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Cabri reactor safety reassessment



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S. Massieux, D. Blanc, J. Couturier,
TRTR-IGORR 2005, Joint Meeting of the National Organization of Test,
Research and Training Reactors and the International Group on Research Reactors, Gaithersburg, Etats-unis, 12-16 septembre 2005,
Rapport DSR 77


The CABRI reactor (P = 25 MW) is a pool type research reactor devoted to study the behaviour of irradiated fuel pins during fast power transients, in particular for reactivity insertion accident (RIA). The test may lead to cladding failure or melting, which may lead to thermomechanical interaction between fuel and the coolant.
A test loop is installed at the centre of the "driver" core. It receives an experimental device containing the irradiated fuel pin and the instrumentation and has its own cooling system. Reactivity insertions are provided by the depressurisation of tubes filled with helium 3. Starting from an initial value of about 100 kW, the core power can reach about 10 GW in few milliseconds.
The tests carried out between 1978 and 1997 were related to fast breeder reactors (FBR). So the coolant of the loop was sodium. Now the sodium loop is removed and will be replaced by a pressurised water loop (T=350°C; P=15.5 MPa) in order to test UO2 or MOX fuel pins with high burn-up.
In the framework of the authorization needed to perform the above modifications, the operator sent to the French Safety Authority a preliminary safety analysis report in concern with:
- the design, manufacturing and installation of the pressurised water loop in the centre of the CABRI core in replacement of the sodium loop,
- a facility safety re-assessment with a particular attention for driver core, equipment and circuits.
Upon a request from the French Safety Authority, the preliminary safety analysis report was reviewed by the IRSN (Institut de Radioprotection et de Sûreté Nucléaire) and the results of this review were presented, in 2004, to the Standing Group of Experts (Groupe permanent), which is an advisory group. The IRSN assessment concerned the following points:
- the design and the sizing of the water loop equipment taking into account the fact that the fuel-water thermomechanical interaction is a normal operating condition,
- the use of Zircalloy for pressure parts of the loop put inside the reactor core. Indeed, the use of Zircalloy, which is a fragile material, is not planned in design codes (ASME, RCC-M, etc.),
- the accidents and the associated radiological consequences,
- the facility safety re-assessment with a particular attention to its seismic behaviour, with an intended lifetime of 20 years.

The preliminary safety analysis report received the agreement of the Standing Group of Experts subject to operator's answers to demands in concern mainly with the Zircalloy sizing criteria, the seismic behaviour of the facility, the pressure increase due to fuel-water interaction in the test loop. A provisional safety analysis report, including operator's answers to the above demands, will be sent to the Safety Authority at the beginning of 2006 in view of a commissioning licence during 2007.