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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 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Nondestructive Examination Guidance for Dry Storage Casks

In this report, an assessment of NDE methods is performed for components of NUHOMS 80 and 102 dry storage system components in an effort to assist NRC staff with review of license renewal applications. The report considers concrete components associated with the horizontal storage modules (HSMs) as well as metal components in the HSMs. In addition, the report considers the dry shielded canister (DSC). Scope is limited to NDE methods that are considered most likely to be proposed by licensees. The document, ACI 349.3R, Evaluation of Existing Nuclear Safety-Related Concrete Structures, is used as the basis for the majority of the NDE methods summarized for inspecting HSM concrete components. Two other documents, ACI 228.2R, Nondestructive Test Methods for Evaluation of Concrete in Structures, and ORNL/TM-2007/191, Inspection of Nuclear Power Plant Structure--Overview of Methods and Related Application, supplement the list with additional technologies that are considered applicable. For the canister, the ASME B&PV Code is used as the basis for NDE methods considered, along with currently funded efforts through industry (Electric Power Research Institute [EPRI]) and the U.S. Department of Energy (DOE) to develop inspection technologies for canisters. The report provides a description of HSM and DSC components with a focus on those aspects of design considered relevant to inspection. This is followed by a brief description of other concrete structural components such as bridge decks, dams, and reactor containment structures in an effort to facilitate comparison between these structures and HSM concrete components and infer which NDE methods may work best for certain HSM concrete components based on experience with these other structures. Brief overviews of the NDE methods are provided with a focus on issues and influencing factors that may impact implementation or performance. An analysis is performed to determine which NDE methods are most applicable to specific components.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Validation Experiments for Spent-Fuel Dry-Cask In-Basket Convection

This work consisted of the following major efforts; 1. Literature survey on validation of external natural convection; 2. Design the experiment; 3. Build the experiment; 4. Run the experiment; 5. Collect results; 6. Disseminate results; and 7. Perform a CFD validation study using the results. We note that while all tasks are complete, some deviations from the original plan were made. Specifically, geometrical changes in the parameter space were skipped in favor of flow condition changes, which were found to be much more practical to implement. Changing the geometry required new as-built measurements, which proved extremely costly and impractical given the time and funds available

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Applications of Analytic Models to Spent Fuel Cask Analysis

The advent of high performance computing has changed the way scientists and engineers solve problems. Scientists previously would make a tremendous effort to simplify a complex problem to it’s essence by reducing geometry, neglecting changes and only preserving first-order effects. Now, scientist are able to leave the details in the problem and computationally arrive at a result. To develop the computational physics programs used by scientists, developers make assumptions and use approximation to reduce complex, continuous calculus to algebraic expressions which computers are able to solve.

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DOE OSTI2023

BRR Cask Use for Pu-238 Isotope Production

US Department of Energy (DOE), in partnership with the National Aeronautics and Space Administration (NASA), needs to produce Pu-238 for space exploration Major activities are focused at Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) Key part of program is movement of targets between INL and ORNL BEA Research Reactor (BRR) package licensed for transport of unirradiated and irradiated targets

42 ENGINEERING↗
DOE OSTI2017

The Gd based Neutron Absorber for Spent Fuel Storage and Transport Cask - 17636

The criticality safety calculations were performed for integral model which has single rack wall of stainless steel added neutron absorber material. The integral model was verified by reference model composed of 3 layer; basket cell, neutron absorber and sheath. For the neutron absorber material, Gd was selected due to its high neutron absorption cross section in thermal region compared to B. The mathematical model based on validated code, KENO-Va in SCALE was used and the data library is ENDF/B-VII 238-Group. The k{sub eff} was calculated for changes in Gd and B contents in flooded condition. The Gd added stainless steel shows higher performance than B at same atomic percent and reactivity reduction decreases as absorber content increases. Therefore, the Gd 0.47 at% was selected for 'optimum Gd content' for effective use of neutron absorber. With Gd 0.47 at%, B, Sm, Er and Eu was considered for additional neutron absorber and B shows the best performance among them. The Gd 0.47 at% with B 2.05 at% was selected for (optimum neutron absorber content'. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Damaged Fuel Rod Service with German Turnkey Technology NCS 45 Transport Casks and Under-Water-Brazing Technology - 18182

Transportation of damaged irradiated fuel rods with the package NCS 45 developed and operated by DAHER Nuclear Technologies GmbH (DNT) saves considerable technical and financial resources in the German phase-out scenario. The technology can be used world-wide to speed up decommissioning of NPPs. In each German reactor (and most likely in all reactors in the world) a small number of damaged fuel rods have accumulated over the operation phase. In general, transport and storage solutions for fuel are concentrated on the bulk of material, i. e. complete fuel assemblies containing undamaged fuel rods. It is hence a rather straightforward process to remove almost all the fuel from the reactor pools. However, damaged fuel rods - which have to be removed from the pool prior to dismantling - can cause considerable delay in dismantling and high costs for safe-guarding and specialized solutions, even though they only contain a comparatively very tiny amount of fuel. DNT has all required tools, equipment and experience to mitigate the described problem: - The package NCS 45 licensed for the transport of damaged Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel rods with a burn-up of up to 120 GWd/MgHM; - A patented underwater brazing (encapsulation) technology to render very high burn-up damaged fuel rods fit for transport; - Qualified personnel to plan, license and operate such encapsulation process and the NCS 45 in a NPP and to carry out the transports; - Proven record of successful removal of damaged fuel rods from several shut-down reactors in Germany. The package NCS 45 is currently licensed in several European countries. Due to its state-of-technique design - type B(U)F-96 approval - it could be licensed and used for world-wide applications. The underwater brazing technology was developed and patented by DNT to address the challenge of transporting high burn-up and/or damaged UOX and MOX fuel rods. Above a burn-up of 55 GWd/MgHM there are neither enough data to assess the behavior of the cladding under stress nor sufficient benchmarks available to validate the source terms of the fuel. Hence, high burn-up fuel rods are encapsulated in the fuel pool by using the underwater-brazing-technology before they are loaded into the package NCS 45. Furthermore, the capacity of the NCS 45 to transport damaged fuel rods can be considerably enhanced by encapsulating these rods. (authors)

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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)

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