Kyoungwoo Seo

INVESTIGATION ON EFFECTS OF ENLARGED PIPE RUPTURE SIZE AND AIR PENETRATION TIMING IN REAL-SCALE EXPERIMENT OF SIPHON BREAKER

To ensure the safety of research reactors, the water level must be maintained above the required height. When a pipe ruptures, the siphon phenomenon causes continuous loss of coolant until the hydraulic head is removed. To protect the reactor core from this kind of accident, a siphon breaker has been suggested as a passive safety device. This study mainly focused on two variables: the size of the pipe rupture and the timing of air entrainment. In this study, the size of the pipe rupture was increased to the guillotine break case. There was a region in which a larger pipe rupture did not need a larger siphon breaker, and the water flow rate was related to the size of the pipe rupture and affected the residual water quantity. The timing of air entrainment was predicted to influence residual water level. However, the residual water level was not affected by the timing of air entrainment. The experimental cases, which showed the characteristic of partical sweep-out mode in the separation of siphon breaking phenomenon [2], showed almost same trend of physical properties.

Generation of Pressure Distribution Inside a Decay Tank in a Research Reactor Using CFD

The Primary Cooling System (PCS) of a research reactor circulates coolant to remove the heat produced in the fuel or irradiation device. The core outlet coolant contains many kinds of radionuclides because it passes the reactor core [1]. As N-16 among them emits a very hard gamma ray, it not only causes radiation damage to some components but also requires very heavy shielding of the primary cooling loop. Since its half-life is 7.13s, its level can be effectively lowered by installing a decay tank including an internal structure to provide a transit time [2]. To ensure that the N-16 activity decreases enough before the coolant leaves the heavily shielded decay tank room, perforated plates are installed inside the decay tank. The perforated plates are designed to disturb and delay the PCS flow. Normally, when a flow from a relative narrow inlet nozzle goes out to an enlarged tank, it becomes a complex turbulent flow inside the tank. In addition, the PCS flow is frequently changed from zero to a normal flow rate owing to the research reactor characteristics. Thus, the integrity of the perforated plate shall be verified with the pump operation and shutdown condition.Copyright

Design of a Discharge Header and a Working Platform for a Research Reactor With a CFD Model

Most research reactors are designed as an open-pool type and the reactor is located on the bottom of the open-pool. The reactor in the pool is connected to the primary cooling system, which is designed for adequate cooling of the heat generated from the reactor core. One of the characteristics of an open-pool type research reactor is that the primary coolant after passing through the reactor core and the primary cooling system (PCS) is returned to the reactor pool. Because the primary coolant contains many kinds of radionuclides, the research reactor should be designed to protect the radionuclides from being released outside the pool by a stratified stable water layer, which is formed between a hot water layer and cold water near the reactor and prevents the natural circulation of water in the pool. In this study, additional components such as a discharge header and a working platform inside the pool were developed to help diminish the radiation level to the pool top. To discharge coolant stably inside the reactor pool, a discharge header was installed at the end of the pool inlet pipe. Many holes were made in the discharge header to discharge the coolant slowly and minimize the disturbance of the hot water layer by the flow inside the pool. The working platform was also equipped inside the reactor pool to remove the convective flow near the pool top.The commercially available CFD code, ANSYS CFD-FLEUNT, was used to specifically design the discharge header and working platform for satisfying the requirement of the pool top radiation level. The computations were conducted to analyze the flow and temperature characteristics inside the pool for several geometries using an SST k-ω turbulent model and cell modeling, which were conducted to isolate the root cause of these differences and the given inlet conditions. The discharge header and working platform were designed using the CFD results.© 2013 ASME

A Study for Heat-Loss Characteristics of Hot-Water Layer by the Increment of Reactor Power

It is necessary to access the pool top area for loading and unloading irradiation test pieces by a required irradiation period during a normal operation of an open-pool-type research reactor installed on the bottom of reactor pool with a depth of about 13m. However, when the reactor pool top radiation level exceeds the limit of radiation level by the rising of reactor coolant contaminated by radioactivity due to a natural convection of the pool water, access to the reactor pool top area is denied. In the case of HANARO, a hot-water layer (HWL, hereinafter) is maintained below a depth of 1.2m from the top of the reactor pool in order to reduce the radiation level of the reactor pool top area by suppressing natural convection of the pool water.After normal operation of the HWL, the pool top radiation level is safely maintained below the limit of the pool top radiation level. For studying the characteristics of the HWL under downward flow pattern of the reactor coolant (hereafter as DFP), the heat loss of the HWL is calculated based on the model for HANARO HWL. The heat loss characteristics of the HWL were reviewed by reactor power level increment.This paper suggests a HWL heat loss relation formula by an increment of the reactor power level. It was confirmed that the total heat loss of the HWL under DFP of reactor coolant linearly increased larger than that of upward flow pattern (hereafter, UFP). The reason is that the bottom convection loss of DFP increased 4.5 times than that of UFP by the 10 times increment of the core bypass flow rate under DFP than that under the UFP.Copyright

A Study on the Convective Mass Transfer of Nitrogen to Water for a Gas Pressurizing System

Mass transfer due to a concentration difference of nitrogen can occur in a nuclear system. Our research work seeks to analyze and understand the mass transfer phenomena of nitrogen in water under the condition of a natural convection using the commercially available CFD computer model, FLUENT 6.3. The maximum solubility was employed to express the boundary condition at an interface between the nitrogen and water. First, the case that nitrogen was transferred to water by only a diffusion was simulated to verify the application of the UDS (User defined scalar) model in FLUENT 6.3 for a mass transfer. Diffusion equation, which was described as a PDE (Partial Differential Equation) with non-homogeneous boundary conditions, was solved and the solved results of the PDE showed a good agreement with those of the FLUENT simulation in the same condition. The same cylinder geometry with that of the diffusion case was used to estimate the convective mass transfer. By the natural convection caused by the thermal boundary condition, the mass transfer of nitrogen had a convection effect. The result of FLUENT 6.3 to compute the convective mass transfer showed that the nitrogen was transferred simultaneously in the entire region by the convection effect and it took about several hours until the mole fraction of nitrogen in the water side reached 50% of the maximum saturated value. The averaged mass transfer coefficient was calculated and compared with the results obtained from the heat and mass transfer analogy. The calculated coefficients showed the lower value than those obtained from the various correlations. When the steam mass transfer toward the gas side was negligible, the pressure drop of the gas side due to the reduced nitrogen caused by a mass transfer was computed using the ideal gas law and the Custom Field Function model in the FLUENT 6.3.Copyright