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58 records · Page 4

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DOE OSTI2023

Formulation and testing of a high-tin borosilicate nuclear waste glass for in-can melting

Here, borosilicate waste glasses were successfully developed to immobilize two high-level waste raffinate streams via an in-can melter process with an Inconel 601 canister at 1050 °C. Measured viscosity and crystallinity thermal profiles were within the targeted processing constraints for the in-can melter process. Measured chemical durability of the glass by ASTM C1285–21 (Method A), ranged from normalized loss of boron, NL(B) = 1.44 – 2.65 g•m -2 , and NL(B) decreased with increased waste loading, accompanied by increased SnO 2 crystallinity. Measured corrosion of the in-can melter canister by a glass melt showed that Inconel 601 performed well at 1100 °C for up to 500 hr. Resistance polarization measurements versus time revealed that Inconel 601 corrosion rates in (and by) glass melts decreased from an initial rate of 63 mm•y -1 down to 10.2 mm•y -1 after 137 h with increased duration, which was attributed to formation of an oxide passivation layer (mainly Cr 2 O 3 ) at the alloy-glass interface.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

The detrimental effect of elemental contaminants when using B additions to improve the creep properties of a Ni-based superalloy

The effect of Si contamination when using B to improve the creep properties of a Ni-based superalloy was investigated using advanced characterization techniques and first-principles simulations on alloys with high and low B levels with varying Si contents. The positive effect of B segregation along grain boundaries on the creep properties was mitigated by the presence of Si which showed a similar segregation preference. Density functional theory calculations were used for validation by calculating grain boundary cleavage energies. Silicon was shown to decrease grain boundary cohesion which offsets the positive effect of B.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Initial development of an in-situ, passive material surveillance test article for monitoring high temperature reactor structural components

This report describes the development of a family of passively-actuated surveillance test articles, suitable for use in an in-situ reactor test program monitoring the degradation in structural material properties caused by the reactor environment. The test articles apply creep-fatigue-type cyclic load to a test section driven by changes in the reactor coolant temperature converted to mechanical force by differential thermal expansion. The report describes three types of test articles, all of which apply the load to a 316H test section using Alloy 617 as the driver material. The three types of specimens each serve a unique purpose: (1) initial proof-of-concept, (2) validation of the models used to size the specimens to achieve a given strain range, and (3) demonstration that the articles can be miniaturized for use in an actual operating reactor. In addition to the specimen design, fabrication, and testing, this report describes the outline of two types of in-situ monitoring programs, using the test article. Monitoring programs such as these could be a critical component in gaining regulator acceptance for Molten Salt Reactor (MSR) designs using structural materials for which there is little long-term experimental environmental testing or operating experience available for calibrating the component structural design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Report on Thermal Cycling Testing and Bimaterial Weld Development for a Passively Actuated Materials Surveillance Test Article

This report describes the continuation and conclusion of a thermal cycling test program designed to evaluate a family of passively actuated material surveillance test articles. Additionally, the report details a preliminary study of bimetallic welding between 316 stainless steel and TZM (tungsten-zirconium-molybdenum) alloy. The test program was ultimately successful in demonstrating the basic viability of the passively actuated test article concept as well as specifically validating the procedures developed to size specimens to mimic the mechanical response of operating components. The report discusses and summarizes the results of the complete test program which tested samples for more than 1,000 controlled thermal cycles. The weld development program included both stir friction and electron beam joining. The stir friction welding was ultimately successful, while the electron beam welding would require additional development. Ultimately, these test programs pave the way for implementing a materials surveillance program in future molten salt reactors (MSRs). The report conclusions outline a plan for transforming the research and developed described here into a definite set of practices and procedures for a surveillance program at a future operating reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗