Investigation of Warm Prestress for the Case of Small ΔT During a Reactor Loss-of-Coolant Accident

1979 ◽  
Vol 101 (4) ◽  
pp. 298-304 ◽  
Author(s):  
F. J. Loss ◽  
R. A. Gray ◽  
J. R. Hawthorne
Keyword(s):  
Experimental Investigation ◽  
Statistical Analysis ◽  
Vessel Wall ◽  
Crack Extension ◽  
Crack Tip ◽  
Material Behavior ◽  
Virgin Material ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Failure Temperature

An experimental investigation was conducted to characterize the benefits of warm prestress (WPS) in limiting crack extension in the wall of a nuclear vessel during a LOCA-ECCS. The present research emphasized material behavior under conditions of a small ΔT between the temperature of WPS and the failure temperature as might occur during a LOCA. The results have demonstrated that fracture will not occur during a simultaneous unloading and cooling of the crack-tip region following WPS even though the critical KIc of the virgin material is achieved. Based on a statistical analysis, it is concluded that WPS produces an “effective” elevation in KIc; furthermore, it is suggested that this elevation will limit crack extension in the vessel wall so as to retain the coolant.

Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

2015 ◽  
Vol 285 ◽  
pp. 1-14 ◽  
Author(s):  
Asko Arkoma ◽  
Markku Hänninen ◽  
Karin Rantamäki ◽  
Joona Kurki ◽  
Anitta Hämäläinen
Keyword(s):  
Statistical Analysis ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

The Application of Best Estimate and Uncertainty Analysis Methodology to Large LOCA Power Pulse in a CANDU 6 Reactor

2002 ◽  
Author(s):  
A. Abdul-Razzak ◽  
J. Zhang ◽  
H. E. Sills ◽  
L. Flatt ◽  
D. Jenkins ◽  
...  
Keyword(s):  
Statistical Analysis ◽  
Uncertainty Analysis ◽  
Large Loss ◽  
Power Pulse ◽  
Analysis Methodology ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Safety Margins ◽  
Loss Of Coolant ◽  
Two Phases

The paper describes briefly a best estimate plus uncertainty analysis (BE+UA) methodology and presents its prototyping application to the power pulse phase of a limiting large Loss-of-Coolant Accident (LOCA) for a CANDU 6 reactor fuelled with CANFLEX® fuel. The methodology is consistent with and builds on world practice [1], [2]. The analysis is divided into two phases to focus on the dominant parameters for each phase and to allow for the consideration of all identified highly ranked parameters in the statistical analysis and response surface fits for margin parameters. The objective of this analysis is to quantify improvements in predicted safety margins under best estimate conditions.

Experimental Investigation of Liquid Entrainment in ADS-4 Branch Line

2014 ◽  
Author(s):  
Ducheng Sun ◽  
Jianchang Liu ◽  
Wenxi Tian ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
...  
Keyword(s):  
Experimental Investigation ◽  
Preliminary Test ◽  
Scaling Analysis ◽  
Entrainment Rate ◽  
Fourth Stage ◽  
Loss Of Coolant Accident ◽  
Test Loop ◽  
Loss Of Coolant ◽  
Vertical Up ◽  
Hydraulic Processes

In order to gain more insights into the system depressurization and entrainment behavior after actuation of the fourth-stage (ADS-4) valves during a loss-of-coolant-accident, the ADS-4 Depressurization and Entrainment TEst Loop (ADETEL) scaled to AP1000 was constructed to simulate the accident scenario with air-water and steam-water. A brief scaling analysis with emphasis on related thermal hydraulic processes was presented. Entrainment phenomena at vertical up tee branch were observed and analyzed. Preliminary test data of onset of entrainment and entrainment rate were collected with air-water tests and relevant conclusions were obtained.

scholarly journals A Study of Prediction Model Improvement for Air-Oxidation Breakaway in a Postulated Spent Nuclear Fuel Pool Complete Loss of Coolant Accident

2021 ◽  
Vol 13 (3) ◽  
pp. 1442
Author(s):  
Sanggil Park ◽  
Jaeyoung Lee ◽  
Min Bum Park
Keyword(s):  
Kinetic Model ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Complete Loss ◽  
Air Oxidation ◽  
Loss Of Coolant Accident ◽  
Spent Nuclear Fuel Pool ◽  
Loss Of Coolant ◽  
Model Derivation ◽  
Model Improvement

The temperature of zirconium alloy cladding on the postulated spent nuclear fuel pool complete loss of coolant accident is abruptly increased at a certain time and the cladding is almost fully oxidized to weak ZrO2 in the air. This abrupt temperature escalation phenomenon induced by the air-oxidation breakaway is called a zirconium fire. Although an air-oxidation breakaway kinetic model correlated between time and temperature has been implemented in the MELCOR code, it is likely to bring about unexpected large errors because of many limitations of model derivation. This study suggests an improved time–temperature correlated kinetic model using the Johnson–Mehl equation. It is based on that the air-oxidation breakaway is initiated by the phase transformation from the tetragonal to monoclinic ZrO2 at the oxide–metal interface in the cladding. This new model equation is also evaluated with the Zry-4 air-oxidation literature data. This equation resulted in the almost similar air-oxidation breakaway timing to the actual experimental data at 800 °C. However, at 1000 °C, it showed an error of about 8 min. This could be inferred from the influence of the ZrN phase change due to the nitrogen existing in air.

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