SEARCH · Engineering Papers

Results for “Spent”

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.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 433 records · Page 24

All sources NASA NTRS DOE OSTI

DOE OSTI2023

COBRA-SFS Version 6.2.1 User Guide: A Thermal Hydraulic Analysis Code for Spent Fuel Storage and Transportation Casks

This document is an updated and reduced version of the complete documentation of the COBRA-SFS thermal analysis code (COolant Boiling in Rod Arrays – Spent Fuel Storage) (Michener et al. 1995). This manual provides the analyst with input instructions and fundamental guidelines for contrasting models using COBRA-SFS. COBRA-SFS is designed to analyze the dry storage and transportation casks that house spent or used nuclear fuel. The primary focus and application is on square light water reactor fuel assemblies, however the code also has the capability to model hexagonal assemblies and is will suited to any geometry that can be defined as an extruded 2-D map.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

FY2024 Mid-Year Report: Verification of Spent Fuel Inside Dry Storage Casks by Cask Top Fast Neutron Mapping

This project is developing a prototype scanner array verification system for detection of missing fuel assemblies in spent-fuel storage casks. The prototype consists of six fast-neutron scintillator detectors mounted to a linear actuator frame that is placed on the top of a spent fuel cask to scan across all fuel assembly positions. The scanner array was assembled and tested at LLNL in prior years. A field test schedule has been requested at the Idaho National Laboratory (INL) Cask Farm site for FY2024. Following the Field Test, we will present results and discuss the technology with the IAEA. The IAEA may have special requirements for portability, shipping, lifting, and installation. We will incorporate additional improvements to the system based on lessons learned from the Field Test and feedback from the IAEA. If successful, the technology can be transferred to the IAEA or other stakeholders for assessment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Advanced Reactors Spent Fuel and Waste Streams Disposition Strategies

This report describes research and development (R&D) activities conducted during Fiscal Year 2023 (FY23) in the Advanced Fuels and Advanced Reactor Waste Streams Strategies work package in the Spent Fuel Waste Science and Technology (SFWST) Campaign supported by the United States (U.S.) Department of Energy (DOE). This report is focused on evaluating and cataloguing Advanced Reactor Spent Nuclear Fuel (AR SNF) and Advanced Reactor Waste Streams (ARWS) and creating Back-end Nuclear Fuel Cycle (BENFC) strategies for their disposition. The R&D team for this report is comprised of researchers from Sandia National Laboratories and Enviro Nuclear Services, LLC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Advanced reactors spent fuel and waste streams disposition strategies

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

SCALE 6.3 Validation: Spent Nuclear Fuel

This report is the fifth volume in a series documenting the validation of SCALE 6.3 with ENDF/B-VII.1 libraries for nuclear criticality safety, reactor physics, radiation shielding, and spent nuclear fuel applications. This fifth volume, which focuses on validating SCALE capabilities that impact spent nuclear fuel applications, provides an update of the similar validation reported for SCALE 6.2.4.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

International Collaboration Activities in Geologic Disposal R&D: Spent Fuel and High-Level Waste Disposition (FY25 Progress Report)

This report describes the FY25 status of international collaboration on geologic disposal research and development (R&D) in the Office of Spent Fuel and High-Level Waste Disposition (SFHLWD) in the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE). The mission of this office is to provide confidence in the safe long-term management of the nation’s spent nuclear fuel and high-level radioactive waste by reducing uncertainty and advancing technology for extended storage, transportation, and geologic disposal. R&D in geologic disposal is administered in SFHLWD’s Disposal R&D Campaign.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Sequential Recovery of Critical Metals from Leached Liquor of Processed Spent Lithium-Ion Batteries

The processing and extraction of critical metals from black mass is important to battery recycling. Separation and recovery of critical metals (Co, Ni, Li, and Mn) from other metal impurities must yield purified metal salts, while avoiding substantial losses of critical metals. Solvent extraction in batch experiments were conducted using mixed metal sulphates obtained from the leach liquor obtained from spent and shredded lithium-ion batteries. Selective extraction of Mn 2+ , Fe 3+ , Al 3+ and Cu 2+ from simulated and real leached mixed metals solution was carried out using di-2-ethylhexylphophoric acid (D2EPHA) and Cyanex-272 at varying pH. Further experiments with the preferred extractant (D2EPHA) were performed under different conditions: changing the concentration of extractant, organic to aqueous ratio, and varying the diluents. At optimum conditions (40% v/v D2EPHA in kerosene, pH 2.5, O:A = 1:1, 25 °C, and 20 min), 85% Mn 2+ , 98% Al 3+ , 100% Fe 3+ , and 43% Cu 2+ were extracted with losses of only trace amounts (<5.0%) of Co 2+ , Ni 2+ , and Li + . The order of extraction efficiency for the diluents was found to be kerosene > Exxal-10 >>> dichloromethane (CH 2 Cl 2 ) > toluene. Four stages of stripping of metals loaded on D2EPHA were performed as co-extracted metal impurities were selectively stripped, and a purified MnSO 4 solution was produced. Spent extractant was regenerated after Fe 3+ and Al 3+ were completely stripped using 1.0 M oxalic acid (C 2 H 2 O 4 ).

25 ENERGY STORAGE↗

DOE OSTI2022

Recycling of Lead Pastes from Spent Lead–Acid Batteries: Thermodynamic Constraints for Desulphurization

Lead–acid batteries are important to modern society because of their wide usage and low cost. The primary source for production of new lead–acid batteries is from recycling spent lead–acid batteries. In spent lead–acid batteries, lead is primarily present as lead pastes. In lead pastes, the dominant component is lead sulfate (PbSO 4 , mineral name anglesite) and lead oxide sulfate (PbO•PbSO 4 , mineral name lanarkite), which accounts for more than 60% of lead pastes. In the recycling process for lead–acid batteries, the desulphurization of lead sulfate is the key part to the overall process. In this work, the thermodynamic constraints for desulphurization via the hydrometallurgical route for recycling lead pastes are presented. The thermodynamic constraints are established according to the thermodynamic model that is applicable and important to recycling of lead pastes via hydrometallurgical routes in high ionic strength solutions that are expected to be in industrial processes. The thermodynamic database is based on the Pitzer equations for calculations of activity coefficients of aqueous species. The desulphurization of lead sulfates represented by PbSO 4 can be achieved through the following routes. (1) conversion to lead oxalate in oxalate-bearing solutions; (2) conversion to lead monoxide in alkaline solutions; and (3) conversion to lead carbonate in carbonate solutions. Among the above three routes, the conversion to lead oxalate is environmentally friendly and has a strong thermodynamic driving force. Oxalate-bearing solutions such as oxalic acid and potassium oxalate solutions will provide high activities of oxalate that are many orders of magnitude higher than those required for conversion of anglesite or lanarkite to lead oxalate, in accordance with the thermodynamic model established for the oxalate system. An additional advantage of the oxalate conversion route is that no additional reductant is needed to reduce lead dioxide to lead oxide or lead sulfate, as there is a strong thermodynamic force to convert lead dioxide directly to lead oxalate. As lanarkite is an important sulfate-bearing phase in lead pastes, this study evaluates the solubility constant for lanarkite regarding the following reaction, based on the solubility data, PbO•PbSO 4 + 2H + ⇌ 2Pb 2+ + SO 4 2– + H 2 O(l).

25 ENERGY STORAGE↗

DOE OSTI2024

Krypton-85 chronometry of spent nuclear fuel

We describe the use of the radionuclide 85 Kr, which is produced by nuclear fission and has a half-life of 10.76 years, to determine the age of spent nuclear fuel. The method is based on mass-spectrometric measurement of the relative abundance of fissiogenic Kr isotopes extracted from a fuel sample, and we show that it can be applied to micron-scale particles of spent fuel that are analogous to particles that have been released into the environment from various nuclear facilities in the past. 85 Kr chronometry is potentially valuable for identifying and attributing nuclear materials, grouping samples into collections of common origin that can be used to reconstruct the origin and irradiation history of the material, and verifying declared nuclear activities in the context of international monitoring programs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

NASA NTRS1972

Refocusing of the spent axisymmetric beam in klystron tubes

Analytic methods were developed and employed to optimize the magnetic field transition region between the output interaction gap of a klystron and a multi-stage depressed potential electrostatic beam collector, in order to enhance the power conversion efficiency of satellite-borne broadcast transmitters. Permanent magnet structures were designed to provide the magnetic field distributions required to expand and recollimate the spent electron beam of the power amplifier klystron for proper entry into the beam collector. These design criteria for magnetic field distributions for expanding and recollimating spent-beam electrons for optimal entry into a multi-stage depressed potential collector are generally applicable to traveling-wave tubes as well as klystrons.

Branch, G. M.↗

NASA NTRS1977

Analysis of spent beam refocusing to achieve optimum collector efficiency

A reasonable criterion for evaluating the effectiveness of spent beam refocusing is the reduction of spent beam turbulence. The rms deviation of particle angles where the angles are calculated from the ratio of radial velocity to axial velocity is one measure of beam turbulence. It is demonstrated that the angular deviation can be reduced by almost half in some magnetic field configurations. Experimental evidence indicates that beam processing of this type is most likely to yield an optimum collector efficiency.

Stankiewicz, N.↗

NASA NTRS1985

Multistage spent particle collector and a method for making same

A description is given of a spent particle collector which maintains structural integrity when raised to a high temperature although constructed of materials having widely different coefficients of expansion. The collector is comprised of one or more axisymmetric stages, each stage comprising a subassembly. A subassembly includes an inner pyrolytic graphite ring, a transition ring, a ceramic insulator ring and an outer metal ring which forms part of the wall of the collector. Each transition is of a ductile metal having high thermal conductivity and is provided with an annular sputter shield wall extending toward the source of spent particles and, where necessary, a trough in the other surface to enclose the sputter shield of the next adjacent transition ring. A plurality of radial extending slots are provided in a transition ring to form segments which are retained in their position by the sputter shield.

Ebihara, B. T.↗

NASA NTRS1988

A re-examination of spent beam refocusing for high-efficiency helix TWT's and small MDC's

The usefulness of the concept of spent-beam refocusing in optimizing the performance of low- and medium-power helix TWTs (traveling-wave tubes) equipped with small multistage depressed collectors (MDCs) is examined. Several direct comparisons of the performance of individually optimized TWT-MDC combinations with and without controlled beam expansion and recollimation are presented. The collector efficiency of a number of representative space and airborne TWT-MDCs, which do not use refocusing, is compared to a measure of quality (standard of excellence) established by an examination of irrecoverable MDC losses. The results suggest that the application of the traditional concept of spent-beam refocusing to most helix TWTs that are equipped with small MDCs in order to obtain maximum efficiency is not required (or even desired). Two effects combine to obviate the need for refocusing: 1) the unexpanded beam more nearly meets the `point source' ideal at the input to the small MDC, and 2) the larger space-charge spreading and the slightly larger (but manageable) injection angles of the unexpanded beam can reduce the amount of backstreaming current and the attendant loss in efficiency.

Ramins, Peter↗

NASA NTRS1990

Spent-beam refocusing analysis and multistage depressed collector design for a 75-W, 59- to 64-GHz coupled-cavity traveling-wave tube

A computational design technique for coupled-cavity tubes (TWTs) equipped with spent-beam refocusers (SBRs) and multistage depressed collectors (MDCs) is described. A large-signal multidimensional computer program was used to analyze the TWT-SBR performance and to generate the spent-beam models used for MDC design. The results of a design involving a 75-W, 59 to 64 GHz TWT are presented. The SBR and MDC designs are shown, and the computed TWT, SBR, and MDC performances are described. Collector efficiencies in excess of 94 percent led to projected overall TWT efficiencies in the 40-percent range.

Wilson, Jeffrey D.↗

DOE OSTI2012

Delayed Gamma-Ray Spectroscopy for Spent Nuclear Fuel Assay

High-energy, beta-delayed gamma-ray spectroscopy is investigated as a non-destructive assay technique for the determination of plutonium mass in spent nuclear fuel. This approach exploits the unique isotope-specific signatures contained in the delayed gamma-ray emission spectra detected following active interrogation with an external neutron source. A high fidelity modeling approach is described that couples radiation transport, analytical decay/depletion, and a newly developed gamma-ray emission source reconstruction code. Initially simulated and analyzed was a “one-pass” delayed gamma-ray assay that focused on the long-lived signatures. Also presented are the results of an independent study that investigated “pulsed mode” measurements, to capture the more isotope-specific, short-lived signatures. Initial modeling results outlined in this paper suggest that delayed gamma-ray assay of spent nuclear fuel assemblies can be accomplished with a neutron generator of sufficient strength and currently available gamma-ray detectors.

Campbell, Luke W↗

DOE OSTI2014

Dose Rate Analysis Capability for Actual Spent Fuel Transportation Cask Contents

The approved contents for a U.S. Nuclear Regulatory Commission (NRC) licensed spent nuclear fuel casks are typically based on bounding used nuclear fuel (UNF) characteristics. However, the contents of the UNF canisters currently in storage at independent spent fuel storage installations are considerably heterogeneous in terms of fuel assembly burnup, initial enrichment, decay time, cladding integrity, etc. Used Nuclear Fuel Storage, Transportation & Disposal Analysis Resource and Data System (UNF ST&DARDS) is an integrated data and analysis system that facilitates automated cask-specific safety analyses based on actual characteristics of the as-loaded UNF. The UNF-ST&DARDS analysis capabilities have been recently expanded to include dose rate analysis of as-loaded transportation packages. Realistic dose rate values based on actual canister contents may be used in place of bounding dose rate values to support development of repackaging operations procedures, evaluation of radiation-related transportation risks, and communication with stakeholders. This paper describes the UNF-ST&DARDS dose rate analysis methodology based on actual UNF canister contents and presents sample dose rate calculation results.

Radulescu, Georgeta↗

DOE OSTI2012

Spent nuclear fuel recycling with plasma reduction and etching

A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF 6 and PuF 4 . During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF 6 (gas) that is formed to decompose back to PuF 4 (solid), and b) to maintain stability of the UF 6 . Uranium (in the form of gaseous UF 6 ) is easily extracted and separated from the plutonium (in the form of solid PuF 4 ). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF 6 (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF 4 or NF 3 as the fluorine sources instead of F 2 or HF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

U and Pu Gamma-Ray Measurements of Spent Fuel Using a Gamma-Ray Mirror Band-Pass Filter

Abstract. We report on the use of grazing incidence gamma-ray mirrors to serve as a narrow band-pass filter for advanced non-destructive analysis (NDA) of spent nuclear fuel. The purpose of the mirrors is to limit the radiation reaching a HPGe detector to narrow spectral bands around characteristic emission lines from fissile isotopes in the fuel. This overcomes the normal rate issues when performing gamma-ray NDA measurements. In a proof-of-concept experiment, a set of simple flat gamma-ray mirrors were used to directly observe the atomic florescence lines from U and Pu from spent fuel pins with the detector located in a shirt-sleeve environment. The mirrors, consisting of highly polished silicon substrates deposited with WC/SiC multilayer coatings, successfully deflected the lines of interest while the intense primary radiation beam from the fuel was blocked by a lead beam stop. The gamma-ray multilayer coatings that make the mirrors work at the gamma-ray energies used here (~ 100 keV) have been experimentally tested at energies as high as 645 keV, indicating that direct observation of nuclear emission lines from 239Pu should be possible with an appropriately designed optic and shielding configuration.

Ziock, Klaus-Peter↗