Zoltán Hózer

Core Loss during a Severe Accident (COLOSS).

KFKI Atomic Energy Research Institute (AEKI), Hungary Electricité de France (EDF), France Ente per le Nuove Tecnologie, l’Energia e l’Ambiente (ENEA) Italy Framatome ANP, France Forschungszentrum Karlsruhe GmbH (FZK), Germany European Commission – JRC/IE, International European Commission – JRC/ITU, International Paul Scherrer Institut (PSI), Switzerland Framatome ANP Gmbh, Germany SKODA-UJP Praha a.s., Czech Republic Universidad Politécnica de Madrid (UPM), Spain Ruhr-Universität Bochum (RUB), Germany Universität Stuttgart (IKE), Germany University Lappeenranta, Finland

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.

Safety analysis of a VVER-440 spent fuel storage pool

Numerical analysis has been carried out for the spent fuel pools of the Paks Nuclear Power Plant (NPP), where the fuel assemblies are stored for several years after discharge from the reactor. The performed calculations covered both severe and design basis accidents. The results indicated that the spent fuel pool accidents may have severe consequences. For this reason new procedures have been introduced at the NPP in order to guarantee sufficient cooling of fuel assemblies even in the case of accidents.

Building Up an On-Line Plant Information System for the Emergency Response Center of the Hungarian Nuclear Safety Directorate

Abstract The main design features, services, and human-machine interface characteristics are described of the CERTA VITA on-line plant information system developed and installed by KFKI AEKI at the Nuclear Safety Directorate (NSD) of the Hungarian Atomic Energy Authority (HAEA) in cooperation with experts from the NSD. The Center for Emergency Response, Training, and Analysis (CERTA) located at the headquarters of NSD, Budapest, Hungary, was established in 1997. The center supports the NSD installation, radiological monitoring, and advisory team in case of nuclear emergencies, with appropriate hardware and software for communication, diagnosis, prognosis, and prediction. The vital information transfer and analysis (VITA) system represents an important part of the CERTA, as it provides for the continuous remote inspection of the four VVER-440/V213 units of the Hungarian Paks nuclear power plant (NPP). The on-line information system maintains a continuous data link with the NPP through a managed leased line that connects CERTA to a gateway computer located at the Paks NPP. The present scope of the system is a result of a 4-yr development project: In addition to the basic safety parameter display functions, the VITA system now includes an on-line break parameter estimation module, an extensive training package based on simulated transients, and on-line data transfer capabilities to feed accident diagnosis/analysis codes.

Activity release from the damaged spent VVER-fuel during long-term wet storage

Abstract An ex-core fuel damage incident took place at Unit 2 of Paks Nuclear Power Plant in Hungary on the 10th April 2003. After this event the damaged fuel assemblies were stored under water for four years. During wet storage a continuous activity release was observed. The evaluation of the measured activity concentration showed that the UO2 mass released from the fuel into the coolant was ≈1.8% of the total fuel mass. Furthermore this paper contains the calculation methods and the calculated activity release of the main analysed isotopes.

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.

Safest Roadmap for Corium Experimental Research in Europe

Severe accident facilities for European safety targets (SAFEST) is a European project networking the European experimental laboratories focused on the investigation of a nuclear power plant (NPP) s ...