Modeling of BWR Inter-Ramp Project experiments by means of TRANSURANUS code

Davide Rozzia, Alessandro Del Nevo, Martina Adorni, Francesco D'Auria

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2 Citations (Scopus)


During the normal operation of a Light Water Reactor (LWR), the fuel-cladding gap may close, as a result of several phenomena and processes, including different thermal expansion and swelling of both the fuel and the cladding (Pellet Cladding Interaction - PCI). In this equilibrium state, a significant increase of local power (like a transient power ramp in the order of 100 kW/m h), induces circumferential stresses in the cladding. In presence of corrosive fission products (i.e. iodine) and beyond specific stress threshold depending on the material, cracks typical of stress corrosion may appear and grow-up (stress corrosion cracking - SCC). These cracks may spread out from the cladding internal surface, causing the fuel failure. Investigations on fuel behavior are carried out in close connection with experimental research, operation feedback and computational analyses. To address the issue of PCI/SCC, the "BWR Inter-Ramp Project" is investigated by means of TRANSURANUS code. The BWR Inter-Ramp Project is part of the OECD/NEA International Fuel Performance Experiments database. It provides the experimental data of 20 BWR fuel rods during power ramp. The objective was to establish the failure-safe operating limits of representative BWR fuel rods when subjected to power ramp tests after short to medium irradiation time. The burn-up ranges from 10 to 20 MWd/kgU. The aim of the activity is to compare, investigate and summarize the main outcomes achieved after the simulations of 20 Zircaloy-2 - UO2rods of various types and designs by TRANSURANUS code. Focus is given to the main variables, which are involved or may influence the simulation of cladding failure and characterize the power ramp tests. Sensitivity analyses are carried out to address the relevance of the knowledge of the boundary conditions, as well as the impact of selected parameters and code options on the simulations. The analyses bring to the conclusion that the code may under-estimate the pellet gaseous swelling (and consequently the cladding failures), of six fuel rods that experienced low power cycles at 10 MWd/kgU. The failure propensity of the remaining fourteen rods is conservatively simulated. © 2012 Elsevier Ltd. All rights reserved.
Original languageEnglish
Pages (from-to)238 - 250
Number of pages13
JournalAnnals of Nuclear Energy
Publication statusPublished - Dec 2012
Externally publishedYes


All Science Journal Classification (ASJC) codes

  • Nuclear Energy and Engineering

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