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INTRODUCTION: The main objective of accident-tolerant fuel (ATF) development is to withstand high-temperature accident conditions in nuclear reactors. An important part of ATF fuel qualification involves experimental testing of fuel components in both in-pile and out-of-pile test facilities. The CODEX experimental program included two tests with ATF fuel materials: the CODEX-ATF test, conducted in a steam atmosphere, and the CODEX-ATF-AIT test, which simulated an air ingress scenario.
1. THE CODEX FACILITY
The CODEX facility had hexagonal bundles with 7 electrically heated fuel rods in both experiments. The test section was similar to the previous CODEX experiments [1][2][3]. The height of test section was 1000 mm, but the length of the rods was only 650 mm (CODEX-ATF) and 910 mm (CODEX-ATF-AIT), so the upper part of the shroud was used for thermocouples, without fuel rods. The fuel rods were filled with zirconia pellets and heated by tungsten wires. The rods were individually pressurised. In the CODEX-ATF test the seven-rod bundle consisted of three uncoated and four Cr coated optZIRLOTM cladding tubes and uncoated spacer grids and shroud were used. In case of CODEX-ATF-AIT tests all rods, spacer grids and the shroud were coated by chromium. The Zr shroud was surrounded by several thermal insulation layers, electrical heaters and a steel tube (Fig. 1). The Cr coating of components was carried out at the Czech Technical University in Prague. The main parameters of the test section are summarised in Table 1.
FIG. 1. Horizontal cross section of the test section with thermal insulations
TABLE 1. MAIN CHARACTERISTICS OF CODEX TEST SECTIONS WITH ATF FUEL
CODEX-ATF CODEX-ATF-AIT
Number of fuel rods 7 7
Cladding of fuel rods No. 2., 4. and 6. optZIRLOTM Cr coated optZIRLOTM
Cladding of fuel rods No. 1., 3., 5. and 7. Cr coated optZIRLOTM
Length of fuel rods 650 mm 910 mm
External diameter of fuel rods 9.1 mm 9.1 mm
Pellet material inside of the rods ZrO2 ZrO2
Pellet material in the bottom of the rods Al2O3 Al2O3
Spacer grid material Zr1%Nb Cr coated Zr1%Nb
Number of spacer grids 2 3
Shroud material Zr2.5%Nb Cr coated Zr2.5%Nb
Shroud thickness 2 mm 2 mm
Length of shroud 1000 mm 1000 mm
The online gas composition measurement at the test section outlet was conducted using a quadrupole mass spectrometer. Gas sampling at the outlet was achieved by inserting a sampling tube into the off-gas pipe. The measurement system recorded system and rod internal pressures, flow rates, outlet gas composition, values of input and output power, coolant inlet and outlet temperatures and rod temperatures. Several high temperature thermocouples were built into the surface of fuel rods, shroud and insulation layers at different elevations.
2. THE CODEX-ATF EXPERIMENT
The CODEX-ATF simulated a high temperature nuclear power plant severe accident terminated by water quench. The test focused on the observation of fuel failure and degradation mechanisms.
In the preparatory phase the facility was heated up to 600 °C in 0.2 g/s steam and 0.2 g/s argon flow rates using both external heaters and fuel rod heaters. The heat-up phase continued with the same flow rates and with 1000 W heating power on the rods and 800 W power of external heaters. The cladding burst took place at ≈900 °C on most of the rods. The temperature increase was very smooth. During the quench phase, room temperature water was injected to the bottom of the test section, when the cladding temperature in the top of the bundle was above 1600 °C. In the upper part of the fuel rods 1400 °C was reached (Fig. 2.).
FIG. 2. Cladding temperatures in the CODEX-ATF test FIG. 3. Cladding temperatures in the CODEX-ATF-AIT test
The total hydrogen production during the experiment was about 3 g, which indicated significant oxidation of the Zr components. Intense Zr-Cr eutectic formation took place at these temperatures on the external surface of Cr coated cladding tubes. The post-test examination showed large deformation and failure of both coated and uncoated cladding tubes (Fig. 4.).
FIG. 4. Bundle cross section (left), damaged uncoated (centre) and Cr-coated (right) cladding tubes removed from the top of CODEX-ATF bundle
FIG. 5. Oxide and nitride formation on the cladding of CODEX-ATF-AIT bundle
- THE CODEX-ATF-AIT EXPERIMENT
The main objective of the CODEX-ATF-AIT test was to check if accident tolerant fuel (ATF) cladding with Cr coating would have a protective role in case of NPP severe accidents with air ingress.
The simulated scenario was a reactor accident with corium melting through the bottom head of the pressure vessel and penetration of air+steam mixture from the reactor cavity into the reactor vessel. The experiment was focused on covering several phenomena of fuel behaviour during accidents (burst, oxidation, nitriding, eutectic formation). Slow cool-down was selected to provide information on
the state of the bundle before quench.
In the preparatory phase the facility was heated up to 600 °C in steam – argon flow using both external heaters and fuel rod heaters. The pre-oxidation phase continued with the same flow rates and with stepwise increased heating power on the rods. During the heat-up phase the rods were pressurised and cladding burst took place between 750-800 °C on most of the rods. The opening allowed the coolant to enter the rods and start chemical reactions on both sides of the cladding. In the intermediate cool-down phase the temperatures were reduced below 750 °C. In the steam-air phase the temperature increase was very smooth compared to the reference experiment CODEX-AIT-3 [2]. The reason of the slower temperature increase was the protective effect of the Cr coating. The duration of air ingress phase in the reference experiment was 1 hour with 1600 °C maximum cladding temperature, and it was 1.5 hour in the CODEX-ATF-AIT with 1545 °C maximum cladding temperature (Fig. 3.).
The temperature profile significantly changed during the air ingress phase similar to the reference experiment: the maximum temperature moved from the upper section of the bundle to lower elevations due to the intense chemical interactions in the less oxidised and/or melted lower part. The maximum shroud temperature was similarly around 1470 °C at 300, 500 and 700 mm elevations.
The outlet gas composition showed that during the air ingress phase steam and oxygen starvation conditions were established. The partial consumption of nitrogen indicated the formation of nitrides as well. The total of 1.4 g hydrogen was produced during the pre-oxidation phase and 3.3 g of hydrogen during the air ingress phase. Slow cool-down of the bundle was performed in argon flow in order to avoid interactions that might take place during water quench. At the lower half of the bundle the main degradation mechanism of rods was the Cr-Zr eutectic melt formation and its fast oxidation/nitriding in the last phase of the experiment. From 600 mm upwards, all the pressurised rods were ballooned and the Cr coating became cracked on their surfaces (Fig.5.).
SUMMARY
Two integral tests with ATF cladding materials were conducted in the CODEX facility under steam and steam-air atmospheres (Table 2). The high-temperature behavior of the cladding materials in both tests demonstrated reduced hydrogen production, a lower heat-up rate, and longer coping times, highlighting the advantages of ATF materials. Cr-Zr eutectic formation was observed in both tests, and its role in cladding failure was identified. The experimental data are available in an electronic database for code development and validation purposes.
TABLE 2. MAIN CHARACTERISTICS OF CODEX TEST SECTIONS WITH ATF FUEL
CODEX-ATF CODEX-ATF-AIT
Max. temperature 1655 °C 1545 °C
Steam atmosphere Yes Yes
Steam+air atmosphere No Yes
Oxidation Yes Yes
Nitriding No Yes
Cr-Zr eutectic Yes Yes
Quench Yes No
ACKNOWLEDGEMENTS
The CODEX-ATF test was conducted within the framework of the IAEA ATF-TS project, while the CODEX-ATF-AIT test was carried out as part of the EU OFFERR project. Both tests were supported by the Paks Nuclear Power Plant. The main parameters of the scenarios were selected based on pre-test calculations performed by Pál Kostka and Gábor Lajtha (NUBIKI), Kirill Dolganov (IBRAE), Thorsten Hollands (GRS) and Líviusz Lovász (GRS).
REFERENCES
[1] HÓZER, Z., MARÓTI, L., WINDBERG, P., MATUS, L., NAGY, I., GYENES, G., HORVÁTH, M., PINTÉR, A., BALASKÓ, M., CZITROVSZKY, A., JANI, P., NAGY, A., PROKOPIEV, O., TÓTH, B. (2006). Behavior of VVER fuel rods tested under severe accident conditions in the CODEX facility. Nuclear Technology, 154(3), 302-317. (2006) https://doi.org/10.13182/NT06-A3735
[2] FARKAS, R., HOZER, Z., NAGY, I., VER, N., HORVATH, M., STEINBRÜCK, M., STUCKERT, J., GROSSE, M. (2022). Effect of steam and oxygen starvation on severe accident progression with air ingress. Nuclear Engineering and Design, 396, 111884. (2022) https://doi.org/10.1016/j.nucengdes.2022.111884
[3] FARKAS, R., HÓZER, Z., NAGY, I., VÉR, N., SZABÓ, P., HORVÁTH, M., KOSTKA, P., LAJTHA, G. (2023). Experimental simulation of selected design extension condition scenarios without core meltdown in the CODEX facility. Progress in Nuclear Energy, 161, 104720. (2023) https://doi.org/10.1016/j.pnucene.2023.104720
