Reactivity Initiated Accident (RIA) leads to an unwanted increase in fission rate and power in a region of the reactor core confined around the position of occurrence. The power excursion due to such events may cause fuel rods failures and a subsequent release of radioactive material into the primary coolant of reactor, in severe cases, this release could damage nearby fuel assemblies. In nuclear power plants, RIAs are due to control system faults, e. g. control elements ejection/insertion, or rapid changes in temperature or pressure of moderator. In Boiling Water Reactors (BWRs), the control rod drop accidents (RDAs) at cold zero power have been deeply investigated, in fact, notwithstanding they are less frequent in comparison with the control rod ejection event in PWRs, in this kind of plant these conditions are the most severe in case of a RIA occurrence. RDA transient, comprised in the design basis events considered in safety analysis, may cause rod failures especially at high burnup. To simulate a RIA, a peaked power pulse is applied to a pre-irradiated and re-instrumented rodlet aiming at investigating the most important phenomena that could lead to the rupture of cladding tubes. This paper is focused on the investigation of the TRANSURANUS fuel performance code capability to predict the thermomechanical state of rodlets subjected to RIA tests. To this purpose the FK-1 test, carried out at the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute (JAERI), was simulated. This experiment is part of a set of 12 tests performed at the NSRR facility to study the performance under a reactivity initiated accident of BWR rodlets with burnup between 41 and 61 MWd/kgHM. In the FK-1 test, a STEP I BWR rodlet, previously irradiated in the Fukushima Daiichi Nuclear Power Station (Unit 3) operated by the Tokyo Electric Power COmpany (TEPCO) up to 45 MWd/kgHM, was subjected to a peak enthalpy insertion of 544 J/g. In this paper the code findings for the FK-1 test are discussed on the basis of the experimental data and the predictions of other stand-alone codes for transient analysis. The FK-1 predictions of FRAPTRAN (2001), FALCON (2003) and SCANAIR (ver. 3–2) are reported. The choice of fuel relocation model and important cladding properties (swelling, thermal expansion, thermal conductivity) was made relying on preliminary calculations whose results are also presented. Notwithstanding a satisfactory agreement between predictions and experimental data and a good agreement in the presented code-to-code comparison were envisaged, these results also emphasized the need to improve the models for FGR, heat transfer to plenum. Investigations are also required to ascertain possible contribution from fission gas to pelles thermal expansion. Ongoing modeling activity, performed at the ITU Joint Research Centre, is focused on a new model for FGR, ENEA (Agenzia nazionale per le nuove tecnologie, l’energia e lo sviluppo economico sostenibile) is expected, in the near term, to give a contribution to refine the model for plenum gas temperature. These activities should improve the description of RIA transient and further investigations on NSRR tests will be performed with newly developed models. The work presented in this paper will be part of ENEA contribution in FUMEX III project leaded by the International Atomic Energy Agency (IAEA) and aimed at the improvement of fuel codes predictions at high burnup.

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