• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.78-84
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• 2016

The FHX (Forced-draft sodium-to-air Heat Exchanger) employed in the ADHRS (active decay heat removal system) is a shell-and-tube type counter-current flow heat exchanger with M-shape finned-tube arrangement. Liquid sodium flows inside the heat transfer tubes and atmospheric air flows over the finned tubes. The unit is placed in the upper region of the reactor building and has function of dumping the system heat load into the final heat sink, i.e., the atmosphere. Heat is transmitted from the primary cold sodium pool into the ADHRS sodium loop via DHX (decay heat exchanger), and a direct heat exchange occurs between the tube-side sodium and the shell-side air through the FHX tube wall. This paper describes the DHRS and the structural design of the FHX.

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.109-114
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• 2014

After the Fukushima accident in 2011, it was revealed that nuclear power plant has the vulnerability to SBO accident and its extension situation without sufficient cooling of reactor core resulting core meltdown and radioactive material release even after reactor shutdown. Many safety systems had been developed like PAFS, hybrid SIT, and relocation of RPV and IRWST as a part of steps for the Fukushima accident, however, their applications have limitation in the situation that supply of feedwater into reactor is impossible due to high pressure inside reactor pressure vessel. The concept of hybrid heat pipe with control rod is introduced for breaking through the limitation. Hybrid heat pipe with control rod is the passive decay heat removal system in core, which has the abilities of reactor shutdown as control rod as well as decay heat removal as heat pipe. For evaluating the cooling performance hybrid heat pipe, a commercial CFD code, ANSYS-CFX was used. First, for validating CFD results, numerical results and experimental results with same geometry and fluid conditions were compared to a tube type heat pipe resulting in a resonable agreement between them. After that, wall temperature and thermal resistances of 2 design concepts of hybrid heat pipe were analyzed about various heat inputs. For unit length, hybrid heat pipe with a tube type of $B_4C$ pellet has a decreasing tendency of thermal resistance, on the other hand, hybrid heat pipe with an annular type $B_4C$ pellet has an increasing tendency as heat input increases.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.10
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• pp.1251-1259
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• 2013

In this study, high temperature design and creep-fatigue damage evaluation of a decay heat exchanger (DHX) in the decay heat removal systems of a sodium-cooled fast reactor (SFR) have been performed. Detail design and 3D finite element analysis have been conducted for the DHXs to be installed in active and passive decay heat removal systems in Korean Generation IV SFR, and the DHX installed in the STELLA-1(Sodium integral effect test loop for safety simulation and assessment) at KAERI (Korea Atomic Energy Research Institute). Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two Mod.9Cr-1Mo steel heat exchangers according to the elevated temperature design codes of ASME Section III Subsection NH and RCC-MR code. Code comparisons were made based on the creep-fatigue damage evaluation and issues on conservatisms of the design codes were discussed.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.276-286
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• 1999

A computer code PARS2 is developed for the analysis of PSDRS, which is the safety grade RHRS of HAMMER, and applied to the investigation of the relation between design parameters and performance of PSDRS. The concept of the heat transfer resistance network is applied in assessing the importance of the various heat transfer modes. From the analysis results, the qualitative relations between the PSDRS performance and design parameters are found and guidelines for the PSDRS design procedures are also proposed.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.654-664
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• 2019

The geometrical and electromagnetic variables of a rectangular-type magnetohydrodynamic (MHD) circulation system are optimized to solve MHD equations for the active decay heat removal system of a prototype Gen-IV sodium fast reactor. Decay heat must be actively removed from the reactor coolant to prevent the reactor system from exceeding its temperature limit. A rectangular-type MHD circulation system is adopted to remove this heat via an active system that produces developed pressure through the Lorentz force of the circulating sodium. Thus, the rectangular-type MHD circulation system for a circulating loop is modeled with the following specifications: a developed pressure of 2 kPa and flow rate of $0.02m^3/s$ at a temperature of 499 K. The MHD equations, which consist of momentum and Maxwell's equations, are solved to find the minimum input current satisfying the nominal developed pressure and flow rate according to the change of variables including the magnetic flux density and geometrical variables. The optimization shows that the rectangular-type MHD circulation system requires a current of 3976 A and a magnetic flux density of 0.037 T under the conditions of the active decay heat removal system.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.156.1-156
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• 2002

• Nuclear Engineering and Technology
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• v.39 no.6
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• pp.683-690
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• 2007

Because of increasing concerns regarding global warming and the longevity of oil and gas reserves, the importance of nuclear energy as a major source of sustainable energy is gaining recognition worldwide. To make nuclear energy truly sustainable, it is necessary to ensure not only the sustainability of the fuel supply but also the sustained availability of waste repositories, especially those for high-level radioactive waste (HLW). From this perspective, the effort to maximize the waste loading density in a given repository is important for easing repository capacity problems. In most cases, the loading of a repository is controlled by the decay heat of the emplaced waste. In this paper, a comparison of the decay heat characteristics of HLW is made among the various fuel cycle options. It is suggested that, for a future fast breeder reactor (FBR) cycle, the removal and burning of minor actinides (MA) would significantly reduce the heat load in waste and would allow for a reduction of repository size by half.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1321-1329
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• 2016

To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from $450^{\circ}C$ to $700^{\circ}C$ and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

• Nuclear Engineering and Technology
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• v.38 no.2
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• pp.129-138
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• 2006

The Gas-cooled Fast Reactor (GFR) is one of the six reactor concepts selected within the framework of the Generation IV initiative and is the reference concept for the Commissariat $\grave{a}$ l'Energie Atomique $(CEA^1)$. Two reactor unit sizes have been considered: 600 MWth and 2400 MWth. As far as thermal-hydraulics is concerned, reactor decay heat removal (DHR) proves to be a major issue. The CEA has conducted exploratory design studies to address this issue and a reference solution for the 600MWth reactor has been recommended.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3106-3116
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• 2022

The present work presents analysis results of the core design optimizations for an annular, prismatic High Temperature Gas-cooled Reactor (HTGR) with passive decay-heat removal features. Its thermal power is 100 MW_t and the operating temperature is 850 ℃ (1123 K). The neutronic calculations are done for the core with heterogeneous distribution of fuel and burnable poison particles (BPPs) to flatten the reactivity swing and power peaking factor (PPF) during the reactor operation as well as for control rod (CR) insertion into the core to restrain a small excess reactivity less than 1$. The next step of the study is done for evaluation of core reactivity coefficient of temperature.