Abstract
Abstract The post-analysis of control rod withdrawal end-of-life experiments carried out in the Phenix reactor during June 2009 is carried out at IGCAR by using the FARCOB and ERANOS 2.1 neutronics code systems. The measured results and the geometrical data required for this study is provided by the IAEA as a part of IAEA-CRP on “Control Rod Withdrawal and Sodium Natural Circulation Tests Performed during the PHENIX End-of-Life Experiments”. Four critical core states have been simulated with the average descriptions of the core. Parameters estimated are absorber rod worth, radial power distribution and the deviation in sodium heating change with respect to the reference state. All the predictions are based on 3-D diffusion theory calculations. This is the first experimental validation of radial power tilt due to control rod movement at full power with FARCOB code system. This study also enabled the benchmarking of calculation procedures and code systems being used at IGCAR for applications to oxide fueled sodium cooled fast power reactors. It is found that there is a good agreement in the predicted results from FARCOB and ERANOS 2.1 systems for this test.
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