• Nuclear Engineering and Technology
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• 제49권1호
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• pp.71-81
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• 2017

When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

• 설비공학논문집
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• 제11권6호
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• pp.835-845
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• 1999

The objective of the present study was to investigate heat and mass transfer characteristics in a vertical falling film type absorber using LiBr-$H_2O$ solution with 6owt%. The experimental apparatus consisted of an absorber with inner diameter of 17.2 mm and length of 1150mm, a generator, an evaporator/condenser, a solution tank, a sampling trap etc. The parameters were solution temperature of 45 and $50^{\circ}C$, coolant temperature of 30 and $35^{\circ}C$, and film Reynolds numbers from 50 to 150. Pressure drop in the absorber increased as solution and coolant temperatures decreased. Pressure drop in the absorber increased up to the film Reynolds number of 90, and then decreased at the further increase of the Reynolds number above 90. The maximum absorption mass flux observed at the film Reynolds number of 90. Absorption mass flukes increased as coolant temperature decreased. Absorption mass fluxes and heat transfer coefficients under subcooled condition were larger than those under superheated condition. Heat transfer coefficients were affected by solution temperature more than coolant temperature. The maximum absorption effectiveness under the subcooled condition was 23% for coolant temperature of $30^{\circ}C$ and 31% for coolant temperature of $35^{\circ}C$ under the present experimental conditions.

• 한국자동차공학회논문집
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• 제24권1호
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• pp.53-58
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• 2016

Modeling of engine coolant temperature was conducted for a series hybrid powertrain system. The purpose of this modeling was a simplification of complex heat transfer process inside a engine cooling system in order to apply it to the vehicle powertrain simulation software. A basic modeling concept is based on the energy conservation equation within engine coolant circuit and are composed of heat rejection from engine to coolant, convection heat transfer from an engine surface and a radiator to ambient air. At the final stage, the coolant temperature was summarized as a simple differential equation. Unknown heat transfer coefficients and heat rejection term were defined by theoretical and experimental methods. The calculation result from this modeling showed a reasonable prediction by comparison with the experimental data.

• 한국전기전자재료학회:학술대회논문집
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• 한국전기전자재료학회 2007년도 하계학술대회 논문집 Vol.8
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• pp.37-38
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• 2007

It needs several large reservoirs and long time to remove suspended solid like steel fines and iron oxide in hot rolling coolant. If removing rate of suspended solid in rolling coolant is improved by using high gradient magnetic separation (HGMS) system, the productivity of working process can be increased and the area of reservoir can be reduced. Pre-treatment process that react magnetic floc with inorganic and organic flocculant in coolant was studied. Horizontal type superconducting HGMS system was manufactured successfully, which was constituted with automatic filter transportation apparatus, 6 T He-free superconducting magnet system and water circulation system. The solid suspension removal ratio in hot rolling coolant was increased by flocculation with inorganic and organic flocculant.

• 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.

• Nuclear Engineering and Technology
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• 제42권4호
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• pp.434-441
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• 2010

The activation products of aluminum and argon are key radionuclides for operational and environmental radiological safety during the normal operation of open-tank-in-pool type research reactors using aluminum-clad fuels. Their activities measured in the primary coolant and pool surface water of HANARO have been consistent. We estimated their sources from the measured activities and then compared these values with their production rates obtained by a core calculation. For each aluminum activation product, an equivalent aluminum thickness (EAT) in which its production rate is identical to its release rate into the coolant is determined. For the argon activation calculation, the saturated argon concentration in the water at the temperature of the pool surface is assumed. The EATs are 5680, 266 and 1.2 nm, respectively, for Na-24, Mg-27 and Al-28, which are much larger than the flight lengths of the respective recoil nuclides. These values coincide with the water solubility levels and with the half-lives. The EAT for Na-24 is similar to the average oxide layer thickness (OLT) of fuel cladding as well; hence, the majority of them in the oxide layer may be released to the coolant. However, while the average OLT clearly increases with the fuel burn-up during an operation cycle, its effect on the pool-top radiation is not distinguishable. The source of Ar-41 is in good agreement with the calculated reaction rate of Ar-40 dissolved in the coolant.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.1017-1023
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• 2019

The increase of the alternate current frequency results in increased rotational speed of the electrical motors and connected pumps. The consequence for the reactor coolant pumps is increased flow in primary coolant system. Increase of the current frequency can be initiated by the subsynchronous resonance phenomenon (SSR). This paper analyses the implications of the SSR and consequential increase of the frequency on the nuclear power plant safety. The Simulink $MATLAB^{(R)}$ model of the steam turbine and governor system and RELAP5 computer code of the pressurized water reactor are used in the analysis. The SSR results in fast increase of reactor coolant pumps speed and flow in the primary coolant system. The turbine trip value is reached in short time following SSR. The increase of flow of reactor coolant pumps results in increase of heat removal from reactor core. This results in positive reactivity insertion with reactor power increase of 0.5% before reactor trip is initiated by the turbine trip. The main parameters of the plant did not exceed the values of reactor trip set points. The pressure drop over reactor core is small discarding the possibility of core barrel lift.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.546-555
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• 2019

Since the first nuclear reactor first critical, nuclear systems has gone through four generations of history, and the fourth generation nuclear system will be truly realized in the near future. The notions of SVBR and lead-bismuth eutectic alloy coolant put forward by Russia were well received by the international nuclear science community. Lead-bismuth eutectic alloy with the ability of the better neutron economy, the low melting point, the high boiling point, the chemical inertness to water and air and other features, which was considered the most promising coolant for the 4th generation nuclear reactors. This study mainly focuses on the structural design optimization of the 4th-generation reactor coolant pump, including analysis of external characteristics, inner flow, and transient characteristic. It was found that: the reactor coolant pump with a central symmetrical dual-outlet volute structure has better radial-direction balance, the pump without guide vane has better hydraulic performance, and the pump with guide vanes has worse torsional vibration and pressure pulsation. This study serves as experience accumulation and technical support for the development of the 4th generation nuclear energy system.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.93-102
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• 2021

The analysis of the fluid flow characteristics in reactor pressure vessel is an important part of the hydraulic design of nuclear power plant, which is related to the structure design of reactor internals, the flow distribution at core inlet and the safety of nuclear power plant. The flow distribution and mixing characteristics in the pressurized reactor vessel for the 1000MWe advanced pressurized water reactor is analyzed by using Computational Fluid Dynamics (CFD) method in this study. The geometry model of the full-scaled reactor vessel is built, which includes the cold and hot legs, downcomer, lower plenum, core, upper plenum, top plenum, and is verified with some parameters in DCD. Under normal condition, it is found that the flow skirt, core plate holes and outlet pipe cause pressure loss. The maximum and minimum flow coefficient is 1.028 and 0.961 respectively, and the standard deviation is 0.019. Compared with other reactor type, it shows relatively uniform of the flow distribution at the core inlet. The coolant mixing coefficient is investigated with adding additional variables, showing that mass transfer of coolant occurs near the interface. The coolant mainly distributes in the 90° area of the corresponding core inlet, and mixes at the interface with the coolant from the adjacent cold leg. 0.1% of corresponding coolant is still distributed at the inlet of the outer-ring components, indicating wide range of mixing coefficient distribution.

• 대한기계학회논문집A
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• 제41권5호
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• pp.427-433
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• 2017

고강도 재료나 열전도성이 낮은 재료의 밀링 가공시 가공 부위에서 발생하는 높은 마찰열로 인해 공구 파손 등이 심각하게 발생할 수 있으므로 절삭유 공급이 매우 중요하다. 기존의 절삭유 공급 방식은 수동식 관절 구조를 활용한 방법으로 과도한 양을 분사함에도 불구하고 효과적인 결과를 얻기 힘들었다. 또한, 작업 중 비산으로 인해 작업장 환경에도 큰 악영향을 미쳤다. 이에 본 연구에서는 자동으로 분사 위치 조절이 가능한 장치에 기반하여 절삭유 분사위치에 따른 STS316L 의 밀링가공 특성을 분석하였다. 그 결과, 절삭유 공급 위치 변화에 따라 공구 마모와 표면 거칠기 변화가 관찰되었으며 이를 통해 제작된 장치의 효용성을 보였다.