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

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.

Code assessment of ATLAS integral effect test simulating main steam-line break accident of an advanced pressurized water reactor

2017 ◽  
Vol 55 (1) ◽  
pp. 104-112 ◽  
Author(s):  
Kyoung-Ho Kang ◽  
Yu-sun Park ◽  
Byoung-Uhn Bae ◽  
Jong-Rok Kim ◽  
Nam-Hyun Choi ◽  
...  
Keyword(s):  
Steam Line ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Integral Effect ◽  
Main Steam Line ◽  
Main Steam ◽  
Effect Test

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.

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.

Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

2003 ◽  
Vol 143 (1) ◽  
pp. 37-56 ◽  
Author(s):  
Horst-Michael Prasser ◽  
Gerhard Grunwald ◽  
Thomas Höhne ◽  
Sören Kliem ◽  
Ulrich Rohde ◽  
...  
Keyword(s):  
Steam Line ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Core Cooling
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