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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 847 records · Page 47
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Fission Product Gas Monitoring During Fuel Drying Operations - 20094

The goal of the project was to devise and implement a method of accurately qualifying any noble gas emitted during fuel drying operations (part of the dry fuel storage process) in order to provide the data needed to validate the off-site dose calculations. The imputes of the project was the NRC publishing Information Notice 18-01, 'Noble Fission Gas Releases During Spent Fuel Cask Loading Operations' in February 2018. The information notice describes some industry events that occurred during vacuum drying operations as part of a dry fuel campaign. The solution was devised by Fort Calhoun Station staff in partnership with Mirion Technologies technical staff. The monitoring was accomplished by directing all the effluent from the fuel storage cask vacuum drying machine through a sample chamber containing a compact ISOCS characterized CZT based gamma spectrometer connected to a Mirion Data Analyst. This resulted near real-time quantification of the radionuclides of concern. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Investigating Wireless Quantum Key Distribution for Advanced Reactor Communications

Remote operation of small modular reactor (SMR) facilities is an appealing prospect for streamlined operation costs, safety concerns, and user convenience. The ability to guarantee security of communication channels between offsite users and nuclear reactor facilities is critical to enabling remote operations, given the sensitive nature of reactor state data. This report proposes a configuration implementing quantum key distribution (QKD) and advanced encryption standard (AES) protocols to establish secure reactor data feeds. The model outlined, which utilizes the Argonne-developed SeQUeNCe software package and custom MATLAB code, determines free-space channel losses for microwave-band communications in a variety of weather conditions and simulates QKD via ground-satellite links. This report demonstrates the feasibility of repeatable, real-time key generation and distribution for this configuration towards enabling secure wireless reactor communications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI1978

An Assessment of the Nevada Test Site for Low-Level Waste Management

This report presents a summary of description information, facilities, and operating practices at the Nevada Test Site. It is concluded that NTS is not only a feasible location for surface and near-surface storage of low-level wastes, but also a potential site for underground storage of higher level wastes.

052002* -- Nuclear Fuels-- Waste Disposal & Storag↗
DOE OSTI2020

TWF Waste Container Intra-Facility Transfer and Outgoing Shipment

This procedure provides instructions for the inspection and transfer of Transuranic (TRU) Waste containers within the Transuranic Waste Facility (TWF) (Waste Storage Buildings [WSBs], Characterization Waste Storage Building [CWSB]) and Characterization Trailers. This also includes the preparation to transfer to other LANL facilities. Performance of this procedure meets the requirements of SR 4.2.1.1, 4.2.1.2, 4.2.1.4, 4.2.1.5, 4.3.4.1, SAC 5.9.1, SAC 5.9.3 and AC 5.8.13 2nd bullet, as well as requirements set forth in the Los Alamos National Laboratory (LANL) Hazardous Waste Facility Permit (HWFP).

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

WSF PISA example for SAFER pilot

On October 15, 2021, the Radioactive and Hazardous Waste Management Facility Manager declared a Potential Inadequacy in the Safety Analysis (PISA) due to a new information relating to an error discovered in calculation AB-WSF-20-002. Calculation AB-WSF-20-002 is used to determine source terms and heat release rates for vehicle and aircraft impacts with fuel fires involving TRU waste containers. The source term for an aircraft’s impact has different source terms factors (i.e., DR, ARF, RF, LPF) that are applied to multiple categories of drums impacted in the accident scenario based on the physical stresses of the aircraft impact and involvement in the subsequent fuel pool fire. The source terms of the multiple categories of affected drums are summed to determine the final source term for the accident scenario. Certain categories of affected drums are modeled to lose their lid, causing material to eject and burn unconfined on the ground. For these specific categories, a lower ARF value was incorrectly applied in AB-WSF-20-002 (i.e., an ARF value of 1E-3 was used instead of the DOE-STD-5506-2007 directed ARF value of 1E-2). For these categories, the ARF is applied to a relatively small number of low-activity drums (i.e., 17 drums at 2.9 PE-Ci each) compared to the total number of drums affected (i.e., 166 drums totaling 621.8 PE-Ci). Therefore, this error is not expected to significantly increase final dose for the accident scenario. No immediate actions or compensatory measures are required to maintain the facility in a safe condition.

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DOE OSTI2022

Project Report for Cold Spray: Phase 2

In support of the ongoing environmental protection mission at the Hanford Site, led by Washington River Protection Services (WRPS), the cold spray team has continued development of the tank refurbishment technology. Orano Federal Services, along with subcontractors VRC Metal Systems (VRC) and Robotic Technologies of Tennessee (RTT), have worked with WRPS experts to refine the tools demonstrated in Phase 1 of this project, with the potential to refurbish double shell tanks (DST)s on the Hanford Site, extending the service lives of these vital waste storage facilities.

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DOE OSTI2020

Start-up of Silo 130 Waste Retrieval at La Hague: first Step Towards Reducing Legacy Inventory - 20020

On June 25, 2019, the first grapple of waste and debris was lifted from Silo 130, one of the most challenging legacy waste storage silos at La Hague. This was the achievement of more than ten years of studies and efforts focused on solving some very unique technical and safety challenges, and on designing an innovative, efficient solution. The waste is from de-cladding materials from past Gas-Cooled Reactor fuel reprocessing operations. It is an unsorted mixture of magnesium cladding pieces, graphite debris, and various other metallic waste types. These materials can be chemically reactive and therefore need specific dispositions for handling and storing. The building is more than 40 years old and does not meet modern standards. French safety authorities have required the silo to be emptied. After about three years of operation, the bulk of this waste will be removed and placed in safe storage, pending the selection of a final conditioning solution. Removal of bottom debris will then be possible. When finished, this operation will bring a conclusion to a significant legacy-related hazard on the site. The silo will then be available for final cleaning and de-activation. This paper will describe the technical, safety, and economical challenges, and the various decisions and optimizations that led to the definition of the process and technology to be implemented. Several innovative features were developed and implemented, in all sections of the project: project organization, progressive approach for the waste disposition strategy, light building design, customized and robust retrieval technology, processing technologies involving the most modern robotic and sorting techniques, storage method... Several of these first-of-a-kind innovative features have been developed and qualified specifically for this application, using the detailed qualification procedure specified for all La Hague projects. The facility was built in less than five years, overcoming various difficulties along the line. Commissioning is now finished with the combined efforts of the project, engineering, start-up, and operating teams assisted by dedicated suppliers. The facility is now ramping-up. The paper will also report on the commissioning and start-up of the facility and describe the first operational results. (authors)

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DOE OSTI2024

Analytical Assessment of Impurities Found in Process Gases using the InfiTOF, a High Resolution, Low Mass, Variable Time-of-Flight Mass Spectrometer

The JEOL InfiToF is a high-resolution, low mass (HRLM) mass spectrometer (MS) with the application potential to meet analytical needs for tritium handling systems, such as those at the Savannah River Tritium Facilities (SRTE) and those used in fusion research. The Service Level Agreement (SLA) #144 line 27 “InfiTOF R&D” has concluded its first year of efforts under Tritium Recycle and Renewal. The project goals for FY22 were: (1) Production of a finalized modification list for tritium hardened components in the InfiTOF, (2) perform an assessment on impurity gases commonly found in process gases from Tritium Facilities and (3) begin developing a set of standardized work instructions that will be used as a basis for operations on the InfiTOF.

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DOE OSTI2025

Incorporating geological structure into sensitivity analysis of subsurface contaminant transport

Simulating subsurface contaminant transport at the kilometer-scale often entails modeling reactive flow and transport within and through complex geologic structures. These structures are typically meshed by hand and as a result geologic structure is usually represented by one or a few deterministically generated geological models for uncertainty studies of flow and transport in the subsurface. Uncertainty in geologic structure can have a significant impact on contaminant transport. In this study, the impact of geologic structure on contaminant tracer transport in a shale formation is investigated for a simplified generic deep geologic repository for permanent disposal of spent nuclear fuel. An open-source modeling framework is used to perform a sensitivity analysis study on transport of two tracers from a generic spent nuclear fuel repository with uncertain location of the interfaces between the stratum of the geologic structure. The automated workflow uses sampled realizations of the geological structural model in addition to uncertain flow parameters in a nested sensitivity analysis. Concentration of the tracers at observation points within, in line with, and downstream of the repository are used as the quantities of interest for determining model sensitivity to input parameters and geological realization. Finally, the results of the study indicate that the location of strata interfaces in the geological structure has a first-order impact on tracer transport in the example shale formation, and that this impact may be greater than that of the uncertain flow parameters.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Sulfur donating extractants for the separation of trivalent actinides and lanthanides

The effective separation of trivalent actinides and lanthanides is capable of reducing the long term radiation hazard associated with used nuclear fuel. Here, this class of separation exploits the tendency of ligands containing large and polarizable soft donor atoms to preferentially bind to the trivalent actinides instead of the lanthanides. Among the soft donors, nitrogen and sulfur based ligands have received the most attention with sulfur donors generally having greater selectivity for the actinides. Herein, the speciation, mechanism, and selectivity for the extraction of lanthanides and actinides from aqueous media into an organic phase by various sulfur containing extractants is reviewed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Mineral-specific heterogeneous neptunium sorption onto geological repository rocks in oxic and anoxic conditions and various temperatures

Neptunium poses challenges for radioactive wastes management due to its long half-life and ingrowth from americium decay, as well as its high mobility in water in its oxidized forms. Np(V)-bearing aqueous solutions were contacted with polished and crushed gneiss and dolerite from the Yeniseisky site, Russia, which is under consideration as a location for an underground repository, both in oxic and anoxic conditions at 20 and 90 °C. Neptunium sorption on crushed gneiss and dolerite in oxic conditions at 20 °C was quite low, but increased substantially in anoxic conditions. Autoradiography on samples treated at 20 °C revealed that sorbed Np was associated mostly with dark colored iron-bearing minerals, and solvent extraction indicated partial reduction of Np(V) to Np(IV). Sorption of Np from aqueous solutions contacted with dolerite and gneiss at 90 °C was in all cases higher than at the lower temperature. Autoradiography of polished rock samples treated in anoxic conditions at 90 °C revealed that Fe-bearing minerals had 3–4 orders of magnitude more Np sorbed on them than the quartz and feldspar grains. X-ray photoelectron spectroscopy indicated this sorbed Np was present as NpO 2 . This study supports the concept that the geologic conditions present at the planned repository site will retard Np migration.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Momentum informed muon scattering tomography for monitoring spent nuclear fuels in dry storage cask

Development of an effective monitoring method for spent nuclear fuel (SNF) in a dry storage cask (DSC) is important to meet the increasing demand for dry storage investigations. The DSC investigation should provide information about the quantity of stored SNF, and quality assurance of materials should be possible without opening the cask. However, traditional nondestructive examination (NDE) methods such as x-rays are difficult to deploy for DSC investigation because a typical DSC is intentionally designed to shield against radiation. To address this challenge, cosmic ray muons (CRMs) are used as an alternative NDE radiation probe because they can easily penetrate an entire DSC system; however, a wide application of muons is often hindered due to the naturally low CRM flux (~10 4 muons/m 2 /min). This paper introduces a newly proposed imaging algorithm, momentum-informed muon scattering tomography (MMST) , and presents how a limitation of the current muon scattering tomography technique has been addressed by measuring muon momentum. To demonstrate its functionality, a commercial DSC with 24 pressurized light water reactor fuel assemblies (FAs) and the MMST system were designed in GEANT4. Three noticeable improvements were observed for MMST system as a DSC investigation tool: (1) a signal stabilization, (2) an enhanced capability to differentiate various materials, and (3) statistically increased precision to identify and locate missing FAs. The results show that MMST improves the investigation accuracy from 79 to 98% when one FA is missing and 51% to 88% when one-half FA is missing. The advancement of the NDE technique using CRM for DSC verification is expected to resolve long-standing problems in increasing demand for DSC inspections and nuclear security.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Low and High LET Degradation Studies of Metal-Loaded Organic Phase Ligands in the ALSEP Process

Organic solutions comprising the Actinide Lanthanide Separation Process (ALSEP) solvent consisting of 0.5 M 2-ethylhexyl phosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) and 0.05 M N,N,N’,N’-tetra(2-ethylhexyl)diglycolamide (T2EHDGA) in n-dodecane were subjected to low LET and high LET irradiation before and after equilibration with an aqueous phase of 3 M HNO 3 . Degradation dose constants revealed greater ligand degradation due to gamma irradiation than alpha irradiation for both ligands. Furthermore, equilibration with nitric acid did not have a significant impact on ligand degradation for either irradiation source. Identified degradation products were similar for both gamma and alpha irradiation and occurred mostly through the rupture of the N–C carbonyl and C–O ether bonds for T2EHDGA and the C–Oether bond in HEH[EHP]. Acid contact appears to alter the degradation pathway by favoring the formation of higher molecular weight recombination products. Furthermore, mixed T2EHDGA-HEH[EHP]-NO 3 complexes were formed with Nd(III) after extraction from 3 M HNO 3 , and low LET gamma irradiation of the Nd(III) loaded organic solution produced similar degradation products as the organic solution absent of Nd(III). Interestingly, and likely due to the greater radiolytic susceptibility of T2EHDGA than HEH[EHP], a HEH[EHP]-Nd(III) complex appears to form as the T2EHDGA degrades with increasing absorbed dose.

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DOE OSTI2020

Preliminary Determination of the Impact of Alkaline Earth Metals on Crystalline Silicotitanate

Savannah River Remediation (SRR) recently deployed the Tank Closure Cesium Removal (TCCR) system which utilizes an ion exchange process to remove 137 Cs from tank waste supernate. In TCCR, salt solution is filtered and then passed through ion exchange columns containing crystalline silicotitanate (CST) media, commercially known as UOP IONSIV™ R9120-B, which is highly selective for removing cesium. The current process focuses on dissolving salt in Savannah River Site (SRS) Tank 10H, followed by processing in the TCCR skid located near the tank. Prior to processing each batch of waste generated in Tank 10H, the projected maximum cesium loading on the CST is determined using an in-tank batch contact equilibrium test. This involves placing a small amount (~0.1 g) of CST in a “teabag” device and submerging it in the tank waste for a period of 10 days. Results from those earlier teabag tests had indicated much lower cesium loading than predicted by the equilibrium sorption model (ZAM). There are numerous possible explanations for this result; however, based on the observed loadings of Ca (and other metals) on the CST from the teabag tests, it was hypothesized that competition from these other ions may be limiting the cesium loadings.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Glycolate Analysis in Tank 22: Developing and Testing Analytical Methods for the Savannah River Site Liquid Waste System

Researchers developed and tested a range of techniques to support a low mg/L Limit of Quantitation (LOQ) of glycolate in radioactive Tank 22 waste solution by Ion Chromatography (IC) and by proton NMR. For the IC method, Dionex OnGuard II cartridges were tested as a means of removing alkali earth and transition metals that can interfere during ion chromatography applications especially when analyzing for low-concentration, chelating analytes. Additionally, concentrations of nitrate in the raw Tank 22 sample (5,000 to 10,000 mg/L) were managed by using reasonable levels of sample dilution. The resulting IC performance quality was enhanced by improving the baseline, increasing sensitivity [Limit of Quantitation (LOQ) of 12 mg/L and Limit of Detection (LOD) of 4.0 mg/L], and resolving analytes into well-defined Gaussian peaks when using the Dionex OnGuard II H⁺ cartridges to remove matrix interferences. High concentrations of nitrate limited the performance of the IC method. To achieve an acceptable baseline, samples with high nitrate content require more dilution, resulting in higher detection limits. Tests on samples with higher concentrations of nitrate (Tank 30 and 32 supernate with approximately 150000 mg/L nitrate) suggested that IC detection limits for glycolate in these samples would be > 500 mg/L. Thus, an LOQ of 12 mg/L is not feasible by this IC method on evaporator feed samples. An alternative method of glycolate analysis, using proton nuclear magnetic resonance (H NMR), was developed by the research team. In initial tests, the H NMR technique provided reasonable quantitation of glycolate in Tank 22 conditions by direct observation of the liquid. The H NMR method may provide improved detection limits for solutions with higher nitrate concentrations. We recommend further development of this analysis for high nitrate LWS samples such as evaporator feed and evaporator drop tank content. A method for pretreatment of samples using crystalline silicotitanate (CST) was developed and tested. The objective of the pretreatment was to facilitate analysis of samples with higher levels of radioactivity. The pretreatment did not influence the glycolate concentration in solution, and the pretreatment is projected to reduce Cs 137 activity in a sample by a factor of 16,200. The double strike CST pretreatment was demonstrated and would allow milliliters of higher activity samples to be transferred from the Shielded Cells and handled in a containment unit for glycolate analysis. The various studies validated IC and H NMR methods for glycolate analysis, defined the range of applicability, and demonstrated key supporting analytical protocols. Based on the results, high quality glycolate analysis of Tank 22 is feasible down to approximately 12 mg/L, with the potential for broader applicability of the methods to other conditions in the Savannah River Site Liquid Waste System (LWS).

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

Results from the Second MCU Flush DSSHT Sample

The Modular Caustic-Side Solvent Extraction Unit (MCU) stopped operations on May 22, 2019. As part of a plan to de-inventory the remainder of the liquids in the system, samples are sent to the Savannah River National Laboratory (SRNL) for analysis. A recent set of samples was pulled from the Decontaminated Salt Solution (DSS) Hold Tank (DSSHT) on 6/25/21, and arrived on the same day. The current DSSHT contents includes heel from the first flush of DSSHT (reference SRNL-L3100-2021-00019), deionized water (DIW) backflush of the DSS Coalescer, DIW flushes of the DSS Hydraulic Accumulator (DSSHA), DIW flush of the Contactor Drain Tank (CDT), and DSS Decanter drain material. The DIW flushes from the coalescer, DSSHA, and CDT were pumped through the DSS Decanter to allow for flushing of the DSS Decanter.

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

Modeling of Glycolate Destruction in the Recycle Collection Tank

The Savannah River Site’s DWPF is being upgraded with the introduction of the NG flowsheet. Glycolic acid has been shown to be a superior alternative to formic acid for sludge processing. The new flowsheet improves or maintains necessary parameters such as 1) reduction of mercury, 2) adjustment of feed rheology, 3) pH stability, and 4) adjustment of melter oxidation/reduction potential. Further, the use of glycolic acid virtually eliminates the potential for catalytic hydrogen generation in DWPF processing DWPF process condensates are collected and returned to the Savannah River Site (SRS) CSTF. The RCT collects off-gas condensate during chemical processing, vitrification, and other unit operations performed in DWPF and is the singular return vessel delivering recycle effluent back to CSTF. Each batch of recycle will have a small amount of glycolate from chemical processing and melter off-gas condensates. To avoid potential flammability issues due to thermolysis of glycolate in the CSTF, Savannah River National Laboratory (SRNL) provided to Savannah River Remediation (SRR) at their request a Task Technical and Quality Assurance Plan (TTQAP) to quantify and mitigate glycolate returns via DWPF’s recycle stream. The request included testing of a process to oxidize glycolate and other organic species that are responsible for hydrogen generation from thermolysis. Following that work SRR provided a Task Technical Request (TTR) that requested process modeling. In 2021 a TTQAP was issued to cover the modeling work. Modeling draws data from laboratory scale studies using chemical simulants and radioactive waste samples. Chemical kinetic modeling was performed to evaluate the feasibility of using sodium permanganate to destroy glycolate in the RCT. The results from the laboratory studies were summarized in a series of reports. Reference 9 is a report of lab scale processing of actual DWPF Slurry Mix Evaporate Condensate Tank (SMECT) and Offgas Condensate Tank (OGCT) samples in the SRNL Shielded Cells. Tests at caustic conditions demonstrated sodium permanganate was effective in converting glycolate to oxalate, and permanganate (Mn 7+ ) is reduced to manganate (Mn 6+ ) with no significant formation of carbon dioxide or carbonate. Equation (1) was found to best describe the observed reaction of glycolate with permanganate under nominal (60 to 145 mg/L in RCT) glycolate entrainment conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Characterization of SDU 6 Sump Water Samples

This report details results from characterization of the Saltstone Disposal Unit 6 (SDU 6) East Sump water samples. In March of 2022, Saltstone Radiation Protection Department (RPD) personnel reported SDU 6 East Sump beta/gamma radiation levels at ~ 33 dpm/mL (~ 15 pCi/mL) beta/gamma in excess of the water sampling limit of 16 dpm/mL (7.2 pCi/mL). Replicate samples obtained from the SDU 6 East Sump were sent to the Savannah River National Laboratory (SRNL) for analysis to include pH, limited anions, toxic metals and radionuclides. Subsequent sump samples associated with SDU 7 that has not yet been filled with any radioactive material were also analyzed at SRNL for both nitrate and nitrite.

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