Rewetting Processes During Top/Bottom Re-Flooding of Heated Vertical Surfaces

2011 ◽  
Author(s):  
Muhammad Ilyas ◽  
Masroor Ahmad ◽  
Colin P. Hale ◽  
Simon P. Walker ◽  
Geoff F. Hewitt
Keyword(s):  
Imaging System ◽  
Heated Surface ◽  
Fast Response ◽  
Light Water Reactors ◽  
Transparent Substrate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Rate Of Progression ◽  
Water Reactors ◽  
Thermal Imaging System

Rewetting of the heated fuel rods is one of the most important phenomena to be considered in analysis of the design basis loss of coolant accident (LOCA) in light water reactors. The rewetting phenomenon is a complex and violent one with the rewetting front moving rather slowly over the heated surface. For water temperature close to saturation, the rate of progression of rewetting front is independent of flow rate of the water approaching the rewetting front; this is an indication of the fact that the rewetting process is governed by events local to the rewetting front [1]. This paper describes an experimental study on the rewetting of heated vertical surfaces during top/bottom reflooding. Through an infrared transparent substrate fixed in the surface, processes occurring locally at the quench front have been studied by using a fast response thermal imaging system (Cedip Titanium 560M). The existence of a cyclic bursting phenomenon at the quench front has been observed. Multiple events of this type gradually remove heat from the metal, allowing the rewetting front to progress slowly over the surface. These intermittent contacts occur over a short axial length. Temperature measurements indicate that the metal surface temperature at the rewetting front is close to the homogeneous nucleation temperature.

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 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.

scholarly journals Fuel R&D Needs and Strategy towards a Revision of Acceptance Criteria

2010 ◽  
Vol 2010 ◽  
pp. 1-7 ◽  
Author(s):  
François Barré ◽  
Claude Grandjean ◽  
Marc Petit ◽  
Jean-Claude Micaelli
Keyword(s):  
Loss Of Coolant Accident ◽  
Safety Margins ◽  
Power Load ◽  
Higher Power ◽  
Loss Of Coolant ◽  
Advanced Fuel ◽  
Water Reactors ◽  
Improved Performance ◽  
Irradiated Fuel ◽  
Cladding Material

The study of fuel behaviour under accidental conditions is a major concern in the safety analysis of the Pressurised Water Reactors. The consequences of Design Basis Accidents, such as Loss of Coolant Accident and Reactivity Initiated Accident, have to be quantified in comparison to the safety criteria. Those criteria have been established in the 1970s on the basis of experiments performed with fresh or low irradiated fuel. Starting in the 1990s, the increased industrial competition and constraints led utilities to use fuel in more and more aggressive conditions (higher discharge burnup, higher power, load follow, etc.) and create incentive conditions for the development of advanced fuel designs with improved performance (new fuel types with additives, cladding material with better resistance to corrosion, etc.). These long anticipated developments involved the need for new investigations of irradiated fuel behaviour in order to check the adequacy of the current criteria, evaluate the safety margins, provide new technical bases for modelling and allow an evolution of these criteria. Such an evolution is presently under discussion in France and several other countries, in view of a revision in the next coming years. For this purpose, a R&D strategy has been defined at IRSN.

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.

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.

Sign in / Sign up

Export Citation Format

Share Document