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|Title:||Assessment of Transuranus Fuel Performance Code against STUDSVIK Inter-ramp BWR Database|
|Authors:||ADORNI Martina; DEL NEVO Alessandro; ORIOLO F.; D'AURIA F.; VAN UFFELEN Paul|
|Citation:||Proceedings of the 17th International Conference on Nuclear Engineering, ISBN: 9780791838525 p. ICONE17-75704 (1-12)|
|Publisher:||American Society of Mechanical Engineers|
|Type:||Articles in periodicals and books|
|Abstract:||The fuel matrix and the cladding constitute the first barrier against radioactive fission product release. Therefore a defense in depth concept requires also the comprehensive understanding of fuel rod behavior and accurate prediction of the lifetime in normal operation and in accident condition as well. Investigations of fuel behavior are carried out in close connection with experimental research operation feedback and computational analyses. In this connection, OECD NEA sets up the ¿public domain database on nuclear fuel performance experiments for the purpose of code development and validation ¿ International Fuel Performance Experiments (IFPE) database¿, with the aim of providing a comprehensive and well-qualified database on UO2 fuel with Zr cladding for model development and code validation. This database includes the data set of the Studsvik Inter-Ramp BWR Project. The objectives of the project are to establish the failure-safe operating limits and the failure mechanism and associated phenomena, during power ramp tests, by varying the design parameters (i.e. cladding heat treatment, gap thickness and fuel density). The experimental data are used for the assessment of the Fission Gas Release (FGR) models implemented in the TRANSURANUS code versions ¿v1m1j07¿ and ¿v1m1j08¿. The starting point of the activity is the transient fission gas release model, the ¿TFGR model¿, specifically implemented in the last code version, to cover power ramp conditions. The paper presents the complete set of simulations of all twenty rods irradiated in the R2 research reactor and the corresponding comparisons with the experimental data. Sensitivity calculations are also performed to address the influence of geometric parameters and the choice of the different code options, relevant to model the FGR, on results.|
|JRC Institute:||Nuclear Safety and Security|
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