Suppression Methods of Acoustic Noise Generated in Main Steam Stop Valve

2018 ◽  
Author(s):  
Shiro Takahashi ◽  
Eiji Ozaki ◽  
Atsuyuki Minenaga
Keyword(s):  
Pressure Loss ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Acoustic Noise ◽  
Acoustic Resonance ◽  
Stop Valve ◽  
Cfd Analyses ◽  
Main Steam Line ◽  
Main Steam

The main steam stop valve (MSSV) is installed in the main steam line in thermal and nuclear power plants. The MSSV is a safety valve that instantaneously shuts off the steam flowing into the steam turbine in an emergency. However, as high-speed steam flow goes through the MSSV during even the rated operation, acoustic sound or noise is generated in the MSSV. Moreover, there is a possibility that flow-induced acoustic resonance occurs in the MSSV. Flow-induced acoustic resonance must be suppressed to decrease the sound noise. Reducing the pressure loss of the MSSV is also an important issue that cannot be neglected with respect to the plant thermal efficiency. Therefore, we have developed the MSSV which can suppress the flow-induced acoustic resonance and its pressure loss. To develop this MSSV, we conducted scale air tests and computational fluid dynamics (CFD) analyses that are described in this paper. Mach and Strouhal number of the test conditions were the same as those of an actual plant. Reynolds number was sufficiently large to obtain the developed turbulent flow. An unsteady compressible CFD analysis was also conducted using large eddy simulation as a turbulence model. We developed new tilted triangular tabs and installed them in the MSSV to suppress the intense vortex generation and pressure loss. As a result, the sound noise due to the flow-induced acoustic resonance was completely attenuated and pressure loss was reduced compared to the case using the current tilted tabs. CFD results also showed that the tilted triangular tabs could suppress the generation of intense vortexes and the flow-induced acoustic resonance.

Investigation of Blast Wave Effects on Containment Wall and Steam Generator

2018 ◽  
Author(s):  
Tae Jin Kim ◽  
Yoon-Suk Chang
Keyword(s):  
Blast Wave ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Criteria ◽  
High Energy ◽  
Element Analysis ◽  
Secondary Damage ◽  
Main Steam Line ◽  
Main Steam

When a sudden rupture occurs in high energy lines such as MSL (Main Steam Line) and safety injection line of nuclear power plants, ejection of inner fluid with high temperature and pressure causes blast wave, and may lead to secondary damage of adjacent major components and/or structures. The objective of this study is to assess integrity of containment wall and steam generator due to the blast wave under a postulated high energy line break condition at the MSL piping. In this context, a preliminary analysis was conducted to examine the blast wave simulation using coupled Eulerian-Lagrangian technique. Subsequently, a finite element analysis was carried out to assess integrity of the structures. As typical results, strain and stress values were calculated at the containment wall and steam generator, which did not exceed their failure criteria.

Evaluation of LBB characteristics of candidate materials for main steam line piping in Korea nuclear power plants

2020 ◽  
Vol 188 ◽  
pp. 104226
Author(s):  
Seokmin Hong ◽  
Jongmin Kim ◽  
Maan-Won Kim ◽  
Hong-Deok Kim ◽  
Bong-Sang Lee ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Steam Line ◽  
Main Steam Line ◽  
Main Steam

An Experimental Measurement of Toughness Locus for a Ductile Material

2005 ◽  
Vol 297-300 ◽  
pp. 2410-2415 ◽  
Author(s):  
Dong Hak Kim ◽  
Jeong Hyun Lee ◽  
Ho Dong Kim ◽  
Ki Ju Kang
Keyword(s):  
Crack Length ◽  
Nuclear Power ◽  
Power Plants ◽  
Test Procedure ◽  
Stress Triaxiality ◽  
Crack Tip ◽  
Potential Drop ◽  
Image Correlation ◽  
Ductile Material ◽  
Main Steam

A toughness locus Jc-Q for a ductile steel, SA106 Grade C used in the main steam piping of nuclear power plants, has been experimentally evaluated. Along with the standard fracture test procedure for J-R curve, Q as the second parameter governing stress triaxiality nearby the crack tip is measured from the displacements nearby the side necking which occurs near the crack tip on the lateral surface of a fracture specimen. The displacements nearby the side necking are measured from the digital images taken during the fracture experiment based on Stereoscopic Digital Photography (SDP) and high resolution Digital Image Correlation (DIC) software. The crack length is monitored by Direct Current Potential Drop (DCPD) method and the J-R curve is determined according to ASTM standard E1737-96. The effects of crack length, specimen geometry and thickness of specimen are studied, which are included in the toughness locus Jc-Q.

LOCA analysis of BWR-4/Mark-I nuclear power plant with TRACE

Kerntechnik ◽  
2021 ◽  
Vol 86 (2) ◽  
pp. 128-142
Author(s):  
J.-J. Huang ◽  
S.-W. Chen ◽  
J.-R. Wang ◽  
C. Shih ◽  
H.-T. Lin
Keyword(s):  
Nuclear Power ◽  
Coupled Model ◽  
Short Term ◽  
Containment System ◽  
Loss Of Coolant Accident ◽  
Main Steam Line ◽  
Main Steam ◽  
Temperature Responses ◽  
Term Analysis

Abstract This study established an RCS-Containment coupled model that integrates the reactor coolant system (RCS) and the containment system by using the TRACE code. The coupled model was used in both short-term and long-term loss of coolant accident (LOCA) analyses. Besides, the RELAP5/CONTAN model that only contains the containment system was also developed for comparison. For short-term analysis, three kinds of LOCA scenarios were investigated: the recirculation line break (RCLB), the main steam line break (MSLB), and the feedwater line break (FWLB). For long-term analysis, the dry-well and suppression pool temperature responses of the RCLB were studied. The analysis results of RELAP5/CONTAN and TRACE models are benchmarked with those of FSAR and RELAP5/GOTHIC models, and it appears that the results of the above four models are consistent in general trends.

Analyzing Droplet Size Distributions Inside a Self-Priming Venturi Scrubber for Filtered Containment Venting Systems

2018 ◽  
Author(s):  
P. Papadopoulos ◽  
T. Lind ◽  
H.-M. Prasser
Keyword(s):  
Droplet Size ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Gas Flow ◽  
Tube Bundle ◽  
Droplet Size Distribution ◽  
Severe Accident ◽  
Design Element ◽  
Venturi Scrubber

After the accident in the Fukushima Daiichi nuclear power plant, the interest of adding Filtered Containment Venting Systems (FCVS) on existing nuclear power plants to prevent radioactive releases to the environment during a severe accident has increased. Wet scrubbers are one possible design element which can be part of an FCVS system. The efficiency of this scrubber type is thereby depending, among others, on the thermal-hydraulic characteristics inside the scrubber. The flow structure is mainly established by the design of the gas inlet nozzle. The venturi geometry is one of the nozzle types that can be found in nowadays FCVS. It acts in two different steps on the removal process of the contaminants in the gas stream. Downstream the suction opening in the throat of the venturi, droplets are formed by atomization of the liquid film. The droplets are contributing to the capture of aerosols and volatile gases from the mixture coming from the containment. Studies state that the majority of the contaminants is scrubbed within this misty flow regime. At the top of the venturi, the gas stream is injected into the pool. The pressure drop at the nozzle exit leads to the formation of smaller bubbles, thus increasing the interfacial area concentration in the pool. In this work, the flow inside a full-scale venturi scrubber has been optically analyzed using shadowgraphy with a high-speed camera. The venturi nozzle was installed in the TRISTAN facility at PSI which was originally designed to investigate the flow dynamics of a tube rupture inside a full-length scale steam generator tube bundle. The data analysis was focused on evaluating the droplet size distribution and the Sauter mean diameter under different gas flow rates and operation modes. The scrubber was operated in two different ways, submerged and unsubmerged. The aim was to include the effect on the droplet sizes of using the nozzle in a submerged operation mode.

Analytical Evaluation of Unsteady Thermal Hydraulic Characteristics in Shell and Tube Heat Exchangers

2006 ◽  
Author(s):  
Shiro Takahashi ◽  
Yuichi Narumi ◽  
Kiyoshi Ishihama ◽  
Akihito Yokoyama ◽  
Toyohiko Tsuge ◽  
...  
Keyword(s):  
Heat Exchangers ◽  
Nuclear Power ◽  
Power Plants ◽  
Tube Bundle ◽  
Fluid Temperature ◽  
Detached Eddy Simulation ◽  
Tube Heat Exchanger ◽  
Eddy Simulation ◽  
Safety And Reliability ◽  
Cfd Analyses

Many shell & tube heat exchangers are used in nuclear power plants. Unsteady thermal hydraulic phenomena have been studied in shell & tube heat exchangers to improve their safety and reliability and to extend their lifetime based on experience obtained from long periods of plant operation. We investigated unsteady flow in shell & tube heat exchangers by using computational fluid dynamics (CFD) analyses. The inlet flow on the shell side was separated and flow in several directions. A large part of the flow crossed over the tube bundle, and some parts of the flow took two circuitous roots (up and down) along the inner surface of the shell. Separated circuitous flows collided again where a baffle plate had been cut off. A pair of symmetric vortexes could be seen in that location. Some parts of the circuitous flow moved backwards into the tube bundle due to vortexes. These vortexes were unstable and changed their size and location. A pair of vortexes changed from symmetric to asymmetric. As a result, the direction of flow in the tube bundle near the vortexes changed continuously. Variations in vortexes simulated through CFD analyses could also be seen in tests on the actual size. Fluid temperature fluctuations around tubes were also evaluated through CFD analyses. Unsteady phenomena with changes from symmetric to asymmetric vortexes could be observed in the shell & tube heat exchanger and were simulated through CFD analyses with a detached eddy simulation (DES) turbulence model.

Experimental Study on the Interaction of Molten Sn With Water

2018 ◽  
Author(s):  
Longkun He ◽  
Pengfei Liu ◽  
Xisi Zhang ◽  
Wenjun Hu ◽  
Bo Kuang ◽  
...  
Keyword(s):  
Water Temperature ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Dynamic Pressure ◽  
Ambient Pressure ◽  
Influence Factors ◽  
Video Camera ◽  
Metallic Oxide ◽  
Set Up

In nuclear power plants, fuel-coolant interaction (FCI) often accompanied with core melt accidents, which may escalate to steam explosion destroying the integrity of structural components and even the containment under certain conditions. In the present study, a new facility for intermediate-scaled experiments named ‘Test for Interaction of MELt with Coolant’ (TIMELCO) has been set up to study FCI phenomena and thermal-hydraulic influence factors in metal or metallic oxide/water mixtures with melt at maximum 2750°C. The first series of tests was performed using 3kg of Sn which was heated to 800°Cand jetted into a column of 1m water depth (300mm in diameter) under 0.1MPa ambient pressure. The main changing parameter was water temperature, at 60 °C and 72 °C respectively. From the high-speed video camera, violent explosion phenomenon occurred at water temperature of 60°C, while no evident explosion observed at 72°C. The size of melt debris at 60°C is smaller than this at 72°C.On the contrary, the dynamic pressure at 60°C is larger. The results indicate that water temperature has an important effect on FCI and decreasing the temperature of the coolant is advantageous to the explosion.

Sign in / Sign up

Export Citation Format

Share Document