Transient tests on fast reactor fuel segments at the TRIGA-ACPR core

The Strategic Research Agenda of the EU Sustainable Nuclear Energy Technical platform requires new large infrastructures for its successful deployment. MYRRHA has been identified as a long term supporting research facility for all ESNII systems and as such put in the high-priority list of ESFRI. MYRRHA, under development by the Belgian Nuclear Research Centre (SCK•CEN), is a fast reactor system with lead-bismuth (LBE) cooling, driven by a particle accelerator. The goal of the FP7 MAXSIMA project is to contribute to the "safety in MYRRHA" assessment. In the frame of MAXSIMA project, transient experiments on MYRRHA type fuel segments have been started at RATEN ICN, aimed to validate the experimental chain and methodology. The task is being carried out in collaboration by RATEN ICN and SCK•CEN.

The irradiation device placed in the TRIGA-ACPR core central dry channel allows the exposure of three fuel test segments in stagnant liquid LBE in a neutron pulse of a few milliseconds. The core pulse configuration is computed according the desired energy deposition. The real energy deposition is determined by fission products from gamma activity measurements. The irradiation device is formed by a stainless steel capsule with electric heater containing the test fuel pins and surrounded by a Be structure aimed to increase the thermal neutron flux. The test fuel and LBE can be heated up to 300 0C before producing the neutron pulse.

SPND signal (neutron flux), temperatures in LBE, as well on the fuel cladding are recorded during the transient by using a Fast Data Acquisition System. The optimal solution for cladding temperature measurement has been not yet established, more tests being required.

The four test experiments that have been done so far allowed to finalize the experimental details (instrumentation, handling) and to develop computational models for pulse configuration computation. The irradiation tests have been done by using UO2 test fuel pins with different 235U enrichment depending on the desired level of energy deposition. With the first series of tests currently completed, no cladding damage has been observed.

The transient tests program will continue to validate the experimental solutions (instrumentation, on-line parameter measurements, post irradiation tests) as well the whole methodology for the determination of the fuel pin failure threshold for lead(-bismuth) cooled fast reactors.


Csaba Roth, RATEN ICN

Brian Boer, Belgian Nuclear Research Centre (SCK•CEN)