Stratified Flow Experiments Pertaining to Advanced LWR Reactor Passive Safety System Designs

2004 ◽  
Author(s):  
Jim C. P. Liou ◽  
Alan G. Stephens ◽  
Richard R. Schultz
Keyword(s):  
Flow Behavior ◽  
Stratified Flow ◽  
Water Hammer ◽  
Test Facility ◽  
Light Water Reactors ◽  
Loss Of Coolant Accident ◽  
Passive Safety System ◽  
Water Reactors ◽  
Lower Temperature ◽  
System Designs

During a loss-of-coolant-accident in advanced light water reactors, outside coolant enters the cold leg by gravity to cool the core. This coolant is at a substantially lower temperature and thus is heavier than the liquid in and from the reactor. Consequently, stratified flow may occur. A stratified flow may cause condensation-induced water hammer, and will influence the coolant flow behavior. Two sets of experiments are in progress to better understand stratified flow conditions that lead to water hammer, and the density stratification behavior. The first set uses air-oil-water as the test media. Its purposes are to conduct exploratory tests and to provide instruction an apparatus for education purposes. The second set of tests will use steam and water and, later, the refrigerant R123. This paper describes the exploratory test facility, gives a brief description of the facility that will be used for the steam-water and refrigerant tests, describes the overall test plan, and finally gives some preliminary results on the intrusion of a lighter liquid into a pipe against flow.

scholarly journals Experimental investigation of LWR passive safety systems performance at the INKA test facility

2019 ◽  
Vol 137 ◽  
pp. 01035
Author(s):  
Rafał Bryk ◽  
Thomas Mull ◽  
Holger Schmidt
Keyword(s):  
Test Facility ◽  
Passive Safety ◽  
Light Water Reactors ◽  
Passive Systems ◽  
Actual Performance ◽  
Passive Safety Systems ◽  
Original Plant ◽  
Water Reactors ◽  
Original Objective ◽  
Safety Systems

INKA is a test facility designed by Framatome and built in the technical center in Karlstein. The original objective for establishing this test rig was the investigation of the performance of the passive safety systems developed in a new Framatome Boiling Water Reactor (BWR) design – KERENA. INKA was constructed in the scale of 1:1 in heights while the total volume of the containment was replicated in 1:24. Since the geometries of particular safety systems are faithfully reflected, their actual performance in the original plant can be investigated at the full scale. Due to the unquestionable interest of the nuclear community in the inherent safety, not only new BWR and PWR designs are equipped with the passive systems, but also particular passive solutions are considered to be applied into the already existing Light Water Reactors (LWR). In this context and due to the fact that both, single component tests and integral tests can be conducted at INKA, the facility can be employed for a demonstration/qualification of a large range of passive safety systems foreseen for quite different types of LWRs. Hence, the goal of the EASY project was the experimental confirmation of the passive systems performance and the analysis of their interactions between each other in the integral tests. Besides, the overarching target of all tests performed at INKA is provision of data for codes validation. This paper presents major outcomes and conclusions drawn on the basis of EASY project results.

Evaluation and Prediction of Critical Overhang Length Under Pipe Whip Accident

1986 ◽  
Vol 108 (2) ◽  
pp. 182-187
Author(s):  
T. Yano ◽  
N. Miyazaki ◽  
S. Miyazono
Keyword(s):  
Estimation Method ◽  
Plastic Collapse ◽  
Light Water Reactors ◽  
Pipe Length ◽  
Loss Of Coolant Accident ◽  
Overhang Length ◽  
Water Reactors ◽  
The Moment ◽  
Accident Conditions ◽  
Good Agreement

When the pipe length between break exit and restraint is long in the pipe whip accident, the pipe will undergo a plastic collapse as the moment increases. The length at which plastic collapse may occur is called the critical overhang length, (OH)cr. The experimental results of (OH)cr show good agreement with the prediction by a static simplified estimation method for (OH)cr although the pipe whipping is a dynamic phenomenon. The diagrams of (OH)cr are also described for a range of sizes of stainless steel pipe under the loss of coolant accident conditions of light water reactors.

scholarly journals Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

2017 ◽  
Vol 2017 ◽  
pp. 1-13 ◽  
Author(s):  
Eltayeb Yousif ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Hao-ran Ju
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Safety Issue ◽  
Primary System ◽  
Light Water Reactors ◽  
Transient Time ◽  
Loss Of Coolant Accident ◽  
Significant Difference ◽  
Water Reactors ◽  
Accident Scenarios

Many reactor safety simulation codes for nuclear power plants (NPPs) have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA) is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR). RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches) is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

Reflood Investigation of Model Fuel Assemblies of Light-Water Reactors for a Loss-of-Coolant Accident

Atomic Energy ◽  
2014 ◽  
Vol 116 (5) ◽  
pp. 343-349 ◽  
Author(s):  
S. S. Bazyuk ◽  
N. Ya. Parshin ◽  
E. B. Popov ◽  
Yu. A. Kuzma-Kichta
Keyword(s):  
Light Water Reactors ◽  
Light Water ◽  
Loss Of Coolant Accident ◽  
Model Fuel ◽  
Fuel Assemblies ◽  
Loss Of Coolant ◽  
Water Reactors

scholarly journals Normal and Accidental Scenarios Analyses with MELCOR 1.8.2 and MELCOR 2.1 for the DEMO Helium-Cooled Pebble Bed Blanket Concept

2015 ◽  
Vol 2015 ◽  
pp. 1-9
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
Severe Accident ◽  
Fusion Reactors ◽  
Transfer System ◽  
Light Water Reactors ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pressure Peak ◽  
Water Reactors ◽  
Heat Transfer System ◽  
Do So

As for Light Water Reactors (LWRs), one of the most challenging accidents for the future DEMOnstration power plant is the Loss of Coolant Accident, which can trigger the pressurization of the confinement structures and components. Hence, careful analyses have to be executed to demonstrate that the confinement barriers are able to withstand the pressure peak within design limits and the residual cooling capabilities of the Primary Heat Transfer System are sufficient to remove the decay heat. To do so, severe accident codes, as MELCOR, can be employed. In detail, the MELCOR code has been developed to cope also with fusion reactors, but unfortunately, these fusion versions are based on the old 1.8.x source code. On the contrary, for LWRs, the newest 2.1.x versions are continuously updated. Thanks to the new features introduced in these latest 2.1.x versions, the main phenomena occurring in the helium-cooled blanket concepts of DEMO can be simulated in a basic manner. For this purpose, several analyses during normal and accidental DEMO conditions have been executed. The aim of these analyses is to compare the results obtained with MELCOR 1.8.2 and MELCOR 2.1 in order to highlight the differences among the results of the main thermal-hydraulic parameters.

scholarly journals Nuclear Power after Fukushima

2011 ◽  
Vol 133 (12) ◽  
pp. 27-29 ◽  
Author(s):  
Gail H. Marcus
Keyword(s):  
Nuclear Power ◽  
Cost Savings ◽  
Nuclear Material ◽  
Light Water Reactors ◽  
Light Water ◽  
Loss Of Coolant Accident ◽  
Major Loss ◽  
Water Reactors ◽  
A Current ◽  
Reactor Pressure

This article discusses advanced reactor technologies that are now getting renewed attention after the Fukushima nuclear plant accident. Interest in smaller reactors has been growing in recent years. Some of these designs have advantages over the traditional large light water reactors (LWRs) for certain applications. The smaller designs carry less of an inventory of nuclear material, so there is less material at risk in an accident involving a release. Proponents of small modular reactors (SMRs) point to cost savings due to the factory fabrication and shorter construction times. They have significant advantages for countries with small grids, where a current 1500 MWe reactor would exceed demand and threaten grid stability. Other designs that are getting the most attention at present are small or medium LWR concepts. In addition to their smaller size, these designs differ from current large, light-water designs in that most of them use an “integral” design. Most major reactor components are inside the reactor pressure vessel, thus significantly reducing the threat of a major loss-of-coolant accident.

Thermal Stratification and Mixing in an Open Water Pool by Submerged Mixtures of Steam and Air

2006 ◽  
Author(s):  
Timothy L. Norman ◽  
Hyun-Sik Park ◽  
Shripad T. Revankar ◽  
Mamoru Ishii ◽  
Joseph M. Kelly
Keyword(s):  
Thermal Stratification ◽  
Open Water ◽  
Mixed Volume ◽  
Steam Injection ◽  
Light Water Reactors ◽  
Reduced Pressure ◽  
Loss Of Coolant Accident ◽  
Mass Flow Rates ◽  
Water Reactors ◽  
Pure Steam

Phenomena associated with jet-plume condensation of steam-air mixtures in a large subcooled pool of water have implications in predicting global system parameters, such as the containment pressure, in light water reactors. A scaled down, reduced pressure suppression pool was designed to study condensation and mixing phenomena using scaled test conditions obtained from RELAP5 code results of a loss of coolant accident in a simplified boiling water reactor. Results from the experiments were compared with the TRACE code predictions which reveal deficiencies in the code to predict the pool thermal stratification as TRACE was not initially developed for predicting such phenomena. A dimensionless boundary map was plotted from several experimental runs of pure steam injection to determine conditions when the pool transits from being a homogeneously mixed volume to being a thermally stratified one. Steam-air mixture injection cases for single horizontal venting indicated that above a pool temperature of 40 °C with air mass flow rates below 0.1 g/s the pool can attain thermal stratification.

scholarly journals Quenching Experiments with CrAl-coated Zircaloy Cladding in Reflooding Water Flows

Energies ◽  
2021 ◽  
Vol 14 (7) ◽  
pp. 1859
Author(s):  
Wang Kee In ◽  
Kwan Geun Lee
Keyword(s):  
Nuclear Power ◽  
Early Stage ◽  
Nucleate Boiling ◽  
Front Velocity ◽  
Light Water Reactors ◽  
Test Conditions ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Film Layer ◽  
Water Reactors

A quenching experiment is performed to investigate the heat transfer characteristics and cooling performance of CrAl-coated Zircaloy (Zr) cladding in a water flow. The CrAl-coated Zr cladding is one of the accident tolerant fuels for light water reactors. The uncoated Zr cladding is also used in this quenching experiment for comparison. This experiment simulates reflood quenching of fuel rod during loss of coolant accident (LOCA) in nuclear power plant. The test conditions were determined to represent the peak cladding temperature, the coolant subcooling and the reflood velocity in the event of LOCA. The flow visualization showed the film boiling during early stage of reflood quenching and the transition to nucleate boiling. The film layer decreases as the coolant subcooling increases and becomes wavy as the reflood velocity increases. The CrAl-coated Zr cladding showed more wavy and thinner film than the uncoated Zr cladding. The rewetting temperature increases as the initial wall temperature and/or the coolant subcooling increases. The quench front velocity increases significantly as the coolant subcooling increases. The reflood velocity has a negligible effect on rewetting temperature and quench front velocity.

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