| Article ID | Journal | Published Year | Pages | File Type |
|---|---|---|---|---|
| 297051 | Nuclear Engineering and Design | 2012 | 9 Pages |
Irradiation behavior of medium and high burnup BWR fuels during base-irradiation and subsequent power ramp test is analyzed by a fuel performance code FEMAXI-7. The code has a 1.5-D cylindrical geometry (4 axial segments) to have a coupled solution of thermal analysis and FEM mechanical analysis. Two kinds of target fuels are selected; one was subjected to a power ramp test in the DR3 reactor at RISO after the base-irradiation in a commercial BWR, and the other was subjected to the power ramp test in the DR3 reactor after the base-irradiation in the Halden boiling water reactor. The calculated values such as fission gas release after the base-irradiation and a cladding diameter profile before and after the ramp test show a reasonable agreement with measured data. In addition, the calculated ridging deformation of the cladding before and after the ramp test, which is obtained by using a local pellet-cladding mechanical interaction (PCMI) analysis geometry in FEMAXI-7, is compared with the measured data, and it is found that the FEMAXI-7 code is applicable to the local PCMI analysis of medium and high burnup rods under normal operation and power ramp conditions.
► Two power ramp experiments of BWR fuels were analyzed by FEMAXI-7 code. ► Calculated FGR and cladding deformation showed reasonable agreement with PIE data. ► High temperature FGR could be predicted by the enhanced Turnbull FG diffusion constant. ► Local PCMI model in the code could reasonably predict cladding ridging deformation.
