Irradiation effects on thermal properties of LWR hydride fuel

Abstract

Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.33% FIMA or ~5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U(30wt%)ZrH1.6) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each minifuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

Performance evaluation and post-irradiation examination of a novel LWR fuel composed of U0.17ZrH1.6 fuel pellets bonded to Zircaloy-2 cladding by lead bismuth eutectic

“…Four fuel rodlets were initially fabricated at UCB, with detailed description of fabrication procedure available in an accompanying publication in this journal issue [5]. Three of these rodlets (rodlets 1, 2, and 3) were of the design shown in Fig.…”

“…The elaborate procedure of LM gap-filling to eliminate voids formation is described in detail by Terrani et. al [5] which include multiple cycles of pressurizing the tube to 0.3 MPa using helium gas and subsequently pumping down to vacuum while the tube was mechanically shaken. While the lead-bismuth is solid at room temperature, it was liquid at irradiation temperatures and will be referred to as "liquid metal" (LM) throughout the rest of this paper.…”

“…An attempt was made to use neutron radiography technique to image imperfections in the fuel and the gap. But its resolution was inadequate for detecting possible void in the gap [5]. Some portions of the inner LM were smeared or dislodged during the sectioning and polishing processes, as shown in Figure 3b.…”

“…Previously FIMA rather than FIHMA has been adopted for hydride fuels [2]. It can conveniently be compared to UO 2 power reactor expressed as MWd/kgU [2,5]. FIMA in the present work is estimated though total fissions obtained from linear heat rate that in turn derived from MCNP calculation on MIT reactor core for 5 MW power [ref.…”

Post irradiation examination of a uranium-zirconium hydride TRIGA fuel element

“…Some of the unique safety features for this reactor type include a prompt negative temperature coefficient of reactivity, high fission product retention, chemical stability when quenched from high temperatures in water, and dimensional stability over large swings of temperature (GA Technologies, 1992). U-ZrH fuel has been evaluated for use in different types of reactors, including space nuclear systems (Zinkle et al, 2002), light water reactors (LWRs) (Olander and Ng, 2005;Olander, 2009;Olander et al, 2012;Terrani et al, 2017), and micro-reactors such as the Microreactor Applications Research Validation and EvaLuation (MARVEL) reactor (American Nuclear Society, 2021). An evaluation of the literature shows that information from detailed microstructural characterization using techniques like electron microscopy of as-irradiated U-ZrH fuel is lacking.…”