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This paper presents simulations of the xM3 power ramp with the fuel performance code ALCYONE performed during an international simulation exercise organized within the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Power to Melt and Maneuverability project. The xM3 test involved a large-grain UO2 fuel from Mitsubishi Heavy Industries cladded with Zirlo and pre-irradiated in a Spanish pressurized water reactor up to an average burnup of 27 GWd/tU−1. It was then submitted to a staircase ramp protocol in the R2 reactor at Studsvik (Sweden) with 10 successive steps of 5 kW·m−1 up to a ramp terminal level of 70 kW·m−1. The fuel rodlet did not fail, and detailed post irradiation examinations performed during the Studsvik Cladding Integrity Project II evidenced recrystallization of the pellet center around a central hole, interpreted as signs of fuel melting. In this paper, simulations with ALCYONE of the xM3 power ramp, including an advanced model for fuel melting based on thermodynamic equilibrium calculations, are detailed. The model relies on the determination of the liquid fuel fraction evolution with temperature that is used to obtain a continuous description of the material properties during phase change. In consequence of the incorporation of rare earths and actinides in the bulk of the fuel, distinct solidus and liquidus temperatures are estimated. It is shown that the observed central hole and recrystallized central part of the pellet could be the consequence of totally melted fuel (liquidus is reached), partially melted fuel (solidus is reached), or pore migration only. [ABSTRACT FROM AUTHOR]

This paper presents the results of the Power To Melt and Maneuverability (P2M) Simulation Exercise on past fuel melting irradiation experiments, organized within the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Framework for IrraDiation ExperimentS (FIDES) framework by the Core Group (CEA, EDF, and SCK‧CEN) and open to all FIDES members. The exercise consisted in simulating two past power ramps where fuel melting was detected: (1) the xM3 staircase power transient [ramp terminal level (RTL) 70 kW‧m−1, average burnup 27 GWd‧tU−1], carried out in 2005 in the R2 reactor at Studsvik (Sweden), where the rodlet maintained its integrity, and (2) the HBC4 fast power transient (RTL 66 kW‧m−1, average burnup 48 GWd‧tU−1), carried out in 1987 in the BR2 reactor at SCK‧CEN (Belgium), where the cladding failed during the experiment. The exercise was joined by 13 organizations from 9 countries using 11 different fuel performance codes. In this paper, the main results of the Simulation Exercise are presented and compared to available postirradiation examinations (PIE) or on-line measurements during the power ramps (fuel and clad diameters, rod elongation, pellet-clad gap, and fission gas release). Since the focus of the Simulation Exercise is on fuel melting assessment, determination of the boundary between melted/nonmelted fuel and the consequent definition of a melting radius from PIE are first discussed. During the HBC4 ramp, fuel melting was predicted by most of the codes despite differences in the melting models. Higher discrepancies were observed for the xM3 rod that can be attributed partly to power uncertainty and partly to the limited capability of the models to describe partial melting of the fuel during this ramp. Finally, possible code developments to improve simulation results are presented. [ABSTRACT FROM AUTHOR]

This paper presents 2D(r, $\it \theta $ θ) simulations of the HBC-4 power-to-melt experiment performed with the fuel performance code ALCYONE. The HBC-4 experiment is one of the two test cases selected for the simulation exercise on past fuel melting experiments of the Power to Melt and Maneuverability (P2M) project. The ramp terminal level (RTL) at peak power node (PPN) has been estimated at 66 kW·m−1 by gamma scanning and 70 kW·m−1 based on online measurements of thermal fluxes. The fuel burnup at PPN was close to 60 GWd/tU−1. The cladding failed during the short holding time at a RTL of 40 s. Fuel melting took place at the pellet center, and in particular, in front of clad cracks. In this paper, simulations of the HBC-4 power-to-melt experiment are performed using an updated version of the 2D(r, $\it \theta $ θ) scheme of ALCYONE where half of the fuel pellet is described. This configuration allows for the modeling of clad failure by iodine stress corrosion cracking and of its consequences on fuel pellet deformation. The modeling of fuel melting relies on thermochemical equilibrium calculations performed with the OpenCalphad Gibbs Energy Minimizer and the Thermodynamics of Advanced Fuels International Database. The simulation without clad failure indicates that the solidus is reached during the HBC-4 experiment but not the liquidus. The simulation with clad failure leads to a small increase in the fuel temperature that is sufficient to reach the liquidus at the pellet center, in agreement with postirradiation examination (PIE). The impact of water ingress in the rod and vaporization at the pellet surface is discussed, showing that it could explain the pronounced swelling of the fuel pellet reported from the PIE. [ABSTRACT FROM AUTHOR]

To refine knowledge about pressurized water reactor fuel melting, the Power to Melt and Maneuverability (P2M) project in the Framework for Irradiation Experiments II (FIDES-II) managed by the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development aims to perform two irradiation tests to reach fuel centerline partial melting using instrumented experimental fuel rodlets irradiated in a pressurized water capsule (PWC) of the Belgian Reactor 2 (BR2) [Studiecentrum voor Kernenergie (SCK CEN), Belgium]. Prior to these experiments, two preliminary tests will be performed in the PWC for qualification purpose toward safety. The experimental rodlets will be instrumented with a fuel centerline thermocouple in the lower part of the fuel and a pressure sensor (PS) in the upper part. An objective of the P2M experiments is to determine the amount of fission gas released from the fuel thanks to the PS measurement. The plenum pressure evolution is known from the measurement, and knowledge of the gas plenum temperature is required to determine the amount of fission gas released from the fuel to the plenum. The gas plenum temperature will depend on natural convection of the PWC coolant, and no temperature measurement will be possible within this gap during the test. This paper describes the setting up of a model of the BR2 PWC equipped with the P2M rodlets based on the coupled NEPTUNE_CFD (multiphase fluid calculation code) and SYRTHES (solid thermal module) simulation tools. Simulations performed thanks to this model allow assessment of the thermal-hydraulic (TH) behavior of the pressurized water in the capsule and the thermal behavior of the rodlet, in particular, regarding the temperature of the gas located in the plenum. For consistency of the results, computations presented in this paper were performed using a single and consistent set of TH models. A mesh sensitivity analysis was carried out for all the studied cases. Simulation results related to the water capsule behavior were found to be in good agreement with the available experimental data. The gas plenum temperature results obtained from this study will be used to assess the fission gas release during the test from the plenum pressure measurements. Overall experimental validation of fission gas release during the test will be possible after the transient test in BR2 based on the postirradiation examination program foreseen on the rodlets at the LECA-STAR facility [Commissariat à l'énergie atomique et aux énergies alternatives (CEA) Cadarache, France]. [ABSTRACT FROM AUTHOR]