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DOE OSTI2020

Uncertainty analysis of steady-state measurements with a hot-filament type calorimetric emissometer

Calorimetric emissometers measure total hemispherical emissivity by measuring the heat transferred from a heated sample to its surroundings under a vacuum. The accuracy of emissometers standardized by the ASTM C835-06 are well understood. This work uses the Guide to the Evaluation of Uncertainty in Measurement (GUM) for the propagation uncertainties for an ASTM compliant emissometer. The GUM method was able to develop a measurement model and expressions to determine the uncertainty for other emissometers of this type. Data on ‘as-received’ Hastelloy X was used to develop a detailed uncertainty analysis of the emissivity measurement. Data on ‘as-received’ SS 347 and sandblasted A387 Gr. 91 and previous data by the group on A508/A533B were used to determine uncertainty over the ranges 0.16 to 0.81. For all samples, relative uncertainties in emissivities varied from 0.77% to 2.5% when using a fluxgate magnetometer sensor (FMS) to measure the DC heating current. Data on Hastelloy X using a Hall-effect sensor for DC current and low alloy steel showed the DC current and voltage across the test section to be dominate sources of uncertainty. When these sources were reduced, the specimen temperature and the surface area of the test sections were main sources of uncertainty in the emissivity, especially at higher temperatures. As thermal expansion of the surface was considered in the calculations, correlation between specimen temperature and surface area was examined. It was found to be a small contribution to emissivity's uncertainty despite the differences in linear CTE and its uncertainty for the materials analyzed in this study. For low temperatures, the chamber temperature can be a significant source of uncertainty if not sufficiently cooled. The GUM was also briefly compared to uncertainty from the 2nd and 3rd expansions of the Taylor series. We found that the results were the same when rounding to two significant figures.

42 ENGINEERING↗

DOE OSTI2024

Instrumentation for the In-Core Real-Time Mechanical Testing of Structural Materials (INCREASE) Project

Idaho National Laboratory (INL), in collaboration with the Electric Power Research Institute (EPRI), the Nuclear Regulatory Commission (NRC), the French Atomic and Alternative Energies Commission (CEA), the Joint Research Center (JRC), the Nuclear Research and Consultancy Group (NRG), and the Research Center Rez (CVR), started a Joint Experimental Program (JEEP) project that operates within the Nuclear Energy Agency’s Framework for Irradiation Experiments (FIDES II) program in order to develop capabilities for the in-core real-time mechanical testing of structural materials. This effort will focus on designing a shared capsule capable of housing a variety of in-core mechanical testing instrumentation allowing enhanced experiments for the material science community. The outcome of the project would be high-priority, stress relaxation data for stainless-steel-based materials provided by EPRI and CEA. Stress relaxation is a major phenomenon that contributes to material degradation in nuclear reactor components. Currently, nuclear material stress relaxation is assessed both before and after irradiation, using complex and costly post-irradiation examination (PIE) activities. In-situ data would support the development of precision modeling and simulation of this degradation phenomena and would provide validation and benchmarking for existing models using the PIE data. As part of the U.S. Department of Energy Advanced Sensor and Instrumentation (ASI) program, INL has fabricated and tested out-of-core mechanical test instrumentation. This instrumentation was designed for easy adaptation to the irradiation capsule proposed under this JEEP, and can be deployed to measure real-time stress relaxation under pressurized-water reactor (PWR) conditions. This initial effort will partially serve to replace the testing capabilities lost because of shutting down the Halden Boiling Water Reactor (HBWR). The project would provide these capabilities to the international community via a shared capsule design that is easily adaptable to additional material test reactors. The design features will incorporate expansion to PWR and non-light-water reactor (LWR) environments that will be developed in future work. The capsule and instrumentation will be demonstrated in the Massachusetts Institute of Technology Reactor (MITR) for phase I and the Petten High Flux Reactor (HFR) for phase II irradiations to deliver real-time stress relaxation data on high priority stainless steel structural materials.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Corrosion of 347 stainless steel in the presence of uranyl sulfate solution and radiation

The U.S. molybdenum 99 (Mo-99) industry is pursuing production of fission-made Mo-99 using a uranium solution such as uranyl sulfate. In this process, uranyl sulfate solution containing low-enriched uranium will be bombarded by neutrons creating Mo-99 and other fission products. During the production, the uranyl sulfate solution will be irradiated until an acceptable activity level of Mo-99 is produced. The uranyl sulfate solution containing Mo-99 and other fission products will then undergo a series of separation steps. First, uranyl sulfate can be separated from Mo-99 using a primary titania column to recover Mo-99, with the uranyl sulfate solution to be used for another irradiation cycle. Then, raffinate from a primary titania column containing Mo-99 can be concentrated and purified using a LEU modified Cintichem process developed by Argonne National Laboratory. During irradiation, the temperature of the uranyl sulfate solution can reach near boiling (up to ~80° C assumed), causing radiolysis of water and the resultant formation of hydrogen peroxide. Because high-radiation fields will be present during each irradiation cycle, it is important to determine the corrosion rates of SS-347 under such conditions to estimate the life cycle of the target solution vessel. The buildup of corrosion products from the SS components in the uranyl sulfate solution also needs to be well understood because potential accumulation of iron, nickel, and other corrosion products may affect the Mo-99 recovery and purification process. To study the corrosion rates of SS-347 material under conditions relevant to future Mo-99 production facility, SS-347 coupons in uranyl sulfate solution at ~80° C were irradiated using Argonne’s Van de Graaff generator, which can generate high-radiation fields without fissioning of uranium or production of activation products.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗