Lajos Matus

Interaction of Failed Fuel Rods Under Air Ingress Conditions

Abstract In the late phase of a severe reactor accident, the molten corium interacts with the vessel wall, and it can lead to the failure of the lower head. Through the failed bottom wall, part of the corium can flow into the cavity, and air can enter the primary circuit. The residual fuel in the core periphery will be further oxidized in air atmosphere. The degradation process will accelerate, and new chemical species will be formed, which can have an impact on the release of radioactive materials. Two experiments were carried out with electrically heated nine-rod pressurized water reactor-type bundles in the CODEX (COre Degradation EXperiment) facility to provide experimental data on the behavior of real fuel bundles under air oxidation conditions. The main objective of the tests was the investigation of oxidation phenomena, and some other important aspects (e.g., enhanced fission product release) were not addressed. The CODEX air ingress tests indicated the acceleration of oxidation phenomena and core degradation processes during the late phase of the vessel melt through accident, when air can have access to the residual fuel bundles in the reactor core. The degradation process was accompanied with zirconium-nitride formation and release of uranium-rich aerosols.

Ballooning experiments with VVER cladding

Abstract The results of single-rod and bundle ballooning tests with VVER (E110 type) cladding are presented. The comparative study of E110 and Zircaloy-4 showed a significant difference in behavior at 800 to 1000°C. The local maximum of mechanical strength was observed at a low oxidation rate. The pressurization rate played a considerable role in the burst conditions. The rate of the temperature increase and the iodine pretreatment did not significantly influence the mechanical behavior of the fuel rods under accident conditions in the investigated range of parameters. The maximum blockage rate observed in the bundle tests remained below 80%. The experimental data were collected into a database for model development and code validation purposes.

Behavior of VVER Fuel Rods Tested Under Severe Accident Conditions in the CODEX Facility

The early phase of severe accidents in VVER reactors was simulated in the CODEX (COre Degradation EXperiment) facility with electrically heated fuel rod bundles. The selected test conditions and applied measurement techniques made possible the observation of some specific phenomena, such as the protective role of oxide scale during quenching of high-temperature bundles, the composition of gases produced during the oxidation of boron-carbide control rods, and the interlink between the aerosol release and the oxidation process. The general behavior of the VVER bundles did not differ significantly from that of the Western-design light water reactor bundles tested under similar high-temperature conditions, but the experiments emphasized that the application of VVER-specific material properties and models is essential for comprehensive numerical simulations.

Investigation of aerosols released at high temperature from nuclear reactor core models

Abstract Two experiments were performed to simulate severe reactor accident with air ingress into the hot reactor core. The model bundles contained nine PWR type fuel rods. Their cladding was pre-oxidised by argon–oxygen (test 1) and steam (test 2). The released aerosol was measured continuously by laser particle counters. Morphology and elemental composition of the aerosol particles were studied on samples collected by impactors and quartz filters. The highest aerosol release was detected at the steepest rise of the bundle temperature. A second increase of the aerosol release appeared at the cooling down period. Because of the higher maximum temperature at test 2 about two orders of magnitude more uranium was released than in test 1. The highest emission was found for tin at test 1 and for zirconium and iron at test 2.

Experimental investigation of the late phase of spent fuel pool accidents

Experimental programmes have been carried out in order to investigate the behaviour of nuclear fuel components in high-temperature air atmosphere, which characterises the main conditions of the late phase of spent fuel pool accidents. The tests provided new data on the oxidation of zirconium cladding in different atmospheres, on the oxidation and release of ruthenium from fuel pellets and on the integral behaviour of fuel bundles. The integral test confirmed that water injection into the spent fuel storage pool is the right measure to terminate a severe accident.