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.

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

Analysis of Possibility of Pressure Tube Cold Pressurization During ECCS Injection for a Small Break LOCA

2002 ◽  
Vol 124 (4) ◽  
pp. 483-486 ◽  
Author(s):  
D. Mukhopadhyay ◽  
S. K. Gupta ◽  
V. Venkat Raj
Keyword(s):  
Control System ◽  
System Design ◽  
Heavy Water ◽  
Pressure Control ◽  
Severe Accident ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Pressure Control System ◽  
Safe Limits ◽  
Water Reactors

ECCS is designed to keep the reactor fuel temperatures within safe limits. The paper describes an additional criterion for Indian pressurized heavy water reactors (IPHWRs) evolving from the need to avoid a small break loss of cooling accident (LOCA) developing into a more severe accident. During a small break loss of coolant accident (LOCA) in PHWRs, the hydro-accumulators ride on the system and inject emergency coolant. The atmospheric steam discharge valves (ASDVs) open and cool the system due to energy discharge. In addition, the pressure control system tends to maintain the pressure. Depending on the system design, this could lead to cold pressurization of the system. This paper examines this issue.

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.

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.

Investigations on Catalyst Systems for Catalytic Recombiners

2002 ◽  
Author(s):  
Ernst-Arndt Reinecke ◽  
Peter Broeckerhoff ◽  
Inga M. Tragsdorf
Keyword(s):  
Catalyst Activity ◽  
Heat Removal ◽  
Gaseous Mixture ◽  
Severe Accident ◽  
Experimental Investigations ◽  
Light Water Reactors ◽  
Reaction Heat ◽  
Conversion Rates ◽  
Water Reactors ◽  
Catalyst Systems

Passive Autocatalytic Recombiners (PARs) are used for hydrogen removal in the containments of Light Water Reactors after a severe accident. These devices make use of the fact that hydrogen and oxygen react exothermally on catalytic surfaces already at low temperatures generating steam and heat. One major concern is the fact that existing recombiners bear the risk of ignition of the gaseous mixture by overheated catalytic substrates or parts of the casing, since the heat generated is not removed by cooling systems. Overheating may occur due to insufficient heat removal. Experimental investigations on existing systems show that the highest temperatures appear near the leading edges of the catalyst sheets. Furthermore, local conversion rates are too high not allowing sufficient reaction heat removal by convection. Possible countermeasures are additional cooling or limiting local conversion rates. At FZJ investigations are made on adapting the catalyst activity according to the requirements by using electro-plating technology instead of washcoat coatings, allowing well defined coating densities. Substrates with corresponding coatings have been tested, proving their ability in mixtures up to oxygen limitation. Different substrate materials and pre-treating measures are investigated for optimizing the surface properties. SEM-studies give insight in the surface structure and allow detailed analysis of the catalyst activity.

scholarly journals Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes: The FPT-0 Test

2017 ◽  
Vol 2017 ◽  
pp. 1-25 ◽  
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
Fission Products ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Light Water Reactors ◽  
Control Rod ◽  
Light Water ◽  
Severe Accidents ◽  
Fuel Bundle ◽  
Water Reactors

The integral Phébus tests were probably one of the most important experimental campaigns performed to investigate the progression of severe accidents in light water reactors. In these tests, the degradation of a PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the results of such tests, numerical codes such as ASTEC and MELCOR have been developed to describe the evolution of a severe accident. After the termination of the experimental Phébus campaign, these two codes were furthermore expanded. Therefore, the aim of the present work is to reanalyze the first Phébus test (FPT-0) employing the updated ASTEC and MELCOR versions to ensure that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The analysis focuses on the stand-alone containment aspects of this test, and the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products behavior. This paper is part of a series of publications covering the four executed Phébus tests employing a solid PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3.

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