Outcomes of the «steady-state crisis» experiment in the MIR reactor channel
To license nuclear fuel for nuclear power plants, data on the behavior of fuel elements under design-basis accidents are required. These data are obtained during tests of FAs and single fuel rods in research reactor channels, followed by post-test studies in protective chambers.
An accident with an unauthorized release of control rods from the reactor core leads to a pulsed channel power increase. It can proceed according to two scenarios: without a departure from nucleate boiling (DNB) on the fuel rod cladding at the final stage and with a DNB of the first type. To date, a series of experiments have been carried out according to the first scenario in the MIR reactor channel and the corresponding data on the behavior of fuel elements have been obtained. Preparation for conducting reactor experiments according to the second scenario is currently an urgent task.
The main experimental parameter that determines the behavior and the final state of fuel rods is their temperature. No experimental data were found on the critical heat flux for the rod bundles in the low coolant mass flow rate region (experiments in the MIR reactor channel can be conducted in the range of 200–250 kg/(m2s)). The available data are in the extrapolation range.
The «Steady-state crisis» experiment was conducted to obtain data on the critical heat flux value within the specified coolant mass rate range in the MIR reactor channel. The test object was a fragment fuel assembly composed of three shortened VVER-1000 fuel rods with a length of 1230 mm (the fuel part length is 1000 mm) installed in a triangular lattice at a pitch of 12.75 mm, which is a cell of the VVER-1000 core. This assembly configuration is used for in-pile tests to study the behavior of fuel rods under emergency conditions.
The results of the in-pile testing are presented. The paper shows the possibility of detecting the start and development of a type I cladding DNB based on the records of thermoelectric converters located inside inside the fuel rod kernel. As a result, the directly measured test parameters were used to determine the critical heat flux value.
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MIR reactor fuel rod experimental fuel assembly (EFA) departure from nucleate boiling (DNB) RIA (reactivity-initiated accident) thermoelectric converter (TEC) temperature coolant flow rate power critical heat flux