• Transactions of the Korean Society of Pressure Vessels and Piping
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• 제12권1호
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• pp.85-92
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• 2016

The operating temperature of VHTR components is much higher than that of conventional PWR due to high core outlet temperature of VHTR. Material requirements and technical issues of VHTR reactor components which are mainly dominated by high temperature service condition were discussed. The codification effort for high temperature material and design methodology are explained. The design class for VHTR components are classified as class A or B according to the recent ASME high temperature reactor design code. A separation of thermal boundary and pressure boundary is used for VHTR components as an elevated design solution. Key design characteristics for reactor pressure vessel, control rod, reactor internals, graphite reflector, circulator and intermediate heat exchanger were analysed. Thermo-mechanical analysis of the process heat exchanger, which was manufactured for test, is presented as an analysis example.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.72-86
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• 2016

A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel-coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD) Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In reactor calculations, the largest uncertainties in the prediction of the steam explosion strength are expected to be caused by the large uncertainties related to the jet breakup. To obtain some insight into these uncertainties, premixing simulations were performed with both available jet breakup models, i.e., the global and the local models. The simulations revealed that weaker explosions are predicted by the local model, compared to the global model, due to the predicted smaller melt droplet size, resulting in increased melt solidification and increased void buildup, both reducing the explosion strength. Despite the lower active melt mass predicted for the pressurized water reactor case, pressure loads at the cavity walls are typically higher than that for the boiling water reactor case. This is because of the significantly larger boiling water reactor cavity, where the explosion pressure wave originating from the premixture in the center of the cavity has already been significantly weakened on reaching the distant cavity wall.

• Journal of Industrial Technology
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• 제21권B호
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• pp.141-147
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• 2001

For the preparation of calcium carbonate particles from aqueous $Ca(OH)_2$ slurry, carbonation reaction of aqueous $Ca(OH)_2$ slurry was carried out by batch method the $CO_2$ into reactor filled with aqueous slurry of $Ca(OH)_2$. The concentration of $Ca(OH)_2$ varies from 1.00 to 7.00wt%, reactor temperature at 20 and $40^{\circ}C$, and reactor pressure from atmospheric pressure to $6.0kg_f/cm^2$. Crystal structure of calcium carbonate was of calcite, the particle size were about $0.05{\sim}2.0{\mu}m$, and the particle shape was cubic and spindle. When reactor temperature was higher, particle size of calcium carbonate was bigger and particle shape was varied, but reaction rate was increased. When reactor pressure was higher, particle size of calcium carbonate was smaller, particle shape was cubic, and reaction rate was increased.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.1020-1028
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• 2021

The nuclear reactor coolant pump transferring heat energy inherently brings with it the unsteady flow and inevitably threatens to the safe operation of the pump unit, especially with the pressure pulsation induced by the rotor-stator interaction. In this paper, the characteristics of pressure pulsation of the diffuser in a nuclear reactor coolant pump were investigated by the numerical simulation with experimental validation. Pressure pulsation signals measured synchronously from sensors mounted on the radial diffuser of a model pump were analyzed via Welch's method. Frequency components induced by the rotor-stator interaction can be revealed by the diameter mode analysis method. The pressure pulsation of the diffuser is dominated by the blade passing frequency and its harmonics, which are free from the effect of flow rate and rotational speed while the corresponding amplitudes are easily affected by different operational conditions and measuring positions. The non-uniformity is much more affected by the rotational speed than the flow rate. This research is helpful for further work to reduce the pressure pulsation for the reactor coolant pump.

• Proceedings of the KSME Conference
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.782-787
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• 2001

This paper is concerned with the global rupture of a nuclear reactor pressure vessel(RPV) in a severe accident. During the severe reactor accident of molten core, the temperature and the pressure in the nuclear reactor rise to a certain level depending on the initial and subsequent condition of a severe accident. While the rise of the temperature cause the thermal softening of RPV material, the rise of the internal pressure could cause failure of the RPV lower head. The global rupture of an RPV is simulated by finite element limit analysis for the collapse pressure and mode and this analysis results have been compared with a variation of the internal pressure of RPV. The finite element limit method is a systematic tool to secure the safety criteria of a nuclear reactor and to evaluate the in-vessel corium retention.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• 제7권1호
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• pp.17-26
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• 2011

CANDU reactor core is composed a few hundreds pressure tubes, which support and locate the nuclear fuels in the reactor. Each pressure tube provides pressure boundary and flow path of primary heat transport system in the core region. In order to guarantee the structural integrity of pressure tube flaws which can be found by in-service inspection, crack growth and fracture initiation assessment have to be performed. Stress intensity factors are important and basic information for structural integrity assessment of planar and laminar flaws (e. g. crack). This paper reviews and confirms the stress intensity factor of axial crack, proposed in CSA N285.8-05, which is an fitness-for-service evaluation code for pressure tubes in CANDU nuclear reactors. The stress intensity factors in CSA N285.8-05 were compared with stress intensity factors calculated by three methods (finite element results, API 579-1/ASME FFS-1 2007 Fitness-For-Service and ASME Boiler and Pressure Vessel Code Section XI). The effects of Poisson's ratio and anisotropic elastic modulus on stress intensity factors were also discussed.

• Nuclear Engineering and Technology
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• 제27권5호
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• pp.710-720
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• 1995

The propagation of pump-induced pressure pulsation in a reactor is important because of the potential for vibration and resultant damage of reactor internals. A hydrodynamic model has been developed to obtain the pressure fluctuation due to the operation of pumps in the annulus(between the core support barrel and reactor vessel of a pressurized water reactor) including the coolant inlet pipe. The mathematical analysis is formulated in accordance with the linearized Navier-Stokes equation by assuming a compressible, inviscid flow. Two regions are considered separately and by coupling the solutions of the inlet pipe and the annulus, the inlet nozzle pressure(pressure at pipe and annulus interface) is to be calculated without assumptions. The geometric parameter effect on the pump-induced pressure pulsation is evaluated. Comparison of predicted and measured inlet nozzle pressure values for each forcing frequency shows good order of magnitude agreement.

• Nuclear Engineering and Technology
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• 제41권7호
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• pp.921-928
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• 2009

As part of an evaluation for an in-vessel severe accident management strategy, a coolant injection into the reactor vessel under depressurization of the reactor coolant system (RCS) has been evaluated in detail using the SCDAP/RELAP5 computer code. A high-pressure sequence of a small break loss of coolant accident (SBLOCA) has been analyzed in the Optimized Power Reactor (OPR) 1000. The SCDAP/RELAP5 results have shown that safety injection timing and capacity with RCS depressurization timing and capacity are very effective on the reactor vessel failure during a severe accident. Only one train operation of the high pressure safety injection (HPSI) for 30,000 seconds with RCS depressurization prevents failure of the reactor vessel. In this case, the operation of only the low pressure safety injection (LPSI) without a HPSI does not prevent failure of the reactor vessel.

• The KSFM Journal of Fluid Machinery
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• 제19권3호
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• pp.10-13
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• 2016

The wet-air oxidation in gravity pressure reactor is effective for organic waste treatment with energy saving under high pressure and high temperature. But its oxidation control is difficulty because its multi-phase flow characteristics is very complicated. The flow characteristics of an oxidizer feed section in the gravity pressure reactor were investigated using numerical method which are verified by comparison with experimental results. In this study, the results showed that the flow rate of oxidizer have an effect on the generation of bubble around feed section.

• Proceedings of the Korean Nuclear Society Conference
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• 한국원자력학회 1996년도 춘계학술발표회논문집(3)
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• pp.290-295
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• 1996

The purpose of this study is to establish the pressure/temperature curves of Reactor Coolant System for brittle fracture prevention. The pressure/temperature curve is the basis to select RC Pump and limits to operate the plant. Based on the plant operation experience, this curve should be re-generated periodically in order to ensure the structural integrity using data from the test of reactor vessel surveilance materials to compensate for the irradiation effects. This study provides the procedure of pressure/temperature curve generation in term of brittle fracture prevention of reactor vessel. Using the UCN 3&4 data, the sample pressure/temperature curve was generated, and it was compared with those of YGN 3&4 based on the stress and $RT_{NDT}$value.