• Proceedings of the KSME Conference
• /
• 2001.06d
• /
• pp.110-114
• /
• 2001

A numerical analysis has been perfonned to estimate the effect of turbulent penetration and thermal stratified flow in the branch lines piping. This phenomenon of thermal stratification are usually observed in the piping lines of the safety related systems and may be identified as the source of fatigue in the piping system due to the thermal stress loading which are associated with plant operating modes. The turbulent penetration length reaches to $1^{st}$ valve in safety injection piping from reactor coolant system (RCS) at normal operation for nuclear power plant when a coolant does not leak out through valve. At the time, therefore, the thermal stratification does not appear in the piping between RCS piping and $1^{st}$ valve of safety injection piping. When a coolant leak out through the $1^{st}$ valve by any damage, however, the thermal stratification can occur in the safety injection piping. At that time, the maximum temperature difference of fluid between top and bottom in the piping is estimated about $50^{\circ}C$.

• Nuclear Engineering and Technology
• /
• v.54 no.1
• /
• pp.152-161
• /
• 2022

We analyzed mitigation strategies for steam generator tube rupture (SGTR) accidents using MARS code under both full-power and low-power and shutdown (LPSD) conditions. In general, there are two approaches to mitigating SGTR accidents: supplementing the reactor coolant inventory using safety injection systems and depressurizing the reactor coolant system (RCS) by cooling it down using the intact steam generator. These mitigation strategies were compared from the viewpoint of break flow from the ruptured steam generator tube, the core integrity, and the possibility of the main steam safety valves opening, which is associated with the potential release of radiation. The "cooldown strategy" is recommended for break flow control, whereas the "RCS make-up strategy" is better for RCS inventory control. Under full power, neither mitigation strategy made a significant difference except for on the break flow while, in LPSD modes, the RCS cooldown strategy resulted in lower break and discharge flows, and thus less radiation release. As a result, using the cooldown strategy for an SGTR under LPSD conditions is recommended. These results can be used as a fundamental guide for mitigation strategies for SGTR accidents according to the operational mode.

• Journal of the Korean Society of Propulsion Engineers
• /
• v.7 no.4
• /
• pp.46-52
• /
• 2003

This paper presents the numerical thermal analysis for regeneratively cooled rocket thrust chambers. An integrated numerical model incorporates computational fluid dynamics for the hot-gas thermal environment, and thermal analysis for the liner and coolant channels. The flow and temperature fields in rocket thrust chambers is assumed to be axisymmetric steady state which is presumed to the combustion liner. The heat flux computed from nozzle flow is used to predict the temperature distribution of the combustion liner As a result, we present the wall temperature of combustion liner and the temperature change of coolant.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.29 no.12
• /
• pp.654-662
• /
• 2017

When emergency core cooling system (ECCS) is operated during loss of coolant accident (LOCA) in a pressurized water reactor (PWR), pressurized thermal shock (PTS) phenomenon can occur as cooling water is injected into a cold leg, mixed with hot primary coolant, and then entrained into a reactor vessel. Insufficient flow mixing may cause temperature stratification and steam condensation. In addition, flow vibration may cause thermal stresses in surrounding structures. This will reduce the life of the reactor vessel. Due to the importance of PTS phenomenon, in this study, calculation was performed for Test 1 among six types of OECD/NEA ROSA tests with ANSYS CFX R.17. Predicted results were then compared to measured data. Additionally, because temperature difference between the hot coolant at the inlet of the cold leg and the cold cooling water at the inlet of the ECCS injection line is 200 K or more, buoyancy force due to density difference might have significant effect on thermal-hydraulic characteristics of flow. Therefore, in this study, the necessity to include buoyancy force term in governing equations for accurate prediction of single phase thermal stratification in both cold legs and downcomer by ECCS injection was numerically studied.

• 한국전산유체공학회:학술대회논문집
• /
• 2011.05a
• /
• pp.269-274
• /
• 2011

A CFD analysis was performed to examine the inner channel blockage of dual-cooled fuel which has being developed for the power uprate of a pressurized water reactor (PWR). The dual-cooled fuel consists of an annular fuel pellet($UO_2$) and dual claddings as well as internal and external cooling channels. The dual-cooled annular fuel is different from a conventional solid 려el by employing an internal cooling channel for each fuel pellet as well as an external cooling channel. One of the key issues is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause the Departure from Nucleate Boiling (DNB) in the inner channel that eventually causes a fuel failure. This paper presents the CFD simulation of the flow through the side holes of the bottom end plug for the complete entrance blockage of the inner channel. Since the amount of coolant supply to the inner channel depends on largely the pressure loss at the side hole, the pressure loss coefficient of the side hole was estimated by the CFD analysis. The CFD prediction of the loss coefficient showed a reasonable agreement with an experimental data for the complete blockage of both the inner channel entrance and the outer channel. The CFD predictions also showed the decrease of the loss coefficient as the outer channel blockage increases.

• Nuclear Engineering and Technology
• /
• v.52 no.5
• /
• pp.937-946
• /
• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

• The KSFM Journal of Fluid Machinery
• /
• v.10 no.1 s.40
• /
• pp.49-55
• /
• 2007

An 'Y' strainer type automatic flow rate regulating valve, which functions are to remove impurities from hot water inside the pipe and to maintain a constant flow rate regardless of variations of the differential pressure between valve inlet and outlet at the same time, is developed for distributing hot water equally to several pipes with district heating or central heating system. Numerical analysis of the three dimensional turbulent flow field in a valve shape is carried out to confirm the flow field whether the designed regulator shape is acceptable or not. The final developed valve improves installation time and cost and maintenance ability comparing with set-up 'Y' strainer and regulator separately. Tolerance for the nominal flow rate is also satisfied within ${\pm}5%$.

• Nuclear Engineering and Technology
• /
• v.30 no.6
• /
• pp.592-608
• /
• 1998

Inspection of RCS flow measurements for the domestic pressurized water reactors has been performed by the Korea Institute of Nuclear Safety (KINS) as one of the authorized periodical inspection activities. The inspection results for the Westinghouse-type plants reveal that 1) the RCS flow instrumentation has been calibrated by using the initial design and commissioning test result, without reflecting the cycle specific reference flow measurements, 2) the loop-to-loop now variation in the actual flow measurement which has not been considered in the safety analysis affects the asymmetric How transient results, and 3) the measured RCS flows in Kori 3 and 4, Yonggwang 1 and 2 do not support the definition of the best estimate RCS flow, approaching the RCS flow limit. In this study, the revealed problems were discussed with review of the design and the RCS flow measurement uncertainty evaluation, and the technical approaches and recommendations for resolving these problems were proposed.

• Proceedings of the Korean Society of Propulsion Engineers Conference
• /
• 2009.11a
• /
• pp.119-122
• /
• 2009

The purpose of this study is to design of film-cooling ring for the small thrust rocket engine using green propellants(Hydrogen peroxide and kerosene). Cold flow test was carried out to measure the mass flow rate and atomizing characteristic. Required mass flow rate was obtained from thermal analysis of the engine, and measured flow rate 42.25g/s was in the range of permissible coolant flow rate. With the same mass flow rate, cooling ring with more hole and high velocity shows better spray pattern. The result of thermal analysis, cooling ring has enough cooling performance.

• The KSFM Journal of Fluid Machinery
• /
• v.9 no.6 s.39
• /
• pp.35-44
• /
• 2006

Operation conditions of a reheat cycle gas turbine for a combined cycle power plant was analyzed. Based on measured performance parameters of the gas turbine, a performance analysis program predicted component characteristic parameters such as compressor air flow, compressor efficiency, efficiencies of both the high and low pressure turbines, and coolant flows. The predicted air flow and its variation with the inlet guide vane setting were sufficiently accurate. The compressor running characteristic in terms of the relations between air flow, pressure ratio and efficiency was presented. The variations of the efficiencies of both the high and low pressure turbines were also presented. Almost constant flow functions of both turbines were predicted. The current methodology and obtained data can be utilized for performance diagnosis.