Alloys and pseudoalloys
Properties of nuclear fuel cladding tubes made of zirconium alloys
In light-water nuclear reactors, fuel element cladding tubes made of Zr alloys constitute the first barrier against the leak of fission products from UO2 pellets into the primary coolant circuit. In the environment of ongoing liberalization of the electricity market, competitiveness and economy of operation in conjunction with a high standard of nuclear safety are gaining in importance in the nuclear power sector. Given the increasing demands for nuclear fuel reliability with increased burnup, we must gain a deeper insight into the behaviour of fuel cladding materials in the reactor core in normal, abnormal and emergency conditions. In this context, the computer codes to model the behaviour of nuclear fuel call for improvement as well.
Research
UJP PRAHA has been engaged in a systematic research into zirconium alloy cladding tubes since 1972. Currently, our focus is on the long-term corrosion properties of such zirconium alloys as are in use at the two Czech nuclear power plants (Dukovany and Temelín), in comparison with alloys employed for Western PWR type reactors. Properties such as oxide layer thickness, oxide properties, and hydrogen content of the alloys are investigated by using state-of-the-art methods within co-operation projects with university-level institutions and the Czech Academy of Sciences, and the data obtained are input to the KOROZE. (CORROSION) Database, which currently contains some 50 000 data (the maximum sample exposure is 1600 d).
Methodologies developed by UJP PRAHA enable us to examine changes in oxide properties during water-steam-water temperature transitions, and the results help us gain a deeper insight into the mechanism of corrosion. Published results have been favourably accepted internationally (USA, Park City, Tahoe). With regard to fuel element behaviour in emergency conditions, research is aimed at oxidation, creep, and thermomechanical properties following high-temperature transitions of the LOCA (loss-of-coolant accident) type. Our efforts should result in a proposal for a conservative oxidation criterion, only dependent on the temperature development of the LOCA, which will assure a minimal plasticity of the cladding (implying that the cladding will not fail during the temperature shock and the fuel can be removed from the reactor).
In the domain of severe accidents, the problems of guide tube-control element interactions, UO2 pellet dissolution in (Zr, Fe) and (Zr, Ag) melts, and oxidation of U-Zr-O alloys that are formed during fuel element meltdown were addressed with success within the 4th and 5th EU Research Programmes.
The results show that significant reactions occur even at temperatures below 1200°C, which is the maximum admissible temperature for LOCA type accidents. As regards oxidation of U-Zr-O alloys, temperature escalation takes place at temperatures as low as 400°C – 500°C, which brings about much faster oxidation than the oxidation of the cladding tubes(as much as 100-fold).