• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2002년도 추계학술대회논문집
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• pp.74-79
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• 2002

The piping verification tests were performed in order to verify the structural integrity during initial operation of the reactor coolant systems and the primary heat transportation systems of nuclear power plants by KEPRI in Korea. The tests were conducted at full operating temperature and pressure. The objective is to evaluate the possibility of excessive load generating on piping, piping supports, and reactor structures etc. in the steady normal operation and expected pump transient conditions. As a result, the measured vibrations have been shown acceptable level according to ASME/ANSI OMa-Standard, Part 3.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2002년도 추계학술대회논문초록집
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• pp.315.1-315
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• 2002

The piping verification tests were performed in order to verify the structural integrity during initial operation of the reactor coolant systems and the primary heat transportation systems of nuclear power plants by KEPRI in Korea. The tests were conducted at full operating temperature and pressure. The objective is to evaluate the possibility of excessive load generating on piping, piping supports, and reactor structures etc. in the steady normal operation and expected pump transient conditions. (omitted)

• 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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.151-156
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• 2001

Experimental frictional and wear characteristics of silicon graphite materials is studied in this paper. Those specimens are lubricated with high temperature and highly pressurized water to simulate the same operating condition for the journal bearing and the thrust bearing on the main coolant pump bearing in the newly developing nuclear reactor named SMART(System-integrated Modular Advanced ReacTor). Operating condition of the bearings is realized by the tribometer and the autoclave. Friction coefficient and wear loss are analyzed to choose the best silicon graphite material. Pin on plate test specimens are used and coned disk springs are used to control the applied force on the specimens. Wear loss ana wear width are measured by a precision balance and a micrometer. The friction force is measured by the strain gauge which can be used under high temperature and high pressure. Three kinds of silicon graphite materials are examined and compared with each other, and each material shows similar but different results on frictional and wear characteristics.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2173-2187
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• 2022

A 3D thermo-elasto-hydrodynamic model is developed to analyze the sealing performance of a notched mechanical seal applied in the reactor coolant pump. In the model, the generalized Reynolds equation, the energy equation coupled with notch heat balance equation, the heat conduction equations, and the deformation equations of the sealing rings are iteratively solved by the finite element method. The film pressure and temperature distribution are obtained, and the deformation of the sealing rings is revealed to study the mechanism of the notched mechanical seals. A parameterized study is conducted to analyze the sealing performance under different operating conditions. As a comparison, the sealing performance of non-notched seals is also studied. The results show that the hydrostatic effect is dominant in the load-carrying capacity of the fluid film due to the radial mechanical and thermal deformations. The notch can cool the fluid film and influence the thermal deformation of seal rings. The sealing performance is sensitive to the pressure difference, ambient temperature, and rotational speed. It is suggested to set the notches on the softer sealing rings to acquire the greater hydrodynamic effect. Compared with the non-notched, the notched end face holds a better lubrication performance, especially under lower rotational speed.

• 대한설비공학회:학술대회논문집
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• 대한설비공학회 2005년도 동계학술발표대회 논문집
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• pp.353-358
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• 2005

The heat transfer and pressure drop of supercritical $CO_2$ cooled in a helically coiled tube was investigated experimentally. The experiments were conducted without oil in the refrigerant loop. The experimental apparatus of the refrigerant loop consist of receiver, a variable speed pump, a mass flowmeter, a pre-heater, a gas cooler(test section) and an isothermal tank. The test section is a helically coiled tube in tube counter flow heat exchanger with $CO_2$ flowed inside the inner tube and coolant( water) flowed along the outside annular passage, It was made of it copper tube with the inner diameter of 4.55[mm]. the outer diameter of 6.35 [mm] and length of 10000 [mm]. The refrigerant mass fluxes were $200^{\sim}600$ [kg/m2s] and the inlet pressure of gas cooler varied from 7.5 [MPa] to 10.0 [MPa]. The main results are summarized as follows : The heat transfer coefficient of supercritical $CO_2$ increases, as the cooling pressure of gas cooler decreases. And the heat transfer coefficient increases with the increase of the refrigerant mass flux. The pressure drop decreases in increase of the gas cooler pressure and increases with increase the refrigerant mass flux.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.233-240
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• 1996

Analyses of the viscous and end effects on electromagnetic (EM) pumps of annular linear induction type for the sodium coolant circulation in Liquid Metal Fast Breeder Reactors have been carried out based on the MHD laminar flow analysis and the electromagnetic field theory. A one-dimensional MHD analysis for the liquid metal flowing through an annular channel has been performed on the basis of a simplified model of equivalent current sheets instead of three-phase currents in the discrete primary windings. The calculations show that the developed pressure difference resulted from electromagnetic and viscous forces in the liquid metal is expressed in terms of the slip, and that the viscous loss effects are negligible compared with electromagnetic driving forces except in the low-slip region where the pumps operate with very high flow velocities comparable with the synchronous velocity of the electromagnetic fields, which is not applicable to the practical EM pumps. A two-dimensional electromagnetic field analysis based on an equivalent current sheet model has found the vector potentials in closed form by means of the Fourier transform method. The resultant magnetic fields and driving forces exerted on the liquid metal reveal that the end effects due to finiteness of the pump length are formidable. In addition, a two-dimensional numerical analysis for vector potentials has been performed by the SOR iterative method on a realistic EM pump model with discretely-distributed currents in the primary windings. The numerical computations for the distributions of magnetic fields and developed pressure differences along the pump axial length also show considerable end effects at both inlet and outlet ends, especially at high flow velocities. Calculations of each magnetic force contribution indicate that the end effects are originated from the magnetic force caused by the induced current ( u x B ) generated by the liquid metal movement across the magnetic field rather than the one (E) produced by externally applied magnetic fields by three-phase winding currents. It is concluded that since the influences of the end effects in addition to viscous losses are extensive particularly in high-velocity operations of the EM pumps, it is necessary to find ways to suppress them, such as proper selection of the pump parameters and compensation of the end effects.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2012년도 춘계학술대회 논문집
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• pp.734-738
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• 2012

A comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals is established in accordance with the United States Nuclear Regulatory Commission Regulatory Guide 1.20 Revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results at the inner barrel assembly top plate in the reactor. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at Inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for the Advanced Power Reactor 1400 reactor vessel internals.

• 한국소음진동공학회논문집
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• 제22권5호
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• pp.474-479
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• 2012

A comprehensive vibration assessment program for the advanced power reactor 1400(APR1400) reactor vessel internals is established in accordance with the united states nuclear regulatory commission regulatory guide 1.20 revision 3. This paper is related to instruments and measurement locations based on the vibration and stress response analysis results of the inner barrel assembly top plate in APR1400. The analysis results of the inner barrel assembly top plate in the reactor show that the deterministic stress and deformation due to the reactor coolant pump induced pressure pulsations are larger than the random stress and deformation induced by the flow turbulence. The selection of the instruments and measurement locations at inner barrel assembly top plate in the reactor is essential requirements and very important study process for the vibration and stress measurement program in comprehensive vibration assessment program for APR1400 reactor vessel internals.