Pre-reactor tests of uranium-zirconium metal fuel rod fragments

1. STATE ATOMIC ENERGY CORPORATION "ROSATOM" PRE-REACTOR TESTS OF FRAGMENTS OF URANIUM-ZIRCONIUM METAL FUEL RODS OBTAINED DURING THE MODERNIZATION OF THE REACTOR CORE IVG.1M K.K. Polunin, A.N. Bakhin, D.A. Zaitsev, D.M. Soldatkin, V.A. Solntsev The 11th Conference on Reactor Materials Science May 27 – 31, 2019

2. INTRODUCTION In accordance with the Reduced Enrichment for Research and Test Reactors (RERTR)Program, in 2014 the project for modernization of IVG.1M reactor had been initiated with the goal of conversion of the reactor to operation on low-enrichment fuel. The design of the two-blade spiral self-positioning fuel rod had been developed. The fuel rod design is a matrix made of E110 alloy with coaxial metallic low-enrichment uranium (19.75% by U-235) fibers dispersed along the cross-section of the fuel rod. Fine-grain uranium growth with strong grain orientation (a bar rolled at 300 °C) after 300 cycles within a temperature range of (50 – 550) °C [1] The goal of this work is to evaluate the thermal swelling of the fuel composite of VOTK-NOU fuel elements of IVG.1M reactor to predict its behavior throughout the entire temperature range of the possible operation modes of the reactor, up to the melting point of uranium. Physical materials science: Textbook for Higher Educational Institutions / Under the general editorship of B.A. Kalin. M.: NRNU MEPHI, 2012. Vol. 7. Nuclear fuel materials / V.G. Baranov, Yu.G. Godin, A.V. Tenishev, A.V. Tenishev, A.V. Khlunov, V.V. Novikov – M.: MEPHI, 2012. – 640 p.

3. PROCEDURE OF WORK Thermal testing modes Design of the work area for thermal cycling testing Post-testing examination: – visual inspection; – geometry measurement; – X-ray phase analysis; – electron and optical microscopy; – microhardness measurement. Typical appearance of the original samples

4. ANALYSIS OF THE ORIGINAL SAMPLE Zr No traces of uranium have been found beyond the fuel fibers (cells, cladding, coating) UZr2 U Fuel fibers in the zirconium matrix of the fuel rod Typical structure of the spiral samples Number of cells – 133 Cell size – 150 μm Uranium fiber size – 50 μm Zirconium cladding thickness – 250 μm Weight fraction of uranium – (20 – 25) % Nickel coating at the butt-end of the fuel rod

5. THE GEOMETRICAL PARAMETERS AND THE DENSITY

6. THERMAL CYCLING – STAGE 1(T = 350 °C; n = 30; τΣ = 200 h) Cross-section of the middle part of the fuel rod after testing Appearance of the sample after testing • The analysis of the samples has shown the following: • iridescence is visible on all the samples; • integrity of the butt-end coatings has not been disrupted; • alteration of the sample diameters and blade thicknesses is within the measurement error margins; • thickness of the intermetallide layer does not exceed 1 μm; • no diffusion of uranium into the zirconium matrix; • the phase composition of the fuel rod corresponds to the initial state.

7. THERMAL CYCLING – STAGE 1(T = 350 °C; n = 30; τΣ = 200 h) Ni-coating Zr,U(O) Zr Zr-U Results of the local electron microprobe analysis of the butt end of the fuel rod State of the nickel coating at the butt end of the fuel rod The oxide film on the surface of E110 cladding and nickel coating is less than 1 μm thick, heavy oxidation begins at the “cladding – coating – melted butt end” interface and continues along the entire surface of the melted butt end. The thickness of Zr,U(O) sublayer is about 35 μm. The samples of this set had thin nickel coatings at the butt ends varying from 25 to 50 μm.

8. THERMAL CYCLING – STAGE 2 (T = 675 °C; n = 5; τΣ = 30 h) Ni-coating Zr Zr-U • The analysis of the samples has shown the following: • iridescence is visible on all the samples; • the thickness of the intermetallide layer has grown to 19 μm; • a grid of lightly colored bars is visible around the fuel fibers which is related to uranium diffusion through the matrix grain boundaries. The zone of uranium diffusion into the cladding did not exceed 15 μm; • the matrix grains (α-Zr) are of polyhedral shape close to equiaxial with the grain size of about 7 μm; • no oxide layers have been found under the nickel coating; • no uranium is present in the coating at a distance of 15 μm.

9. THERMAL CYCLING – STAGE 3 (T = 780 °C; n = 5; τΣ = 30 h) • The analysis of the samples has shown the following: • uniform occurrence of iridescence on all the samples; • local spalling of the coating from the fuel rod surface; • formation of a 7 μm sublayer at the “coating – cladding” interface (point 2); • no diffusion of uranium into the coating; • The X-ray diffraction spectrum obtained from the longitudinal section of the fuel rod confirms the absence of metallic uranium.

10. THERMAL CYCLING – STAGE 3 (T = 780 °C; n = 5; τΣ = 30 h) Formation of channels with diameters of about 20 μm occurred in the center of some fuel fibers consisting of UZr2 intermetallide and U-Zr solid solution. The structure of both the zirconium matrixand the fuel fiber is homogeneous and notable for an equiaxial polyhedral grain. A grain size of the matrix is (10 – 15) μm and the fuel fiber isabout 60 microns. No traces of uranium have been found at a distance of 20 µm from the fuel rod surface.

11. THERMAL CYCLING – STAGE 4(T = 1140 °C; n = 2; τΣ = 6 h) The fuel rod microstructure consists of mostly large polyhedral grains with the size of up to 350 μm and grains with the size of (30 – 35) μm. The composition of the fuel rod grains is non-uniform. Local electron microprobe analysis has shown that the lightly-colored areas are those of the higher uranium concentration (up to 23 wt. %), while the darker ones are those where zirconium content is increased (up to 98 wt. %).

12. EVALUATION OF THE MECHANICAL PROPERTIES Mechanical testing was conducted on both the original (79 – round, 22 – spread) and tested (9 – round, T = 350 °C; 20 – spread, T = 350 °C; 18 – spread, T = 675 °C) Annealing in the temperature range of (350 – 675) °C left practically no effect on the ductile properties of the samples. A dramatic (almost twofold) increase of bend ductility after spreading process operation is worth noting.

13. PHASE COMPOSITION AND MICROHARDNESS

14. CONCLUSIONS The testing performed on the samples made of the fuel composition in the form of E110 alloy matrix with metallic uranium fuel fibers almost uniformly dispersed along the transversal section of the fuel rod revealed no considerable thermal swelling in the employed time-temperature range. It has been shown that after testing at T = 350 °Cthe quantity of UZr2intermetallide phase which forms at the “fuel fiber – matrix” interface, does not change as compared with the initial state (the intermetallide layer thickness does not exceed 1.0 μm). Intense formation of the intermetallide phase has been observed after testing at T = 675°C (intermetallide layer thickness comes up to 19 μm), and uranium diffusion into the cladding is also observed (diffusion zone did not exceed 30 μm).

15. CONCLUSIONS After testing at T = 780 °C virtually no metallic uranium has been found in the fuel rods. The fuel composite consists of the intermetallide and U-Zr solid solution, only the surface layer of the cladding about 20 μm thick doesn’t contain the uranium phase. At T = 1140 °C uranium diffusion occurred throughout the entire cladding thickness, though α-radiation at the surface of the samples doesn’t exceed 1.0 particle / (cm2·min). The analysis of mechanical stress diagrams typical for VOTK (stretching, compression, bending) has shown that testing within the temperature range of (350 – 675) °C left virtually no effect on the ductile properties of the samples. The stability of microhardness of the zirconium matrix and the sample cladding after testing makes it possible to anticipate rather good mechanical properties of this uranium-zirconium compound.

16. STATE ATOMIC ENERGY CORPORATION "ROSATOM" THANK YOU FOR ATTENTION!