scholarly journals Regulatory analysis for the resolution of Generic Safety Issue 105: Interfacing system loss-of-coolant accident in light-water reactors

1993 ◽  
Author(s):  
Keyword(s):  
Safety Issue ◽  
Light Water Reactors ◽  
Light Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Water Reactors ◽  
Regulatory Analysis

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 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.

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.

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.

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