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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Waste Shielded Casks Transfer by Funicular from the Surface to Disposal Level in Cigeo - 20007

Andra is completing the detailed engineering phase of Cigeo - the French HLW and ILW-LL DGR. A funicular system is dedicated to the transfer of the waste disposal containers from the surface to the underground infrastructures down the nuclear access ramp. This paper explains the rationale having led to the choice of a funicular and presents the advantages of this system. Main technical components and mechanical performance of the funicular system are then detailed. Lastly, the ongoing two stages qualification testing program managed by Andra (in order to reach a level 6 on the international Technology Readiness Level scale) is discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Tank Side Cesium Removal IXC-150 Cask Detonation Report

A set of experiments were conducted at Los Alamos National Laboratory’s firing point 88 to determine the blast pressure and effects of a stoichiometric gas mixture of Hydrogen (H 2 ) and Nitrous Oxide (N 2 O) on the IXC-150 Storage Pad Vent Stack assemblies. The Lab was asked to design, build, and test the vent stack assemblies using a reaction gas volume of 340 in 3 that mimics the Tank Side Cesium Removal (TSCR) Ion Exchange Column (IXC) assembly. The experimental system with vent stack assembly was designed with input from the Washington River Protection Services (WRPS) to ensure an accurate test setup. The tests were conducted at LANL on the newly built filter/gas assembly at the end of April 2021 timeframe, using WRPS supplied vent stack assemblies and internal HEPA filters. The objective of these tests was to observe that the structural integrity of the vent stack assemblies was maintained and determine the resulting blast overpressures at a distance of 16 and 32 inches from the assembly. All work was performed under the LANL Quality Assurance Program (SD330), using a graded application of ASME NQA-1- 2008/NQA-1a-2009. To ensure that all client quality assurance (QA) expectations were addressed, evaluation of the end data needs was performed and the appropriate controls applied for this work.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Commissioning of a Cask Enabling Characterization of Irradiated Nuclear Fuels with Pulsed Neutrons

Irradiation tests are a key component of nuclear fuel development and identifying typical and atypical regions in the irradiated fuel volume relies on very few characterization techniques. The goal of the effort reported here is to provide complementary measurements adding to the available parameter space for post irradiation examination as well as to inform subsequent hot cell PIE examinations by identifying typical and atypical regions with respect to microstructure, tomographic data, or isotope densities. Pulsed neutrons, enabling diffraction as well as energy-resolved neutron imaging and neutron absorption resonance spectroscopy, offer unique capabilities for this purpose. Time-of-flight neutron diffraction has the potential to offer efficient, non-destructive and non-contact microstructural characterization of irradiated fuel pins with spatial resolution of 1 mm 3 to 1 cm 3 while for energy-resolved neutron imaging (and by extension tomography) a resolution of 100 μm was demonstrated. The potential results include crystallinity vs. amorphous volumes and microstructural information such as phase compositions, lattice strains (indicative of residual stresses or chemistry variations) and textures from the diffraction data as well as distances (e.g. pellet to cladding), cracks, and isotopic distributions of minor actinides, fission products as well as fission gas partial pressures e.g. in the plenum from energy resolved neutron imaging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Feasibility Study for Disposal Control Rod Assemblies Using UNF-ST&DARDS As-Loaded Zion Dual Purpose Cask Models

This report documents an initial evaluation to support future use of disposal control rod assemblies (DCRAs) for post-closure criticality control in dual purpose canisters (DPCs). The work described herein is an extension of previous efforts performed by Walker using inputs generated by the Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNFST&DARDS) for the Zion site DPCs with as-loaded isotopic compositions. The results of this analysis demonstrate that there are multiple pathways to using DCRA for post-closure criticality control. Various configurations of DCRA material, diameter, number of rods per DCRA, and number and location of DCRAs within a DPC were shown to be effective in varying degrees for the set of DPCs analyzed by Walker. Because of the variations in DPC as-loaded isotopic compositions considered in an array of DCRA parametric sweeps, it can be concluded that a DPC-specific methodology is feasible (i.e., a one-size-fits-all approach may not be needed). Instead, the utility program created for this work can be expanded to develop capabilities to provide DPC-specific DCRA arrangements to limit cost and weight and to allow for operational considerations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

UNF-ST&DARDS Enhancements for RCCA Data in As-loaded Dual Purpose Cask Models

This report summarizes the work performed to enable detailed modeling of rod cluster control assembly (RCCAs) in dual purpose canisters (DPCs) in as-loaded configurations using the Used Nuclear Fuel – Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). The goal of this project was to evaluate the reactivity impact that RCCAs have on k eff of DPCs with pressurized water reactor (PWR) fuel to potentially use the additional margin in future post-closure criticality safety analysis. This preliminary evaluation determined the number of DPCs that currently require compensatory actions prior to emplacement in a repository because of their high reactivity under post-closure criticality scenarios, which could be made acceptable by including the as-loaded RCCAs as specified in the Unified Database (UDB). This report briefly describes the modeling methods currently used within UNF-ST&DARDS and the modifications made to automate inclusion of the as-loaded RCCAs in the DPCs for post-closure criticality calculations for the loss of neutron absorber (NA) scenarios, or NA models. For the Zion site, the loss of basket scenarios, or degraded basket (DB) models, are also included. Discussion is also provided regarding the UDB data compared to available site-specific loading map data for the sites, which are specifically evaluated herein. This report compares the modifications to a similar evaluation for the Zion by site conducted by Walker as an initial validation. After the partial validation, all existing PWR sites with applicable NA models within the UDB were evaluated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗