The CORA out-of-pile experiments are part of the international Severe Fuel Damage (SFD) Program. They were performed to provide information on the damage development of Light Water Reactor (LWR) fuel elements in loss-of-coolant accidents in the temperature range 1200 C to 2400 C. Test CORA-30 should investigate the fuel element damage behaviour with reduced initial heat-up rate of about 0.2 K/s compared to the normally used value of 1 K/s. The lower initial heat-up rate is representative for a loss-of-coolant accident which may develop after a shutdown of the reactor. The initial increase rate of the cladding temperature influences the thickness of the oxide layer reached at a given temperature. The thickness of the protective oxide layer i...
Die Quenchexperimente CORA-12, CORA-13 und CORA-17 zeigten ebenso wie die Inpile-Experimente LOFT LP...
Excessive oxidation, hydriding, and extensive irradiation damage occur in high-burnup fuel cladding,...
International audienceIn the event of a severe core meltdown accident in a pressurized water reactor...
The CORA-Out-of-pile experiments are part of the international Severe Fuel Damage (SFD) prgram. They...
The CORA out-of-pile experiments are part of the international Severe Fuel Damage (SFD) Program. The...
Die CORA-Out-of-pile-Experimente wurden im Rahmen des internationalen 'Severe Fuel Damage' Programms...
The CORA out-of-pile experiments were part of the international Severe Fuel Damage (SFD) program. Th...
The PWR-type assemblies usually consist of 25 rods with 16 electrically heated fuel rod simulators a...
An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental compari...
The 'Severe Fuel Damage' (SFD) experiments of the Kernforschungszentrum Karlsruhe (KfK), Federal Rep...
The major objectives of the experiment were to investigate the behavior of PWR fuel elements during ...
Copy held by FIZ Karlsruhe; available from UB/TIB Hannover / FIZ - Fachinformationszzentrum Karlsruh...
A loss-of-coolant accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechani...
Das Verhalten der Unfallsequenz ist durch folgende Schritte gekennzeichnet: 1. Der totale Ausfall al...
The Fukushima-Daiichi nuclear accident has highlighted the importance of analyzing the meltdown of a...
Die Quenchexperimente CORA-12, CORA-13 und CORA-17 zeigten ebenso wie die Inpile-Experimente LOFT LP...
Excessive oxidation, hydriding, and extensive irradiation damage occur in high-burnup fuel cladding,...
International audienceIn the event of a severe core meltdown accident in a pressurized water reactor...
The CORA-Out-of-pile experiments are part of the international Severe Fuel Damage (SFD) prgram. They...
The CORA out-of-pile experiments are part of the international Severe Fuel Damage (SFD) Program. The...
Die CORA-Out-of-pile-Experimente wurden im Rahmen des internationalen 'Severe Fuel Damage' Programms...
The CORA out-of-pile experiments were part of the international Severe Fuel Damage (SFD) program. Th...
The PWR-type assemblies usually consist of 25 rods with 16 electrically heated fuel rod simulators a...
An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental compari...
The 'Severe Fuel Damage' (SFD) experiments of the Kernforschungszentrum Karlsruhe (KfK), Federal Rep...
The major objectives of the experiment were to investigate the behavior of PWR fuel elements during ...
Copy held by FIZ Karlsruhe; available from UB/TIB Hannover / FIZ - Fachinformationszzentrum Karlsruh...
A loss-of-coolant accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechani...
Das Verhalten der Unfallsequenz ist durch folgende Schritte gekennzeichnet: 1. Der totale Ausfall al...
The Fukushima-Daiichi nuclear accident has highlighted the importance of analyzing the meltdown of a...
Die Quenchexperimente CORA-12, CORA-13 und CORA-17 zeigten ebenso wie die Inpile-Experimente LOFT LP...
Excessive oxidation, hydriding, and extensive irradiation damage occur in high-burnup fuel cladding,...
International audienceIn the event of a severe core meltdown accident in a pressurized water reactor...