• Nuclear Engineering and Technology
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• 제49권8호
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• pp.1617-1628
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• 2017

In the passive residual heat removal system of a molten salt reactor, one of the residual heat removal methods is to use the thimble-type heat transfer elements of the drain salt tank to remove the residual heat of fuel salts. An experimental loop is designed and built with a single heat transfer element to analyze the heat transfer and flow characteristics. In this research, the influence of the size of a three-layer thimble-type heat transfer element on the heat transfer rate is analyzed. Two methods are used to obtain the heat transfer rate, and a difference of results between methods is approximately 5%. The gas gap width between the thimble and the bayonet has a large effect on the heat transfer rate. As the gas gap width increases from 1.0 mm to 11.0 mm, the heat transfer rate decreases from 5.2 kW to 1.6 kW. In addition, a natural circulation startup process is described in this paper. Finally, flashing natural circulation instability has been observed in this thimble-type heat transfer element.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.654-664
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• 2019

The geometrical and electromagnetic variables of a rectangular-type magnetohydrodynamic (MHD) circulation system are optimized to solve MHD equations for the active decay heat removal system of a prototype Gen-IV sodium fast reactor. Decay heat must be actively removed from the reactor coolant to prevent the reactor system from exceeding its temperature limit. A rectangular-type MHD circulation system is adopted to remove this heat via an active system that produces developed pressure through the Lorentz force of the circulating sodium. Thus, the rectangular-type MHD circulation system for a circulating loop is modeled with the following specifications: a developed pressure of 2 kPa and flow rate of $0.02m^3/s$ at a temperature of 499 K. The MHD equations, which consist of momentum and Maxwell's equations, are solved to find the minimum input current satisfying the nominal developed pressure and flow rate according to the change of variables including the magnetic flux density and geometrical variables. The optimization shows that the rectangular-type MHD circulation system requires a current of 3976 A and a magnetic flux density of 0.037 T under the conditions of the active decay heat removal system.

• 한국환경보건학회지
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• 제21권4호
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• pp.17-23
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• 1995

As a basic experiment to develope biological pretreatment proces~ in water treatment, the experiments on biodegradability and isothermal adsorption of activated carbon were performed on refractories such as humic acid, $NH_3-N$, phenol and ABS which caused the problems in drinking water treatment. Also, the treatabilities on humic acid were examined in the continuous flow type reactors. The removal efficiencies of humic acid, $NH_3-N$, phenol and ABS in the biodegradable experiments for 5 days were 20.1%, 73.4%, 91.7% and 97.5%, respectively. In the isothermal adsorption test of refractories on activated carbon to be used as a media in the continuous flow type reactors, ABS and phenol are adsorbed easily, but humic acid and $NH_3-N$ are difficult to be done. The removal efficiencies of humic acid in granular activated carbon(GAC) reactor were about 7-8% higher than in biological activated carbon(BAC) reactor. The removal efficiencies of humic acid in biological fluidized bed(BFB) reactor were about 30% in GAC media, but were almost zero in sea sand media.

• 에너지공학
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• 제25권1호
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• pp.153-162
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• 2016

요르단 수출용 연구용 원자로(JRTR, Jordan Research and Training Reactor)에 사용되는 판형핵연료(Plate-type-fuel)의 부수로를 모의하는 좁은 사각 유로에서 하향류 유동 조건의 임계열유속(CHF, Critical Heat Flux) 실험 연구가 수행되었다. 실험은 연구용 원자로 인허가 요건을 만족하는 유동 조건에서 수행되었으며 크기가 다른 두 가지 시험부를 이용하였다. 두 시험부는 각각 피션몰리(Fission moly) 우라늄 타겟의 부수로와 판형핵연료의 부수로를 모의하며 모두 원형과 같은 크기로 제작되었다. 각 시험부에 대해 임계열유속 실험이 수행되었으며 이를 통해 임계열유속에 영향을 주는 변수를 분석하였다. 그리고 실험결과를 기존의 임계열유속 모델과 비교하였으며 모델의 예측 정확도를 제시하였다. 이를 통해 좁은 사각 유로 내 하향류 유동 조건에서의 임계열유속 예측에 대한 기존 모델의 적용성을 분석하였다.

• KSBB Journal
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• 제11권6호
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• pp.669-675
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• 1996

간단한 조작으로 시안화물을 신속하게 계측할 수 있는 센서의 개발을 목적으로 하여 미생물과 산소전극을 이용한 시안센서의 개발을 연구하였다. 산소전극에 S. cerevisiae를 흡착 고정화한 막을 부착하여 막형 시안센서를 제작하고 시료용액중의 시 안이온농도를 O~1.00ppm이 되게 시안을 첨가하면 서 전류값 변화를 살펴본 결과 시안농도 0.10~1.00ppm의 범위에서 응답이 확인되었으며 고정화 직후의 호흡활성은 약 하루정도 유지되었다. S. ceremSlae를 고정화한 키토-펼 HP-5020을 사용하여 반응기형 시안센서를 제작하고 막형 시안센서와 동일하게 시안이온농도 O~1.00ppm 용액의 전 류값 변화를 살펴본 결과 0.10~ 1.00ppm의 범위에서 응답이 확인되었으며 고정화 직후 90%의 호흡활 성이 16일간 유지되었다. 반응기형 시안샌서의 반응기 크기의 영향을 검토 한 결과, 반응기의 최적 체적이 존재한다는 사실을 알 수 있었다.

• 한국가스학회지
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• 제17권2호
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• pp.78-84
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• 2013

수증기 개질 반응로 설계를 위해 수증기 개질 반응로에 대한 열유동 해석을 수행하였다. 6개의 이중관형 개질기 튜브들과 1개의 버너로 구성된 반응로에 대해 개질반응과 열유동 해석을 연계하여 수치계산을 수행하였다. 버너 형상 변화에 따른 반응로 내에서의 유동구조를 계산하여 버너 형상을 선택하였다. 개질반응 통합해석 결과, 반응로 상하단내에서 온도구배가 크게 나타났으며 개질기 튜브 내의 조성 변화도 반응로 온도분포의 영향을 받는 것으로 나타났다. 또한, 운전조건인 SCR 및 GHSV 변동에 따른 개질 반응특성이 변화함을 확인할 수 있었다.

• 한국전산유체공학회지
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• 제9권2호
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• pp.11-17
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• 2004

CFD analysis was performed to figure out flow behavior in the inlet plenum of new research reactor where coolant is injected to the flow tubes with the fuel assembly. The computation results showed that large-scale vortices are generated in the inlet plenum by flow stream injected from inlet pipe. These vortices are divided into small vortices and reversed their revolution. They may lead to flow-induced vibration of fuel assembly, moreover, which has been regarded as a cause of fretting wear of fuel assembly. Also there is an another important thing that average velocity of each flow-tube is uneven showing difference in maximum 18%. So it has been suggested that perforated baffle will be installed to prevent the formation of vortex in the inlet plenum. Two perforated baffles, one is flow skirt and the other is muffler type flow straightener, were proposed and their effect was evaluated using commercial CFD code, Fluent. According to CFD analysis for two perforated baffles, it was confirmed that both of them can prevent or reduce vortex formation in the inlet plenum and make average velocity of each flow tube more even.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.257-272
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• 2007

Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once-through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.

• 한국유체기계학회 논문집
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• 제15권5호
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• pp.5-10
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• 2012

An experimental test facility for the complete characteristics of pumps is constructed at KIMM(Korea Institute of Machinery and Materials). All sensors instrumented in test facility for measuring flow rate, pressure, force and moment are in-situ calibrated by primary method. This paper describes the test facility and test technique of the complete characteristics of pumps, together with an experimental test results for a reactor coolant pump which is designed at KIMM for the first time in Korea. The test results for the mixed-flow type pump of $n_s$=1.425 are presented by three curves: constant head, torque, and speed.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.937-946
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• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.