Keramische Brennstoffkapselungen für Druckwasserreaktoren mit verbesserten Sicherheitsmerkmalen (English shop)

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In order to maintain the peaceful use of nuclear energy as an option for future energy supply, a marked increase in the safety characteristics of nuclear facilities is required. Catastrophic core meltdown accidents must either be ruled out or, if this is not possible, their consequences must remain confined to the plant itself.

This means that the four stability criteria of reactor safety must be fulfilled at all times and in every plant condition. In this context, long-term residual heat removal following an extreme accident constitutes a central problem. For this reason, passive plant concepts are increasingly being employed in nuclear engineering, which ensure long-term residual heat removal by means of self-acting mechanisms such as heat conduction, thermal radiation and natural convection, so that the use of active systems, which always entail a potential probability of failure, can be dispensed with.

Within the framework of this thesis, a concept for a modified pressurised water reactor was presented which exhibits markedly improved safety characteristics compared with conventional systems. The essential modifications consist, on the one hand, in an innovative fuel element concept in which the uranium dioxide fuel is encapsulated with pressureless sintered silicon carbide (SSiC) and, on the other hand, in the filling of the reactor core with lead in the event of an extreme accident, whereby residual heat removal can be considerably improved.

The focus of the investigations lay in demonstrating the chemical and mechanical stability of the fuel element concept presented. To this end, in the experimental part of the work the high-temperature joining technique for SSiC described by Henkel was first optimised in such a way that encapsulation of the UO2 was achieved with reproducibly high quality. In this process, the joining surfaces of the SSiC bodies are provided with a high-carbon-content interlayer and subsequently positioned against one another. A silicon donor compound is applied around the joining zone. In the subsequent high-temperature process, the silicon diffuses into the joining zone and reactive siliconisation of the interlayer takes place. By varying the main influencing parameters temperature, holding time, composition of the carbon slurry and composition of the silicon donor compound, material composites of reproducibly high quality could be produced with this technique. At a temperature of 1650 °C and a holding time of 15 minutes, cross-section analyses demonstrated a largely pore-free joining zone with a maximum free silicon content of 5 %. The mechanical strength was determined in 4-point bending tests as 389 MPa with a Weibull modulus of 6.5.

Using the optimised test parameters, several batches of SSiC capsules with internal graphite and/or UO2 pellets were joined and investigated with regard to their behaviour under reactor-typical normal conditions and under accident loads.

For the specimens exposed to a radiation exposure corresponding to that of one year of operation in a PWR, the gas tightness of the capsules could be confirmed by helium leak tests. The strength decreased by approximately 35 % to 252 MPa with a Weibull modulus of 5. The losses in the mechanical properties of the SSiC capsules were in the same range as those of the monolithic reference specimens.

To assess the chemical stability of the fuel element under accident conditions, SSiC, Zircaloy and the cladding materials proposed as alternatives, Nicrofer 3220 H and Nicrofer 6025 HT, were exposed to lead melts in a temperature range up to 1100 °C with a maximum holding time of 5 hours. With the exclusion of air, only extremely slight corrosive damage was observed for SSiC up to the maximum temperature. For the nickel-base alloys, the corrosion resistance decreased markedly from 1000 °C upwards.

For the automated quality control of the joined SSiC capsules, computed tomography (CT) has proved to be fundamentally suitable as a non-destructive method with a view to production on an industrial scale. Deliberately introduced defects could be clearly detected. For optimal use of this method, however, special adaptations of the equipment technology should be carried out in order to be able to exploit the resolving power of the CT method in full.

FEM analyses of the fuel element concept provided results regarding the resulting temperature fields and the thermally induced stresses arising from them. The maximum loads occur when using Nicrofer 6025 HT as cladding material in the hot channel in the irradiated condition. Within the UO2 pellet, the temperatures rise to 1283 °C in the steady state. Under these conditions the ceramic encapsulation is subject to a tensile stress of 308 MPa. This value corresponds to approximately 90 % of the maximum permissible equivalent stress.

On the basis of the results obtained, it can be stated that the fuel element concept presented possesses markedly improved safety characteristics compared with conventional systems. Encapsulation of the fuel with SSiC introduces a highly corrosion-resistant fission product barrier which withstands the highest thermal and mechanical loads.

If the reactor core is designed in such a way that in the event of an accident the fuel elements are completely covered by the injected lead granulate or by the resulting lead melt, thus ensuring the exclusion of oxygen, the mechanical stability of the SSiC encapsulation can be considerably increased.

Further investigations should address the following points in greater detail:

The SSiC capsules used can be improved with regard to their geometry. Especially in the region of the transition from the cylindrical wall to the end face, the encapsulation must be designed in such a way that the resulting thermal stresses are reduced and thus a greater safety margin against mechanical failure of the ceramic is achieved.

The substitution of the metallic cladding materials by SSiC should be investigated more thoroughly. Since SSiC is highly corrosion-resistant to water vapour, its use could considerably reduce the risk of hydrogen formation in the containment during an accident and thus the possibility of a deflagration.

Finally, an assessment of the concept with regard to the final disposal of the spent fuel elements must be carried out. Although SSiC possesses excellent resistance to acids and alkalis, experiments on the long-term behaviour of SSiC under the chemical conditions of a final repository are necessary.