[en] Experience with Type C thermocouples operating for long periods in the 1400 to 1600 C temperature range indicate that significant decalibration occurs, often leading to expensive downtime and material waste. As part of an effort to understand the mechanisms causing drift in these thermocouples, the Idaho National Laboratory conducted a long duration test at 1500 C containing eight Type C thermocouples. As report in this document, results from this long duration test were adversely affected due to oxygen ingress. Nevertheless, results provide key insights about the impact of precipitate formation on thermoelectric response. Post-test examinations indicate that thermocouple signal was not adversely impacted by the precipitates detected after 1,000 hours of heating at 1,500 C and suggest that the signal would not have been adversely impacted by these precipitates for longer durations (if oxygen ingress had not occurred in this test)

[en] Because of the impact that melt relocation and vessel failure may have on subsequent progression and associated consequences of a Light Water Reactor (LWR) accident, it is important to accurately predict heating and relocation of materials within the reactor vessel, heat transfer to and from the reactor vessel, and the potential for failure of the vessel and structures within it. Accurate predictions of such phenomena require high temperature thermal and structural properties. However, a review of vessel and structural steel material properties used in severe accident analysis codes reveals that the required high temperature material properties are extrapolated with little, if any, data above 1000 K. To reduce uncertainties in predictions relying upon extrapolated high temperature data, Idaho National Laboratory (INL) obtained high data for two metals used in LWR vessels: SA 533 Grade B, Class 1 (SA533B1) low alloy steel, which is used to fabricate most US LWR reactor vessels; and Type 304 Stainless Steel SS304, which is used in LWR vessel piping, penetration tubes, and internal structures. This paper summarizes the new data, and compares it to existing data

[en] To evaluate the performance of new fuel, cladding, and structural materials for use in advanced and existing nuclear reactors, robust instrumentation is needed. Changes in material deformation are typically evaluated out-of-pile, where properties of materials are measured after samples were irradiated for a specified length of time. To address this problem, a series of tests were performed to examine the viability of using pushrod dilatometer techniques for in-pile instrumentation to measure deformation. The tests were performed in three phases. First, familiarity was gained in the use and accuracy of this system by testing samples with well defined thermal elongation characteristics. Second, high temperature data for steels, specifically SA533 Grade B, Class 1 (SA533B1) Low Alloy Steel and Stainless Steel 304 (SS304), found in Light Water Reactor (LWR) vessels, were acquired. Finally, data were obtained from a short pushrod in a horizontal geometry to data obtained from a longer pushrod in a vertical geometry, the configuration likely to be used for in-situ measurements. Results of testing show that previously accepted data for the structural steels tested, SA533B1 and SS304, are inaccurate at high temperatures (above 500 C) due to extrapolation of high temperature data. This is especially true for SA533B1, as previous data do not account for the phase transformation of the material between 730 C and 830 C. Also, comparison of results for horizontal and vertical configurations show a maximum percent difference of 2.02% for high temperature data