• Proceedings of the Korea Institute of Applied Superconductivity and Cryogenics Conference
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• 1999.02a
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• pp.177-181
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• 1999

Cold Neutron Source(CNS) facility comprises moderator circulation system, helium cooling system, neutron guide and auxiliary sistems. To increase the amount of cold neutron, the thermal neutron should pass cold moderator at cryogenic temperature. As cold moderator in Hanaro, the liquid hydrogen or liquid deuterium will be used and the temperature in operation will be used and the temperature in operation will be maintained to be $250^{\circ}C$ below zero. To maintain the moderator at this cryogenic temperature. He refrigerator is used to cool it down in thermosiphon having natural circulation. As a part of the conceptual design of Hanaro CNS, study on the characteristics of moderators, design of moderator chanmber and cooling method were done through the collaboration of Korea Atomic Energy Research Institute and Petersburg Nuclear Physics Institute. During the collaboration, a program for the design of moderator cooling system design concept through the parametric study using this program. In the parametric study, the effect of the moderator type on the design parameters was investigated. Also, the requirements on the performance test for the cooling system, which will be made before the basic design, were investigated.

• Proceedings of the Korean Vacuum Society Conference
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• 2015.08a
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• pp.104-104
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• 2015

Due to the demand of the cold neutron flux in the neutron science and beam utilization technology, the cold neutron source (CNS) has been constructed and operating in the nuclear research reactor all over the world. The majority of the heat load removal scheme in the CNS is two-phase thermosiphon using the liquid hydrogen as a moderator. The CNS moderates thermal neutrons through a cryogenic moderator, liquid hydrogen, into cold neutrons with the generation of the nuclear heat load. The liquid hydrogen in a moderator cell is evaporated for the removal of the generated heat load from the neutron moderation and flows upward into a heat exchanger, where the hydrogen gas is liquefied by the cryogenic helium gas supplied from a helium refrigeration system. The liquefied hydrogen flows down to the moderator cell. To keep the required liquid hydrogen stable in the moderator cell, the CNS consists of an in-pool assembly (IPA) connected with the hydrogen system to handle the required hydrogen gas, the vacuum system to create the thermal insulation, and the helium refrigeration system to provide the cooling capacity. If one of systems is running out of order, the operating research reactor shall be tripped because the integrity of the CNS-IPA is not secured under the full power operation of the reactor. To prevent unscheduled reactor shutdown during a long time because the research reactor has been operating with the multi-purposes, the introduction of the standby cooling system (STS) can be a solution. In this presentation, the design considerations are considered how to design the STS satisfied with the following objectives: (a) to keep the moderator cell less than 350 K during the full power operation of the reactor under loss of the vacuum, loss of the cooling power, loss of common electrical power, or loss of instrument air cases; (b) to circulate smoothly helium gas in the STS circulation loop; (c) to re-start-up the reactor within 1 hour after its trip to avoid the Xenon build-up because more than certain concentration of Xenon makes that the reactor cannot start-up again; (d) to minimize the possibility of the hydrogen-oxygen reaction in the hydrogen boundary.

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.459-468
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• 2013

During a station blackout (SBO), the initiating event is a loss of Class IV and Class III power, causing the loss of the pumps, used in systems such as the primary heat transporting system (PHTS), moderator cooling, shield cooling, steam generator feed water, and re-circulating cooling water. The reference case of the SBO case does not credit any of these active heat sinks, but only relies on the passive heat sinks, particularly the initial water inventories of the PHTS, moderator, steam generator secondary side, end shields, and reactor vault. The reference analysis is followed by a series of sensitivity cases assuming certain system availabilities, in order to assess their mitigating effects. This paper also establishes the strategies to mitigate SBO accidents. Current studies and strategies use the computer code of the Integrated Severe Accident Analysis Code (ISAAC) for Wolsong plants. The analysis results demonstrate that appropriate strategies to mitigate SBO accidents are established and, in addition, the symptoms of the SBO processes are understood.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.161-167
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• 1994

This paper presents a concept of a pressure tube-type water-cooled reactor without the emergency core cooling system. It adopts an innovative fuel channel design using metallic fuel matrix to improve heat transfer from fuel to moderator at loss of coolant cooling. The heat produced in the fuel is cooled by the coolant system during normal operation, but by the passive moderator system at loss of coolant cooling including the loss-of-coolant accident(LOCA). Simple analysis shows that the fuel channel temperature can be maintained within the permissible range for both normal operation and a complete LOCA.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.9 s.252
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• pp.825-833
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• 2006

To investigate the moderator coolability for CANDU-6 reactors, a test facility (HU-KINS) has been manufactured as a 1/8 scaled-down of a calandria tank. In the design of the test facility, a scaling law was developed in such a way to consider the thermal-hydraulic characteristics of a CANDU-6 moderator. The proposed scaling law takes into consideration of the energy conservation, the dynamic similitude such as dimensionless numbers, Archimedes number (Ar) and Reynolds number (Re), and thermal-hydraulic properties similitude. Using this proposed scaling law, the thermal-hydraulic scaling analyses of similar test facilities such as the SPEL (1/10 scale) and the STERN (1/4 scale), have been identified. As a result, in the case of the SPEL, while the energy conservation is well defined, the similarities of Ar and the heat density are not well considered. As for the similarity of the STERN, while both the energy conservation and the characteristics of Ar are well defined, the heat density is not. In the meanwhile, the HU-KINS test facility with 1/8 length scaled-down is well similitude in compliance with all similarities of the energy conservation, the fluid dynamics and thermal-hydraulic properties. To verify the adequacy of the similarities in terms of thermal-hydraulics, a computational fluid dynamic (CFD) analysis has been conducted using the CFX-5 code. As the results of the CFD analyses, the predicted flow patterns and variation of axial properties inside the calandria tank are well consistant with those of previous studies performed with FLUENT and this implies that the present scaling method is acceptable.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.2
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• pp.189-196
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• 2012

The heavy moderator acts as the ultimate heat-sink in an operating CANDU reactor. HUKINS has been developed to investigate moderator flow patterns. HUKINS consists of a 38.4-mm-thick cylindrical shell with a 0.95 m inner diameter and 88 sus-tubes that produce a total heat of 10 kW. A chemical visualization method was selected to estimate the occurrence of typical moderator flow patterns. Momentum-dominated flow, mixed flow, and buoyancy-dominated flow are detected under conditions of a heat load of 7.7 kW and input mass flow rates of 4, 7, and 11 L/min. The experimental results are similar to the results of a CFD simulation that consisted of approximately 1.9 million grids and was conducted using the k-${\varepsilon}$ turbulence model. Therefore, both the present experiments and simulations using HUKINS, a 1/8-scale model, represent all three important flow patterns expected in the real CANDU6 reference reactor. Thus, it has been demonstrated that HUKINS could be useful in the study of CANDU6 moderator circulation.

• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.831-846
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• 2012

KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for local boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a system-scale code, MARS.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.678-699
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• 2015

Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for nearterm human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of $100MW_{th}$ and an electricity generation mode of $100MW_{th}$, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and thermohydraulics was carried out. The result indicates that the innovative design has great potential for high propellant efficiency and thrust-to-weight of engine ratio, compared with the existing NTR designs. However, the build-up of fission products in fuel has a significant impact on the bimodal operation of the moderated reactor such as xenon-induced dead time. This issue can be overcome by building in excess reactivity and control margin for the reactor design.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.547-556
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• 2014

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, $UO_2$ pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a $2{\times}10$ cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the $2{\times}10$ cask array, dose rates at the center point of the array and at the center of the casks' height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the $2{\times}10$ cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.