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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Systems Analysis of hiRX for Improved Reprocessing Safeguards

The goal of this work was to examine the potential safeguards improvement of the High Resolution X-ray (hiRX) technology for aqueous reprocessing plants. Existing plants utilize KEdge and Hybrid K-Edge Densitometry for routine accountancy measurements for the key input and output accountability tanks. The IAEA also use these technologies for verification measurements. The hiRX technology has the potential to significantly reduce the measurement uncertainty with an equally simple but less costly instrument that uses a much smaller sample size. Such a technology could be highly desirable for both the operator and inspector. This work utilizes a Separations and Safeguards Performance Model of a PUREX reprocessing plant to determine the safeguards improvement offered by the reduced measurement uncertainty of hiRX. Results show that a significant improvement in the ability to detect a protracted diversion may be possible.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Multiscale models of nuclear waste reprocessing: from the mesoscale to the plant-scale

Nuclear waste reprocessing and nonproliferation models are needed to support the renaissance in nuclear energy. This report summarizes an LDRD project to develop predictive capabilities to aid the next-generation nuclear fuel reprocessing, in SIERRA Mechanics, Sandia’s high performance computing multiphysics code suite and Cantera, an open source software product for thermodynamics and kinetic modeling.

Rao, Rekha Ranjana↗
DOE OSTI2017

Insight into “Changing-look” AGN Mrk 1018 from the Fe Kα Line: The Reprocessing Gas Has Yet to Fully Respond to the Fading of the AGN

Mrk 1018 is a “changing-look” active galactic nucleus (AGN) whose optical spectrum transitioned from a Type 1.9 to a Type 1 between 1979 and 1984, and then back to a Type 1.9 in 2015. This latest transition was accompanied by a decrease in X-ray flux. We analyze the Chandra spectra from 2010 and 2016 and NuSTAR spectra from 2016, with a careful treatment of pileup in the Chandra spectrum from 2010, and self-consistently model absorption, reflection, and Fe Kα line emission in the X-ray spectra from 2016. We demonstrate that while the 2–10 keV X-ray flux decreased by an order of magnitude (1.46{sub −0.13}{sup +0.10}×10{sup −11}–1.31{sub −0.04}{sup +0.09}×10{sup −12} erg s{sup −1} cm{sup −2}), the Fe Kα equivalent width (EW) increased from 0.18{sub −0.12}{sup +0.17} to 0.61{sub −0.25}{sup +0.27} keV due to a depressed AGN continuum. We jointly fit the Chandra and NuSTAR spectra from 2016 using the physically motivated MYTorus model, and find that the torus orientation is consistent with a face-on geometry and that lines of sight intersecting the torus are ruled out. While we measure no line-of-sight absorption, we do measure a column density of N{sub H}=5.38{sub −4.0}{sup +14}×10{sup 22} cm{sup −2} for gas out of the line of sight that reprocesses the X-ray emission. We find a high relative normalization between the Compton-scattered emission and transmitted continuum, which is indicative of time lags between the primary X-ray source and reprocessing gas. We predict that the Fe Kα line will respond to the decrease in AGN flux, which would manifest as a decrease in the Fe Kα EW.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2018

Reprocessed MPL data sets

This dataset contains reprocessed micropulse lidar (MPL) data (equivalent to the ARM cloud mask VAP's lidar fields) which were carefully corrected for each site. These reprocessed data were used in Silber et al. (JGR, 2018). The data extend up to a height of 15 km and have vertical and temporal resolutions of 15 m and 30 s, respectively.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2013

Novel Sorbent Development and Evaluation for the Capture of Krypton and Xenon from Nuclear Fuel Reprocessing Off-Gas Streams

The release of volatile radionuclides generated during Used Nuclear Fuel reprocessing in the US will most certainly need to be controlled to meet US regulatory emission limits. A US DOE sponsored Off-Gas Sigma Team has been tasked with a multi-lab collaborative research and development effort to investigate and evaluate emissions and immobilization control technologies for the volatile radioactive species generated from commercial Used Nuclear Fuel (UNF) Reprocessing. Physical Adsorption technology is a simpler and potential economical alternative to cryogenic distillation processes that can be used for the capture of krypton and xenon and has resulted in a novel composite sorbent development procedure using synthesized mordenite as the active material. Utilizing the sorbent development procedure, INL sigma team members have developed two composite sorbents that have been evaluated for krypton and xenon capacities at ambient and 191 K temperature using numerous test gas compositions. Adsorption isotherms have been generated to predict equilibration and maximum capacities enabling modeling to support process equipment scale-up.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Novel Sorbent Development and Evaluation for the Capture of Krypton and Xenon from Nuclear Fuel Reprocessing Off-Gas Streams

The release of volatile radionuclides generated during Used Nuclear Fuel reprocessing in the US will most certainly need to be controlled to meet US regulatory emission limits. A US DOE sponsored Off-Gas Sigma Team has been tasked with a multi-lab collaborative research and development effort to investigate and evaluate emissions and immobilization control technologies for the volatile radioactive species generated from commercial Used Nuclear Fuel (UNF) Reprocessing. Physical Adsorption technology is a simpler and potential economical alternative to cryogenic distillation processes that can be used for the capture of krypton and xenon and has resulted in a novel composite sorbent development procedure using synthesized mordenite as the active material. Utilizing the sorbent development procedure, INL sigma team members have developed two composite sorbents that have been evaluated for krypton and xenon capacities at ambient and 191 K temperature using numerous test gas compositions. Adsorption isotherms have been generated to predict equilibration and maximum capacities enabling modeling to support process equipment scale-up.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Separation of Xenon from Krypton in UNF Reprocessing Off-Gas

Reprocessing of used nuclear fuel (UNF) results emits several volatile radionuclides that must be removed from the bulk off gas prior to release in order to meet U.S. regulatory emission requirements. After other volatile radionuclides have been removed, krypton (Kr) remains, along with associated with xenon (Xe) in the gas stream. The half-life of Xe is short enough that by the time UNF is sufficiently cooled to be reprocessed Xe is no longer a radiological concern and it may be release to atmosphere or collected for other purposes. Kr, however, must be placed in long term storage. In order to ensure the smallest practical volume for long term storage, the Xe, which is nearly ten times the volume of Kr, must be separated from the Kr. Although cryogenic distillation has been used successfully for this purpose, high capital and operational cost, along with safety considerations, made development of gas adsorption processes for separation of Xe and Kr attractive. INL developed an engineered form sorbent consisting of silver mordenite in a polyacrylonitrile matrix (AgZ-PAN) capable of complete separation of Xe from Kr in a single pass, allowing Kr to be captured downstream with no detectable Xe present.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Multiscale Evaluation of Acetohydroxamic Acid (AHA) Radiolysis Under Used Nuclear Fuel Reprocessing Solvent System Conditions

Acetohydroxamic acid (AHA) has been proposed as an alternative agent for the selective separation of plutonium and neptunium from co-extracted uranium during the reprocessing of used nuclear fuel. However, the fundamental radiolytic behavior of this molecule under envisioned process conditions – i.e., acidic biphasic solvent systems – is not sufficiently understood to support process applications. Here we present a systematic irradiation study (steady-state gamma and time-resolved pulsed electron) into the radiolytic integrity of AHA and formation of degradation products in aqueous nitric acid (HNO3) solutions (0.2 M) in presence and absence of an organic phase, comprising current (tri-butyl phosphate - TBP) and future (N,N-di-(2-ethylhexyl)butyramide - DEHBA and di-2-ethylhexylisobutyramide - DEHiBA) reprocessing ligands dissolved in n-dodecane diluent. Experimental data are complimented by predictive multiscale model calculations for the elucidation of underpinning mechanisms.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Classic Nuclear Fuel Reprocessing Flowsheet

This is a flowsheet as well as a series of subsheets to be used for discussion on the standard design of a reprocessing plant. This flowsheet consists of four main sections: offgas handling, separations, solvent wash, and acid recycle. As well as having the main flowsheet, subsections have been broken off into their own sheets to provide for larger font and ease of printing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Recovery from loss of reductant feed to a U/Pu separation contactor, magnox reprocessing plant, Sellafield - 14217

Following a revealed sustained loss of reductant feed to the Primary Separation 4 (PS4) U/Pu separation contactor in the Magnox Reprocessing Plant in 2014 the solvent extraction cycles were safely shutdown. However, initial mass balance calculations indicated that over four times the typical mass of Pu could be present in PS4. Due to the loss of reductant feed, much of this Pu was expected to be in the solvent phase of Tributyl Phosphate and Odourless Kerosene (TBP/OK) where it would be difficult to observe with the neutron monitors located under PS4. Samples were taken of the solvent in PS4 which confirmed the presence of Pu in solvent; normally almost all of the Pu in PS4 is in the aqueous phase. The volume ratio of solvent to aqueous in the PS4 contactor is high. This meant that simply re-dosing PS4 with reductant could not be supported as a recovery option due to the potential to realise high Pu concentrations in the aqueous phase. This paper details the recovery options which were considered and the ultimate decision to recover PS4 using Gadolinium (Gd) poisoning and Gd-dosed reductant feeds. The recovery included: - The need to consider the impact of Gd poisoning on monitoring of the recovery with the PS4 neutron monitors; - Close support from criticality safety personnel on the plant at all times during the recovery; - Production of operator aids and short fit for purpose safety documentation to assist with the recovery. The recovery of PS4 proceeded safely and smoothly. There were no losses of Pu to raffinate streams despite the prolonged shutdown. The plant was then safely rundown to flush out residual Pu and Gd. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Initial recovery of the magnox reprocessing plant following a loss of reductant feed in the solvent extraction process - 14429

A recent event in the Magnox Reprocessing Plant involved the loss of reductant feed to a contacting vessel and required the plant to be shutdown. The criticality team attended the site out-of-hours to facilitate the production of a recovery plan. This initial recovery plan involved manual dosing of the contactor with reductant but was heavily dependent upon initial information regarding the length of time the maloperation had been occurring. Key assumptions fundamental to maintaining criticality safety during the planned recovery were clearly highlighted in the assessment that was written to support the operation, which allowed informed decisions to be made regarding the adequacy of the plan, once further information emerged suggesting the fault had been ongoing for much longer than originally assumed and therefore the amount of Pu in the contactor was much higher than initially thought. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Burnup credit for BWR UOX used fuels: criticality-safety sensitivity studies for reprocessing - 14609

Actinides Only Burnup Credit (BUC) has been successfully used for years for PWR UOX fuels at AREVA La Hague reprocessing plant. For BWR UOX fuel assemblies, the application of the methodology developed for PWR fuel is not straightforward. Indeed, BWR UOX fuel assemblies have specific and various design characteristics and progress can be made on our understanding of the impact of such characteristics on the rotary dissolver k{sub eff} value, taking into account various irradiation conditions. Moreover, the void fraction introduces a new variable, compared to PWR fuels. The purpose of this paper is to present, for a reference BWR UOX fuel assembly, the impact of design characteristics and irradiation conditions on the rotary dissolver k-effective. The impacts of each parameter are studied in an uncorrelated way and are defined according to burnup and void fraction. The design characteristics and irradiation conditions of the Fukushima-Daini benchmark are taken into account as reference values for all these parameters. A range of values is defined for each of these parameters in order to consider a large variety of design characteristics and irradiation conditions Depletion, with APOLLO2-A code, and criticality, with CRISTAL V1.2 package, calculations are successively performed to define impact of each parameter on dissolver reactivity. Then a ranking of these parameters according to their impact on the dissolver's reactivity is done. This ranking could be useful in a general BWR UOX burnup credit approach to cover a collection of BWR UOX fuel assemblies by a penalizing fuel assembly defined with a particular set of parameters. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Adsorption of molecular iodine and alkyl iodides from spent-nuclear-fuel-reprocessing off-gas using reduced silver mordenite

Radioactive iodine is a hazardous byproduct of spent-nuclear-fuel reprocessing that must be removed from the off-gas stream before it can be discharged. Reduced silver mordenite (Ag 0 Z) is currently the baseline material for iodine capture in the U.S. Although the performance characteristics and capture mechanisms of I 2 and CH 3 I have been established, similar investigations into long-chain organic iodides have yet to be performed. In this study, thin beds of Ag 0 Z were loaded with I 2 , CH 3 I, C 4 H 9 I, and C 12 H 25 I (5 ppm to 50 ppm) carried in a dry air stream at 150°C. The maximum iodine capacity was 105 ± 5 mg I/g Ag 0 Z for all species. Saturated Ag 0 Z samples were characterized using scanning electron microscopy, X-ray fluorescence, X-ray photoelectron spectroscopy, diffuse reflectance UV–visible spectroscopy, pair distribution function analysis, and thermogravimetric analysis. Further, near-complete Ag utilization and similar physical/chemical properties were observed for all samples. Through a comparison with previous studies and an investigation of aged Ag 0 Z, we propose that iodine species react with Ag + at exchange sites, forming α-AgI within the mordenite channels, and with surface Ag 0 nanoparticles, yielding β-/γ-AgI. The available Ag sites in the interior (Ag + ) and exterior (Ag 0 ) of mordenite determine the adsorption capacity since α-AgI formation is limited by the total pore volume. Potential iodine uptake routes were summarized for aging and non-aging environments. A scalable predictive model was implemented for deep-bed iodine removal, and predictions were in good agreement with experimental data. Sensitivity analysis suggests that iodine uptake kinetics is governed by pore diffusion.

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

Molten Salt Reactors and Electrochemical Reprocessing: Synthesis and Chemical Durability of Potential Waste Forms for Metal and Salt Waste Streams

The molten salt reactor (MSR) is one of the leading advanced nuclear reactor candidates to replace current nuclear reactor technologies in the U.S. Besides having more economical and reliable designs, MSRs have are amenable to a closed fuel cycle, in which electrochemical reprocessing can be performed to recycle the used nuclear fuel. This review intends to provide information about potential waste forms for metal and salt waste streams from these salt-based nuclear processes. Metal waste streams arise from reactor components and structural materials. Salt waste streams are generated during reactor operations as fission products build up in salt-fueled systems. Waste forms that have the highest waste loading and/or have shown the most commercial promise are discussed with an emphasis on the current state of efforts to understand the synthesis and chemical durability of metal and ceramic waste forms.

molten salt reactors, electrochemical reprocessing↗
DOE OSTI2013

Know-how and experience of the decommissioning of the Eurochemic reprocessing plant

Belgoprocess started the industrial decommissioning of the main process building of the former Eurochemic reprocessing plant in 1987. In 2008 the eastern part was torn down and in 2010 the central part of the building was removed. The demolition of the third and last part is planned in 2014. At that point the area will be restored to brown field conditions. During the 25 years of decommissioning the Eurochemic installations a lot of effort has been invested in improving the efficiency. The Belgian decommissioning policy is based on maximizing effort for decontamination with the goal of unconditional release and minimization of nuclear waste. As a result a number of installations have been implemented that reduce the processing costs, increase the output of decontamination and increase the speed of release measurements. - The abrasive blasting installation is used to remove a thin layer from metal components and concrete objects. - The concrete crushing and sampling installation is used as the final measurement before unconditional release of concrete debris. - The concrete spec installation is used to separate the contaminated part from the non contaminated part of concrete debris. As the end of the decommissioning project is approaching the time is right to look back and evaluate the course of the project. Looking back what were the key decisions during the project, what were important milestones and what could or should have been done differently knowing what we know now. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Storage and Reprocessing of Spent Nuclear Fuel

Addressing the problem of waste, especially high-level waste (HLW), is a requirement of the nuclear fuel cycle that cannot be ignored. We explore the two options employed currently, long-term storage and reprocessing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗