operating performance of the sodium heat pipe was critically affected by the inner wall temperature of the condenser region where the working fluid may be changed to a solid phase unless the temperature was higher than its melting point. The maximum temperature difference across the bottom surface was observed to be $114^{\circ}C$ for 850 W thermal load and $100^{\circ}C$ coolant inlet temperature. The effects of fill charge ratio, coolant inlet temperature and operating temperature on thermal performance of heat pipe were analyzed and discussed.

• Journal of the Korean Chemical Society
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• v.27 no.4
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• pp.262-267
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• 1983

A study has been carried out for a scheme of the selective extraction and determination of cesium, sodium and nuclides by the ion association with dicyclohexyl-24-crown-8 and sodium tetraphenylborate from primary coolant of a pressurized water reactor. For that purpose, the effects of hydrogen, cesium and borate ions on the extraction have been investigated. Interferences of iodine and xenon nuclides were found but could be removed by reducing with sodium thiosulfate and back extraction with 1 N hydrochloric acid solution, respectively.