Comparative Study of Loss-of-Coolant Accident Using MAAP4.03 and RELAP5/MOD3.2.2

2002 ◽  
Author(s):  
Chang Hwan Park ◽  
Doo Yong Lee ◽  
Ik Jeong ◽  
Un Chul Lee ◽  
Kune Y. Suh ◽  
...  
Keyword(s):  
Flow Model ◽  
Correction Term ◽  
Sensitivity Study ◽  
Severe Accident ◽  
Primary System ◽  
Two Phase ◽  
Flow Rates ◽  
Flow Models ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Analysis was performed for a large-break loss-of-coolant accident (LOCA) in the APR1400 (Advanced Power Reactor 1400 MWe) with the thermal-hydraulic analysis code RELAP5/ MOD3.2.2 and the severe accident analysis code MAAP4.03. The two codes predicted different sequences for essentially the same initiating condition. As for the break flow and the emergency core cooling (ECC) flow rates, MAAP4.03 predicted considerably higher values in the initial stage than RELAP5/ MOD3.2.2. It was considered that the differing break flow and ECC flow rates would cause the LOCA sequences to deviate from one another between the two codes. Hence, the break flow model in MAAP4.03 was modified with partly implementing the two-phase homogeneous critical flow model and adopting a correction term. The ECC flow model in MAAP4.03 was also varied by changing the hardwired friction factor through the sensitivity study. The modified break flow and ECC flow models yielded more consistent calculational results between RELAP5/MOD3.2.2 and MAAP4.03. It was, however, found that the resultant effect is rather limited unless more mechanistic treatments are done for the primary system in MAAP4.03.

scholarly journals ASTEC–MAAP Comparison of a 2 Inch Cold Leg LOCA until RPV Failure

2018 ◽  
Vol 2018 ◽  
pp. 1-24
Author(s):  
J. C. de la Rosa Blul ◽  
S. Brumm ◽  
F. Mascari ◽  
S. J. Lee ◽  
L. Carenini
Keyword(s):  
Failure Time ◽  
Hydrogen Generation ◽  
Severe Accident ◽  
Primary System ◽  
Similar Amount ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Expansion Model ◽  
Gas Expansion ◽  
Comparison Exercise

A 2 inch, cold-leg loss-of-coolant accident (LOCA) in a 900 MWe generic Western PWR was simulated using ASTEC 2.1.1 and MAAP 5.02. The progression of the accident predicted by the two codes up to the time of vessel failure is compared. It includes the primary system depressurization, accumulator discharge, core heat-up, hydrogen generation, core relocation to lower plenum, and lower head breach. The purpose of the code comparison exercise is to identify modelling differences between the two codes and the user choices affecting the results. The two codes predict similar primary system depressurization behaviour until the accumulation injection, confirming similar break flow and primary system thermal-hydraulic response calculations between the two codes. The choice of the accumulator gas expansion model, either isentropic or isothermal, affects the rate and total amount of coolant injected and thereby determines whether the core is quenched or overheated and attains a noncoolable geometry during reflooding. A sensitivity case was additionally simulated by each code to allow comparisons to be made with either accumulator gas expansion models. The two codes predict similar amount of in-vessel hydrogen generated and core quench status for a given accumulator gas expansion model. ASTEC predicts much larger initial core relocation to lower plenum leading to an earlier vessel failure time. MAAP predicts more gradual core relocation to lower plenum, prolonging the lower plenum debris bed heat-up and time to vessel failure. Beside the effect of the code user in conducting severe accident simulations, some discrepancies are found in the modelling approaches in each code. The biggest differences are found in the in-vessel melt progression and relocation into the lower plenum, which deserve further research to reduce the uncertainties.

scholarly journals Multiphysical Simulations for the IAEA/ISCP Benchmark Model on the Contact of Pressure Tube and Calandria Tube in the Moderator System of CANDU-6 PHWR

2018 ◽  
Vol 2018 ◽  
pp. 1-8
Author(s):  
Hyoung Tae Kim ◽  
Se-Myong Chang ◽  
Young Woo Son ◽  
Taegee Min
Keyword(s):  
Outer Boundary ◽  
Pressure Tube ◽  
Severe Accident ◽  
Nuclear Facilities ◽  
Two Phase ◽  
Loss Of Coolant Accident ◽  
Commercial Code ◽  
Loss Of Coolant ◽  
Heavy Water Reactor ◽  
Possible Cause

The LOCA (loss of coolant accident) is a kind of severe accident in the operation of PHWR (pressurized heavy water reactor) as well as other nuclear facilities, and possible cause of LOCA can be counted on the ballooning of pressure tube (PT) contacted to the outer calandria tube (CT) in the moderator system of CANDU-6 reactors. In the paper, we simulated the 150-kW experimental facility proposed by IAEA/ISCP, modeling the transient creeping behavior of pressurized tube heated with thermal radiation between the gaps of two concentric pipes. The outer boundary is simplified with a switched model that depends on the local temperature. With a multiphysical model supported by a commercial code, COMSOL multiphysics, the unsteady phenomena are simulated with models concerning various kinds of mechanics such as thermodynamics, nonlinear structural dynamics, and two-phase boiling heat transfer models.

scholarly journals Analysis of Containment Pressure and Temperature Changes Following Loss of Coolant Accident (LOCA)

2015 ◽  
Vol 5 (4) ◽  
pp. 1-8
Author(s):  
Van Thai Nguyen ◽  
Ngoc Dung Kieu
Keyword(s):  
High Pressure ◽  
Cooling System ◽  
Primary System ◽  
Similar Work ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pressure Water ◽  
Main Steam Line ◽  
Main Steam ◽  
Work Done

This paper present a preliminary thermal-hydraulics analysis of AP1000 containment following loss of coolant accident events such as double-end cold line break (DECLB) or main steam line break (MSLB) using MELCOR code. A break of this type will produce a rapid depressurization of the reactor pressure vessel (primary system) and release initially high pressure water into the containment followed by a much smaller release of highly superheated steam. The high pressure liquid water will flash and rapidly pressurize the containment building. The performance of passive containment cooling system for steam removal by condensation on large steel containment structure is a major contributing process, controlling the pressure and temperature maximum reached during the accident event. The results are analyzed, discussed and compared with the similar work done by Sandia National Laboratories.

A Simulation of Small Break Loss of Coolant Accident (SB-LOCA) in AP1000 Nuclear Power Plant Using RELAP5-MV

2017 ◽  
Author(s):  
Eltayeb Yousif ◽  
Zhang Zhijian ◽  
Tian Zhao-fei ◽  
A. M. Mustafa
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Systems ◽  
Nuclear Power ◽  
Power Plants ◽  
Computational Simulation ◽  
Primary System ◽  
Transient Time ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

To ensure effective operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, Loss of Coolant Accident (LOCA) is one of the main accidents. RELAP-MV Visualized Modularization software technology is recognized as one of the best estimated transient simulation programs of light water reactors, and also has the options for improved modeling methods, advanced programming, computational simulation techniques and integrated graphics displays. In this study, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SB-LOCA accident in AP1000 nuclear power plant (NPP) is carried out by Relap5-MV thermo-hydraulics code. While focusing on LOCA analysis in this study, effort was also made to test the effectiveness of the RELAP5-MV software already developed. The accuracy and reliability of RELAP5-MV have been successfully confirmed by simulating LOCA. The calculation was performed up to a transient time of 4,500.0s. RELAP5-MV is able to simulate a nuclear power system accurately and reliably using this modular modeling method. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially for the beginners.

Parameter Sensitivity Study of Boiling and Two-Phase Flow Models in CFD

2012 ◽  
Vol 4 (4) ◽  
pp. 411-425 ◽  
Author(s):  
Timothy J. Drzewiecki ◽  
Isaac M. Asher ◽  
Timothy P. Grunloh ◽  
Victor E. Petrov ◽  
Krzysztof J. Fidkowski ◽  
...  
Keyword(s):  
Two Phase Flow ◽  
Sensitivity Study ◽  
Parameter Sensitivity ◽  
Phase Flow ◽  
Two Phase ◽  
Flow Models

scholarly journals A revised Tesla Turbine concept for 2-phase applications

2021 ◽  
Vol 238 ◽  
pp. 10006
Author(s):  
L. Talluri ◽  
P. Niknam ◽  
A. Copeta ◽  
M. Amato ◽  
P. Iora ◽  
...  
Keyword(s):  
Shear Stress ◽  
Flow Model ◽  
Preliminary Design ◽  
Labyrinth Seals ◽  
Two Phase ◽  
Flow Rates ◽  
Real Fluid ◽  
New Concepts ◽  
Torque Transmission ◽  
Original Concept

The Tesla turbine is an original expander working on the principle of torque transmission by wall shear stress. The principle – demonstrated for air expanders at lab scale has some attractive features when applied to two-phase expanders: it is suitable for handling limited flow rates (as is the case for machines in the range from 500W to 5 kW), it can be developed to a reasonable size (rotor of 0.1 to 0.25 m diameters), with acceptable rotational speeds (which range from 500 to 10000 rpm). The original concept was revisited, designing it for two-phase operation and considering not only the rotor configuration but the whole machine. The flow model was developed using complete real fluid assumptions including several new concepts such as bladed channels for the stator, labyrinth seals, and a rotating diffuser. Preliminary design sketches are presented, and results discussed and evaluated.

Applicability of Choking Flow Models for Subcooled Flashing Flow Through Tube Cracks

2011 ◽  
Author(s):  
Brian Wolf ◽  
Shripad T. Revankar ◽  
Jovica R. Riznic
Keyword(s):  
Mechanistic Model ◽  
Channel Length ◽  
Two Phase ◽  
Experimental Conditions ◽  
Flow Rates ◽  
Flow Models ◽  
Model Predictions ◽  
Flow Through ◽  
The Difference ◽  
Non Equilibrium

Recently there is some database available on choking flow through cracks relevant to steam generator (SG) tubes to model the critical flow. These data are used in assessing the key choking flow models. Based on this assessment a mechanistic choking model is developed. The model is used to predict the choking flow rates for various experimental conditions for subcooled flashing flow through narrow slits with L/D varying from small values (∼5) to large values (100). Results are presented on the effects of thermal and mechanical non-equilibrium on the choking flow for small L/D channels. A mechanistic model was developed to model two-phase choking flow through slits. A comparison of model results to experimental data shows that the homogeneous equilibrium based models markedly under predict choking flow rates in such geometries. As subcooling increases, and channel length decreases the non-equilibrium effects play a greater role in the choking phenomenon, therefore the difference in model predictions and experimental results increases.

Sensitivity Analysis of the Response of PWR Containment During a Loss of Coolant Accident Using RELAP5-3D and MELCOR

2014 ◽  
Author(s):  
Rodolfo Vaghetto ◽  
Andrew Franklin ◽  
Alessandro Vanni ◽  
Yassin A. Hassan
Keyword(s):  
Sensitivity Analysis ◽  
Discharge Coefficient ◽  
Physical Models ◽  
Geometrical Parameters ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Flow Models ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Selection Of

The prediction of specific parameters for the reactor containment, such as pressure and sump pool temperature, is of paramount importance when studying the thermal-hydraulic phenomena involved in the debris generation, transport, and accumulation during Loss of Coolant Accidents (LOCA). The response of the reactor containment during these events may significantly vary depending of several factors such as break size and location, and other plant-specific features. When modeling the reactor and containment response using systems codes, the predictions may also depend on the selection of physical models, correlations and their coefficients. A sensitivity analysis of the response of a typical Pressurized Water Reactor (PWR) 4-loop reactor system and associated containment during a large break LOCA was conducted using RELAP5-3D and MELCOR to investigate the influence of geometrical parameters (break location), physical models (chiked flow models), and related coefficients (discharge coefficient at the break), on the containment response. The simulation results showed how the containment response changed by varying the selected parameters and confirmed the importance of identifying and studying the factors triggering the containment engineered features (containment sprays) when simulation the containment response.

Steady-State Tilting-Pad Bearing Performance Under Reduced Oil Supply Flow Rates

2018 ◽  
Vol 140 (5) ◽  
Author(s):  
Bradley R. Nichols ◽  
Roger L. Fittro ◽  
Christopher P. Goyne
Keyword(s):  
Experimental Data ◽  
Steady State ◽  
Flow Rate ◽  
Flow Model ◽  
Leading Edge ◽  
Input Power ◽  
Flow Rates ◽  
Flow Models ◽  
Oil Supply ◽  
Tilting Pad

Reduced oil supply flow rates in fluid film bearings can cause cavitation, or lack of a fully developed hydrodynamic film layer, at the leading edge of the bearing pads. Reduced oil flow has the well-documented effects of higher bearing operating temperatures and decreased power losses and is commonly referred to as starvation. This study looks at the effects of oil supply flow rate on steady-state bearing performance and provides increased experimental data for comparison to computational predictions. Tests are conducted on a five-pad tilting-pad bearing positioned in a vintage, flooded housing with oil supply nozzles. Pad temperatures, sump temperature, journal operating position, and motor input power are measured at various operating speeds ranging from 2000 to 12,000 rpm and various oil supply flow rates. Predicted results are obtained from bearing modeling software based on thermoelastohydrodynamic (TEHD) lubrication theory. A starved flow model was previously developed as an improvement over the original flooded flow model to more accurately capture bearing behavior under reduced flow conditions. Experimental results are compared to both flow models. The starved bearing model predicts significantly higher journal operating positions than the flooded model and shows good correlation with the experimental data. Predicted pressure profiles from the starved bearing model show cavitation of the upper unloaded pads that increase in severity with increasing speed and decreasing oil supply flow rate. The progressive unloading of these top pads explains the rise in shaft centerline position and helps further validate the starvation model.

The Development and Application of SCDAP-3D©

2002 ◽  
Author(s):  
E. W. Coryell ◽  
E. A. Harvego ◽  
L. J. Siefken
Keyword(s):  
Computer Code ◽  
Alternative Fuel ◽  
Severe Accident ◽  
Initial Development ◽  
Standard Problem ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Environmental Laboratory ◽  
Reactor Accidents ◽  
Thermal Hydraulic Analysis

The SCDAP-3D© computer code (Coryell 2001) has been developed at the Idaho National Engineering & Environmental Laboratory (INEEL) for the analysis of severe reactor accidents. A prominent feature of SCDAP-3D© relative to other versions of the code is its linkage to the state-of-the-art thermal/hydraulic analysis capabilities of RELAP5-3D©. Enhancements to the severe accident models include the ability to simulate high burnup and alternative fuel, as well as modifications to support advanced reactor analyses, such as those described by the Department of Energy’s Generation IV (GenIV) initiative. Initial development of SCDAP-3D© is complete and two widely varying but successful applications of the code are summarized. The first application is to large break loss of coolant accident analysis performed for a reactor with alternative fuel, and the second is a calculation of International Standard Problem 45 (ISP-45) or the QUENCH 6 experiment.

Sign in / Sign up

Export Citation Format

Share Document