Facilities

The activities covered the following three areas:

• Thermal hydraulic calculations described the cooling conditions possibly established during the incident.
• Simulation of fuel behaviour described the oxidation and degradation mechanisms of fuel assemblies.
• The release of fission products from the failed fuel rods was estimated and compared to available measured data.

The produced numerical results improved the understanding of the causes and mechanisms of fuel failures during the Paks-2 incident and provided new information on the behaviour of nuclear fuel under accident conditions.

The RUSET experimental programme was launched in 2002 at the Hungarian Academy of Sciences KFKI Atomic Energy Research Institute (AEKI). The aim of the program was to get data for assessment of ruthenium release at severe accident with air ingress. More than forty small scale tests have been performed with mixed powder components of inactive materials and with short fuel rods. The influence of temperature, air flow rate and the presence of other fission products on the gaseous Ru release and the retention role of fuel pellets and cladding have been investigated. The test series indicated that if an air ingress type severe accident occurs most of the initial Ru mass can be released from the reactor core to the containment or environment. Some part of the released gaseous Ru undergoes precipitation and deposits on the cold surfaces, another part is released in gaseous form. The deposited Ru oxides can serve as a secondary source for further gaseous Ru release.

Since the beginning of the 1990s’, several experimental series have been performed at the Hungarian Academy of Sciences KFKI Atomic Energy Research Institute (AEKI) with E110 and Zircaloy claddings. The aims of these experiments were to study and to compare the mechanical properties of the cladding materials and to investigate the effect of oxidation and hydrogen uptake on the mechanical performance of the claddings under accident conditions. The objectives have been achieved through separate effect tests with well defined conditions.



The main results of the tests have been collected into the Experimental Database of E110 Claddings under Accident Conditions. The database involves the data of several experimental series in the following main directories:

• Cladding ballooning tests (more than 170 tube tests).
• Tensile tests with tube and sheet specimens (more than 100 samples).
• Oxidation tests between 500-1200 °C.
• Ring compression tests (more than 100 samples).

Most of the tests were carried out with E110 cladding, but several experimental points were produced with Zircaloy-4 cladding as well for comparison purposes. The database includes not only the on-line measured data, but the results of post-test examinations (visual observations, metallographic analysis, SEM analysis). Experimental technical reports and some selected papers describing the tests are also available in the database.

The high temperature interaction of reactor core materials during a severe accident leads to the oxidation and melting of metal components. The interactions of molten Zr cladding with uranium-dioxide and zirconium-dioxide are important factors in the determination of fuel failure conditions and play a role in the loss of fuel rod like geometry and in the formation of debris bed and molten pool in the core.



The dissolution of uranium-dioxide and zirconium-dioxide by molten Zircaloy were investigated in earlier separate effect tests. UO2 dissolution experiments were carried out with UO2 crucible and Zr charge. The simultaneous UO2-ZrO2 molten Zircaloy dissolution was investigated in UO2 crucibles with Zircaloy charge and ZrO2 central rod [1]. The analysis of different experiments showed some discrepancy between the results, which was connected with different crucible sizes, UO2/Zr mass ratios and the melt surface to volume ratios [2]. The comparison of simultaneous dissolution tests of UO2/ZrO2 and separate UO2 and ZrO2 showed faster dissolution and larger extent of dissolution in the case of simultaneous tests. These observations emphasized the importance of prototypic conditions on the dissolution process. For these reasons experiments with short fuel rod segments were carried out in the KFKI Atomic Energy Research Institute, Budapest. The development of dissolution models needs more prototypical experiments, this task was addressed in the COLOSS project of the 5th FWP. The analytical support of the AEKI short fuel rod dissolution tests was provided by COLOSS partners [3],[4]. The results of the experimental series are expected to make possible the further model development and code validation. 2. OBJECTIVES

Complex measurements were performed at the integral high temperature test facility CODEX (COre Degradation Experiment) between 1995-2002 with electrically heated UO2 fuel rod bundles. The test matrix included the first VVER-440 type integral severe accident experiment. The results of a quench test with pre-oxidised bundle indicated the protective role of the external oxide scale. Unique experiments were performed with PWR bundles under air ingress conditions. The last test of the current series helped to resolve the methane production issue during the oxidaton of a boron-carbide control rod in a severe accident. Some experiments were related to the preparation of PHEBUS tests, and some others were performed parallel with similar QUENCH tests. The experimental results contributed to the general understanding of severe accident progression in the loss of rod-like geometry phase and the test data have been used and are available for model development and code validation purposes.

The CODEX out-of-pile integral test facility was built and put into operation in 1995 at the KFKI Atomic Energy Research Institute, Hungary in order to investigate some specific aspects of core degradation and to extend the experimental database for code valiadation and development. Some of the experiments were VVER specific, while others were of general interest for any light water reactor. The comparison of CODEX exepriments with CORA and QUENCH tests can help to sift out the effects related to the specific features or scaling of the facilities. Some new techniques (e.g. aerosol measurements) applied in the test facility provide additional information on the high temperature behaviour of core materials. For the investigation of the aerosol release a cascade impactor system is connected to the upper plenum of the cooler and two pipelines allowes the continuous measurement of aerosols by means of laser particle counters. The gas concentration in the off-gas system is measured using a quadropole mass spectrometer. The instrumentation of the facility consists of the measurements of the operational parameters as heating power, flowrates, temperatures, levels and pressures. Thermocouples are placed in several positions in the heat insulation material, on the heat shield, on the external surface of the shroud, on the fuel rods and inside of the central (unheated) rod. Two pyrometers and a video camera are located at three windows in the upper part of the bundle. After the experiments the post-test examination of the bundle and aerosol samples is carried out with several techniques, including metallography, SEM, microprobe analysis, X-ray radiography and mass spectrometry.
Facility is not in operation.

Reactor cavity flooding is a cornerstone of severe accident management strategy in Swedish type BWRs. In a hypothetical severe accident with core melting and reactor vessel melt-through, it is assumed that the melt ejected into a deep water pool will fragment, quench and form a porous debris bed coolable by natural circulation. If natural circulation cannot remove decay heat produced by the debris, then dryout, reheating and remelting of the debris bed is expected to occur. Attack of molten core materials on the reactor containment base-mat presents a threat to containment integrity. Amount of the heat which can be removed by natural circulation from the debris bed is contingent, among other factors, upon the properties of the bed as porous media. Debris agglomeration and especially formation of “cake” regions can significantly increase hydraulic resistance for the coolant flow and thus negatively affect coolability of the debris bed. If melt is not completely solidified prior to settlement on top of the debris bed, then agglomeration of the debris and even “cake” formation is quite possible.

The SIMECO (SImulation of MElt COolability) test facility consists of a slice type vessel, which includes a semi-circular section and a vertical section, representing the lower head of the reactor vessel. The size of the facility is scaled to be 1/8 of prototypic PWR type reactors. Fig.1 shows a schematic of the facility and Fig.2 shows the main dimensions of the vessel test section. The diameter and height of the test section are 620 mm and 530 mm, respectively. The width of the test section is 90 mm. The front and back faces of the facility are insulated in order to decrease heat losses. The vessel’s wall, represented by a 23-mm thick brass plate, is cooled by a regulated water loop. On the top of the vessel a heat exchanger with regulated water loops is employed to measure the upward heat transfer. The sideways and downward heat fluxes are measured by employing array of thermocouples at several different angular positions. Practically isothermal boundary conditions are provided at pool boundaries. A cable-type heater 3 mm in diameter and 4 m in length is submerged in the pool and provides internal heating. A heat exchanger mounted on the exit of cooling water, is employed to maintain the cooling capacity of the water. The isothermal bath is designed to provide constant temperature. A circulation pump was mounted in order to establish necessary flow rate. One digital and one analog flowmeter were mounted to measure water flow through the wall of the slice, while one analog flowmeter is used to measure the flow in the upper heat exchanger.

Facility is in operation.

FOREVER program at KTH was concerned with the vessel integrity under the molten corium attack in the reactor lower plenum during a severe accident. Total 9 tests were performed in the FOREVER program, to simulate the behavior of the lower head of a reactor pressure vessel (RPV) under different conditions: French steel/American steel, with/without penetrations, with/without gap cooling.



The facility employs a 1/10th scaled lower head (hemispherical in shape and made of SA533B, American reactor steel) of 400 mm outer diameter and 15 mm wall thickness. A cylindrical shell of 15Mo3 German steel, of 400 mm height and thickness of 15 mm, was welded to hemispherical lower head to make a complete vessel.

Facility is in operation.