Compatibility of fuel-element cladding materials with gallium-containing uranium and uranium–plutonium oxide fuel

Abstract

Powders of uranium oxide powder and mixed fuel containing 5 and 20 mass % plutonium and 0.4 and 5 mass % gallium were prepared by coprecipitation from nitrate solutions. Pelleted samples for testing were made by cold pressing and sintering. The compatibility of uranium oxide fuel and mixed uranium-plutonium fuel, containing 0.4 and 5 mass % gallium, with the zirconium alloy E-110 at 400 and 500°C and ChS-68 corrosion-resistant steel at 650 and 750°C over periods of 1000, 2000, and 3000 h was investigated. Metallographic and x-ray spectral microprobe analyses of diffusion samples established that there was no interaction and penetration of gallium into the zirconium alloy and steel. In addition, the diffusion coefficient of metallic gallium in zirconium alloy and the distribution of the elements on interaction of gallium with ChS-68 steel were evaluated.The nuclear fuel obtained by reprocessing weapons plutonium can contain more or less of the gallium as an impurity depending on the reprocessing technology used. It is well known that gallium is quite high reactive and actively interacts with many metals [1]. Consequently, it has not been ruled out that gallium present in fuel can interact with fuel-element cladding. Such an interaction will negatively affect the serviceability of fuel elements. This concerns cladding made of zirconium alloys for the fuel elements in water cooled and moderated reactors as well as cladding made of corrosion-resistant steel for fast reactors. This makes it necessary to investigate the effect of residual gallium in sintered oxide fuel pellets on the long-time compatibility of this fuel with the materials used for fuel-element cladding.In the present investigations, various samples of oxide fuel into which gallium was purposefully introduced were fabricated and the fuel-cladding diffusion pairs made were subjected to long-time annealing. The alloy E-110 (Zr-1%Nb) and ChS-68 austenitic corrosion-resistance steel containing 16.4% Cr and 14.3% Ni and additionally alloyed with Mo, Mn, Si, and Ti were used as the materials for the fuel-element cladding. Preparation of Samples of Mixed Urnaium-Plutonium Oxide Fuel for Compatibility TestsPreparation of oxide powders by the GRANAT method. This process is based on coprecipitation of uranium and plutonium hydroxides from nitrate solutions in the presence of surfactants followed by heat-treatment in oxidizing and reducing atmospheres [2]. Initial nitrate solutions of uranium (C U = 98-107 g/liter, C HNO 3 = 0.85 mole/liter), plutonium (C Pu = = 28.46-31.15 g/liter, C HNO 3 = 3.5-3.7 mole/liter), and gallium (C Ga = 70.8 g/liter, C HNO 3 = 3.1 mole/liter) were used to prepare batches of uranium, uranium-plutonium, and uranium-plutonium-gallium powders.The mixed uranium, plutonium, and gallium hydroxides were precipitated at temperature ~30°C by adding NH 4 OH to pH = 8.6-9.3 and polyacrylamide to 20-25 mg surfactant per 1 g total uranium and plutonium. The precipitation and gran-