SCANAIR-BISON BENCHMARK ON CIP0-1 RIA TEST
In the frame of their research programs on fuel safety, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) and the Idaho National Laboratory (INL) have developed respectively SCANAIR and BISON codes to describe the thermo-mechanical behavior of irradiated fuel rods during Reactivity Initiated Accidents (RIA) in LWR. A RIA, characterized by a very rapid increase of reactivity and power in some rods of the reactor, can be schematically represented in two main phases. First, the energy deposition leads to a rapid rise of the fuel temperature which induces thermal swelling of the fuel pellets. During this phase, the Pellet Clad Mechanical Interaction (PCMI) leads to clad deformation and potentially to failure depending on fuel enthalpy increase and on the level of clad embrittlement. After the PCMI phase, the increase of clad temperature can lead to the boiling crisis of the water surrounding the rod. During the film boiling phase, the clad to coolant heat transfer becomes very low and the clad can reach high temperature (>700°C). Depending on the internal gas pressure, the ductile clad can undergo large deformation and possible failure. The RIA codes have to predict properly these two phases of the transient in order to be used, in particular, to assess current fuel safety criteria but also to be able to predict the behavior of upcoming fuel (such as accident tolerant fuel). In this paper, codes computation comparison against experimental data obtained on CIP0-1 test is done for the first phase of the transient, the PCMI phase, in order to point out the strengths and weaknesses of the codes. The CIP0-1 experiment was performed with UO2 high burnup fuel in the CABRI sodium loop facility, at 280°C and low pressure (~3 bars) without boiling crisis and no failure of the fuel rod. The initial state of the rod calculated by irradiation codes are compared with measurements before the transient. Then global rod thermal behavior during the CIP0-1 test is assessed with the sodium coolant temperature measurements. The rod mechanical behavior is analyzed with the fuel and clad elongation and the clad residual hoop strain. Finally comparison of the fission gas behavior between codes is performed and compared to the measurements of the fission gas released in the experiment.
- Research Organization:
- Idaho National Lab. (INL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- DOE Contract Number:
- DE-AC07-05ID14517
- OSTI ID:
- 1409689
- Report Number(s):
- INL/CON-17-41110
- Resource Relation:
- Conference: Water Reactor Fuel Performance Meeting 2017, Jeju Island, Korea, September 10–14, 2014
- Country of Publication:
- United States
- Language:
- English
Similar Records
High-burnup Experiments in Reactivity Initiated Accidents (HERA)
Reactivity Initiated Accident Test Series RIA Scoping Test Experiment Predictions