• 한국진공학회지
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• 제13권2호
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• pp.47-53
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• 2004

국내 진공기술 기반을 구축하는 국가사업의 일환으로 게터펌프 성능평가장치를 개발하고 있다. 본격적으로 장치를 구성하기 전에 평가절차의 개발, 장치 설계요건 및 사양을 확정하기 위해 예비실험장치를 만들었다. 이 장치를 이용하여 밀봉형 기기에 들어가는 비교적 활성화 온도가 낮은 게터 소자들의 특성평가 절차를 만들고 이를 적용하여 수소, 일산화탄소, 질소 등에 대해 배기속도와 배기용량을 측정해 보았다. 주로 고순도 기체 공급장치용으로 국내에서 개발된 게터 소자와 램프용으로 널리 쓰이는 외국제품의 배기성능을 비교해 보았다.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.435-440
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• 1996

Large Loss-of-Coolant Accidents analyzed in Final Safety Analysis Reports are usually covered by Reactor Inlet Header. Reactor Outlet Header and Primary Pump Suction breaks as representative cases. In this study we analyze the total (guillotine) break of an Emergency Core Cooling System (ECCS) pipe located at the ECCS injection point into the Primary Heat Transport System (PHTS). It was expected that thermal-hydraulic behaviors in the PHT and ECC systems are different from those of a Reactor Inlet Header break, having an equivalent break size. The main purpose of this study is to get insights on the differences occurred between the two cases and to assess these differences from the phenomenon behavior point of view. It was also investigated whether the ECCS line break analysis results could be covered by header break analysis results. The study reveals that as the intact loop has almost the same behavior in both analyzed cases. broken loop behavior is different mostly regarding sheath temperature in the critical core pass and pressure decrease in the broken Reactor Inlet Header. Differences are also met in the ECCS behavior and in event sequences timings.

• 에너지공학
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• 제17권3호
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• pp.125-138
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• 2008

에너지원의 다양화와 효율적인 이용을 위하여 광역지역의 통합에너지관리 시스템 구축이 요구되고 있다. 본 보문에서는 이러한 필요성에 부응하여 광역에너지 네트워크 구축에 필요한 핵심기술을 발굴하여 이를 종합적으로 현장에 적용하는 기술개발 을 수행하고 있다. 이를 위해 IT기술과 접목한 광역 네트워크 열공급 최적화 통합시스템 구축 기술, 발생되는 배열을 이용한 열펌프와 같은 미활용에너지 이용기술, 다양한 열원을 이용한 열수송 및 축열 기술 등 새로운 시스템을 개발하고 이를 현장적용을 통해 경제성을 평가하여 최종적으로 실용화할 수 있는 사업화 모델에 대하여 기술하였다.

• 한국항공우주학회지
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• 제35권9호
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• pp.805-815
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• 2007

단일추력의 정상 작동 상태의 액체 로켓 엔진 시스템 모듈화 프로그램을 작성하기 위한 선행 연구로 엔진 주요 구성품들에 대한 성능설계 프로그램을 작성하였다. 주요 구성품으로는 추력실, 원심형 펌프, 충동형 터빈, 재생 냉각 채널 등이 고려되었다. 복잡성을 피하기 위하여 열역학적 관계식 및 비점성 이론을 바탕으로 한 여러가지 관계식들과 간략한 수학적 모델을 사용하였다. 본 논문에서는 도출된 결과를 정성적으로 살펴보고, 주요 설계 파라미터를 바꿔가면서 구성품의 작동특성 변화에 대한 경향성을 검토함으로써 일반적인 구성품 설계 이론에 부합하는가를 확인하였다.

• Nuclear Engineering and Technology
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• 제41권10호
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• pp.1263-1274
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• 2009

Several experimental tests to simulate a reflood phase of a cold-leg LBLOCA of the APR1400 have been performed using the ATLAS facility. This paper describes the related experimental results with respect to the thermal-hydraulic behavior in the core and the system-core interactions during the reflood phase of the cold-leg LBLOCA conditions. The present descriptions will be focused on the LB-CL-09, LB-CL-11, LB-CL-14, and LB-CL-15 tests performed using the ATLAS. The LB-CL-09 is an integral effect test with conservative boundary condition; the LB-CL-11 and -14 are integral effect tests with realistic boundary conditions, and the LB-CL-15 is a separated effect test. The objectives of these tests are to investigate the thermal-hydraulic behavior during an entire reflood phase and to provide reliable experimental data for validating the LBLOCA analysis methodology for the APR1400. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results for the LBLOCA scenario of the APR1400. The ECC water flow rate from the safety injection tanks and the decay heat were simulated from the start of the reflood phase. The simulated core power was controlled to be 1.2 times that of the ANS-73 decay heat curve for LB-CL-09 and 1.02 times that of the ANS-79 decay curve for LB-CL-11, -14, and -15. The simulated ECC water flow rate from the high pressure safety injection pump was 0.32 kg/s. The present experimental data showed that the cladding temperature behavior is closely related to the collapsed water level in the core and the downcomer.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2002년도 학술대회지
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• pp.423-426
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• 2002

A custom micro-PIV optics assembly has been used to measure the flow field inside a T-shaped microchannel. The micro-PIV system consists of microscope objectives of various magnifications, a dichroic cube, and an 8-bit CCD camera. Fluorescent particles of diameters 620nm have been used with a Nd:YAG laser and color filters. A programmable syringe pump with Teflon tubings were used to inject particle-seeded distilled water into the channel at flow rates of $420,\;40,\;60{\mu}L/hr$. The microchannels are fabricated with PDMS with a silicon mold, then $O_2-ion$ bonded onto a slide glass. Results show differences in flow characteristics and resolution according to fluid injection rates, and magnifications, respectively. The results show PIV results with vector-to-vector distances of $2{\mu}m$ with 32 pixel-square interrogation windows at $50{\%}$ overlap.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2743-2759
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• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

• 한국안전학회지
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• 제31권6호
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• pp.129-134
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• 2016

Standard design of APR+(advanced power reactor plus) was certified at 2014 by Korea regulatory body. Based on the experience gained from OPR1000 and APR1400, the APR1400 was being developed as a 1,500MWe class reactor using Korean technologies for design code, reactor coolant pump, and man-machine interface system. APR+ has been basically designed to have the seismic design basis of safe shutdown earthquake (SSE) 0.3g, a 4-train safety concept based on N+2 design philosophy, and a passive auxiliary feedwater system (PAFS). Also, safety issues on the Fukushima-type accidents have been extensively reviewed and applied to enhance APR+ safety. APR+ provides higher reliability and safety against tsunami and earthquake. The purpose of this paper is to implement probabilistic safety assessment considering these design features and to analyze sensitivity of core damage frequency for large loss of coolant accident of APR+.

• 대한기계학회논문집B
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• 제33권4호
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• pp.272-277
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• 2009

The liquid phase LPG injection (LPLi) system (the 3rd generation technology) has been considered as one of the most promising fuel supply systems for LPG vehicles. To investigate the reaction characteristics of LPG with rubber parts in LPLi system, various rubbers were tested. The results showed that the amount of residue from the cover rubber of a fuel pump was increased about 10 times after testing. Furthermore, the amount of sulfur and nitrogen species which are considered as main sources of deposit formation in LPLi fuel injectors were also found to be higher than those in original LPG fuel. In addition, these residues made the core parts of LPLi injector such as needle and nozzle, partially worn, which eventually causes leakage in LPLi injectors.