• 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.

• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.559-570
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• 2004

A computational fluid dynamics model for predicting moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard $k-{\varepsilon}$ turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by DFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are "mined-type", as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is $82.9^{\circ}C$ at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is $24.8^{\circ}C$.

• Journal of the Korean Society of Visualization
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• v.12 no.1
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• pp.18-24
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• 2014

The Korea Atomic Energy Research Institute (KAERI) has a scaled-down moderator test program to simulate the CANDU6 moderator circulation phenomena during steady state operation and accident conditions. In the present work a preliminary experiment using a 1/8 scaled-down moderator tank has been performed to identify the potential problems of the flow visualization and measurement in the scaled-down moderator test facility. With a transparent moderator tank model, a velocity field is measured with a Particle Image Velocimetry (PIV) technique under an isothermal state. The flow patterns from the inlet nozzles to the top region of the tank are investigated using PIV for a 1/8 scale moderator tank.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.116-116
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.488-489
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• 2005

• Journal of the Korean Society of Visualization
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• v.10 no.3
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• pp.30-36
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• 2012

We are planning to construct a scaled-down moderator facility to simulate the CANDU-6 moderator circulation phenomena during steady state operating and accident conditions. In the present work a preliminary experiment using a 1/40 scaled-down moderator tank has been performed to investigate the anticipated problems of the flow visualization and measurement in the planning scaled-down moderator facility. We shortly describe CFD analysis result for the 1/40 scaled-down test model and the flow measurement techniques used for this test facility under isothermal flow conditions. The Particle Image Velocimetry (PIV) method is used to visualize and measure the velocity field of water in a transparent Plexiglas tank. Planar Laser Induced Fluorescence (PLIF) technique is used to evaluate the feasibility of temperature field measurement in the range of $20-40^{\circ}C$ of water temperature using an one-color method.

• 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.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.602-603
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• 2005

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.900-906
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• 2024

The Passively-Cooled Molten Salt Fast Reactor (PMFR) is one of the advanced design concepts of the Molten Salt Fast Reactor (MSFR) which utilizes a natural circulation for the primary loop and aims to attain a long-life operation without any means of fuel reprocessing. For an extended operation period, it is necessary to have enough fissile material, i.e., high excess reactivity, at the onset of operation. Since the PMFR is based on a fast neutron spectrum, direct implementation of a burnable absorber concept for the control of excess reactivity would be ineffective. Therefore, a localized moderator concept that encircles the active core has been envisioned for the PMFR which enables the effective utilization of a burnable absorber to achieve low reactivity swing and long-life operation. The modified PMFR design that incorporates a moderator and burnable absorber is presented, where depletion calculation is performed to estimate the reactor lifetime and reactivity swing to assess the feasibility of the proposed design. All the presented neutronic analysis has been conducted based on the Monte Carlo Serpent2 code with ENDF/B-VII.1 library.