Engineering topics

Momozaki, Yoichi

Publications and source records attributed to Momozaki, Yoichi.

6 records

DOE OSTI2024

FRIB Beam Power Ramp-up: Status and Plans

After project completion on scope, on cost, and ahead of schedule, the Facility for Rare Isotope Beams began operations for scientific users in May of 2022. The ramp-up to a beam power of 400 kW is planned over a six-year period; 1 kW was delivered for initial user runs from in 2022, and 5 kW was delivered as of February 2023. Test runs with 10 kW 36Ar and 48Ca beams were conducted in July 2023. Upgrade plans include doubling the primary-beam energy to 400 MeV/nucleon for enhanced discovery potential (¿FRIB 400¿). This talk reports on the strategic plans towards high power operations emphasizing challenges and resolutions in beam-interception devices and targetry systems, radiation protection and controls, and legacy system renovation and integration.

43 PARTICLE ACCELERATORS↗

DOE OSTI2023

FRIB Transition to User Operations, Power Ramp Up, and Upgrade Perspectives

After project completion on scope, on cost, and ahead of schedule, the Facility for Rare Isotope Beams began operations for scientific users in May of 2022. During the first 12 months of user operations, the FRIB accelerator complex delivered 5250 beam hours, including 1528 hours to nine science experiments conducted with primary beams of 36Ar, 48Ca, 70Zn, 82Se, 124Xe, and 198Pt at beam energies >200 MeV/u; 2724 hours for beam developments, studies, and tuning; and 998 hours to industrial users and non-scientific programs using the FRIB Single Event Effect (FSEE) beam line. The ramp-up to a beam power of 400 kW is planned over a six-year period; 1 kW was delivered for initial user runs from in 2022, and 5 kW was delivered as of February 2023. Upgrade plans include doubling the primary-beam energy to 400 MeV/nucleon for enhanced discovery potential (¿FRIB 400¿). This talk reports on FRIB status and progress since SRF2021, emphasizing lessons learned during the transition from beam commissioning to machine operations, challenges and resolutions for the power ramp-up, progress with accelerator improvements, and R&D for the energy upgrade.

43 PARTICLE ACCELERATORS↗

DOE OSTI2023

Progress Report of Alloy 709 Performance in Liquid Sodium

With improved high-temperature properties, Alloy 709 shows great potential for advanced reactor applications. To understand the effects of sodium on A709, small tensile specimens were machined from two commercial heats and subject to sodium exposure in two forced convection loops operated by Argonne. Thermal aging experiments were also conducted in parallel to isolate any sodium-specific effects. An increasing trend in yield strength and a decrease trend in ultimate tensile strength can be seen with the increasing exposure time, suggesting microstructural changes at these temperatures. Nonetheless, the thermally aged and sodium exposed samples behaved similarly, implying no sodium-specific effect at these temperatures. In contrast, 316H SS exposed to sodium at 650°C showed a strong sodium effect. The different responses of A709 and 316H SS to sodium exposure is of interest and requires additional effort to understand. The cold traps of the sodium loop were replaced early in FY23. During a maintenance operation, an unexpected sodium leak incident occurred at one of the sodium loops, triggering an emergency response. An investigation into the incident revealed that the leak happened at a VCR fitting on the recently replaced cold trap. The metal gasket of the leaked VCR fitting was not compressed sufficiently. A dimensional mismatch between the new component and old piping configuration was responsible for a significant misalignment. A new cold trap with improved design is in production. The new design provides additional flexibility and an access point for post-installation helium leak check and for venting of trapped gas. These design features will make the system more error-tolerant and the installation quality verifiable. A thermodynamic analysis was performed to understand the carburization-decarburization behavior of A709 in sodium environment. The equilibrium carbon activity in the alloy was evaluated with M 23 C 6 , TiC, and NbC. The carburization-decarburization boundary was calculated with different methods developed previously. The results suggests that A709 will be carburized in SMT-1 and SMT-2, and can be decarburized in low-carbon-activity environments such as EBR-II at temperatures above 620-650°C. Nonetheless, A709 should be less likely to be decarburized than 316H SS when exposed to the same low-carbon-activity environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Effects of Sodium Exposure on the Tensile Properties of Grade 91 steel

Two heats of G91 steel have been investigated in liquid sodium at 550- 650°C to understand their corrosion behaviour, microstructural evolution, and tensile properties. Sodium exposure experiments of subsized, sheet-type tensile specimens of G91 were performed in Argonne’s forced convection sodium materials testing loops, SMT-1 and SMT-2. Maximum exposure times of ~49,000 h at ~38,000 h, and ~20,000 h have been achieved at 550, 600, and 650°C, respectively. Thermal aging experiments of G91 steel were also conducted in parallel, and the results were compared with those of the sodium-exposed specimens to isolate the effect of thermal aging from the effect of sodium exposure. It was found that the yield stress and ultimate tensile strength of G91 were reduced by more than 50% after sodium exposures at 650°C due to decarburization. The reduction in tensile strength was accompanied by the dissolution of M 23 C 6 carbides and excessive grain growth. The effect of sodium at 550 and 600°C was insignificant for exposure times up to ~49,000 h and ~38,000 h, respectively. The effect of sodium on the microstructural stability and tensile properties of G91 is largely dependent on the carbon mass transfer in the alloy-sodium system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Surveillance Test Articles Development

Material degradation in Advance Test Reactors (ATR) is governed by irradiation, corrosion, elevated temperature exposure and cyclic mechanical creep-fatigue loads. This degradation information during reactor operation condition is limited. Hence, material damage monitoring is a key aspect of the design, analysis and licensing of ATR components. The idea is to monitor material component operation conditions of component by using a surveillance test article. This test article is fabricated with bi-metal configuration with two different thermal expansion coefficients, and design is motivated by Simplified Model Test (SMT) specimen which can capture structure-like mechanical response. Upon raising temperature of the bi-metal test article configuration, expansion mismatch results tensile load on specimen. Thus, temperature dependent passively actuated loading is achieved. The idea is to place this surveillance test article in reactor at location ‘x’ to surveil the mechanical response at critical location ‘y’. By calibrating the test article design, material degradation at critical location can be surveilled through assessing the degradation in surveillance test article. This study presents test article development with different material combinations and follow-up experimental testing work through passively loading test article with temperature history. The test article geometry and observed test results are presented in presentation slides.

36 MATERIALS SCIENCE↗

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↗