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At least 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Temperature-dependent microbial reactions by indigenous microbes in bentonite under Fe(III)- and sulfate-reducing conditions

Bentonite is a promising buffer material for constructing spent nuclear fuel (SNF) repositories. However, indigenous microbes in bentonite can be introduced to the repository and subsequent sealing of the repository develops anoxic conditions over time which may stimulate fermentation and anaerobic respiration, possibly affecting bentonite structure and SNF repository stability. Moreover, the microbial activity in the bentonite can be impacted by the heat generated from radionuclides decay. Therefore, to investigate the temperature effect on microbial activities in bentonite, we created microcosms with WRK bentonil (a commercial bentonite) using lactate as the electron donor, and sulfate and/or ferrihydrite (Fe(III)) as electron acceptors with incubation at 18 °C and 50 °C. Indigenous WRK microbes reduced sulfate and Fe(III) at both temperatures but with different rates and extents. Lactate was metabolized to acetate at both temperatures, but only to propionate at 18 °C during early-stage microbial fermentation. More Fe(III)-reduction at 18 °C but more sulfate-reduction at 50 °C was observed. Thermophilic and/or metabolically flexible microbes were involved in both fermentation and Fe(III)/ sulfate reduction. In conclusion, our findings illustrate the necessity of considering the influence of temperature on microbial activities when employing bentonite as an engineered buffer material in construction of SNF repository barriers.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2021

Micrometer-sized Magnetite Synthesis using Fe(OH)2(s) as a Precursor for Technetium Sequestration from Liquid Nuclear Waste Streams

Systematic batch experiments under variable adjusted physicochemical conditions were conducted to explore optimization of micrometer-sized magnetite synthesis for Tc sequestration from radionuclide waste streams using Fe(OH)2(s) as the precursor. Extensive solid characterization using x-ray diffraction and spectroscopic methods was performed to assess changes in particle morphology and size distribution, as well as Tc speciation and incorporation, in the produced mineral phases. The results show that the solution pH, temperature, and oxidation kinetics play key roles in the final mineral products. Micrometer-sized magnetite crystals (0.62-0.96 µm on average) with well-defined dodecahedral or octahedral structures were synthesized under near neutral (~pH 8) or alkaline (~pH13) conditions at 75 °C, respectively; whereas goethite dominated the end products at room temperature. An increase in pH at 75 °C improved Tc removal from 27% (near neutral pH) to 42% (alkaline pH), but the removal process remained inhibited by redox competitive Cr(VI) present in the waste streams. By adding additional Fe(II) to the system, Tc sequestration was dramatically improved to up to 87% without observable changes in the solid product. The sequestrated Tc existed as TcO2·2H2O and/or Tc(IV) incorporated into magnetite, where extended X-ray absorption fine structure (EXAFS) spectroscopy showed that more Tc was incorporated into magnetite at elevated temperatures and pH conditions, with complete Tc(IV) incorporation into magnetite occurring under 75 °C-pH 13 conditions. Our results indicate that optimal micrometer-sized magnetite can be produced for Tc sequestration by reacting Fe(OH)2(s) with a waste stream simulant under elevated pH (~13) and temperature (75 °C) conditions. The incorporation of reduced Tc(IV) into stable micrometer-sized magnetite provides a viable supplemental immobilizing technology that may be used to improve nuclear waste treatment and disposal needs.

Wang, Guohui↗
DOE OSTI2024

The effect of cerium, neodymium, and ytterbium doping on UO 2 dissolution

Here, the dissolution rate of spent nuclear fuel has been studied heavily to understand the impacts of a failed waste package scenario on potential radionuclide release from a geologic repository. Multiple countries have evaluated these scenarios in oxidizing and reducing environments respective to their relevant repository conditions. The repository environment (e.g., water chemistry, temperature, oxygen content) and the fuel itself (e.g., chemical content of the fuel) both heavily impact the dissolution rate of the spent fuel. This work examined the impact that rare earth element dopants (Ce, Nd, Yb) have on the fuel dissolution under repository relevant conditions with decreasing oxidizing conditions using a single pass flowthrough system. UO 2 samples were doped with Ce, Nd, or Yb with concentrations between 1 and 5 at%. The addition of dopants to the samples reduced the dissolution rates on most samples relative to pure UO 2 samples. Scoping experiments that occurred in a less oxidizing environment showed a reduction in dissolution rate compared to fully oxidizing conditions. The results within highlight the importance of dopant behavior in used fuel dissolution modeling.

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

Glass-bonded ceramic waste forms for immobilization of radioiodine from caustic scrubber wastes

Glass-bonded sodalite composite waste forms have been developed for the immobilization of liquid radioactive wastes resulting from off-gas treatment during aqueous reprocessing of used nuclear fuel, with a particular focus on 129I. The proposed composite waste form is comprised of aluminosilicate ceramic phases containing volatile radionuclides bonded with a glassy matrix. In this work, a suite of ten candidate low-temperature glass binders (ZnO-Bi2O3-based glasses and a Na2O-B2O3-SiO2 glass) were examined. Six glasses were mixed with caustic scrubber waste simulant previously converted into a sodalite-rich material (to provide glass fractions of 10 and 20 wt.%), uniaxially pressed into pellets, and sintered at 350 °C or 550 °C for 8 h in air. Iodine retention after heat treatment was assessed by neutron activation analysis, showing retention of 67-100 % of expected iodine. The aqueous durabilities of the resulting materials were then determined, following the ASTM C1308 standard test, showing iodine releases of 1 to 23 g m-2 after 4 d. The cumulative iodine release for the best performing system (a zinc-bismuth-borate glass binder) was <1 g m-2, and its iodine retention from processing was 67 %. The iodine releases compared favorably with other waste forms. In parallel, this best-performing composition was also consolidated via hot isostatic pressing (HIP) in a stainless-steel canister at 550 °C for 2 h under 100 MPa pressure. The HIPed sample was produced at the ~20 g scale and showed improved densification and minimal reaction with the canister.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Rehydration of metastudtite in the alteration kinetics of α– and β–U 3 O 8 in dilute aqueous solutions of hydrogen peroxide

The formation of alteration phases on uranium ore concentrates and used nuclear fuels under oxidizing conditions is key to understanding the potential mobility of radionuclides in the environment and designing optimal storage conditions of materials. However, the time-dependent distribution of alteration phases on α– and β–U 3 O 8 under oxidizing conditions has yet to be explored. Here, in this study, crystalline powders of α– and β–U 3 O 8 were submerged in aqueous solutions of hydrogen peroxide (1.6 × 10 −1 to 5.4 × 10 −5 M) with aliquots of solution and solid removed for analysis at 1, 8, 15, 22, 29, 36, 46, 58, 71, and 83 days. Within one day there is significant alteration of U 3 O 8 to the uranyl peroxide metastudtite, [(UO 2 )(O 2 )(H 2 O) 2 ], that is replaced by studtite, [(UO 2 )(O 2 )(H 2 O) 2 ]·2H 2 O, within a week regardless of the polymorph of U 3 O 8 or the initial concentration of H 2 O 2 in solution, as determined by partial least squares regression (PLSR) of Raman spectra collected from the solids. A dissolution/reprecipitation mechanism is proposed for both the alteration of U 3 O 8 to metastudtite and the subsequent alteration of both U 3 O 8 and metastudtite to studtite. The two polymorphs of U 3 O 8 exhibit similar rates and extents of alteration over time. The rehydration of metastudtite to studtite has not been previously reported and highlights the need for future work to determine the mechanism by which metastudtite is converted to studtite and what other conditions facilitate this rehydration.

Alteration of U3O8↗
DOE OSTI2025

Evaluation of AGR-3/4 In-pile Silver Release Predictions Against Post-Irradiation Examination Measurements

Fuel performance modeling codes that accurately predict the transport of radionuclides in high-temperature gas-cooled reactors that utilize tristructural isotopic (TRISO) fuel particles are an important aspect of reactor safety analyses. One objective of the Advanced Gas Reactor (AGR)-3/4 experiment was to assess the transport of fission products through fuel particles and their subsequent release into the compact matrix and structural graphite materials. This was accomplished by irradiating uranium oxycarbide (UCO) driver fuel particles and designed-to-fail (DTF) particles to serve as known sources of fission products. The fission product of particular interest when it comes to such transport is silver (Ag-110 m), as it has a 250-day half-life and has relatively high mobility in the TRISO coating layers. Furthermore, to assess the current modeling capabilities and diffusion parameters employed in the fuel performance codes PARFUME and BISON, the fractional release of silver release predicted by the two codes were compared against post-irradiation examination measurements from the AGR-3/4 experiment.

AGR-3/4 Experiment↗
DOE OSTI2021

Nuclear Data Sheets for A=64

The evaluated experimental data are presented and evaluated for 13 known nuclides of mass 64 (Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Ga, Ge, As, Se). For each nuclide, the best values combining all available data are recommended for spectroscopic properties. No excited states have been identified in 64 Ti, 64 As, and 64 Se. Only one excited state in 64 V as an isomer, and three in 64 Cr have been identified. Data for excited states in 64 Mn, 64 Fe, and 64 Co remain limited. 64 Ni, 64 Cu, and 64 Zn are the most studied nuclides through various reactions and decays, followed by 64 Ga and 64 Ge. In the opinion of the evaluators, there are several incomplete or discrepant aspects of the high-spin portion of the level scheme for 64 Zn above ≈5 MeV excitation which need to be resolved in further experiments using large γ-detector arrays. The decay schemes of 64 Ti β - , 64 As ε and 64 Se ε are unknown, while very little information is available for 64 V β - and 64 Cr β - . The decay schemes of 64 Mn β - , 64 Fe β - , 64 Co β - and 64Ge ε are somewhat better known, but still considered incomplete by evaluators. The decay scheme of 64 Ga ε decay is known in detail, however there is the possibility of additional levels populated above 4713 keV, as the Q(ε) value is 7171 keV. The β - , β + and ε decay modes of 64 Cu are well known, as this radionuclide is of great importance in applications, for example, as a dosimeter for neutron flux determination in nuclear reactors, and as a radiopharmaceutical for positron emission tomography (PET). This work supersedes earlier full evaluations of A=64 by 2007Si04, 1996Si12, 1991Si03, 1979Ha35 and 1974Au04.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Branching ratio and γ -ray emission probabilities in the decay of the J π = 13 / 2 + isomer in 197Hg

Gamma rays following the decay of the 197 g Hg and 197 m Hg radionuclides were measured using chemically-purified sources and a planar HPGe detector. In the case of 197 m Hg, precise values of 94.68 (9)% and 5.32 (9)% were confirmed for the IT and EC branches, respectively, the latter discovered to differ significantly from previously published data. The half-lives of 197 g Hg and 197 m Hg were also measured as 64.81 (24) h and 23.82 (4) h.

197𝑔Hg↗
DOE OSTI2021

Calculations versus measurements for residual dose rates from SNS spent structures

Here, residual dose rate measurements were conducted on target vessel #13 and proton beam window #5 after extraction from their service locations. These measurements are used to verify calculation methods of radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation of these structures. Neutronics analyses for predicting residual dose rates are carried out using the transport code MCNPX and the transmutation code CINDER90. For transport analyses a complex and rigorous geometry model of the structures and their surroundings are applied. The neutronics analyses are carried out using the Bertini and CEM high energy physics models for simulating particles interactions above the table-based cross section range. Finally, obtained calculational results are analyzed and compared to the measured dose rates and overall show good agreement within 25%, which shows applicability of the methods used in analyses.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

A safeguards approach for molten salt reactors: Direct gamma-spectrometry of 233 U

Multidimensional gamma-spectrometry is a potential solution for safeguarding future thorium-fueled molten salt reactors (MSRs). A high-sensitivity system developed at Pacific Northwest National Laboratory (PNNL), the Advanced Radionuclide Gamma spectrOmeter (ARGO), has demonstrated direct 233 U assay in freshly irradiated fuel salt samples using a dual-detector configuration and the 97.1 keV gamma-emission. These measurements were made using 1.4 x 10 6 Bq of 233 U activity (at 1.4 x 10 8 fissions), at relatively high dead times (10.4%), and using short count durations of ~800 s. Other major emissions were also evaluated (42.5 keV, 54.7 keV, 146.4 keV), in additional to limited (1.96%) Compton suppression, but provided less consistent determination of the 233 U activity. Accurate 233 U measurement was also demonstrated using its four coincidence gamma-emissions (42.5 keV, 54.7 keV; 42.5 keV, 278.1 keV; 146.4 keV, 174.2 keV; 146.4 keV, 219 keV). Further improvements in coincidence measurements may be achievable with improvements in the detector configuration and nuclear data. Whilst this demonstration utilized the ARGO laboratory system, a similar dual-detector configuration could be utilized for online safeguards measurements.

Gamma-spectrometry↗
DOE OSTI2023

Development and performance evaluation of a novel γ-γ coincidence analysis software

Processing of large amounts of data from a γ-γ coincidence system is a practical challenge in trace-level radionuclide measurements. Given the benefits and growing number of multidimensional gamma spectrometers worldwide, this is a major restriction in wide-spread applicability of coincidence techniques. This study demonstrates a novel γ-γ coincidence analysis software developed at Pacific Northwest National Laboratory, USA. The software utilizes advanced search techniques and Python libraries to readily identify coincidence signatures in the large data. Further, using both synthetic and experimental data, the study uncovers various strategies/libraries employed during software’s development. This work aims to provide a reference document in handling large data routinely encountered in trace-level measurements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Durability test of a flowing-water target for isotope harvesting

A high intensity proton irradiation was performed with the flowing-water isotope harvesting target at the University of Wisconsin-Madison Cyclotron Laboratory to measure the rate of degradation of the target shell during irradiation conditions. The beam reached an intensity of 34 µA by the end of the irradiation and covered an area of 0.7 cm 2 on the target. Radiolysis products, such as H 2 O 2 , H 2 , and O 2 , were measured in the bulk water of the system and found to be present at much lower levels than predicted by literature escape yields. Radionuclides formed in the target shell were measured in the system water as a radiotracer for target degradation. Using a simple, beam intensity dependent model, a corrosion rate of 1.5E-6 μm/(μA*s) was found to match the measured radiotracer activities at various points in the irradiation. Furthermore, this rate was used to extrapolate the lifetime of future isotope harvesting targets at the NSCL and FRIB, using the areal power density of different ion beams to scale the corrosion rate.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

10 Be at iThemba LABS using a silicon nitride membrane stack as absorber for isobar suppression

10 Be is an important isotope for accelerator mass spectrometry (AMS) because of the demand for cosmogenic radionuclide dating methods in the earth science and paleo-sciences community. At the iThemba Laboratory for Accelerator Based Science (iThemba LABS) we implemented full suppression of the interfering isobar 10 B using a silicon nitride foil-stack, utilizing the 2+ charge state for high efficiency. We demonstrate the performance of this newly established AMS system using standards and test samples. Here we further present the results of an inter-comparison between iThemba LABS and the Center for Accelerator Mass Spectrometry/Lawrence Livermore National Laboratory, on AMS samples prepared from purified quartz at the University of Vermont. The results for 10 Be from the laboratories are in close agreement, fully consistent with cross-calibration between them. AMS results for 26 Al are in similarly good agreement, demonstrating the performance and accuracy of iThemba LABS for the most commonly measured in situ produced cosmogenic nuclides.

10-Be↗
DOE OSTI2024

FLUKA simulation validation to measure the beam energy of a cyclotron

FLUKA is a fully integrated particle physics MonteCarlo simulation package which has many applications in high energy experimental physics and engineering. When using a cyclotron for medical radionuclide production, it is important to know that the beam energy is suitable to cause the intended nuclear reaction and minimize unwanted impurities. Here, this work focuses on the comparison of FLUKA as a tool to measure the beam energy of a cyclotron with experimental results using the method of stacked copper foils interspaced by an aluminum degrader. We compared two methods to determine the activity ratio: determining the activity in each foil using the experimentally measured cross-sections from the National Nuclear Data Center (NNDC) cross-section data and calculating the activity in the two Cu foils using the FLUKA program. These results were compared to those reported by Gagnon et al., 2011. The FLUKA results gave energy measurements that were different than the true energy by a percentage in the range of 2.4% to 13%, suggesting that FLUKA is not recommended for pre-calculating the cyclotron beam energy using these low Z materials without experimental confirmation. It also confirms that the experimental cross-section data from the NNDC database is superior for this application giving energy values that ranged from 0.1% to 1.6% from the true energy.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Analysis of fission product distribution and composition in the TRISO layers of AGR-2 fuel

The Advanced Gas Reactor Fuel Qualification and Development Program's second irradiation experiment focuses on the performance of fuel particles produced in an industrial-scale coating system. Ongoing post-irradiation examination efforts include exploration of tristructural-isotropic (TRISO)-coated particle fuel in the as-irradiated state and subjected to elevated temperatures to simulate accident conditions (safety testing). Scanning electron microscopy (SEM) analysis with energy dispersive x-ray spectroscopy (EDS) was used to show the distribution of fission products within the TRISO layers. The SEM analysis showed a variation of fission product distribution within the TRISO layers based on irradiation temperature, burnup, individual particle fission product inventory retention, safety testing history and kernel composition. Finally, a key observation is that particles which likely experienced higher temperature showed a greater diversity and distribution of fission products and actinides in the IPyC and SiC layers, implying that temperature is a dominant variable influencing the transport of radionuclides in TRISO particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Burnable absorbers in nuclear reactors – A review

Burnable absorbers can benefit nuclear reactors of virtually any design by providing reactivity control for extended fuel cycles, tritium production, burning of long-lived radionuclides, and reactor safety. When selecting the ideal burnable absorber type and its design, one must consider the resulting impact on the reactor’s fuel cycle design and cost, reactivity, thermal hydraulics, manufacturing, and radiation response. These selection criteria, as well as neutronic and thermophysical material property requirements, may be vastly different depending on whether the burnable absorber is intended for use in a commercial water-cooled reactor, a research reactor, or a next-generation advanced reactor system. A recent integration, synthesis, description of past and present technologies, and identification of existing gaps and areas of future research is lacking on these important topics. Here, this paper includes a fundamental description of the use of burnable absorbers and their impacts on reactivity, absorber depletion, self-shielding, basic thermophysical properties, and the use of burnable absorbers in next-generation nuclear applications..

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗