Michigan State University (MSU) in East Lansing, MI was selected by the U.S. Department of Energy (DOE) to design and establish a Facility for Rare Isotope Beams (FRIB), a cutting-edge research facility to advance the understanding of rare nuclear isotopes and the evolution of the cosmos. The research conducted at the FRIB will involve experimentation with intense beams of rare isotopes within a well-shielded target cell that will result in activation and contamination of components. The target cell is initially hands-on accessible after shutdown and a brief cool-down period. Personnel are expected to have hands-on access to the tops of shielded component modules with the activated in-beam sections suspended underneath. The modules are carefully designed to include steel shielding for protecting personnel during these hand-on operations. However, as the facility has greater levels of activation and contamination, a bridge mounted servomaniputor may be added to the cell, to perform the disconnecting of services to the component assemblies. Dexterous remote handling and exchange of the modularized activated components is completed at a shielded window workstation with a pair of master-slave manipulators. The primary components requiring exchange or maintenance are the production target, the beam wedge filter, the beam dump, and themore » beam focusing and bending magnets. This paper provides an overview of the FRIB Target Facility remote handling and maintenance design requirements, concepts, and techniques.« less
The availability of future fusion devices, such as a fusion nuclear science facility or demonstration fusion power station, greatly depends on long operating lifetimes of plasma facing components in their divertors. ORNL is designing the Material Plasma Exposure eXperiment (MPEX), a superconducting magnet, steady-state device to address the plasma material interactions of fusion reactors. MPEX will utilize a new highintensity plasma source concept based on RF technology. This source concept will allow the experiment to cover the entire expected plasma conditions in the divertor of a future fusion reactor. It will be able to study erosion and redeposition for relevant geometries with relevant electric and magnetic fields in-front of the target. MPEX is being designed to allow for the exposure of a priori neutron-irradiated samples. The target exchange chamber has been designed to undock from the linear plasma generator such that it can be transferred to diagnostics stations for more detailed surface analysis. MPEX is being developed in a staged approach with successively increased capabilities. After the initial development step of the helicon source and electron cyclotron heating system, the source concept is being tested in the Proto-MPEX device. Proto-MPEX has achieved electron densities of more than 4×10 19 m -3 with a large diameter (13 cm) helicon antenna at 100 kW power. First heating with microwaves resulted in a higher ionization represented by higher electron densities on axis, when compared with the helicon plasma only without microwave heating.
Oak Ridge National Laboratory (ORNL) successfully demonstrated the Versatile Remediation Module (VRM), a prototype module designed and built by ORNL for on-site remote repair of welded stainless steel storage containers for spent nuclear fuel and high-level radioactive waste. This paper describes the VRM prototype and its design features and components to support continued long-term storage or off-site transportation of spent nuclear fuel and high-level radioactive waste currently stored in storage containers. A remote (100 ft away from the simulated radiative environment) demonstration of the VRM was successfully performed on a full-scale mock-up welded stainless steel canister. The VRM is designed with features to accommodate remediation techniques beyond those currently selected and described in this paper. Therefore, many of the VRM's features may benefit other remote nuclear or nonnuclear applications. The VRM is envisioned to serve as a development center to facilitate and enhance further development of new remediation technologies.
Fusion energy is the most promising energy source for the future, and one of the most important problems to be solved progressing to a commercial fusion reactor is the identification of plasma-facing materials compatible with the extreme conditions in the fusion reactor environment. The development of plasma–material interaction (PMI) science and the technology of plasma-facing components are key elements in the development of the next step fusion device in the United States, the so-called Fusion Nuclear Science Facility (FNSF). All of these PMI issues and the uncertain impact of the 14-MeV neutron irradiation have been identified in numerous expert panel reports to the fusion community. The 2007 Greenwald report classifies reactor plasma-facing materials (PFCs) and materials as the only Tier 1 issues, requiring a “. . . major extrapolation from the current state of knowledge, need for qualitative improvements and substantial development for both the short and long term.” The Greenwald report goes on to list 19 gaps in understanding and performance related to the plasma–material interface for the technology facilities needed for DEMO-oriented R&D and DEMO itself. Of the 15 major gaps, six (G7, G9, G10, G12, G13) can possibly be addressed with ORNL’s proposal of an advanced Material Plasma Exposure eXperiment. Establishing this mid-scale plasma materials test facility at ORNL is a key element in ORNL’s strategy to secure a leadership role for decades of fusion R&D. That is to say, our end goal is to bring the “signature facility” FNSF home to ORNL. This project is related to the pre-conceptual design of an innovative target station for a future Material–Plasma Exposure eXperiment (MPEX). The target station will be designed to expose candidate fusion reactor plasma-facing materials and components (PFMs and PFCs) to conditions anticipated in fusion reactors, where PFCs will be exposed to dense high-temperature hydrogen plasmas providing steady-state heat fluxes of 5–20 MW/m2 and ion fluxes up to 1024 m-2s-1. Since PFCs will have to withstand neutron irradiation displacement damage up to 50 dpa, the target station design must accommodate radioactive specimens (materials to be irradiated in HFIR or at SNS) to enable investigations of the impact of neutron damage on materials. Therefore, the system will have to be able to install and extract irradiated specimens using equipment and methods to avoid sample modification, control contamination, and minimize worker dose. Included in the design considerations will be an assessment of all the steps between neutron irradiation and post-exposure materials examination/characterization, as well as an evaluation of the facility hazard categorization. In particular, the factors associated with the acquisition of radioactive specimens and their preparation, transportation, experimental configuration at the plasma-specimen interface, post-plasma-exposure sample handling, and specimen preparation will be evaluated. Neutronics calculations to determine the dose rates of the samples were carried out for a large number of potential plasma-facing materials.
We have evaluated candidate technologies for in-core fission chambers for high-temperature reactors to monitor power level via measurements of neutron flux from start-up through full power at up to 800°C. This research is important because there are no commercially available instruments capable of operating above 550 °C. Component materials and processes were investigated for fission chambers suitable for operation at 800 °C in reactors cooled by molten fluoride salt (FLiBe) or flowing He, with an emphasis placed on sensitivity (≥ 1 cps/nv), service lifetime (2 years at full power), and resistance to direct immersion in FLiBe. The latter gives the instrument the ability to survive accidents involving breach of a thimble. The device is envisioned to be a two-gap, three-electrode instrument constructed from concentric nickel-plated alumina cylinders and using a noble gas–nitrogen fill-gas. We report the results of measurements and calculations of the response of fill gasses, impurity migration in nickel alloy, brazing of the alumina insulator, and thermodynamic calculations.
Abstract The back end of the nuclear fuel cycle focuses on the interim storage, transportation, and final disposition of the spent nuclear fuel (SNF) from reactors. Commercial light-water nuclear power stations across the United States operate with fuel assemblies that are irradiated for up to 6 years (planned) in the reactor pressure vessel. After their planned irradiation, the fuel assemblies are moved to a spent fuel pool within the facility complex. After the SNF is removed from the fuel pool, it is typically inserted into a welded metal canister that can be transferred between overpacks for storage, transportation, and possibly disposal. Most dry storage systems used in industry today use dual-purpose canisters (DPCs), which designed for use in storage and transportation overpacks but are not specifically designed for disposal. Triple-purpose canisters, designed for disposal in addition to storage and transportation, have also been researched. Traditional manufacturing methods for spent fuel canisters involve fusion welding along the length or circumference of the canister, resulting in high-tensile residual stresses in the joint weld zone (WZ) and heat-affected zone (HAZ). This paper documents work in which spent fuel canister designs were printed by wire arc additive manufacturing (AM) using the 316L SS welding wire to demonstrate (1) the feasibility of SNF canister fabrication using this advanced manufacturing method and (2) the dynamic response of the AM canister design when subjected to the federally mandated Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) physical tests for Type B packages. This paper focuses on the canister printing design and structural tests. The AM 3D-printed design, regulatory testing, and post-test evaluation of the canister tested to the 10 CFR 71.71 and 10 CFR 71.73 requirements are presented herein. One AM canister design was subjected to the penetration, free drop, and puncture test. Before and after the dynamic structural tests, the AM canister design was scanned with a handheld scanner to capture a 3D CAD geometry to compare to the 3D-printed canister design in the deformed shape. The scanned geometry was sectioned in areas with deformation, and the cross section profile was measured to determine accurate and repeated results of the deformed shape of the AM canister design.
