• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.8 s.251
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• pp.818-824
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• 2006

The hydrodynamic performance analysis of an axial-flow main coolant pump for the modular nuclear reactor has been carried out using a commercial computational fluid dynamics (CFD) software. The prediction capability of the CFD software adopted in the present study was validated in comparison with the experimental data. Predicted performance curves agree satisfactorily well with the experimental results for the main coolant pump over the normal operating range. π Ie prediction method presented herein can be used effectively as a tool for the hydrodynamic design optimization and assist the understanding of the operational characteristics of general purpose axial-flow pumps.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.110-115
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• 2014

The nuclear power plant was launched by Kori unit 1 in 1978 years. Currently, 23 nuclear power plants have been operating in Korea since 1978 years. The localization was completed for most of the reactor facility from Hanbit(Youngkwang) unit 3&4. However, RCP(Reactor Coolant Pump) and MMIS(Man Machine Interface System) is an important technology that has been excluded from the scope of the technical transfer has been dependent on a specific overseas vendor. Recent success in RCP development through co-operation with government and industries. Developed RCP will be applied to Shin-Hanul unit 1&2 nuclear power plants. The RCP operates in high speed and high pressure condition and only rotating component in the NSSS(Nuclear Steam Supply System). Therefore, the problem of vibration has arisen caused by the hydraulic forces of the working fluid. These forces can influence on the stability characteristics for entire RCS(Reactor Coolant System) loop, and can act as significant destabilizing forces. In this study, vibration evaluation of the pump shaft of development RCP estimated under normal operation and over speed conditions. In order to predict the vibration characteristics and dynamic behavior, modal analysis, critical speed analysis and unbalance response spectrum analysis were performed.

• The KSFM Journal of Fluid Machinery
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• v.15 no.5
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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.

• Transactions of the Korean Society of Mechanical Engineers
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• v.14 no.6
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• pp.1629-1638
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• 1990

Using a 1/10 scale model pump, which is designed and manufactured to simulate two phase performance of reactor coolant pump of Y.G.N

• 유체기계공업학회:학술대회논문집
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• 1999.12a
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• pp.245-251
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• 1999

The structural analysis of a reactor coolant pump(RCP) of a nuclear power plant is very important for the safety assessment of the plant. Accurate boundary conditions for the heat transfer coefficient are required for reliable thermal stress analysis of the pump casing, especially in transient operations of the pump since the coolant properties are largely dependent on operational conditions. In the present study, a 3D mixed flow type coolant pump was modeled from the RCP drawings and analyzed in the steady state and number of transient flow conditions by using a commercial code STAR-CD. From the result of the computation, it is seem that the average heat transfer coefficients for the cases considered are found to be the suggested values of the manufacturer, Westinghouse Energy System. The unevenness in local heat transfer coefficients, however, is found to be considerable so that the use of average heat transfer coefficients in all boundaries might not give reliable thermal stresses.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.179-189
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• 1994

A performance of reactor coolant pump in two-phase flow is examined using the pump geometric conditions and the performance of the pump in single-phase flow. Wall friction loss of the reactor coolant pump in single-phase flow is prdicted using the Truckenbrodt boundary layer theory, and the head loss in two-phase flow is predicted with calculated well friction loss and separation loss coefficients. The analysis results are compared with the Combustion Engineering pump test data. The effect of two-phase multiplier on the peak clad temperature in Loss-of-Coolant Accident is also examined using the RELAP5 and the results indicate the importance of its accuracy.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2540-2549
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• 2022

For a pressurized water reactor power plant, the reactor coolant pump (RCP) is a kernel component. And for a canned motor RCP, the rotor system's properties determines its safety. The liquid coolant inside the canned motor RCP fills clearance between the metal shields of rotor and stator, forming a lengthy clearance flow. The influence of inlet preswirl on rotordynamic coefficients of clearance flow in canned motor RCP and their effects on the rotordynamic characteristics of the pump are numerically and experimentally investigated in this work. A quasi-steady state computational fluid dynamics (CFD) method has been used to investigate the influence of inlet preswirl. A vertical experiment rig has also been established for this purpose. Rotordynamic coefficients on different inlet preswirl ratios (IR) are obtained through CFD and experiment. Results show that the cross-coupled stiffness of the clearance flow would change significantly with inlet preswirl, but other rotordynamic coefficients would not change significantly with inlet preswirl. For the case of clearance flow between the stator and rotor cans, influence of inlet preswirl is not so significant as the IR is not large enough.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4571-4584
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• 2022

The analysis of rapid flow transients in Reactor Coolant Pumps (RCP) is essential for a reactor safety study. An accurate and precise analysis of the RCP coastdown is necessary for the reactor design. The coastdown of RCP affects the coolant temperature and the colloidal crud in the primary coolant. A realistic and kinetic model has been used to investigate the behavior of activated colloidal crud in the primary coolant and steam generator that solves the pump speed analytically. The analytic solution of the non-dimensional flow rate has been determined by the energy ratio β. The kinetic energy of the coolant fluid and the kinetic energy stored in the rotating parts of a pump are two essential parameters in the form of β. Under normal operation, the pump's speed and moment of inertia are constant. However, in a coastdown situation, kinetic damping in the interval has been implemented. A dynamic model ACCP-SMART has been developed for System Integrated Modular and Advanced Reactor (SMART) to investigate the corrosion due to activated colloidal crud. The Fickian diffusion model has been implemented as the reference corrosion model for the constituent component of the primary loop of the SMART reactor. The activated colloidal crud activity in the primary coolant and steam generator of the SMART reactor has been studied for different equilibrium corrosion rates, linear increase in corrosion rate, and dynamic RCP coastdown situation energy ratio b. The coolant specific activity of SMART reactor equilibrium corrosion (4.0 mg s^-1) has been found 9.63×10^-3 µCi cm^-3, 3.53×10^-3 µC cm^-3, 2.39×10^-2 µC cm^-3, 8.10×10^-3 µC cm^-3, 6.77× 10^-3 µC cm^-3, 4.95×10^-4 µC cm^-3, 1.19×10^-3 µC cm^-3, and 7.87×10^-4 µC cm^-3 for ²⁴Na, ⁵⁴Mn, ⁵⁶Mn, ⁵⁹Fe, ⁵⁸Co, ⁶⁰Co, ⁹⁹Mo, and ⁵¹Cr which are 14.95%, 5.48%, 37.08%, 12.57%, 10.51%, 0.77%, 18.50%, and 0.12% respectively. For linear and exponential coastdown with a constant corrosion rate, the total coolant and steam generator activity approaches a higher saturation value than the normal values. The coolant and steam generator activity changes considerably with kinetic corrosion rate, equilibrium corrosion, growth of corrosion rate (ΔC/Δt), and RCP coastdown situations. The effect of the RCP coastdown on the specific activity of the steam generators is smeared by linearly rising corrosion rates, equilibrium corrosion, and rapid coasting down of the RCP. However, the time taken to reach the saturation activity is also influenced by the slope of corrosion rate, coastdown situation, equilibrium corrosion rate, and energy ratio β.

• The KSFM Journal of Fluid Machinery
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• v.14 no.1
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• pp.18-23
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• 2011

The hydraulic part in reactor coolant pump consists of suction nozzle, impeller, diffuser, and discharge nozzle. Among them, impeller is required to be designed to satisfy performance requirements such as head, NPSHR, and head curve slope at design point. Present study is intended to suggest the preliminary design method sizing the impeller size to satisfy the design requirement particularly including head curve slope at design point. On a basis of preliminary design result, hydraulic components have been designed in detail by CFD and then manufactured in a reduced scale. Experiment in parallel with computational analysis has been executed in order to confirm the hydraulic performance. Comparison results show good agreement with design result, confirming the validity of design method suggested in this study.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2011.11a
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• pp.671-674
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• 2011

An experimental test facility for the complete characteristics of pumps is constructed at KIMM(Korea Institute of Machinery and Materials). This paper describes the test facility and test technique of the complete characteristics of pumps, together with a experimental test result for a reactor coolant pump.