Juhyeon Yoon

Development of a Computer Code, ONCESG, for the Thermal-Hydraulic Design of a Once-Through Steam Generator

Development of the conceptual design of a 300 MWt integral reactor, SMART (System-integrated Modular Advanced ReacTor), for utilization in nuclear cogeneration plants has been completed at the Korea Atomic Energy Research Institute (KAERI). The major primary components of the SMART such as modular helical steam generators, main circulation pumps and a self regulating pressurizer are integrated into a reactor vessel. It is a common practice to employ a once-through steam generator in integral reactor designs because of its advantages in compactness and simplicity of the flow path arrangements. In this study, a thermal hydraulic design and performance analysis computer code for a once-through steam generator using helically coiled tubes, ONCESG, is developed. To benchmark the developed physical models and computer code, once-through steam generators developed by other designers are simulated and ONCESG calculated results are compared with the design data. The overall characteristics of heat transfer area, pressure and temperature distributions calculated by ONCESG showed general agreements with the published data and it is demonstrated that the ONCESG code can be utilized for diverse purposes, such as, sensitivity analyses and optimum thermal design of a once-through steam generator.

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.

Structural Integrity Confirmation of a Once-through Steam Generator from the Viewpoint of Flow Instability

Helically-coiled once-through steam generators have been utilized for an integral type reactor showing several benefits such as high quality steam generation, geometric compactness, and compensation for a thermal expansion. Steam generator operations with unstable two-phase flow conditions on the tube-side may cause degradation of the tube materials and curtail the lifetime of the component. Based on existing experimental results for a once-through steam generator, its structural integrity was confirmed from the viewpoint of flow instability. The work was composed of three items, the prevention of static instability between the module steam/feedwater pipes, tube inlet orifice sizing against a dynamic instability between the heated coils, and a thermal-cyclic stress analysis for an overall component lifetime evaluation. The static thermo-hydraulic calculation for the steam generator cassette showed that while the prevention of the static instability was satisfied for the power operational mode, special care should be taken during the startup/cooling operational modes. The tube inlet orifice size was determined based on the orifice coefficient concept and existing experimental data for once-through steam generators. The thermal-cyclic stress evaluation for the heated tube revealed that the maximum alternating stress intensity was lower than the allowable fatigue limit value of the tube material.

Design evaluation of decay tank for a pool-type research reactor from the required minimum flow residence time point of view

A decay tank shall be designed to provide enough flow residence time to ensure that the N-16 activity decreases before the coolant leaves the decay tanks shielding room. However, when a proper criterion for the flow residence time in a decay tank is not presented, the tank would be oversized/undersized. In this paper, design evaluation for a decay tank is performed by investigating the effect of the fluid distribution along the residence time on the total dose rate and the required minimum flow residence time. The evaluation is also carried out to resize the predesigned decay tank.

Analytic estimation of the impact pressure in a beam-tube due to water-hammer effect

Generated neutrons are transferred through the beam-tubes of a research reactor. When a beam-tube rupture occurs inside the pool, the pool water enters quickly into the tube, and hits the flange foil at the other end. Unless the foil endures the impact pressure, a multiple beam-tube-rupture accident will occur. Therefore, to determine the thickness of the end flange foil, the impact pressure has to be estimated. In this study, the maximum impact pressure is estimated analytically by considering the water hammer phenomenon. The result is then validated with computational fluid dynamic (CFD) simulations. For the simulations, a commercially available CFD code, ANSYS CFX, is used. The analytic solutions show good agreement with the CFD results.

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

Aperiodic Instability of a Once-Through Steam Generator with a Feedwater Line

Aperiodic (static) flow instability is an instability related to the change of a flow direction in individual steam generating U-shaped channels operating at given pressure difference. The nature of an aperiodic instability is close to a Ledinegg instability [1] related to the presence of multiple flows at the full hydraulic curve of a U-shaped channel. In this paper, the conditions for a reverse flow for a once-through steam generator (OTSG) with U-shaped modular feedwater line (MFL) are studied. From the results of the studies, it is revealed that the change of a flow direction in the MFL is due to the boiling of the feedwater in the downcomer branch of the U-shaped MFL and that multiple flows start in an area of the extremes corresponding to the minimum pressure difference of the hydraulic curves. Calculation models for predicting a threshold of an aperiodic instability for the OTSG of interest is proposed and the analysis results are compared with the experimental data.Copyright