Enthalpy measurement on (U1−xPux)O2 (x = 0, 0.18, 0.45, and 1) and analysis of heat capacity
Source abstract
Abstract
The heat capacity of oxide nuclear fuels in high-temperature regions, which is critical for evaluating fuel thermal performance, requires extended studies due to disagreements in the literature. Furthermore, mixed-oxide (MOX) and PuO 2 data are scarce compared with UO 2 data, making it challenging to understand the dependence of Pu content on the properties. Herein, we measured the UO 2 , MOX, and PuO 2 enthalpies up to 2200 K by drop calorimetry. The heat capacity was evaluated by calculating the temperature differential of the enthalpy data. The obtained enthalpy and heat capacity increased almost linearly with the increasing temperature <2000 K and continued to increase significantly at 2000–2200 K. Additionally, heat capacity behavior at higher temperatures, having a peak known as the Bredig transition, is discussed.
Source: expertdata · Quality: candidate unverified · Source record ↗
Research context
OpenAlex topics: Nuclear Materials and Properties · Radioactive element chemistry and processing · Nuclear reactor physics and engineering
SciVX discovery tags are terminology-based. OpenAlex topics are source-provided classifications.
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