• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1802-1813
• /
• 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.

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

• Nuclear Engineering and Technology
• /
• v.41 no.6
• /
• pp.785-796
• /
• 2009

A KALIMER-600 concept which is a type of sodium-cooled fast reactor, has been developed at KAERI. It uses sodium as a primary coolant and is a pool-type reactor to enhance safety. Also, a supercritical carbon dioxide ($CO_2$) Brayton cycle is considered as an alternative to an energy conversion system to eliminate the sodium water reaction and to improve efficiency. In this study, a simplified model for analyzing the thermodynamic performance of the KALIMER-600 coupled with a supercritical $CO_2$ Brayton cycle was developed. To develop the analysis model, a commercial modular modeling system (MMS) was adopted as a base engine, which was developed by nHance Technology in USA. It has a convenient graphical user interface and many component modules to model the plant. A new user library for thermodynamic properties of sodium and supercritical $CO_2$ was developed and attached to the MMS. In addition, some component modules in the MMS were modified to be appropriate for analysis of the KALIMER-600 coupled with the supercritical $CO_2$ cycle. Then, a simplified performance analysis code was developed by modeling the KALIMER-600 plant with the modified MMS. After evaluating the developed code with each component data and a steady state of the plant, a simple power reduction and recovery event was evaluated. The results showed an achievable capability for a performance analysis code. The developed code will be used to develop the operational strategy and some control logics for the operation of the KALIMER-600 with a supercritical $CO_2$ Brayton cycle after further studies of analyzing various operational events.

• Nuclear Engineering and Technology
• /
• v.47 no.3
• /
• pp.253-259
• /
• 2015

The fast neutron reactor BOR-60 is one of the key experimental facilities worldwide to perform large-scale tests of fuel, absorbing, and structural materials for advanced reactors. The BOR-60 reactor was put into operation in December 1969, and by the end of 2014 it had been operating on power for ~265,000 hours. BOR-60 still demonstrates potential capabilities to extend the lifetime of sodium-cooled fast reactors. The BOR-60 lifetime should have expired at the end of 2014. Over the past few years, a great scope of work has been performed to justify the possibility of extending its lifetime. The work included inspection of the equipment conditions, calculations and experimental research on operating parameters and the conditions of nonremovable components, investigation of the structural material samples after their long-term operation under irradiation, etc. Based on the results of the work performed, the residual lifetime was evaluated and the reactor operator made a decision to extend the lifetime period of the BOR-60 reactor. After considering both a set of documents about the reactor conditions and the positive decision of independent experts, the Regulatory Authority of the Russian Federation extended the BOR-60 operating license up to 2020.

• Nuclear Engineering and Technology
• /
• v.41 no.4
• /
• pp.427-446
• /
• 2009

In order to meet the increasing demand for electricity, Korea has to rely on nuclear energy due to its poor natural resources. In order for nuclear energy to be expanded in its utilization, issues with uranium supply and waste management issues have to be addressed. Fast reactor system is one of the most promising options for electricity generation with its efficient utilization of uranium resources and reduction of radioactive waste, thus contributing to sustainable development. The Korea Atomic Energy Research Institute (KAERI) has been performing R&Ds on Sodium-cooled Fast Reactors (SFRs) under the national nuclear R&D program. Based on the experiences gained from the development of KALIMER conceptual designs of a pool-type U-TRU-10%Zr metal fuel loaded reactor, KAERI is currently developing Advanced SFR design concepts that can better meet the Generation IV technology goals. This also includes developing, Advanced SFR technologies necessary for its commercialization and basic key technologies, aiming at the conceptual design of an Advanced SFR by 2011. KAERI is making R&D efforts to develop advanced design concepts including a passive decay heat removal system and a supercritical $CO_2$ Brayton cycle energy conversion system, as well as developing design methodologies, computational tools, and sodium technology. The long-term Advanced SFR development plan will be carried out toward the construction of an Advanced SFR demonstration plant by 2028.

• Nuclear Engineering and Technology
• /
• v.39 no.3
• /
• pp.193-206
• /
• 2007

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

• Food Science of Animal Resources
• /
• v.43 no.2
• /
• pp.232-244
• /
• 2023

We investigated the effects of different phosphate replacements on the quality of ground pork products cured with sodium nitrite or radish powder to determine their potential for achieving clean-label pork products. The experimental design was a 2×5 factorial design. For this purpose, the ground meat mixture was assigned into two groups, depending on nitrite source. Each group was mixed with 0.01% sodium nitrite or 0.4% radish powder together with 0.04% starter culture, and then processed depending on phosphate replacement [with or without 0.5% sodium tripolyphosphate; STPP (+), STPP (-), 0.5% oyster shell calcium (OSC), 0.5% citrus fiber (CF), or 0.5% dried plum powder (DPP)]. All samples were cooked, cooled, and stored until analysis within two days. The nitrite source had no effect on all dependent variables of ground pork products. However, in phosphate replacement treatments, the STPP (+) and OSC treatments had a higher cooking yield than the STPP (-), CF, or DPP treatments. OSC treatment was more effective for lowering total fluid separation compared to STPP (-), CF, or DPP treatments, but had a higher percentage than STPP (+). The STPP (+) treatment did not differ from the OSC or CF treatments for CIE L^* and CIE a^*. Moreover, no differences were observed in nitrosyl hemochrome content, lipid oxidation, hardness, gumminess, and chewiness between the OSC and STPP (+) treatments. In conclusion, among the phosphate replacements, OSC addition was the most suitable to provide clean-label pork products cured with radish powder as a synthetic nitrite replacer.

• Journal of Advanced Marine Engineering and Technology
• /
• v.21 no.5
• /
• pp.527-534
• /
• 1997

Freezing of an aqueous sodium cWoride solution (Nacl- H20) saturating a packed bed with ini¬tial salt concentrations of 5, 10, 15(k by weight is investigated experimentally in a rectangular cav¬ity. The system was cooled from the top, bottom and a vertical side wall. For the freezing experi¬ments from below, there was little effect of the initial salt concentration throughout the freezing process, and the global freezing rate was not affected by the initial salt concentration. For the freezing from above, the size of the mush region decreased and the mushlliquid interface became flatter as the initial liquid concentration is decreased. For the freezing from vertical side wall, reheating of the mixture was intensified with an increase in the initial salt concentration. For Ci= 5%, supercooling was observed only at the early times of freezing process, but for Ci= 15% small supercooling was observed throughout the freezing process.

• Journal of Energy Engineering
• /
• v.12 no.4
• /
• pp.281-288
• /
• 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.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2009.11a
• /
• pp.51-52
• /
• 2009

Recently, several industrial countries using the fission energy have given attention to the Gen-IV SFR (sodium-cooled fast reactor) for achieving sustainable nuclear energy systems. In this context, an SFR is currently developed at the design concepts study stage in the Republic of Korea [Kim & Hahn 200909]. The sustainability of systems means economic, environment-friendly, proliferation-resistant, and safer systems. More specifically, this sustainability can be accomplished in terms of resource recycling and radioactive waste reduction. In the present work, the objective of fuel cycle cost modeling is to identify the impact of various conceptual options as a cost reduction measure for the Gen-IV SFR at the design concepts study stage. It facilitates the selection of several reasonable fuel cycle pathways for the future Gen-IV SFR from an economic viewpoint.