• Nuclear Engineering and Technology
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• 제24권3호
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• pp.243-251
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• 1992

소형 냉각재 상실사고시 루프밀봉 형성 및 제거에 대하여 LSTF에서 수행된 실험 SB-CL-18의 결과를 RELAP5/MOD2와 /MOD3를 이용하여 예측하였다. 본 연구는 증기발생기 상향 및 하향 유동에서의 비대칭 냉각재수용에 따른 마노메트릭 유동에 의해 노심노출의 조기발생을 야기시키는 열수력학적 현상을 예측하기 위하여 수행되었다. RELAP5/MOD2를 이용한 해석결과는 루프밀봉 형성 및 제거를 포함하여 감압사고시의 주요 현상을 전반적으로 잘 예측하고 있으나 기초 계산외 결과를 볼 때 현상 및 시간적 순서에 관련하여 몇 가지의 차이가 있었다. RELAP5/MOD3는 RELAP5/MOD2보다 전반적인 현상, 특히 증기발생기 액체수용을 보다 잘 예측하고 있으며, 또 한 RELAP5/MOD3를 이용하여 증기발생기 U자관과 펌프 흡입관의 nodalization수를 늘린 경우는 루프 밀봉제거현상과 시간적 순서를 잘 예측할 수 있었다.

• 한국항공우주학회지
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• 제43권9호
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• pp.822-829
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• 2015

본 논문에서는 압전재료를 이용하여 중소형 무인기 브레이크 시스템에 적용 가능한 압전유압펌프를 설계 및 제작하고, 제작된 압전유압펌프의 성능검증 실험을 수행하였다. 중소형 무인기급 목표 항공기를 선정하여 브레이크 시스템의 요구조건을 분석하였으며, 이를 바탕으로 압전유압펌프의 성능 요구조건을 선정하였다. 요구조건을 만족하는 압전유압펌프의 형상설계를 수행하였으며, 고속 동작 조건에서도 유체의 역류를 효과적으로 차단시킬 수 있는 체크밸브를 비롯한 모든 구성품을 제작하였다. 제작된 압전유압펌프의 성능검증을 위해 실험장치를 구성하여 무부하 토출특성, 부하 시 압력형성 특성실험을 수행하였다. 실험결과, 무부하 최대 토출유량은 80 Hz에서 120.04 cc/min이고, 부하 시 최대 토출압력은 140 Hz에서 783.17 psi이고, 압력형성 반응속도는 약 30 ms 이내임을 확인하였다. 이는 설계 제작된 압전유압펌프가 펌프성능 요구조건을 충족하고 있다고 판단된다.

• 설비공학논문집
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• 제13권11호
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• pp.1065-1071
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• 2001

To investigate hydraulic and thermal characteristics of ice slurries in a circular tube, ice slurries were tested in a flow loop with a constant heat flux test section, for ranges of flow velocity, ice fraction and heat flux. Heat transfer coefficients and friction factors of ice slurries were calculated by measuring the outer wall temperatures of the test section and the pressure drops over the test section. Heat transfer coefficients of ice slurries were 9% higher than the heat transfer coefficients expected by Petukhov. Friction factors were about 4% lower than the friction factors expected by Petukhov. The effective thermal capacity of ice slurry with 12.8% ice fraction, was found to be about 3 times higher than the thermal capacity of water.

• 대한기계학회논문집B
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• 제40권3호
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• pp.165-172
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• 2016

노심보충탱크(Core Makeup Tank, CMT), 안전주입탱크(SafetyInjection Tank, SIT)와 자동감압계통(Auto Depressurization System, ADS)로 구성된 1 계열의 SMART 피동안전주입계통의 주입특성을 파악하기 위한 소형냉각재상실사고(SBLOCA) 모의에 대한 실험적 연구가 수행되었다. SBLOCA의시험은 0.4 인치 안전주입수 배관파단에 대해 수행되었으며, 정상상태 조건은 실험요건서에 제시된 시험 초기 조건을 만족시키도록 746초 동안 운전되었다. 노심 출력 및 안전주입 유량 등의 경계 조건도 적절히 모의되었으며, 안전주입계통 배관에서의 파단, 히터 트립 및 잔열곡선 인가, 원자로냉각재펌프 관성서행(Coastdown), 급수 중단, CMT 및 SIT의 주입, ADS #1 개방이 SBLOCA 시나리오에 따라 적절히 모의되었다. 노심지지원통 내부의 액체환산수위는 파단 초반에 감소하다가 CMT와 SIT가 주입되면서 서서히 회복되었으며, 피동안전주입계통의 주입유량이 노심 수위를 회복하기에 충분한 것으로 판단할 수 있다.

• 대한기계학회논문집A
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• 제33권2호
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• pp.179-184
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• 2009

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.928-940
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• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.829-841
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• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.745-750
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• 1996

An analysis of a loss of residual heat removal system (RHRS) event during midloop operation after reactor shutdown was performed using the RELAP5/MOD3 thermal-hydraulic computer code. The experimental data of a 5% cold leg break test conducted at the ROSA-IV Large Scale Test Facility (LSTF) to simulate a main coolant pump shaft seal removal event during midloop operation of a Westinghouse-type PWR were used in the analysis. The predicted core boiling time and the peak primary system pressure showed good agreements with the measured data. Some differences between the calculational results and the experimental results were, however, found in areas of the timing of loop seal clearing and the temperature distribution in a pressurizer. Other calculational problems identified were discussed as well.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.719-726
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• 2012

The object of this work is the validation and assessment of the TRACE v5.0 code using the scaled test ATLAS1 facility in the context of a DVI2 line break. In particular, the experiment selected models the 50%, 6-inch break of a DVI line. The same experiment was also adopted as a reference test in the ISP-503. The ISP-50 was proposed to, and accepted by, the OECD/NEA/CSNI due to its technical importance in the development of a best-estimate of safety analysis methodology for DVI line break accidents. In particular, the behavior of the two-phase flow in the upper annulus downcomer was expected to be complicated. What resulted was the need for relevant models to be implemented into safety analysis codes, in order to predict these thermal hydraulic phenomena correctly.

• International Journal of Automotive Technology
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• 제8권2호
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• pp.203-209
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• 2007

As the vehicle electronic control technology quickly grows and becomes more sophisticated, a more efficient means than the traditional in-vehicle driving test is required for the design, testing, and tuning of electronic control units (ECU). For this purpose, the hardware-in-the-loop simulation (HILS) scheme is very promising, since significant portions of actual driving test procedures can be replaced by HIL simulation. The HILS incorporates hardware components in the numerical simulation environment, and this yields results with better credibility than pure numerical simulations can offer. In this study, a HILS system has been developed for ESP (Electronic Stability Program) ECUs. The system consists of the hardware component, which that includes the hydraulic brake mechanism and an ESP ECU, the software component, which virtually implements vehicle dynamics with visualization, and the interface component, which links these two parts together. The validity of HIL simulation is largely contingent upon the accuracy of the vehicle model. To account for this, the HILS system in this research used the commercial software CarSim to generate a detailed full vehicle model, and its parameters were set by using design data, SPMD (Suspension Parameter Measurement Device) data, and data from actual vehicle tests. Using the developed HILS system, performance of a commercial ESP ECU was evaluated for a virtual vehicle under various driving conditions. This HILS system, with its reliability, will be used in various applications that include durability testing, benchmarking and comparison of commercial ECUs, and detection of fault and malfunction of ESP ECUs.