• 한국공작기계학회:학술대회논문집
• /
• 한국공작기계학회 2001년도 춘계학술대회 논문집(한국공작기계학회)
• /
• pp.441-446
• /
• 2001

This paper presents the thermal characteristics analysis of a high-speed HMC spindle system with angular contact ball bearings, built-in motor, oil-jet lubrication method, oil jacket cooling method, and so on. The spindle system is composed of the main spindle and sub-spindle which are mechanically connected by a flexible coupling. The spindles are supported by two front and rear bearings, and the built-in motor is located between the front and rear bearings of the sub-spindle. The thermal analysis model of spindle system is constructed by the finite element method, and the thermal characteristics in the design stage are estimated based on temperature distribution and heat flow under the various testing conditions related to material of bearing ball, spindle speed and coolant temperature.

• 한국공작기계학회논문집
• /
• 제13권1호
• /
• pp.69-75
• /
• 2004

This paper presents the thermal characteristics analysis of a high-speed motor-separated spindle system consisted of angular contact ball bearings and built-in motor with oil-jet lubrication. The spindle system is composed of the main spindle and sub-spindle which are mechanically connected by a flexible coupling. The spindles are supported by two front and rear bearings, and the built-in motor is located between the front and rear bearings of the sub-spindle. The thermal analysis model of spindle system is constructed by the finite element method, and the thermal characteristics in the design stage are estimated based on temperature distribution and heat flow under the various testing conditions related to material of bearing ball, spindle speed and coolant temperature.

• 에너지공학
• /
• 제10권3호
• /
• pp.294-298
• /
• 2001

사용 후 핵연료의 고독성 장수명 핵종을 저독성 단수명 핵종으로 변환시키기 위한 미임계 핵변환로 연구가 진행중이다. 본 논문에서는 이러한 미임계 핵변환로에서 사용될 표적 시스템을 설계하기 위하여 표적시스템에 대한 열적, 구조적 분석을 수행하였다. 표적시스템의 열수력 분석에서는 diffuse plate를 삽입함으로써 빔창의 냉각효과를 증대시킬 수 있었다. 또한, 주요 인자인 빔창두께, 빔출력, 냉각재 유량 변화에 따른 빔창의 열적, 구조적 건전성 분석을 수행하여 표적시스템의 설계치를 설정하였다. 본 설계조건 하에서 빔창의 최대 온도 및 음력은 허용가능한 범위에 있음을 확인하였다.

• 대한인간공학회:학술대회논문집
• /
• 대한인간공학회 1993년도 추계학술대회논문집
• /
• pp.21-30
• /
• 1993

This paper describes the development of a simulation model of nuclear power plant operators including cognitive aspects by using a network modeling soft ware, Micro-SAINT (System Analysis of Integrated Networks of Tasks) for the analysis of operator performance. Network model description based on Micro-SAINT includes tasks, resources, precedence relations among tasks, flow of information and PSFs (Performance Shaping Factors) on task performance. We have tried to evaluate the performance with several performance measures such as the number of tasks allocated, relative time presure among operators within a shift, for the selected test accident scenarior; small-break LOCA (Loss of Coolant Accident) in a PWR (Pressurized Water Reactor) type nuclear power plant.

• 한국자동차공학회논문집
• /
• 제14권5호
• /
• pp.9-16
• /
• 2006

The present study is to simulate coolant flow in IC engine cooling passages under subcooled nucleate boiling conditions and investigate thermal stress analysis of the solid part. To consider nucleate boiling heat transfer effect, Chen's empirical formula is used through user subroutine programing in CFD code and then nucleate boiling model is compared with Robinson's experimental results, which shows reasonable agreement. This Chen's nucleate boiling model is applied to single cylinder IC engine model and we do cylinder liner thermal stress analysis using commercial FEM code.

• 한국압력기기공학회 논문집
• /
• 제16권1호
• /
• pp.11-21
• /
• 2020

This study aims to develop and advance the evaluation technology for assessing PHWR safety. For this purpose, the complete loss of AC power or station blackout (SBO) was selected as a target accident scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes the main features of the primary heat transport system with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was achieved successfully by running the present model to check out the stable convergence of the key parameters. Subsequently, through the SBO transient analyses two cases with and without the coolant leakage via the PHTS pumps were simulated and the behaviors of the major parameters were compared. The sensitivity analysis on the amount of the coolant leakage by varying its flow area was also performed to investigate the effect on the system responses. It is expected that the results of the present study will contribute to upgrading the evaluation technology of the detailed thermal hydraulic analysis on the SBO transient of the operating PHWRs.

• Nuclear Engineering and Technology
• /
• 제53권12호
• /
• pp.3990-4002
• /
• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• 한국초전도ㆍ저온공학회논문지
• /
• 제18권1호
• /
• pp.64-68
• /
• 2016

High Temperature Superconductor power cable systems are being developed actively to solve the problem of increasing power demand. With increases in the unit length of the High Temperature Superconductor power cable, it is necessary to develop highly efficient and reliable cryogenic pumps to transport the coolant over long distances. Generally, to obtain a high degree of efficiency, the cryogenic pump requires a high pressure rise with a low flow rate, and a partial emission type pump is appropriate considering its low specific speed, which is different from the conventional centrifugal type, full emission type. This paper describes the design of a partial emission pump to circulate subcooled liquid nitrogen. It consists of an impeller, a circular case and a diffuser. The conventional pump and the partial emission pump have different features in the impeller and the discharge flow passage. The partial emission pump uses an impeller with straight radial blades. The emission of working fluid does not occur continuously from all of the impeller channels, and the diffuser allows the flow only from a part of the impeller channels. As the area of the diffuser increases gradually, it converts the dynamic pressure into static pressure while minimizing the loss of total pressure. We used the known numerical method for the optimum design process and made a CFD analysis to verify the theoretical performance.

• Nuclear Engineering and Technology
• /
• 제54권2호
• /
• pp.554-564
• /
• 2022

The natural circulation phenomena occurring in fully integrated nuclear reactors are associated with a unique formation mechanism. The phenomenon results from a structural feature of these reactors involving upward flow from the core, located in the central-bottom region of a single vessel, and downward flow to the steam generator in the annulus region. In this study, to understand the natural circulation in a single vessel involving a multi-layered flow path, single-phase and two-phase natural circulation tests were performed using the SMART-ITL facility, and validation analysis of the TASS/SMR-S code was performed by comparing the corresponding test results. Three single-phase natural circulation tests were sequentially conducted at 15%, 10%, and 5% of full-scaled core-power without RCP operation, following which a two-phase natural circulation test was successively conducted with an artificial discharge of coolant inventory. The simulation capability of the TASS/SMR-S code with respect to the natural circulation phenomena was validated against the test results, and somewhat conservative but reasonably comparative results in terms of overall thermalhydraulic behavior were shown.

• Nuclear Engineering and Technology
• /
• 제56권10호
• /
• pp.4397-4403
• /
• 2024

The multi-dimensional thermal-hydraulic phenomena in the downcomer of advanced pressurized water reactor with direct vessel injection system are the key points for the safety analysis during a loss of coolant accident. In order to improve the accuracy of LOCUST code for the predictions of thermal-hydraulic phenomena in downcomer region, some newly correlations have been implemented into LOCUST code. The wall friction model of LOCUST code was modified based on the correlations which developed by Yang. The interfacial friction models in LOCUST code have been modified as Hibiki-Ishii correlations. In addition, in order to simulate the upward flow of recirculation flow in downcomer region, the Kinoshita-Hibiki correlations have been also implemented into LOCUST code for better simulating the recirculation flow in downcomer region. The modified code was validated with experimental data of DOBO facility. Five tests of DOBO facility have been calculated by LOCUST, and the calculated axial void fraction distributions have been compared with the measurements. The results show that the modified LOCUST with new correlations of distribution parameter and drift velocity shows better accuracy than the original code. The deviations of the modified LOCUST code are less than the original code and are almost within ±20 %.