15th International Conference on 
Structural Mechanics in Reactor Technology

Severe Accident Research Activities in Korea: In-Vessel Retention of Molten Core Material
K.Y. Suh, S.H. Chang, H.D.Kim, S.J. Oh and J.L. Lee

EU Research on Severe Accident Phenomenology: Results of the Current EURATOM Framework Programme and Prospects for the Future
G. Van Goethem, G. Keinhorst, J.Martin Bermejo and A. Zurita

Experimental Program to Quantify the Effects of Severe Accidents for PWR Design
H. Plitz, F. Hofmann and W. Scholtyssek

CANDU Severe Accident Resistance
M. Bonechi and V.G. Snell

FOREVER Experimental Program on Reactor Pressure Vessel Creep Behavior and Core Debris Retention
B.R. Sehgal, R.R. Nourgaliev, T.N. Dinh and A. Karbojian

Thermal-Mechanical Behaviour of the Reactor Pressure Vessel and Corium Molten Pool in a Severe Accident with Core Melt Down
Cataldo Caroli, Adio Miliozzi and Francesco Milillo

Thermal and Mechanical Behaviorof Hemispherical Vessel Lower Head
K.Y. Suh, K.H. Kang, J.H. Kim, S.B. Kim, H.D. Kimn, J.S. Cho and J.E. Chang

Themmal and Mechanical Behavior of Metal Layer in Reactor Vessel Lower Head
J.S. Cho, K.Y. Suh, C.H. Chung, R.J. Park and S.B. Kim

COMETA/SEURBNUK-EURDYN: A Couplcd Code for Prediction of Reactor Vessel Response due to Fue1 Coolant Interactions
Arnold Yerkess, Alessandro Annunziato and Carmelo Addabbo

Reactor Pressure Vessel Head Louded by a Corium Slug Results of Model Experiments BERDA I
R. Krieg, B. Goller, G. Hailfinger, G. Messemer and G. Vorberg

FASTHER Failure under Strong Thermal Gradient of RPV Steel
C. Sainte Catherine, Ph. Mongabure, L. Le Ber, J. Devos and S. Bhandari

Experimental Study of the RBMK Fuel Channel Pressure Tube Behaviour under Accident Conditions
Vladimir D. Loktionov, Vladimir A. Gashenko, Olga N. Abakumova, Nikolay I. Yaroshenko and Viktor D. Tsirenov

Finite Element Analysis for Creep Damage of Coolant Piping in Light Water Reactor due to Local Heating under Severe Accident Condition
Seiya Hagihara, Noriyuki Miyazaki and Jun Sugimoto

Study for Therrnal and Structural Behavior of Reactor Pressure Vessel under External Reactor Vessel Cooling
Jong-Sung Kim and Tae-Eun Jin

Estimation of Therrnal and Mechanical Margin to Failure of Reactor Vessel Lower Head
H. J. Yoon, Y.H. Kim and K.Y.Suh

Advanced In- Vessel Retention Design against Water Reactor Core Melt Accidents
I.S. Hwang. K.Y. Suh, K.J. Jeong and S.D. Park

An Experimental Study of Dynamic Response of Model Reactor Structure Subjected to Internal Explosion
S.S. Dhadwal, H. Lal, V.S. Sethi, P. Chellapandi and S.B. Bhoje

Theoretical and Computational Models of the GASFLOW-II Code
John Travis, Jay Spore, Peter Royl and Gottfried Necker

Application and Limitation of Lumped Parameter and 3-D Codes in Nuclear Safety Research
M. A. Movahed

A Study on the Evaluation Methodology of DDT Potential in a PWR Containment during Severe Accidents
Seong-Wan Hong, Soo-Yong Park, Moonkyu Hwang and Hee-Dong Kim

Experimental Study on Hydrogen Behavior at a Subcompartment in the Containment Building
Un-Jang Lee and GOon-Cherl Park

Three-Dimensional Analysis of Steam/Hydrogen Transport with Catalytic Recombinersip Nuclear Reactor Containments Using the Computer Code GAS FLOW
P. Royl, H. Rochholz, J.R. Travis, G. Necker, W. Breitung, L. Seyffarth and W. Gehrig

Evaluation of the Containment Integrity due to Direct Containment Heating in Korean 1300 MWe PWR
Sang-Baik Kim, Rae-Jun Park, Kyoung-Ho Kang and Soo-Yong Park

Sensitivity Studies on Molten Core Concrete Interaction Using MELCOR Code
Seong Wan Hong, Dong Ha Kim, Jong Kyun Park, Song Won Cho and Soo Yong Park

A Comparative Analysis of CANDU Reactor Severe Accidents with PWR Using PSA Method
Y. Choi, Y.M. Song, S.D. Kim, D.H. Kim and Y. Jin