Jae-Hyuk Eoh

SODIUM-CO2 INTERACTION IN A SUPERCRITICAL CO2 POWER CONVERSION SYSTEM COUPLED WITH A SODIUM FAST REACTOR

Abstract For the supercritical CO2 Brayton cycle of a sodium-cooled fast reactor, we carried out surface reaction tests for sodium temperatures ranging from 200 to 600°C. Based on the test results, we found that the reaction kinetics over the sodium temperature range of 300 to 550°C depends heavily on the temperature but is not sensitive to the velocity of CO2 flowing over the gas-liquid reacting interface explored in this study. Gaseous and nongaseous reaction products were sampled and analyzed quantitatively. The rates of the chemical reaction were determined by measuring the gas concentration of the CO/CO2 mixture. Then, we proposed a two-zone reaction model with a threshold temperature of 460°C. The kinetic parameters for each reaction zone were experimentally obtained.

Wastage and Self-Plugging by a Potential CO2 Ingress in a Supercritical CO2 Power Conversion System of an SFR

For a CO2 ingress accident into liquid sodium in a supercritical CO2 power conversion system coupled with a sodium-cooled fast reactor, we investigated two major design issues: i) a wastage phenomenon in regard to structural damage adjacent to the leaking position, and ii) potential channel plugging due to the formation of a particulate reaction product. In order to understand the factors affecting the occurrence of these issues, two kinds of experiments were carried out: a wastage effect test and a self-plugging test. All experimental conditions were chosen to reasonably represent the normal operating conditions and realistic design parameters of the reference plant. The test results indicate the absence of wastage, which will not lead to additional tube ruptures and damage propagation. In the current experiment, the self-plugging of PCHE channels only took place under two limited conditions: i) the sodium temperature is over 500°C and ii) the equivalent diameter of the crack opening is less than 1.5mm with a small leakage rate of far less than 1 g/s of CO2 ingress.

New Design Options Free from a Potential Sodium Freezing Issue for a Passive DHR System of KALIMER

Abstract To enhance the operational reliability of a purely passive decay heat removal system in KALIMER, which is named PDRC, three design options to prevent sodium freezing in an intermediate decay heat removal circuit were proposed, and their feasibilities have been studied for an entire plant operation mode. The potential candidates for the new design options are (a) the partially immersed DHX concept, (b) the cavity air cooling system–coupled PDRC concept, and (c) the advanced PDRC concept with alternative cooling medium. The design features of each concept are quantitatively evaluated in this study. For all the options, more specific design considerations were made to confirm their feasibility to properly materialize their concepts in a practical system design procedure, and the general definitions for a purely passive concept and its design features have been discussed as well.

High-Temperature Design of Sodium-to-Air Heat Exchanger in Sodium Test Loop

KAERI가 개발 중인 소듐냉각 원형로의 붕괴열 제거를 위한 잔열제거계통(Decay heat Removal Circuit: DRC)은 안전등급으로 분류되며, 이는 설계의 다양성 확보를 위해 Fig. 1 에서와 같이 피동형 잔열제거계통(Passive DRC : PDRC)과 능동형 잔열제거계통(Active DRC : ADRC)으로 구성된다. 소듐대 공기 열교환기(sodium-to-air heat exchanger) 에는 ADRC 에 설치되는 핀형(finned) 소듐대 공기 열교환기인 FHX(Finned-tube Sodium-to-Air Heat Exchanger)와 PDRC에 설치되는 헬리컬형 소듐대 공기 열교환기인 AHX(Helical-coil Air Heat Exchanger)가 있다. 여기서는 이미Fig. 2 의 STELLA-1 소듐 시험루프 내에 설치된 AHX 와 KAERI 부지 내에 설치 예정인 Fig. 3 의 강제통풍형 소듐대 공기열교환기 Key Words: Sodium Test Loop(소듐 시험루프), Sodium-to-Air Heat Exchanger(소듐-공기열교환기), High Temperature Design(고온 설계), Creep-Fatigue(크리프-피로) 초록: 제4세대 소듐냉각 고속로에는 중간열교환기(IHX), 붕괴열제거 열교환기(DHX), 공기 열교환기(AHX), 핀형 소듐-공기 열교환기(FHX) 및 증기발생기(SG)를 포함한 다양한 열교환기들이 설치된다. 본 연구에서는 STELLA-1 시험루프에 설치된 소듐-공기 열교환기인 AHX와 SELFA 시험루프에 설치될 핀형(finned) 소듐-공기 열교환기인 FHX 등 2 기의 열교환기 설계에 대해 3D 상세 유한요소해석을 수행하고, 동 결과에 기초하여 고온설계 기술기준을 따라 크리프-피로 손상평가를 수행하였다. 손상 평가결과 AHX 와 FHX 는 의도하는 크리프 피로 손상 하중 하에서 구조 건전성을 유지하는 것으로 확인되었다. Abstract: In a Korean Generation IV prototype sodium-cooled fast reactor (SFR), various types of high-temperature heat exchangers such as IHX (intermediate heat exchanger), DHX (decay heat exchanger), AHX (air heat exchanger), FHX (finned-tube sodium-to-air heat exchanger), and SG (steam generator) are to be designed and installed. In this study, the high-temperature design and integrity evaluation of the sodium-to-air heat exchanger AHX in the STELLA-1 (sodium integral effect test loop for safety simulation and assessment) test loop already installed at KAERI (Korea Atomic Energy Research Institute) and FHX in the SEFLA (sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger) test loop to be installed at KAERI have been performed. Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two heat exchangers according to the high-temperature design codes, and the integrity of the high-temperature design of the two heat exchangers was confirmed.

Feasibility study of a passive DHR system with heat transfer enhancement mechanism in a lead-cooled fast reactor

The KLFR is a pool-type lead-cooled fast reactor, which has a core thermal output of 900 MW(thermal), and a reactor vessel auxiliary cooling system (RVACS) is employed to secure reliable decay heat removal (DHR) during the worst anticipated design-basis condition. Since the RVACS design is based on reliable and economic considerations, a sufficiently large DHR capacity and compact reactor vessel size are desirable. However, these two requirements compete with each other because a sufficient DHR capacity can be achieved by a larger vessel size with a consequential heavy lead coolant weight. An advanced RVACS concept that has a larger capacity with a more compact vessel size was developed. To increase the DHR capacity of the KLFR, which uses natural-air circulation cooling, the feasibility of heat transfer enhancement by introducing new design concepts to essentially reduce the heat transfer resistance of the radial heat transfer elements was investigated. As a result of this work, the parametric analysis results showed that the passive DHR capacity of the KLFR can be substantially increased by up to 24% when compared with the classical RVACS concept, and this feature makes a compact reactor vessel very feasible. With the proposed advanced RVACS concept, one could expect that the heat removal capacity of an RVACS-type passive DHR system will be increased.

Development and Experimental Verification of the Numerical Simulation Method for the Quasi-Steady SWR Phenomena in an LMR Steam Generator

Abstract A quasi-steady system analysis of the sodium-water reaction (SWR) phenomena in a liquid-metal reactor (LMR) was performed using the Sodium-water reaction Event Later Phase System Transient Analyzer (SELPSTA) computer simulation code. The code has been formulated by implementing various physical assumptions to simplify the complex SWR phenomena, and it adopts the long-term mass and energy transfer (LMET) model developed in the present study. The LMET model is based on the hypothesis that the system transient can be described by the pressure and temperature transient of the cover gas space, and it can be applied only to the reaction period characterized by bulk motion. To evaluate the feasibility of the physical model and its assumptions, a scale-down mock-up test was carried out, and it was demonstrated that the numerical simulation using the LMET model adequately replicates the overall phenomena of the experiment with reasonable understanding. Based on the findings, as a numerical example, the long-term system transient responses during the SWR event of the Korea Advanced LIquid MEtal Reactor (KALIMER) were investigated, and it was found that the long-term dynamic responses are strongly dependent on the design parameters and operational strategies. As a result, the numerical simulation method developed in the present study is practicable; furthermore, the SELPSTA code is useful to resolve the risk for the SWR event.

Numerical investigation on the long-term System Transient response of a SWR Event in a LMR

In order to investigate the later phase of a sodium-water reaction (SWR) event, the code SELPSTA (Sodium-water reaction Event Later Phase System Transient Analyzer) has been developed and the analysis for the long-term system dynamic responses of a SWR event in KALIMER (Korea Advanced Liquid MEtal Reactor) has been made. The SELPSTA code uses the very simple analysis model applied only to the reaction period characterized by a bulk motion, and makes the very quick and concise computation possible. The code reasonably predicts the quasi-steady system transients and has the superiority in the aspect that the various design parameters or operational characteristics are flexibly applicable. In the long-term period of a SWR event, the system dynamic responses analyzed by the code totally depend on the system design parameters such as the breaking pressure of the rupture disk, the variation of the steam injection rate and the sodium drain tank pressure,etc. Based on these analyses results, it is expected that the numerical quantification method of the SELPSTA code is practicable for the long-term system transient analysis and also makes the design of a pressure relief system against a SWR event in a liquid metal reactor (LMR) possible.

High Temperature Design and Damage Evaluation of Mod.9Cr-1Mo Steel Heat Exchanger

High temperature design and evaluation of creep-fatigue damage for sodium-sodium heat exchanger, DHX (Decay heat exchanger) in a sodium test loop have been conducted. The DHX is a shell- and tube-type heat exchanger with outer diameter of 21.7mm, thickness of 1.65mm and effective length of 1.73m. The DHX shell and tube materials were Mod.9Cr-1Mo steel. The temperatures of shell inlet and shell outlet in the DHX are 510°C and 308°C, respectively, while the temperatures of tube inlet and outlet are 254°C and 475°C, respectively. Three dimensional finite element analysis was conducted for the DHX and evaluation of creep-fatigue damage at several critical locations of the heat exchanger was carried out according to the elevated temperature design codes of the ASME Section III Subsection NH and RCC-MR. Evaluations on the integrity of the DHX and code comparisons were carried out for the critical locations of the DHX.Copyright

Transient Performance Analysis of the Reactor Pool in KALIMER-600 with an Inertia Moment of a Pump Flywheel

The effect of an inertia moment of a pump flywheel on the thermal-hydraulic behaviors of the KALIMER-600(Korea Advanced LIquid MEtal Reactor) reactor pool during an early-phase of a loss of normal heat sink accident was investigated. The thermal-hydraulic analyses for a steady and a transient state were made by using the COMMIX-1AR/P code. In the present analysis a quarter of the reactor geometry was modeled in a cylindrical coordinate system, which includes a quarter of a reactor core and a UIS, a half of a DHX and a pump and a full IHX. In order to evaluate the effects of an inertia moment of the pump flywheel, a coastdown flow whose flow halving time amounts to 3.69 seconds was supplied to a natural circulation flow in the reactor vessel. Thermal-hydraulic behaviors in the reactor vessel were compared to those without the flywheel equipment. The numerical results showed a good agreement with the design values in a steady state. It was found that the inertia moment contributes to an increase in the circulation flow rate during the first 40 seconds, however to a decrease of it there after. It was also found that the flow stagnant region induced by a core exit overcooling decelerated the flow rate. The appearance of the first-peak temperature was delayed by the flow coastdown during the initial stages after a reactor trip.

Design methodology for insulating and cooling of a small modular reactor head by high temperature structural-thermal-fluid analysis

ABSTRACT The design of a sodium-cooled fast reactor (SFR) head can be complicated due to its shape and functions. The head is a component placed in the pressure boundary to shield nuclear radioactive radiation. At the same time, it needs to seal the reactor vessel, support penetrating components, and minimize heat losses. This paper presents a new insulating and cooling design concept of a small SFR head. For a new design, this study shows a comprehensive design approach considering fluid-thermal-structural computations. The interactive design approach refers to dependent simulation steps of three-dimensional (3D) thermal-structural, one-dimensional (1D) heat-transfer, and 3D computational fluid dynamics (CFD) analysis. This multi-domain approach was applied to the head of the large sodium integral effect test facility called sodium test loop for safety simulation and assessment (STELLA-2). And the STELLA-2 head design was proposed as a thick plate with a sandwich type of insulation, cooling the perimeter annulus of the round head-top surface. For the structural design, the ASME design code was utilized, and the head temperature of 346 °C was calculated as its initial design temperature target. In an axial heat-transfer mode from the in-vessel to the head, a 1D finite element model gave 57 and 75 mm insulation thicknesses with a thermal conductivity of 0.07 W/m/K. The cooling effectiveness of the proposed head design was shown through a commercial CFD package.