• Corrosion Science and Technology
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• v.12 no.2
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• pp.85-92
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• 2013

This work is concerned with an evaluation of dynamic boric acid corrosion (BAC) of low alloy steel for reactor pressure vessel of a pressurized water reactor (PWR). Mockup test method was newly established to investigate dynamic BAC of the low alloy steel under various conditions simulating a primary water leakage incident. The average corrosion rate was measured from the weight loss of the low alloy steel specimen, and the maximum corrosion rate was obtained by the surface profilometry after the mockup test. The corrosion rates increased with the rise of the leakage rate of the primary water containing boric acid, and the presence of oxygen dissolved in the primary water also accelerated the corrosion. From the specimen surface analysis, it was found that typical flow-accelerated corrosion and jet-impingement occurred under two-phase fluid of water droplet and steam environment. The maximum corrosion rate was determined as 5.97 mm/year at the leakage rate of 20 cc/min of the primary water with a saturated content of oxygen within the range of experimental condition of this work.

• Journal of the Korean Society of Propulsion Engineers
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• v.21 no.2
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• pp.48-59
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• 2017

In this study, an analysis of the thermo-fluid dynamic and missile-motion performance was carried out through a numerical simulation inside the missile canister. Calculation was made in an analytical volume using dynamic grid and evaporated water was used as a coolant. To analyze the interaction among the hot gas, coolant, and mixture flow, Realizable $k-{\varepsilon}$ turbulence and VOF (Volume Of Fluid) model were chosen and a parametric study was performed with the change of coolant flow rate. As a result of the analysis, pressure of the canister showed a large difference depending on the presence or absence of the coolant, and also showed a dependancy on the amount of coolant. Velocity and acceleration were dependent on the canister pressure.

• Korean Chemical Engineering Research
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• v.57 no.1
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• pp.28-41
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• 2019

We carried out performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor. We analyzed transient-dynamic behavior of fluids inside the steam generator to vent into a sodium dump tank or a water dump tank when tubes in the steam generator were broken to cause a large-water-leak accident. Accordingly, we preliminarily evaluated design requirements of our system. Our results showed that sodium in the shell side of the steam generator and in Intermediate Heat Transport System was completely vented within 50 s and feed water in the tube side of the steam generator was completely vented within 2.5 s. It was analyzed that pressure of the tube side of the steam generator was higher than pressure of the shell side of the steam generator, which showed that sodium in the shell side did not flow into the tube side. Our results are expected to be used as basis information to performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.322-326
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• 1997

The condensate in a single vertical reflux condenser with a tube of the large L/D ratio could carried over in both ways of fill-and-dump and the annular occurrent to steam flow. From the experimental observation made, a theoretical model based on the lumped parameter method is made to understand the dynamics of the reflux condenser. The present model predicts well the time period of fill-and-dump model and the natural vibrational frequency of the water column. This could be a first step to understand the complex phenomena in the reflux condenser such as itd improved thermal performance due to the well controlled pulsation in steam flow and the tube-to tube effect in the multi tube reflux condenser.

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.835-838
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• 2002

An experiment has been performed using a facility, which simulates the safety depressurization system (SDS) and in-containment refueling water storage tank (IRWST) of APR1400, an advanced PWR being developed in Korea, to investigate the dynamic load resulting from the blowdown of steam from a steam generator through a sparser. The influence of the key parameters, such as air mass, steam pressure, submergence, valve opening time, and pool temperature, on frequency and peak toads was investigated. The blowdown phenomenon was analyzed to find out the real cause of the initiation of bubble oscillation and discrepancy in frequencies between the experiment and calculation by conventional equation for bubble oscillation. The cause of significant damping was discussed and is presumed to be the highly tortuous flow path around bubble. The Rayleigh-Plesset equation, which is modified by introducing method of image, reasonably reproduces the bubble oscillation in a confined tank. Right after the completion of air discharge the steam discharge immediately follows and it condenses abruptly to provide low-pressure pocket. It may contribute to the negative maximum being greater than positive maximum. The subsequently discharging steam does not play as at the driving force anymore.

• Journal of the Korean Society of Safety
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• v.11 no.2
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• pp.52-59
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• 1996

This paper presents a new approach for assessing the dynamic reliability in a complex system such as a nuclear power plant. The method is applied to a dynamic analysis of the potential accident sequences which may occur during mid-loop operation. Mid-loop operation is defined as an operation to make RCS water level below the top of the flow area of the hot legs at the junction with the reactor vessel for repairs and maintenance of steam generators and reactor coolant pumps for a specific time. The Idea behind this approach consists of both the use of the concept of the performance achievement/requirement correlation and of a dynamic event tree generation method. The assessment of the system reliability depends on the determination of both the required performance distribution and the achieved performance distribution. The quantified correlation between requirement and achievement represents a comparison between two competing variables. It is demonstrated that this method is easily applicable and flexible in that it can be applied to any kind of dynamic reliability problem.

• Journal of Korean Society on Water Environment
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• v.28 no.3
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• pp.384-393
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• 2012

Saemangeum lake is an artificial lake created by reclamation works and an estuary embankment since 2006. The sea water flows into the lake by the operation of two sluice gates, and the freshwater enters into the lake by the upper streams. For the reflection of hydrology and hydrodynamics effects in Saemangeum area, a hydrodynamics model was developed by connecting Hydrological Simulation Program with Fortran (HSPF) and Environmental Fluid Dynamic Code (EFDC). The HSPF was applied to simulate the freshwater discharge from the upper steam watershed, and the EFDC was performed to compute water flow, water temperature, and salinity based on time series from 2008 to 2009. The calibration and validation are performed to analyze horizontal and vertical gradients. The horizontal trend of model simulation results is reflected in the trend of observed data tolerably. The vertical trend is conducted an analysis of seasonal comparisons because of the limitation of vertically observed data. Water temperature reflects on the seasonal changes. Salinity has an effect on the near river input spots. The impact area of salinity is depending on the sea water distribution by gate operation, mainly.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.641-646
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• 1998

In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUIN $T^{™}$ has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA).

• Proceedings of the KSME Conference
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• 2001.06d
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• pp.813-820
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• 2001

The Safety depressurization System(SDS) of KNGR prevents RCS from overpressurization by discharging high pressure and temperature coolant through the I-sparger into the IRWST during an accident. If IRWST water temperature rise locally, around the sparger, beyond $200_{\circ}$2000 F by the discharged coolant, unstable steam condensation can cause large pressure load on the IRWST wall. To investigate whether this condition can be avoided for the design basis event IOPOSRV(Inadvertent Opening of one Pilot Operated Safety Relief Valve), the flow and temperature distribution of water in the IRWST is calculated by using CFX 4.3 computational fluid dynamic code. According to the results, since pool water temperature does not exceeds temperature limit within 50 seconds after the opening of one POSRV, it can be assured that the integrity of IRWST wall is maintained.

• 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.