Flow Instabilities and Main Steam Line Vibrations in a Pressurized Water Reactor

2002 ◽  
Author(s):  
Mats Henriksson ◽  
Johan Westin ◽  
Tord Granha¨ll ◽  
Lars Andersson ◽  
Lars-Erik Bjerke
Keyword(s):  
Steam Generator ◽  
Model Testing ◽  
Steam Line ◽  
Scale Model ◽  
Outlet Nozzle ◽  
Pressurized Water ◽  
Structural Dynamic ◽  
Pressure Pulses ◽  
Main Steam ◽  
Steam System

Severe vibrational problems occurred in the main steam system of a PWR nuclear power plant, about 18 months after a steam generator replacement had been carried out. The magnitude of the vibrations reached levels at which the operators had to reduce power in order to stay within the operating limits imposed by the nuclear inspectorate. To solve the problem the following analyses methods were employed: • Testing the influence on vibration level from different modes of plant operation; • Analyses of plant measurement data; • Calculations of: – hydraulic behaviour of the system, – structural dynamic behaviour of the system, – flow at the steam generator outlet; • Scale model testing of the steam generator outlet region. Hydraulic flow disturbances in the main steam system were measured using pressure and strain gauges, which made it possible to track individual pressure pulses propagating through the main steam system. Analyses showed that the pressure pulses causing the vibration originated from the vicinity of the steam generator outlet. By using computer codes for network fluid flow analyses the pressure pulses found in the measurement traces could be generated in calculations. Careful studies of the flow at the steam generator outlet region, using model testing in a 1:3 scale model as well as transient 3D CFD calculations, gave clear indications that flow separation occurred at the steam generator outlet nozzle and at the first bend. Finally, by substituting the outlet nozzle for a different design with a multiport nozzle, the steam line vibration problem has been solved.

Transient Thermal-Hydraulic Responses of the Nuclear Steam Generator Secondary Side to a Main Steam Line Break1

2015 ◽  
Vol 137 (4) ◽  
Author(s):  
Jong Chull Jo ◽  
Frederick J. Moody
Keyword(s):  
Steam Generator ◽  
Steam Line ◽  
Analysis Model ◽  
Pressurized Water ◽  
Line Pipe ◽  
Short Period ◽  
Main Steam Line ◽  
Main Steam ◽  
Transient Thermal ◽  
Secondary Side

This paper presents a multidimensional numerical analysis of the transient thermal-hydraulic response of a steam generator (SG) secondary side to a double-ended guillotine break of the main steam line attached to the SG at a pressurized water reactor (PWR) plant. A simplified analysis model is designed to include both the SG upper space, which the steam occupies and a part of the main steam line between the SG outlet nozzle and the pipe break location upstream of the main steam isolation valve. The transient steam flow through the analysis model is simulated using the shear stress transport (SST) turbulence model. The steam is treated as a real gas. To model the steam generation by heat transfer from the primary coolant to the secondary side coolant for a short period during the blow down process following the main steam line break (MSLB) accident, a constant amount of steam is assumed to be generated from the bottom of the SG upper space part. Using the numerical approach mentioned above, calculations have been performed for the analysis model having the same physical dimensions of the main steam line pipe and initial operational conditions as those for an actual operating plant. The calculation results have been discussed in detail to investigate their physical meanings and validity. The results demonstrate that the present computational fluid dynamics (CFD) model is applicable for simulating the transient thermal-hydraulic responses in the event of the MSLB accident including the blowdown-induced dynamic pressure disturbance in the SG. In addition, it has been found that the dynamic hydraulic loads acting on the SG tubes can be increased by 2–8 times those loads during the normal reactor operation. This implies the need to re-assess the potential for single or multiple SG tube ruptures due to fluidelastic instability for ensuring the reactor safety.

Transient Thermal-Hydraulic Responses of the Nuclear Steam Generator Secondary Side to a Main Steam Line Break

2014 ◽  
Author(s):  
Jong Chull Jo ◽  
Frederick J. Moody
Keyword(s):  
Steam Generator ◽  
Steam Line ◽  
Analysis Model ◽  
Pressurized Water ◽  
Line Pipe ◽  
Short Period ◽  
Main Steam Line ◽  
Main Steam ◽  
Transient Thermal ◽  
Secondary Side

This paper presents a multi-dimensional numerical analysis of the transient thermal-hydraulic response of a steam generator secondary side to a double-ended guillotine break of the main steam line attached to the steam generator at a pressurized water reactor plant. A simplified analysis model is designed to include both the steam generator upper space where steam occupies and a part of the main steam line between the steam generator outlet nozzle and the pipe break location upstream of the main steam isolation valve. The transient steam flow through the analysis model is simulated using the shear stress transport turbulence model. The steam is treated as a real gas. To model the steam generation by heat transfer from the primary coolant to the secondary side coolant for a short period during the blow down process following the main steam line break accident, a constant amount of steam is assumed to be generated from the bottom of the steam generator upper space part. Using the numerical approach mentioned above, calculations have been performed for the analysis model having the same physical dimensions of the main steam line pipe and initial operational conditions as those for an actual operating plant. The calculation results have been discussed in detail to investigate their physical meanings and validity. The results demonstrate that the present CFD model is applicable for simulating the transient thermal-hydraulic responses in the event of the MSLB accident including the blowdown-induced dynamic pressure disturbance in the SG. In addition, it has been found that the dynamic hydraulic loads acting on the SG tubes can be increased by 2 to 8 times those loads during the normal reactor operation. This implies the need to re-assess the potential for single or multiple SG tube ruptures due to fluidelastic instability for ensuring the reactor safety.

Thermal-Hydraulic Response of the Secondary Side of a PWR Steam Generator With an Internal Structure to a Main Steam Line Break

2015 ◽  
Author(s):  
Jong Chull Jo
Keyword(s):  
Internal Structure ◽  
Steam Generator ◽  
Steam Line ◽  
Operating Conditions ◽  
Analysis Model ◽  
Pressurized Water ◽  
Main Steam Line ◽  
Main Steam ◽  
Secondary Side ◽  
Hydraulic Response

This study addresses a numerical analysis of the thermal-hydraulic response of the secondary side of a steam generator (SG) model with an internal structure to a main steam line break (MSLB) at a pressurized water reactor (PWR) plant. The analysis model is comprised of the SG upper space where steam occupies and the part of the main steam pipe between the SG outlet nozzle and the broken pipe end upstream of the main steam isolation valve. To investigate the effects of the presence of the SG internal structure on the thermal-hydraulic response to the MSLB, the numerical calculation results for the analysis model having a perforated horizontal plate as the SG internal structure are compared to those obtained for a simple analysis model having no SG internal structure. Both analysis models have the same physical dimensions except for the internal structure. The initial operating conditions for both SG models are identical to those for an actual operating plant. To simplify the analyses, it is assumed that steam is constantly generated from the bottom of the SG secondary side space during the blowdown process. As the results, it has been found that the pressure wave significantly attenuates as it passes through the perforated internal structure and as time elapses. This leads to reduction in instantaneous hydraulic load on the internal structure including tubing. However, it is seen that the presence of the internal structure does not affect the transient velocities of steam passing through the SG tube bundle during the blowdown, which are 2 to 8 times the velocities during the normal reactor operation as in the case for the empty SG. Consequently, the present findings should be considered for the design of the steam generator to ensure the reactor safety as such elevated high steam velocities can cause fluidelastic instability of tubes which results in high cycle fatigue failure of the tubes.

Effects of a Venturi-Type Flow Restrictor on the Thermal-Hydraulic Response of the Secondary Side of a Pressurized Water Reactor Steam Generator to a Main Steam Line Break

2016 ◽  
Vol 138 (4) ◽  
Author(s):  
Jong Chull Jo ◽  
Frederick J. Moody
Keyword(s):  
Steam Generator ◽  
Steam Line ◽  
Cfd Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Main Steam Line ◽  
Main Steam ◽  
Secondary Side ◽  
Hydraulic Response

A numerical analysis has been performed to simulate the transient thermal-hydraulic response to a main steam line break (MSLB) for the secondary side of a steam generator (SG) model equipped with a venturi-type SG outlet flow restrictor at a pressurized water reactor (PWR) plant. To investigate the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB, numerical calculation results for the SG model equipped with the flow restrictor are compared to those obtained for an SG model without the restrictor. Both analysis models contain internal structures. The present computational fluid dynamics (CFD) model has been examined by comparing to a simple analytical model. It is confirmed from the comparison that the CFD model simulates the transient response of the SG secondary to the MSLB physically plausibly and minutely. Based on the CFD analysis results for both cases with or without the restrictor, it is seen that the intensities of the steam velocity and dynamic pressure are considerably attenuated in the SG model equipped with the restrictor comparing to the case in the SG model without the restrictor.

Study of Steam Hammer in the Main Steam System

2017 ◽  
Author(s):  
Yemin Dong ◽  
Cuiyun Wang ◽  
Jiaming Zhao ◽  
Pei Yu ◽  
Bin Zhao
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Normal Operation ◽  
Saturated Steam ◽  
Pressurized Water ◽  
Main Steam ◽  
Safety Operation ◽  
Steam System

The main steam system plays an important role to transfer the saturated steam generated from the steam generator to the main turbine and other steam consumed devices in a pressurized water reactor. During the normal operation, the main steam system transfers the high temperature and pressure steam generated from the steam generator in the nuclear power plant. Once there is an accident situation delivering the main steam isolation valve fast close signal or an unexpected main steam isolation valve close signal, the steam hammer phenomenon will be induced in the main steam system by the main steam isolation valve fast close incident. The steam hammer phenomenon might induce pressure rise rapidly in the main steam system and generate unintended transient load on the main steam system which might have effect on the safety operation of the nuclear power plant. Therefore the steam hammer phenomenon in the main steam system should be studied. The study creates the main steam system model which includes the main steam pipes, devices and connected system based on PIPENET software following the engineering data by a third generation nuclear power plant. The study takes the advantage of the transient mode in PIPENET to simulate the steam hammer phenomenon in the main steam system. The study simulates different boundary conditions and device parameters in order to analyze the different effects on the steam hammer phenomenon in the main steam system. The simulation model could calculate the pressure, load and other parameters in the main steam system during the main steam isolation valve fast close period. The effects of the steam hammer phenomenon could be analyzed through these characteristic parameters. The PIPENET model could simulate the main steam system action during the main system isolation valve fast close incident which helps the study to master the operation and function of the main steam system and verify the integrality of the main steam system in the steam hammer phenomenon. With the simulation and analysis of the steam hammer phenomenon in the main steam system simulated in the PIPENET, the pressure raise which induced by the steam hammer wouldn’t threat the integrality of the main steam system. And the main steam system could ensure the safety operation by steam discharge through the steam dump valves and main steam safety valves.

Validation of a CFD Analysis Model for the Thermal-Hydraulic Response of PWR Steam Generator to a Steam Line Break

2016 ◽  
Author(s):  
Jong Chull Jo ◽  
Bok Ki Min ◽  
Jae Jun Jeong
Keyword(s):  
Steam Generator ◽  
Steam Line ◽  
Cfd Analysis ◽  
Analysis Method ◽  
Pressurized Water Reactor ◽  
Analysis Model ◽  
Pressurized Water ◽  
Calculation Results ◽  
Main Steam Line ◽  
Main Steam

This paper presents a validation of a computational fluid dynamics (CFD) analysis method for a numerical simulation of the transient thermal-hydraulic responses of steam generator (SG) secondary side to blowdown following a main steam line break (MSLB) at a pressurized water reactor (PWR). To do this, the CFD analysis method was applied to simulate the same blowdown situation as in an experimental work which was conducted for a simplified SG blowdown model, and the CFD calculation results were compared with the experimental results. As the result, both are in reasonably good agreement with each other. Consequently, the present CFD analysis model has been validated to be applicable for numerical simulations of the transient phase change heat transfer and flow situations in PWR SGs during blowdown.

Main Steam Line Break Calculations Using a Coupled RELAP5/PARCS Model for the Ringhals-3 Pressurized Water Reactor

2008 ◽  
Author(s):  
Mathias Sta˚lek ◽  
Jo´zsef Ba´na´ti ◽  
Christophe Demazie`re
Keyword(s):  
Steady State ◽  
Fuel Assembly ◽  
Thermal Power ◽  
Steam Line ◽  
Pressurized Water ◽  
The Core ◽  
Different Types ◽  
Main Steam Line ◽  
Main Steam ◽  
Control Rods

A Main Steam Line Break (MSLB) is an important transient for Pressurized Water Reactors (PWR) due to the strong positive reactivity introduced by the over-cooling of the core. Since this effect is stronger when the Moderator Temperature Coefficient (MTC) has a large amplitude, a conservative result will be obtained for a high burnup of the fuel due to the more negative MTC late in the cycle. The calculations have been performed at a cycle burnup of 12.9742 GWd/tHM. The Swedish Ringhals-3 PWR is a three loop Westinghouse design, currently with a thermal power of 3000 MW. The PARCS model has 157 fuel assemblies of 8 different types. Four different types of reflector are used. The cross sections, and kinetic data were obtained from CASMO-4 calculations, using a cross section interface developed at the department. There are 24 axial nodes, and 2×2 radial nodes for each assembly. The transient option for calculating the effect of poisoning was used. The PARCS model has been validated against steady-state measurements from Ringhals-3 of the Relative Power Fraction (RPF) and of the core criticality. The RELAP5 model has 157 channels for the core which means that there is a one to one correspondence between the thermal hydraulics model and the neutronics model. There is eight axial nodes. Originally, the intention was to have 24 axial nodes but this proved not to work because of some limitation in RELAP5. There is currently no mixing between the different channels in the core. The feedwater, and turbines are modelled as boundary conditions. The stand-alone RELAP5 model has been validated against steady state measurements from Ringhals-3. A number of different cases were considered. In the first case, both the isolation of the feedwater for the broken loop, and all the control rods were assumed to work properly. For the second case one of the control rods was assumed to be stuck. The stuck rod was located in the fuel assembly with the highest power. This rod has also one of the highest rod worths. In the final case, the feedwater control valve for the broken loop was fully open. None of the cases led to any recriticality. The increase in power for each fuel assembly was also investigated. With the control rod located in the assembly with the highest power, the maximum power increase before scram turned out to be about 25% compared to the initial power.

Analysis of the Reactivity during a Pressurized Water Reactor Main-Steam-Line-Break Transient

1998 ◽  
Vol 124 (3) ◽  
pp. 284-290 ◽  
Author(s):  
Garry C. Gose ◽  
Thomas J. Downar ◽  
Karl O. Ott
Keyword(s):  
Steam Line ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Main Steam Line ◽  
Main Steam

Identification of Quad Cities Main Steam Line Acoustic Sources and Vibration Reduction

2007 ◽  
Author(s):  
Guy DeBoo ◽  
Kevin Ramsden ◽  
Roman Gesior ◽  
Brian Strub
Keyword(s):  
Safety Valve ◽  
Thermal Power ◽  
Fatigue Cracking ◽  
Steam Line ◽  
Acoustic Properties ◽  
Relief Valve ◽  
Acoustic Sources ◽  
Main Steam ◽  
Steam System ◽  
Extended Power

The Quad Cities Nuclear Power Station, Units 1 and 2, have a history of steam line vibration issues. The implementation of an Extended Power Uprate resulted in significant increases in steam line vibration as well as acoustic loading of the steam dryers, which led to equipment failures and fatigue cracking of the dryers. This paper discusses the extensive data collection on the Quad Cities Unit 2 replacement dryer and the Main Steam Lines. This data was taken with the intent of identifying acoustic sources in the steam system. Review of the data confirmed that vortex shedding coupled column resonance in the relief and safety valve standpipes were the principal sources of large magnitude acoustic loads in the main steam system. Modifications were developed in subscale testing to alter the acoustic properties of the valve standpipes and add acoustic damping to the system. The modifications developed and installed consisted of acoustic side branches that were attached to the Electromatic Relief Valve (ERV) and Main Steam Safety Valve (MSSV) attachment pipes. Subsequent post-modification testing was performed in plant to confirm the effectiveness of the modifications. The modifications have demonstrated a reduction in the acoustic pressure loads at full Extended Power Uprate (EPU) conditions to levels below those at Original Licensed Thermal Power (OLTP).

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