Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “radionuclide”

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.

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

At least 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Measurement of the 70 Ge( d , n ) 71 As cross section from 2.1 MeV to 10 MeV

This work reports experimental 70 Ge(d,n) 71 As cross sections producing Arsenic-71 (t 1/2 = 65.3 h, 28% β + ), a potentially useful diagnostic radionuclide. Target stacks containing two nat Ge foils, a nat Ni monitor foil, and an Al degrader were irradiated with 5.5–10 MeV deuterons using an FN Tandem Van de Graaff accelerator. 71 As and 61 Cu, 56 Co, and 58 Co (from nat Ni(d,x)) were quantified in the foils by high purity germanium gamma spectrometry. Furthermore, the measured cross sections are larger than previously reported and suitable for producing preclinical quantities of diagnostic radiopharmaceuticals.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

48 Ca(p,pn) 47 Ca linac production for the preparation of a 47 Ca/ 47 Sc generator

Here, the production of 47 Sc via the nuclear reaction 48 Ca(p,pn) 47 Ca→(β-, T 1/2 = 4.54 d) 47 Sc and the assembly of a 47 Ca/ 47 Sc generator system has been investigated at Brookhaven National Laboratory. To produce 47 Ca, a pressed nat CaCl 2 target was irradiated with a proton beam at energy on target of 27.5 MeV at Brookhaven Linac Isotope Producer. Irradiated targets were dissolved in water and then acidified with concentrated HCl before separation using extraction chromatography through a diglycolamide (DGA) resin column. After ingrowth, 47 Sc was collected at >90 % with low contamination from both nonradioactive metals and coproduced radionuclides. A total of four 47 Ca/ 47 Sc separations were performed with 6–7 days between the irradiation and the separation to allow the equilibration of the ingrowing 47 Sc.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

The simultaneous removal of technetium and iodine from Hanford tank waste

The simultaneous removal of radionuclides technetium-99 and iodine-129 from an actual decontaminated Hanford tank waste sample (a mixture of decontaminated waste from tanks 241-AP-105 and 241-AP-107) was demonstrated for the first time in this work. A series of commercially available ion exchange resins were evaluated in batch contact tests in the tank waste, and all showed removal of both Tc and I. The highest Tc removal was observed for Purolite A530e while the highest iodine removal was observed for ResinTech SIR-110-MP. Batch tests in simulated tank waste with these two resins showed that the SIR-110-HP-MP had consistently higher K d for both pertechnetate and iodide and much higher K d than previous works on Tc removal from Hanford waste. As such, the SIR-110-MP was evaluated in a dual -column (lead/lag) test processing 5.2L of the tank waste mixture showing 60% breakthrough of Tc on the lead column and no significant breakthrough on the lag after 625 bed volumes (BV, 6 mL size) while significant iodine breakthrough (>50%) occurred after 28 BV. The limited iodine uptake was attributed to the column conditions generating mass transfer limitations. A fraction of the Tc and I was not captured by the resin (<10%) in either the batch tests or column tests. The iodine fraction was identified to be an iodide, likely organo-iodide. The fraction of the Tc was identified as a non-pertechnetate species, which is the first time non-pertechnetate has been identified in AP-105 and AP-107 tanks, although the exact species is still unknown.

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

Simulation of radiation damage via alpha decay in BFS:PC grouts using 4 He 2+ ion acceleration

The impact of alpha radiation on cements used to encapsulate intermediate-level waste (ILW) is not well understood. ILW wastes can contain high levels of alpha-emitting radionuclides, meaning that the grouts used to encapsulate them are exposed to significant ionising radiation. Thus, a damaged region could develop in the grout adjacent to the alpha-emitting species. This work attempted to recreate this behaviour through nonradioactive 4 He 2+ ion-accelerator experiments, which have not previously been applied to common encapsulation grouts. The influence of this irradiation on a slag-Portland cement was investigated at different ages via transmission electron microscopy energy-dispersive x-ray spectroscopy (TEM-EDX) and supporting techniques, to assess whether 4 He 2+ irradiation caused textural or chemical zonation. No significant changes in hydrate phases or textures were observed, other than minor variations associated with carbonation. This paper provides a proof of concept for using ion acceleration techniques on cements and furthers knowledge on their radiation response.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Plutonium mobilization from contaminated estuarine sediments, Esk Estuary (UK)

Since 1952, liquid radioactive effluent containing 238-242 Pu, 241 Am, 237 Np, 137 Cs, and 99 Tc has been released with authorization from the Sellafield nuclear complex (UK) into the Irish Sea. This represents the largest source of plutonium (Pu) discharged in all western Europe, with 276 kg having been released. In the Eastern Irish Sea, the majority of the transuranic activity has settled into an area of sediments (Mudpatch) located off the Cumbrian coast. Radionuclides from the Mudpatch have been re-dispersed via particulate transport in fine-grained estuarine and intertidal sediments to the North-East Irish Sea, including the intertidal saltmarsh located at the mouth of the Esk Estuary. Saltmarshes are highly dynamic systems which are vulnerable to external agents (sea level change, erosion, sediment supply, and freshwater inputs), and their stability remains uncertain under current sea level rise projections and possible increases in storm activity. In this work, we examined factors affecting Pu mobility in contaminated sediments collected from the Esk Estuary by conducting leaching experiments under both anoxic and oxic conditions. Here, leaching experiments were conducted over a 9-month period and were periodically sampled to determine solution phase Pu via multicollector-inductively coupled plasma-mass spectrometry (MC-ICP-MS), and to measure redox indicators (Eh, pH and extractable Fe(II)). Microbial community composition was also characterized in the sediments, and at the beginning and end of the anoxic/oxic experiments. Results show that: 1) Pu leaching is about three times greater in solutions leached under anoxic conditions compared to oxic conditions, 2) the sediment slurry microbial communities shift as conditions change from anoxic to oxic, 3) Pu leaching is enhanced in the shallow sediments (0–10 cm depth), and 4) the magnitude of Pu leached from sediments is not correlated with total Pu, indicating that the biogeochemistry of sediment-associated Pu is spatially heterogeneous. These findings provide constraints on the stability of redox sensitive Pu in biogeochemically dynamic/transient environments on a timescale of months and suggests that anoxic conditions can enhance Pu mobility in estuarine systems.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Separation of gold from irradiated actinide material utilizing a 2- thenoyltrifluoroacetone extraction resin

Post-detonation nuclear forensics capabilities depend on the ability to rapidly isolate radionuclides to improve measurement quality. In this work an extraction chromatography resin was developed utilizing the extractants thenoyltrifluoroacetone and 1-octanol supported on Eichrom prefilter resin. Here the resin was tested in nitric and hydrochloric acid matrices. In nitric acid matrices the resin is able to extract zirconium, while in hydrochloric acid matrices it is possible to extract iron and gallium. In all acid conditions tested gold is retained but can be removed from the column with a 10% thiourea rinse.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Development of a SnO 2 -based 44 Ti/ 44 Sc generator for medical applications

Towards application of 44 Sc for diagnostic nuclear medicine, a 44 Ti/ 44 Sc generator based on an inorganic resin has been evaluated. Unlike other radionuclide generators used for medical applications, the long-term retention of the parent 44 Ti is vital due to its long half life. In this work, tin dioxide (SnO 2 ), a robust inorganic-based resin, has been synthesized and used as the stationary phase for a 44 Ti/ 44 Sc generator. The sorption behavior of 44 Ti/ 44 Sc was tested on SnO 2 with varying acids, concentrations, and times. Preliminary batch study results showed >88 % 44 Ti retention to the resin at lower acid concentrations (0.05 M HNO 3 and 0.05 M HCl). A pilot generator was evaluated for a year, demonstrating 85.3 ± 2.8 % 44 Sc elution yields and 0.71 ± 0.14 % 44 Ti breakthrough in 5 M HNO 3 . Based on capacity studies, a 7.4 MBq (200 µCi) upscaled generator system was constructed for further evaluation of the SnO 2 resin stability and the efficacy of the eluted 44 Sc for radiolabeling. 44 Sc could be regularly eluted from this generator in 5 M HNO 3 with an overall average radiochemical yield 84.7 ± 9.5 %. Post-elution processing of the 44 Sc with DGA-normal resin removed all 44 Ti present and allowed for high 44 Sc-DOTA labeling yields of 94.2 ± 0.5 %. Overall, SnO 2 has been shown to be a viable material for a 44 Ti/ 44 Sc generator.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Machine learning-enabled prediction of chemical durability of A 2 B 2 O 7 pyrochlore and fluorite

Pyrochlore-structure type and its derivative in a general formula A 2 B 2 O 7 (A = rare earth elements and actinides; B = Ti, Sn, Zr, Hf, Pb, Si, etc.) display excellent structural flexibility and rich crystal chemistry as promising nuclear waste form materials capable of immobilizing actinides and fission products. It is essential to understand these materials’ chemical durability and element release of radionuclides in order to evaluate their performance in near-field environment. However, it is a formidable grand technological challenge to experimentally perform durability testing across hundreds of thousands of possibilities resulting from their extreme compositional complexities due to cation substitutions at both A and B-sites. In this work, we demonstrate a machine learning approach to determine the key materials parameters and structural characteristics governing the leaching behaviors from a small set of selected compositions as model systems, enabling a science-based prediction of their chemical durability that can be extended to a wide range of chemical compositions. The combination of four key structural characteristics and materials parameters, including ionic radius size difference , ionic potential difference , electronegativity difference , and lattice parameter , creates features an optimized prediction of the chemical durability. Two machine learning models, linear regression and Kernel ridge regression models, are trained on the randomly-split training dataset derived from the experimentally-determined elemental release rates, and subsequently tested on the testing dataset. The predicted leaching rates from both machine learning models show an excellent agreement with the experimental data, demonstrating the feasibility of rapidly evaluating the material properties of new compositions. These results highlight the immense potential of synergizing informatics through machine learning-based models and well-controlled experiments of selected model systems to accelerate materials design and discovery with optimized compositions and performance of promising materials for effective nuclear waste management.

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

Binding of uranyl cations to a Zr-based metal-organic framework by density functional theory

Zr-based metal–organic frameworks (Zr-MOFs) have been widely used as ion adsorbents for the removal or extraction of toxic and/or radionuclide species from aqueous solutions. However, the mechanisms by which uranyl cations (UO 2 2+ ) interact with Zr-MOFs have not been established. In this work, the nature of the bonding of uranyl cations with a Zr-MOF was determined using density functional theory for nineteen structurally distinct candidate complexes. Further, the results showed that in all cases the binding energy was of the order of 1.5 eV, but depended on the specific bonding site. The most stable structure involved coordination of the uranyl cation and two structurally distinct oxygens in the Zr-MOF metal node. It was also found that higher degree of deprotonation in Zr-MOF correlated with higher binding energy between the Zr-MOF and uranyl cations. These insights can aid in the design of Zr-MOFs with optimized features for efficient capture of uranyl cations.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Estimation of stress and stress-induced permeability change in a geological nuclear waste repository in a thermo-hydrologically coupled simulation

Geologic disposal is a promising solution for a safe permanent isolation of accumulated high-level nuclear waste from nuclear powerplants. The behavior of host rock is highly coupled thermally, hydromechanically, and chemically. Numerical simulations of such coupled phenomena for the extremely long term (>100,000 years) and large length scale (>1 km) of geologic disposal remain to be computationally challenging. In this study, a methodology has been developed to approximate stress and stress-induced permeability change in host rock using only thermo-hydrological (TH) variables. A coupled thermo-hydromechanical (THM) simulation is carried out using TOUGH-FLAC simulator to model THM behaviors of a generic nuclear waste repository, in order to evaluate the performance of the developed methodology, which is implemented in a coupled TH simulation. Results show that stress and permeability change estimated by the developed methodology in the TH-coupled simulation match those calculated in the THM-coupled simulation over the simulated timeframe of over 10,000 years. Details about errors in stress and permeability estimates accrued by the developed methodology are discussed here. The developed methodology will help incorporate stress-induced permeability change into existing TH simulators for the long-term radionuclide transport in geologic disposal.

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

Reduced-order modeling of near-field THMC coupled processes for nuclear waste repositories in shale

Performance assessment (PA) of geologic radioactive waste repositories requires three-dimensional simulation of highly nonlinear, thermo-hydro-mechanical-chemical (THMC), multiphase flow and transport processes across many kilometers and over tens to hundreds of thousands of years. Integrating the effects of a near-field geomechanical process (i.e. buffer swelling) into coupled THC simulations through reduced-order modeling, rather than through fully coupled geomechanics, can reduce the dimensionality of the problem and improve computational efficiency. In this study, PFLOTRAN simulations model a single waste package in a shale host rock repository, where re-saturation of a bentonite buffer causes the buffer to swell and exert stress on a highly fractured disturbed rock zone (DRZ). Three types of stress-dependent permeability functions (exponential, modified cubic, and Two-part Hooke’s law models) are implemented to describe mechanical characteristics of the system. Overall, our modeling study suggests that compressing fractures reduces DRZ permeability, which could influence the rate of radionuclide transport and exchange with corrosive species in host rock groundwater that could accelerate waste package degradation. Less permeable shale host rock delays buffer swelling, consequently retarding DRZ permeability reduction as well as chemical transport within the barrier system.

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

Public Health Response and Medical Management of Internal Contamination in Past Radiological or Nuclear Incidents: A Review

Following a radiological or nuclear emergency, workers, responders and the public may be internally contaminated with radionuclides. Screening, monitoring and assessing any internal contamination and providing necessary medical treatment, especially when a large number of individuals are involved, is challenging. Experience gained and lessons learned from the management of previous incidents would help to identify gaps in knowledge and capabilities on preparedness for and response to radiation emergencies. In this paper, eight largescale and five workplace radiological and nuclear incidents are reviewed cross 14 technical areas, under the broader topics of emergency preparedness, emergency response and recovery processes. The review findings suggest that 1) new strategies, algorithms and technologies are explored for rapid screening of large populations; 2) exposure assessment and dose estimation in emergency response and dose reconstruction in recovery process are supported by complementary sources of information, including ‘citizen science’; 3) surge capacity for monitoring and dose assessment is coordinated through national and international laboratory networks; 4) evidence-based guidelines for medical management and follow-up of internal contamination are urgently needed; 5) mechanisms for international and regional access to medical countermeasures are investigated and implemented; 6) long-term health and medical follow up programs are designed and justified; and 7) capabilities and capacity developed for emergency response are sustained through adequate resource allocation, routine nonemergency use of technical skills in regular exercises, training, and continuous improvement.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Decontamination of urban surfaces contaminated with radioactive materials and consequent onsite recycling of the waste water

Enhancing rapid remediation strategies is paramount for recovery after a large-scale nuclear contamination event in an urban environment. Some current strategies recommend use of readily available equipment, materials, and facilities to expedite recovery. For example, applying pressurized water to contaminated surfaces may effectively remove radioactive contamination. In this study, a commercial power washer removes soluble forms of 152 Eu 3 + , 85 Sr 2 + , and 137 Cs + contamination from common porous building materials, and computer simulations characterize the recycling of the resultant contaminated wash water. Pressure washing the porous building materials under spray conditions typical with do-it-yourself units improved decontamination factors (DFs) for 152 Eu compared to low-pressure application of tap water (majority of two-tailed t-test p-values < 0.1), but pressure did not improve DFs for 137 Cs or 85 Sr. For both pressurized and low-pressure applications, adding potassium ions (K + ) to promote ion exchange reactions produced significantly higher DFs for tested radionuclides on asphalt, brick, and concrete. The resultant contaminated wash water can be processed through self-prepared chemical filtration beds of clay and sand. Modeled in a prior study, the beds yielded linear trends (R 2 > 0.98) in sensitivity analyses between most bed configuration variables and bed performance variables, permitting flexible ad-hoc bed design. The experimental and simulation results led to estimates of the remediation rate and waste generated after cleaning 250 m 2 of cesium-contaminated concrete from the combined deployment of a power washer and two different mobile treatment beds. Furthermore, the first treatment bed was designed to reduce treatment time and processed 1900 L of wash solution in 70 min using 880 kg of clay/sand infill material. Designed to reduce the solid waste generated, the second bed processed the same solution volume in 1040 min (17 h) using 170 kg of clay/sand infill material. The results of this analysis warrant further investigation of power washing with recycled salt solution as an effective rapid decontamination method with manageable waste.

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

IUPAC-IUGS recommendation on the half-lives of 147 Sm and 146 Sm

In this paper, the IUPAC-IUGS joint Task Group “Isotopes in Geosciences” recommends a value of (106.25 ± 0.38) Ga for the half-life of 147 Sm, and a corresponding decay constant λ 147 = (6.524 ± 0.024) × 10 –12 a –1 , both with a coverage factor k = 2. For the extinct radionuclide 146 Sm two very different half-lives are used in the scientific community ( c . 68 and 103 Ma), to such a degree that no consensus value can be endorsed at present by the Task Group. Pending dedicated re-investigations it is recommended that papers using the 146 Sm decay to quantify the cosmo/geological evolution of (extra)terrestrial samples perform a twin set of calculations using both proposed half-lives.

146Sm↗
DOE OSTI2021

On the dissolution of a borosilicate glass with the use of isotopic tracing – Insights into the mechanism for the long-term dissolution rate

An understanding of the mechanisms responsible for controlling the long-term corrosion rate of nuclear waste glass is paramount if reliable glass dissolution models are to be used to calculate the controlled release of radionuclides from nuclear waste glass under geologic disposal conditions. Additionally, understanding silicate glass dissolution is also important for natural glasses to understand the role from the dissolution of these glasses has on the composition of natural aquatic systems. Two general mechanisms appear to be responsible for the elemental release from glass to the surrounding biosphere - ion exchange (release principally of alkalis) and matrix dissolution (release of the structural components of the glass). Key unknowns related to these mechanisms are the impact of surface layers on the altered glass and how these layers participate in corrosion. To better understand the impact of these layers on glass corrosion, we made two glasses of the same elemental composition, but with selected elements enriched in specific isotopes, e.g. more 10 B than 11 B. We reacted these glasses for 1 y then the solutions from each were exchanged and the chemistry of the solutions and the solids were followed for another 2 y and in some cases 3 y. This allowed us to follow changes in the glasses with respect to these elements and how these elements were transported through the alteration layers with time and, for at least one element, into the glass. Results from these experiments demonstrate that the release behavior of different elements is strongly dependent on their structural role in the glass (e.g. network formers and modifiers), but, more importantly, the role of water transport and subsequent ion exchange has in the long-term dissolution of the glass studied here. In this article, we highlight the behavior of four elements: lithium, sodium, silicon, and boron. Lithium and sodium, network modifying elements, have similar chemistries. The behavior of Li could be tracked in more detail by following both 6 Li and 7 Li; Li penetrates through the gel layer in both directions without hinderance and into and released from the glass deeper than sodium. Silicon, a network forming element, reacts with the silica-rich alteration layer. Boron, a network forming element, does not accumulate in the gel or the pristine glass and has a very sharp elemental profile between the pristine glass and alteration layer. Boron, lithium, and sodium elemental profiles suggest that there is little if any transport control within the alteration layer. The correlation between the Li and Na profiles and that of a water species suggest that the limiting release is a steady state between the diffusion of a water species and matrix dissolution that results in a low release rate for alkali and this provides a steady driving force for the long-term dissolution of glass.

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

Synthesis of results for Brine Availability Test in Salt (BATS) DECOVALEX-2023 Task E

In the 2023 phase of the international collaborative DECOVALEX modeling project, Task E focused on understanding thermal, hydrological, and mechanical (THM) processes related to predicting brine migration in the excavation damaged zone around a heated excavation in salt. Salt is attractive as a disposal medium for radioactive waste because it is self-healing and is essentially impermeable and non-porous in the far field. Investigation of the short-term, near-field behavior is important for radioactive waste disposal because this early period strongly controls the amount of inflowing brine. Brine leads to corrosion of waste forms and waste packages, and possible dissolution of radionuclides with brine transport being a potential transport vector to the accessible environment.The Task was divided into steps. Step 0 included matching unheated brine inflow data from boreholes at the Waste Isolation Pilot Plant (WIPP) and matching temperature observations during a Brine Availability Test in Salt (BATS) heater test. Step 1 included validation of models against a thermo-poroelastic analytical solution, and two-phase flow around an excavation. Finally, Step 2 required all the individual components covered in steps 0 and 1 to come together to match observed brine inflow behavior during the same BATS heater test.There were a range of approaches from the teams, from mechanistic to prescriptive. Given the uncertainties in the problem, some teams used one- or two-dimensional models of the processes, while other teams included more geometrical complexity in three-dimensional models. Task E was a learning experience for the teams involved, and feedback from the modeling teams has led to changes in follow-on BATS experiments at WIPP. The primary Task E lessons learned were the impact of hydrologic initialization methods (wetting up vs. drying down), the difference between confined and unconfined thermal expansion, and the large changes in permeability associated with heating and cooling.

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

Resilient design in nuclear energy: Critical lessons from a cross-disciplinary analysis of the Fukushima Dai-ichi nuclear accident

This paper presents a multidisciplinary analysis of the Fukushima Dai-ichi Nuclear Power Plant accident. Along with the latest observations and simulation studies, we synthesize the time-series and event progressions during the accident across multiple disciplines, including in-plant physics and engineering systems, operators’ actions, emergency responses, meteorology, radionuclide release and transport, land contamination, and health impacts. We identify three key factors that exacerbated the consequences of the accident: (1) the failure of Unit 2 containment venting, (2) the insufficient integration of radiation measurements and meteorology data in the evacuation strategy, and (3) the limited risk assessment and emergency preparedness. We conclude with new research and development directions to improve the resilience of nuclear energy systems and communities, including (1) meteorology-informed proactive venting, (2) machine learning-enabled adaptive evacuation zones, and (3) comprehensive risk-informed emergency planning while leveraging the experience from responses to other disasters.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2020

Evaluation of Gaseous Substrates for Microbial Immobilization of Contaminant Mixtures in Unsaturated Subsurface Sediments

Extensive vadose zone metals and organic contamination remains at many former industrial and defense manufacturing sites, and effective remedial solutions are needed to slow or prevent its migration to groundwater. In this study, the application of gaseous substrates to stimulate microbial respiratory reduction of comingled radioisotopes and nitrate under unsaturated conditions was examined for possible application at the Hanford Site, a former nuclear production facility in southeastern WA, USA. First, screening studies were performed to qualitatively measure the sediment respiratory response to 14 gaseous or volatile organic substrates at two moisture contents, 4% and 8%. Volatile substrates produced the strongest respiratory response, among them were butyrate, pentane, butyl acetate. Ethane and butane were the most effective gaseous substrates but only at 8% water content. Hanford sediment from two waste sites with distinctive chemistries were wetted to 7% moisture content, packed into columns, and treated with ethane or butane. After 4 weeks, columns were then leached to quantify retardation in the mobility of aqueous contaminant concentrations compared to no gas control columns. Treatment with both gases resulted in >80% removal of Cr from the aqueous phase. However, NO3 concentration and a waste sites exposure history to NO3 had a major effect on U and Tc reduction. Incomplete nitrate reduction outcompeted U and Tc in waste site sediments having limited prior exposure to NO3. Conversely, waste site sediments co-contaminated with NO3 were able to achieve highly reduced conditions resulting in complete denitrification of NO3, and delayed leaching of U and Tc. This implied effective reduction of both contaminants to less mobile species. This study demonstrates that unsaturated vadose sediments at Hanford waste sites have the capacity for a sustained respiratory response to gaseous substrate injection, which could potentially be deployed as part of an overall strategy to reduce the flux of long-lived radionuclides to groundwater at Hanford and other legacy waste sites.

Bagwell, Christopher E.↗