Multidimensional TMI-1 Main-Steam-Line-Break Analysis Methodology Using TRAC-PF/NEM

2001 ◽  
Vol 133 (2) ◽  
pp. 169-186 ◽  
Author(s):  
Kostadin N. Ivanov ◽  
Tara M. Beam ◽  
Anthony J. Baratta ◽  
Ardesar Irani ◽  
Nicholas G. Trikouros
Keyword(s):  
Steam Line ◽  
Analysis Methodology ◽  
Main Steam Line ◽  
Main Steam

Fracture properties evaluation of carbon steel piping for main steam line

1995 ◽  
Vol 158 (2-3) ◽  
pp. 241-251 ◽  
Author(s):  
Y.J Kim ◽  
C.S Seok ◽  
Y.S Chang ◽  
J.O Kim ◽  
K.M Yang ◽  
...  
Keyword(s):  
Carbon Steel ◽  
Steam Line ◽  
Fracture Properties ◽  
Main Steam Line ◽  
Main Steam ◽  
Steel Piping

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.

scholarly journals Analysis of the OECD Main Steam Line Break Benchmark Problem Using the Refined Core Thermal-Hydraulic Nodalization Feature of the MARS/MASTER Code

2003 ◽  
Vol 142 (2) ◽  
pp. 166-179 ◽  
Author(s):  
Han Gyu Joo ◽  
Jae-Jun Jeong ◽  
Byung-Oh Cho ◽  
Won Jae Lee ◽  
Sung Quun Zee
Keyword(s):  
Steam Line ◽  
Benchmark Problem ◽  
Main Steam Line ◽  
Main Steam

Steam Dryer Acoustic Load Predictions in the Main Steam Line Break Event

2008 ◽  
Author(s):  
Jin Yan ◽  
Francis Bolger ◽  
Guangjun Li ◽  
Weimin Dai ◽  
Lev Klebanov
Keyword(s):  
Pressure Difference ◽  
Acoustic Pressure ◽  
Steam Line ◽  
Expansion Wave ◽  
Distribution Area ◽  
Ambient Environment ◽  
Design Basis ◽  
Acoustic Load ◽  
Main Steam Line ◽  
Main Steam

In nuclear reactor design, significant acoustic pressure loads impact the steam dryer hood as a result of the main steam line break outside containment (MSLB) event. When a main steam line breaks, it is assumed that the pipe instantaneously breaks completely open to the ambient environment (double-ended guillotine break). Due to the huge pressure difference between the inside of the reactor pressure vessel (RPV) and surrounding ambient environment, a shock wave will form at the break point and burst into the surrounding environment. At the same time, an expansion wave will travel upstream through the main steam line to the RPV, which results in a pressure reduction on the outside of the steam dryer hood. This expansion wave will create a substantial pressure difference between the two sides of the steam dryer hood with a resultant high stress on the hood. This differential pressure load is the acoustic load used in the structure design evaluations for this event. A key design basis requirement for the steam dryer is to maintain structural integrity during transient, and accident conditions. Demonstration that the steam dryers meet this design basis requires a calculation of the magnitude of the acoustic load on the steam dryer during a MSLB. In this study, Computational Fluid Dynamics (CFD) is used as an alternate calculation method to investigate the phenomenon of MSLB. Transient simulations with fine time steps were carried out. The results show that CFD is a useful tool to provide additional information on the acoustic load as compared to the traditional methods. From the CFD results, the minimum pressure value and its distribution area at different flow times was identified. Through the modeling, an understanding of the detailed transient flow field, particular the acoustic pressure field near the dryer hood during the MSLB was achieved.

Calculation of BWR Shroud Head Loads and Axial Lift during a Main Steam-Line Break

1998 ◽  
Vol 121 (3) ◽  
pp. 313-323
Author(s):  
John M. Sorensen ◽  
Nicholas G. Trikouros
Keyword(s):  
Steam Line ◽  
Main Steam Line ◽  
Main Steam

Main-Steam-Line-Break Analysis of TMI-1 Using RETRAN-3D

1999 ◽  
Vol 128 (2) ◽  
pp. 233-244 ◽  
Author(s):  
Craig E. Peterson ◽  
John G. Shatford ◽  
Ardesar Irani ◽  
Nicholas G. Trikouros ◽  
Antonio F. Dias ◽  
...  
Keyword(s):  
Steam Line ◽  
Main Steam Line ◽  
Main Steam

Structural Assessment of Main Steam Line and Containment Building Under Steam Explosion Conditions

2015 ◽  
Author(s):  
Seung-Huyn Kim ◽  
Yoon-Suk Chang ◽  
Sungchu Song ◽  
Yong-Jin Cho
Keyword(s):  
Steam Explosion ◽  
Failure Criteria ◽  
Steam Line ◽  
Dynamic Loads ◽  
Steam Generators ◽  
Structural Assessment ◽  
Concrete Failure ◽  
Nuclear Component ◽  
Main Steam Line ◽  
Main Steam

The steam explosion is a fuel-coolant interaction process where the heat transfer from the melt to water is quite intense and rapid. This phenomenon may threat integrity of nuclear components and structures. For instance, the dynamic loads on the reactor cavity and the reactor lower plenum could potentially lead to failure of the main steam lines connected to the steam generators. In addition, since the main steam line extends to the containment wall, failure of the containment building may occur. The object of present study is to examine characteristics of Main Steam Line (MSL) and containment building under the steam explosion conditions. In this context, the corresponding FE models were generated and previously determined displacements of penetration piping were used as loading conditions. Subsequent FE analyses were conducted for the main steam line and containment building to calculate stresses and crack evaluations. Finally, structural assessment of nuclear component and structure combined with concrete failure criteria was performed and their results were discussed.

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