Analysis of the FPT-0, FPT-1 and FPT-2 experiments of the Phebus FP program investigating in-vessel phenomena during a LWR accidents
G. Repetto, B. Clément (1), S. Ederli (2). Analysis of the FPT-0, FPT-1 and FPT-2 experiments of the Phebus FP program investigating in-vessel phenomena during a LWR accidents.
Since the Three Mile Island accident in 1979, a worldwide effort has been undertaken to understand and model severe accident phenomena in nuclear reactors in case of a hypothetical loss of core cooling. The international Phebus Fission Product (FP) program, dealing with light water reactor source term research, was initiated in 1988 by the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN), and the Joint Research Centre of the Commission of European Communities (CEC), with contribution of most of the countries which are using nuclear power: France (Electricité de France “EDF”), USA (Nuclear Regulatory Commission “NRC”), Canada (CANDU Owners Group “COG”), Japan (Nuclear Power Engineering Corporation “NUPEC” and Japan Atomic Energy Research Institute “JAERI”), Korea (Korea Atomic Energy Research Institute “KAERI”) and Switzerland (Swiss Federal Nuclear Safety Inspectorate “HSK” and Paul-Scherrer-Institute “PSI”). The aim of the experimental program is to study the degradation phenomena and the behaviour of the fission products (FP) released in the reactor coolant system and the containment building. The program consists of six in-pile tests, performed under different conditions concerning the thermal hydraulics and the environment of fuel rods, in particular the amount of steam (strongly or weakly oxidising atmosphere). Four experiments have been successfully performed so far in 1993 (FPT-0), 1996 (FPT-1), 1999 (FPT-4) and 2000 (FPT-2).
This paper describes the results and the current status of the analysis of the core degradation aspects for the FPT-0 FPT-1 and FPT-2 tests, using the mechanistic ICARE/CATHARE code system developed by IRSN. The objective of those experiments was to get a significant FP release induced by the fuel rod degradation and fuel melting in a prototypical way using real materials as present in a PWR fuel assembly. The test section was made of a stainless steel cladded Ag-In-Cd absorber rod, surrounded by 20 PWR-typical fuel rods, 1 meter long, 18 of which being-irradiated (23 to 32 GWd/tU). A few days irradiation in situ allowed to produce a representative short-lived fission products inventory.
During the transient, most of the phenomena that could occur in-vessel during a PWR severe accident (thermo mechanical fuel rod rupture, absorber rod degradation, steam-zircaloy chemical reaction and hydrogen generation, fuel dissolution and molten pool formation) have been observed. They are quite well simulated with the ICARE2 V3mod1 code version, in particular the hydrogen generation.
The experiments showed a core degradation far beyond any other integral experiment (PBF SFD, Phebus-SFD, CORA, FLHT, and LOFT-FP-2). The severe damage observed in the bundle seems to be due to significant material interactions, initiated by structural materials possibly enhanced by the fuel swelling and changes in stoichiometry. The fuel burn-up and the oxygen potential during cladding oxidation are probably important factors. As a lesson learnt, fuel liquefaction and transition from rod like geometry to molten pool could occur at a temperature (2600+/-200K) largely below the actual melting point of the pure UO2 (3110K). Though the detailed modelling of such interactions has still to be improved, the ICARE2 code simulates fairly well the observed fuel degradation. Various types of nuclear power plants (recent PWRs, BWRs, VVERs) use a different neutron absorber material, namely the boron carbide (B4C), which might also largely influence both fuel degradation and fission product behaviour. Those phenomena will be investigated in the next FPT-3 experiment scheduled in 2004. The future Phebus-2K program, under consideration and foreseen from 2007, will investigate the degradation phenomena and FPs release on High Burn-up and MOX (mixed Pu-U oxide) fuel, the effect of air ingress and the consequences of reflooding a degrading core.
(1) : IRSN
(2) : ENEA