• Transactions of the Korean Society of Automotive Engineers
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• v.3 no.2
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• pp.51-61
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• 1995

A coast down test mothod has been used to determine the resistance forces on running vehicle due to the aerodynamic drag, rolling resistance and driveline resistance. Most of the tests, however, are based on the Velocity-Time measurements, which require a sophisticated velocity measuring device and contain much error by nature. In the present study a coast down test method based on Distance-Time measurements is introduced, which contains the original idea of Russian scientist Prof. Petrushov along with the suggestions for improvement of the accuracy.

• Transactions of the Korean Society of Automotive Engineers
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• v.3 no.3
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• pp.1-8
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• 1995

In the companion paper of the present paper, a coast down test method to determine the resistance forces on running vehicle based on the distance-time measurement was explained along with the suggestions to improve its accuracy and testing methodology. In the present paper some of the suggestions discussed previously are implemented and actually road tested to see the applicability of the improved method(short distance method) in the field. From the results. it is shown that the short distance method which requires only 600m long proving ground road gives at least comparable results on the accuracy compared to the original S-t method which requires 2000m. It is hoped that the present method be further refiend to give more accurate results.

• 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.

• Tribology and Lubricants
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• v.35 no.2
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• pp.123-131
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• 2019

This study characterizes the coastdown performances of two small electric motors supported on high-speed ball bearings (BBs) and gas foil bearings (GFBs), and it predicts their acceleration performances. The two motors have identical permanent magnetic rotors and mating stators. However, the shaft of the GFBs has a larger mass and polar/transverse moments of inertia than that of the BBs. Motor coastdown tests demonstrate that the rotor speed decreases linearly with the BBs and nonlinearly with the GFBs. A simple model for the BBs predicts a constant drag torque and linear decay of speed with time. The test data validate the model predictions. For the GFBs, the hydrodynamic lubrication model predictions reveal that the drag torque increases linearly with speed, and the speed decreases exponentially with time. The predictions agree very well with the test data in the speed range of 100-30 krpm. The boundary lubrication model predicts a constant drag torque and linear decay of speed with time. The predictions agree well with the test data below 15 krpm. Mixed lubrication occurs in the speed range of 30-15 krpm. Rotor acceleration performances are predicted based on the characteristics of deceleration performances. The GFBs require more time to reach 100,000 krpm than the BBs because of their larger shaft polar moment of inertia. However, predictions for the assumed identical polar moment of inertia reveal that the GFBs have a nearly identical acceleration performance to that of the BBs with a motor torque greater than $0.03N{\cdot}m$.

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.382-395
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• 2002

An experimental study on transient critical heat flux (CHF) under flow coastdown has been performed for the water flow in a non-uniformly heated vertical annulus under low flow and a wide range of pressure conditions. The objectives of this study are to systematically investigate the effect of the flow transient on the CHF and to compare the transient CHF with steady-state CHF The transient CHF experiments have been performed for three kinds of flow transient modes based on the coastdown data of a nuclear power plant reactor coolant pump. At the same inlet subcooling, system pressure and heat flux, the effect of the initial mass flux on the critical mass flux can be negligible. However, the effect of the initial mass flux on the time-to- CHF becomes large as the heat flux decreases. The critical mass flux has the largest value for slow flow reduction rate. There is a pressure effect on the ratio of the transient CHF data to steady-state CHF data. Except under low system pressure conditions, the flow transient CHF was revealed to be conservative compared with the steady-state CHF data. Bowling CHF correlation and thermal hydraulic system code MARS show promising results for the prediction of CHF occurrence .

• Nuclear Engineering and Technology
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• v.15 no.3
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• pp.188-196
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• 1983

A simplified fast-running system code is developed to simulate loop transients such as pump coastdown, loop failures and natural circulation. Special emphasis is put on the numerical investigation of the natural circulation system with multiloop. For this purpose, 5 governing equations are derived, and they are discretized by the space-time integration technique. The developed computer program is applied to three sample problems; transition from 2-loop to 1-loop operation, transition from 1-loop to 2-loop operation, and the transient behavior with decay power in the case of 2-loop operation.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.33 no.6
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• pp.418-426
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• 2009

The effect of an inertia moment of a pump flywheel on the thermal-hydraulic behaviors of the KALIMER-600(Korea Advanced LIquid MEtal Reactor) reactor pool during an early-phase of a loss of normal heat sink accident was investigated. The thermal-hydraulic analyses for a steady and a transient state were made by using the COMMIX-1AR/P code. In the present analysis a quarter of the reactor geometry was modeled in a cylindrical coordinate system, which includes a quarter of a reactor core and a UIS, a half of a DHX and a pump and a full IHX. In order to evaluate the effects of an inertia moment of the pump flywheel, a coastdown flow whose flow halving time amounts to 3.69 seconds was supplied to a natural circulation flow in the reactor vessel. Thermal-hydraulic behaviors in the reactor vessel were compared to those without the flywheel equipment. The numerical results showed a good agreement with the design values in a steady state. It was found that the inertia moment contributes to an increase in the circulation flow rate during the first 40 seconds, however to a decrease of it there after. It was also found that the flow stagnant region induced by a core exit overcooling decelerated the flow rate. The appearance of the first-peak temperature was delayed by the flow coastdown during the initial stages after a reactor trip.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.191-202
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• 2013

A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (~300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy may be due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.

• Nuclear Engineering and Technology
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• v.16 no.2
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• pp.47-57
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• 1984

Mathematically-rigorous time-volume averaged conservation equations were simplified to established the differential equations of THERMIT-6S, which is a two-fluid 3-D code. The difference equations of THERMIT-6S were obtained by discretizing the proceeding set of differential equations. The spatial discretization is characterized by a first-order spatial scheme, donor cell method, and staggered mesh layout. For time discretization, a first order semi-implicit scheme treats implictly sonic terms and terms relating to local transport phenomena and explicitly convective terms. The results were linearized by the Newton-Raphson method. In order to construct the reduced pressure equation, the linearized equations were manipulated so that all variables are coupled between mesh cells through only the pressure variable. By simulating numerically the OPERA-15 experiment, it was found that THERMIT-6S is a very powerful code in predicting reactor behavior after sodium boiling including flow coastdown, reversal flow and flow oscillation.

• Journal of computational fluids engineering
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• v.17 no.4
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• pp.41-48
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• 2012

A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (~300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy is due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.