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The OVER-RAMP Project investigated the failure popensity of typical PWR fuel in the form of test rods when subjected to fast power ramps.
Thirty nine rods, of two different origins and designs, forming nine groups of rods in regard to variations in design and material parameters, were base irradiated in a power reactor environment(the Kernkraftwerk Obrigheim and the Mol reactor BR-3) at heat rates in the range of 14 to 25 kW/m to burnups in the range of 12 to 31 MWd/kg U and subsequently ramp tested in the research reactor R2 at Studsvik, Sweden.
The rods underwent a thorough examination program, comprising characterization prior to the base irradiation, examinations interim to base and ramp irradiations and examinations after the ramp irradiation.
The principal objective of the OVER-RAMP Project was to make a substantial contribution to the understanding of the pellet-clad interaction (PCI) for commercial type PWR reactor fuel under power ramp conditions so as to acquire greater ability to assess its potential limitations.
The main technical objectives of the OVER-RAMP Technical Program were the following:
Establish through experiments the power ramp fuel failure threshold as a function of fuel burn-up.
Determine the influence on the failure threshold of various design and material parameters.
Determine the influence on the failure threshold of various ramp parameters (for fuel rods with identical sets of design parameters).
Identify fuel failure modes and mechanisms, as well as characteristic stages of fuel failure.
Provide data for PCI failure analysis and for predictive fuel modelling by recording changes in rod profiles, dimensions, structures, properties, etc.
Five rods were ramp tested with ramp rates lower than about 100 W/cm x min. The influence of the lower ramp rates was not conclusive.
In order to meet these objectives, much effort went into recording the power history in detail, and to determine the fuel rod changes based on a thorough pre-irradiation characterization, non-destructive examinations before and after the power ramping, and selected destructive examinations of rods following the power ramping irradiation.
The program power ramped 39 individual test fuel rods of 2 different origins and designs.
24 of the rods were of KWU/CE design and were provided by KWU. They were delivered to the Project following base-irradiation in a power reactor environment (Kernkraftwerk Obrigheim, FR Germany).
15 of the rods were of Westinghouse design and were delivered by W after irradiation in the BR-3 reactor at Mol, Belgium. The burn-up levels ranged from 12 to 31 MWd/kgU achieved at heat ratings of 14-25 kW/m (average over time, at axial peak position).
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Keywords: experimental data, fuel elements, fuel rods, fuel-cladding interaction.