• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.476-483
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• 2017

The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of $0.22^{\circ}C/W$, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

• Journal of the Korean Society of Visualization
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• v.14 no.3
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• pp.15-21
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• 2016

The circular and flat heat pipe were experimentally investigated by using neutron imaging technique. This experimental study was performed at the DINGO of OPAL research reactor, Australia. The diameter of the circular heat pipe is 10 mm and the dimension of flat is $10(width){\times}3(thickness)mm2$, respectively. We used the distilled water as a coolant. The coolant distributions and 2-phase flow patterns were measured under heating conditions. Experimental results show that neutron imaging technique is a good tool to visualize the 2-phase flow and phenomena in the heat pipe. The coolant distributions and 2-phase flow patterns depend on installation posture of the heat pipe and volume ratio of the coolant. Finally, it was discussed to calculate the void fraction by neutron imaging technique.

• Nuclear Engineering and Technology
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• v.55 no.5
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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.

• Proceedings of the KSME Conference
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• 2001.06b
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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.

• The KSFM Journal of Fluid Machinery
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• v.10 no.5
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• pp.9-19
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• 2007

Fluid temperature fluctuations in a mixing tee pipe were numerically analyzed by LES model in order to clarify internal turbulent flows and to develope an evaluation method for high-cycle thermal fatigue. Hot and cold water with an temperature difference $40^{\circ}C$ were supplied to the mixing tee. Fluid temperature fluctuations in a mixing tee pipe is analysed by using the computational fluid dynamics code, FLUENT, Temperature fluctuations of the fluid and pipe wall measured as the velocity ratio of the flow in the branch pipe to that in the main pipe was varied from 0.05 to 5.0. The power spectrum method was used to evaluate the heat transfer coefficient. The fluid temperature characteristics were dependent on the velocity ratio, rather than the absolute value of the flow velocity. Large fluid temperature fluctuations were occurred near the mixing tee, and the fluctuation temperature frequency was random. The ratios of the measured heat transfer coefficient to that evaluated by Dittus-Boelter's empirical equation were independent of the velocity ratio, The multiplier ratios were about from 4 to 6.

• Applied Chemistry for Engineering
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• v.16 no.3
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• pp.434-439
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• 2005

The reaction of hydration proceeded at the same time the vapor was introduced into the reactor that was filled with calcined dolomite. It has shown that the temperature has begun to fall from the bottom of reactor after increase of temperature by the heat of hydration reaction. The reaction initiated at the pipe wall and the heat was transfer to the center of block between the fins. The results show that the use of copper fin in the reactor reducted the hydration reaction time by half when compared to the case without using the fins.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1409-1416
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• 2020

The study of heat transfer coefficient of supercritical water plays an important role in improving the heat transfer efficiency of the reactor. Taking the supercritical natural circulation experimental bench as the research object, the effects of power, flow, pipe diameter and mainstream temperature on the heat transfer coefficient of supercritical water were studied. At the same time, the experimental data of Chen Yuzhou's supercritical water heat transfer coefficient was collected. Through the factorial design method, the influence of different factors and their interactions on the heat transfer coefficient of supercritical water is analyzed. Through the corresponding analysis method, the influencing factors of different levels of heat transfer coefficient are analyzed. It can be found: Except for the effects of flow rate, power, power-temperature and temperature, the influence of other factors on the natural circulation heat transfer coefficient of supercritical water is negligible. When the heat transfer coefficient is low, it is mainly affected by the pipe diameter. As the heat transfer coefficient is further increased, it is mainly affected by temperature and power. When the heat transfer coefficient is at a large level, the influence of the flow rate is the largest at this time.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2315-2324
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• 2023

Space nuclear reactors are becoming popular in deep space exploration owing to their advantages of high-power density and stability. Following the fourth-generation nuclear reactor technology, a conceptual design of the dual drum-controlled space molten salt reactor (D²-SMSR) is proposed. The reactor concept uses molten salt as fuel and heat pipes for cooling. A new reactivity control strategy that combines control drums and safety drums was adopted. Critical physical characteristics such as neutron energy spectrum, neutron flux distribution, power distribution and burnup depth were calculated. Flow and heat transfer characteristics such as natural convection, velocity and temperature distribution of the D²-SMSR under low gravity conditions were analyzed. The reactivity control effect of the dual-drums strategy was evaluated. Results showed that the D²-SMSR with a fast spectrum could operate for 10 years at the full power of 40 kWth. The D²-SMSR has a high heat transfer coefficient between molten salt and heat pipe, which means that the core has a good heat-exchange performance. The new reactivity control strategy can achieve shutdown with one safety drum or three control drums, ensuring high-security standards. The present study can provide a theoretical reference for the design of space nuclear reactors.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.435-440
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• 1996

Large Loss-of-Coolant Accidents analyzed in Final Safety Analysis Reports are usually covered by Reactor Inlet Header. Reactor Outlet Header and Primary Pump Suction breaks as representative cases. In this study we analyze the total (guillotine) break of an Emergency Core Cooling System (ECCS) pipe located at the ECCS injection point into the Primary Heat Transport System (PHTS). It was expected that thermal-hydraulic behaviors in the PHT and ECC systems are different from those of a Reactor Inlet Header break, having an equivalent break size. The main purpose of this study is to get insights on the differences occurred between the two cases and to assess these differences from the phenomenon behavior point of view. It was also investigated whether the ECCS line break analysis results could be covered by header break analysis results. The study reveals that as the intact loop has almost the same behavior in both analyzed cases. broken loop behavior is different mostly regarding sheath temperature in the critical core pass and pressure decrease in the broken Reactor Inlet Header. Differences are also met in the ECCS behavior and in event sequences timings.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.793-802
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• 2024

In the framework of the European project SAMOSAFER, this numerical study focuses on some thermal aspects of the Emergency Draining Tank (EDT) located underneath the core of a Molten Salt Reactor. In case of an emergency, this tank passively receives the liquid fuel salt and is designed to ensure a subcritical state. An important requirement is that the fuel does not overheat to maintain the EDT Hastelloy container integrity. The present EDT is based upon a group of hexagonal cooling assemblies arranged in a hexagonal grid and cooled down thanks to conduction through the inert salt layer up to an air flow in charge of removing the heat. This numerical thermal study relies on a conjugated heat transfer analysis coupling a Finite Element solid thermal code (SYRTHES) and two instances of a Finite Volume CFD codes (Code_Saturne). Calculations on an initial design suggest that a simple center airpipe flow is likely to not sufficiently cool the device. Alternative solutions have been evaluated. Introduction of fins to enhance the heat transfer do not bring a noticeable improvement regarding maximum temperature reached. However, a solution in which the central pipe air flow is replaced by several cooling channels located closer to the fuel is investigated and suggests a better cooling.