Márta Horváth

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 effects of long-term thermal aging on magnetization process in low-alloy pressure vessel steels using first-order-reversal-curves

We have investigated effects of long-term thermal aging at 550°C up to 10000 h on major-loop coercivity, hysteresis scaling of minor loops, and first-order reversal curves (FORCs) for low-alloy pressure vessel steels with low and high Ni contents. While major-loop coercivity and minor-loop coefficient of the scaling exhibit a gradual decrease with aging for high-Ni steel, those for low-Ni one are very weakly dependent on aging time. On the other hand, we found that FORC distribution becomes steep along both axes of interaction and switching fields and the peak shifts toward a lower switching field for both steels. Considering that there is no significant development of nanoscale precipitates during the aging as revealed with small-angle neutron scattering experiments, a relaxation of lattice strain in a matrix, possibly associated with diffusion of Ni atoms, may dominate magnetic properties at 550°C.

Insight on the inconsistencies of Barkhausen signal measurements for radiation damage on nuclear reactor steel

This paper focuses on the use of magnetic measurements, using Barkhausen signals to determine the irradiation effects, attempting to predict fracture toughness changes on nuclear reactor structural materials and correlating these measurements to mechanical testing and microstructure. For this study, two types of nuclear reactor materials were investigated: one sensitive to irradiation effects, the JRQ IAEAs reference material (A533B- -type); and one resistant material, 15KH2MFA WWERs reactor pressure vessel steel. The samples were carefully identified within the original heat block, i.e. forged or rolled plate. These calibrated samples were irradiated at different neutron fluences up to 1023 n/m2. We show how microstructural anisotropy can mask the irradiation effects in the magnetic measurements. A correlation between irradiation effects and the magnetic measurements is explained based on this study.