To get an improved understanding and knowledge of the melt pool convection and the vessel creep and possible failure processes and modes occurring during the late phase of a core melt down accident the FOREVER-experiments are currently underway at the Division of Nuclear Power Safety of the Royal Institute of Technology Stockholm. These experiments are simulating the behaviour of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences in a depressurization scenario. Due to the multi axial creep deformation of the vessel with a non-uniform temperature field these experiments are on the one hand an excellent source of data to validate numerical creep models which are developed on the basis of uniaxial creep tests. On the other hand the results of pre-test calculations can be used to optimize the experimental procedure and by supporting decision making during the experiment. For that, a Finite Element model is developed based on a multi-purpose code. After post-test calculations for the FOREVER-C2 experiment, pre-test calculations for the forthcoming experiments are performed. Additionally metallographic post test investigations of the experiments are conducted to improve the numerical damage model and to adjust the correlation between the metallographic observations and the calculated damage. Taking into account both — experimental and numerical results — gives a good opportunity to improve the simulation and understanding of real accident scenarios. After analysing the calculations, it seems to be advantageous to introduce a vessel support which can unburden the vessel from a part of the mechanical load and, therefore, avoid the vessel failure or at least prolong the time to failure. This can be a possible accident mitigation strategy. Additionally, it is possible to install an absolutely passive automatic control device to initiate the flooding of the reactor pit to ensure external vessel cooling in the event of a core melt down.
Skip Nav Destination
Close
Sign In or Register for Account
10th International Conference on Nuclear Engineering
April 14–18, 2002
Arlington, Virginia, USA
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
0-7918-3595-2
PROCEEDINGS PAPER
Analysis and Insights About FE-Calculations of the EC-Forever-Experiments
H.-G. Willschuetz,
H.-G. Willschuetz
Forschungszentrum Rossendorf e.V., Dresden, Germany
Search for other works by this author on:
E. Altstadt,
E. Altstadt
Forschungszentrum Rossendorf e.V., Dresden, Germany
Search for other works by this author on:
F.-P. Weiss,
F.-P. Weiss
Forschungszentrum Rossendorf e.V., Dresden, Germany
Search for other works by this author on:
B. R. Sehgal
B. R. Sehgal
Royal Institute of Technology, Stockholm, Sweden
Search for other works by this author on:
H.-G. Willschuetz
Forschungszentrum Rossendorf e.V., Dresden, Germany
E. Altstadt
Forschungszentrum Rossendorf e.V., Dresden, Germany
F.-P. Weiss
Forschungszentrum Rossendorf e.V., Dresden, Germany
B. R. Sehgal
Royal Institute of Technology, Stockholm, Sweden
Paper No:
ICONE10-22262, pp. 595-601; 7 pages
Published Online:
March 4, 2009
Citation
Willschuetz, H, Altstadt, E, Weiss, F, & Sehgal, BR. "Analysis and Insights About FE-Calculations of the EC-Forever-Experiments." Proceedings of the 10th International Conference on Nuclear Engineering. 10th International Conference on Nuclear Engineering, Volume 1. Arlington, Virginia, USA. April 14–18, 2002. pp. 595-601. ASME. https://doi.org/10.1115/ICONE10-22262
Download citation file:
- Ris (Zotero)
- Reference Manager
- EasyBib
- Bookends
- Mendeley
- Papers
- EndNote
- RefWorks
- BibTex
- ProCite
- Medlars
Close
Sign In
3
Views
0
Citations
Related Proceedings Papers
Related Articles
Creep-Damage Analysis: Comparison Between Coupled and Uncoupled Models
J. Pressure Vessel Technol (November,2000)
The Influence of Crust Layer on Reactor Pressure Vessel Failure Under Pressurized Core Meltdown Accident
ASME J of Nuclear Rad Sci (October,2018)
Interaction Between Secondaries in a Thermal-Hydraulic Network
J. Dyn. Sys., Meas., Control (December,2006)
Related Chapters
Application of Probabilistic Methods for the Evaluation of Deterministic Deviations from Technical Specifications (PSAM-0277)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Insights and Results of the Shutdown PSA for a German SWR 69 Type Reactor (PSAM-0028)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
PSA Level 2 — NPP Ringhals 2 (PSAM-0156)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)