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All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Mechanical Property Measurements of HFIR-irradiated FM and ODS Alloys

This report provides a summary of the post-irradiation examination of the tensile properties of three ODS ferritic alloys, 14YWT-SM13, 9YWTV-PM2 and OFRAC, and two HT9 ferritic/martensitic steels, one with low N content and the other with high N content. The neutron irradiations were conducted in the flux trap of the High Flux Isotope Reactor as part of the Additive Manufacturing, Oxide Dispersion Strengthened Alloys and Wrought Steels (AMOW) irradiation campaign. Sub-sized tensile specimens fabricated from each material were exposed to nominal target temperatures and neutron doses of 300°C, 385°C, and 525°C and 8 dpa, 16 dpa, and 50 dpa, respectively.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Morphology of Vapor-Deposited Acetonitrile Films

Crystalline acetonitrile has two polymorphs, a high temperature (HT) phase that is stable between 217 K and its melting point at 229 K, and a low temperature (LT) phase that is stable below 217 K. Solid acetonitrile films can be prepared by vapor deposition in an ultra-high vacuum chamber. To prevent sublimation of the film temperatures are often kept below 150 K. While the LT phase is thermodynamically favored at these low temperatures, such preparation usually results in the formation of the metastable HT polymorph. In this work we use reflection adsorption infrared spectroscopy (RAIRS) and temperature programmed desorption (TPD) experiments to investigate the effects of the deposition temperature and underlying substrate on the morphology of acetonitrile films prepared with molecular beam deposition. We obtained the elusive LT phase when dosing at 120 K on a graphene substrate and on a crystalline decane layer. Dosing acetonitrile on other surfaces produced the HT phase, as did annealing of amorphous films. We used TPD experiments to determine the Gibbs energy difference between the HT and LT phases. Our ΔG values agree with extrapolation of equilibrium calorimetry data. Furthermore, we also observed that acetonitrile films were amorphous when dosed at temperatures ≤ 60 K, and porous for temperatures ≤ 50 K.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Biomethanation to Upgrade Biogas to Pipeline Grade Methane

NREL is working closely with DOE, Electrochaea GmbH, and Southern California Gas Company (SoCalGas) to reduce costs of a biomethanation process capable of megawatt-scale deployment that upgrades organic biogas waste streams to produce pipeline quality renewable natural gas (RNG). Biomethanation is a two-step process using a single-celled methanogenic archaea that converts low-carbon low-cost hydrogen (H2) and waste carbon dioxide (CO2) to produce renewable methane (CH4). The process upgrades the biogenic CO2 - while allowing the CH4 to pass through - from biogas sources like dairies, wastewater treatment plants, and landfills. The CH4 produced is a drop-in direct replacement fuel and producers can participate in the growing number of carbon markets; like California's Low Carbon Fuel Standard and the Federal Renewable Fuel Standard. NREL and Argonne National Laboratory have completed a life cycle analysis using the GREET model to show that the biomethanation process produces RNG that is carbon negative even when H2 production via low-temperature water electrolysis is driven by the existing carbon intensity of California's electricity grid. And of course, even further carbon negative (-233 kg CO2e/kWh) when the electricity is produced from low-carbon sources like wind and solar. Leveraging lessons learned from operating SoCalGas' 700L 18-bar bioreactor system, NREL is designing and building a flexible RD&D platform that will enable field trials at biogas and other CO2 sources. A custom 16' long trailer will house a 20L 18-bar bioreactor, 3 - 25 kW proton exchange membrane electrolyzer, and dosing, thermal, and controls systems to support operations with only power, biogas, and water feedstocks required by the field locations. The end-of-project goal is to demonstrate pipeline quality RNG production (> 95% CH4, < 4% H2, <1% CO2, < 0.2% O2 and < 4 parts per million H2 sulfide) using real biogas feedstocks - thereby recycling both greenhouse gases for injection into the natural gas network or to be used onsite.

biogas upgrading↗
DOE OSTI2020

A quantitative study of retention and release of deuterium and tritium during irradiation of y-LiAlO2 pellets

This study reports on the quantification of deuterium in ion-irradiated gamma-LiAlO2 pellets as a function of dose and temperature. The gamma-LiAlO2 pellets were sequentially irradiated with He+ and D2+ ions to the same fluences of 5E16, 1E17 and 2E17 He++D+/cm2 at 188 K. Additional irradiation was performed to 1E17, 2E17 and 4E17 He++D+/cm2 at 573 K. A set of the pellets irradiated at 188 K was shipped and stored at low temperatures from 80 to 132 K and characterized using time-of-flight secondary ion mass spectrometry at ~173 K. The deuterium depth profiles show a Gaussian-like distribution in the low-temperature pellets. The total deuterium retention is found to be directly proportional to the ion fluence. About 27 at.% of the implanted deuterium atoms were released from the pellet irradiated to 2E17 He++D+/cm2 at 188 K during storage at room temperature for ~1 month. Retention of the trapped or bound deuterium during ion irradiation at 573 K increases initially with ion fluence and tends to saturate at a high fluence. The amount of the released deuterium is observed to be quantitatively consistent with that of the released tritium from similar standard pellets during neutron irradiation at 573 K.

Deuterium retention and release, ion irradiation, ↗
DOE OSTI2023

Health Hazards of Exposures to Radioiodine

Iodine is a chemical element with atomic number 53. Iodine-127 is stable (non-radioactive) and commonly found in nature. Elemental iodine is a purple-colored solid at room temperature and pressure, but spontaneously sublimates (turns into vapor). Iodine is an essential element for life, and is required for proper functioning of the thyroid. Iodine is present in many foods, and is readily absorbed by the body and concentrated in the thyroid gland. A fraction of iodine ingested or inhaled is rapidly removed by the kidneys. The rest of the inhaled or ingested iodine is absorbed the by thyroid and retained for many months. Iodine has a biological half-life of approximately 120 days in health individuals. The biological half-life can be shorter in individuals with hyperthyroidism, and longer in individuals with hypothyroidism. Iodine has a number of radioactive isotopes, most of which have relatively short half-lives (days or weeks). Short half-life iodine isotopes are useful for a variety of medical applications, including imaging and cancer therapy. For example, Iodine-123 (half-life 13 hours) is commonly used for medical imaging of the thyroid, while iodine-131 (half-life 8 days) is used for suppressing thyroid function in individuals with hyperthyroidism or ablating (killing) thyroid cells to treat thyroid cancer. Iodine-125 (half-life 59 days) is produced in nuclear reactors, and has medical uses. Although iodine-125 can be used for thyroid imaging, Iodine-123 is more commonly used for that purpose because of its shorter half-life and higher-energy emissions. Iodine-125 is more commonly used for cancer treatment, and can be processed into small metal pellets (seeds) inserted directly into a tumor. Iodine-125 emits low-energy x-rays which can kill tumor cells and generally cannot escape the tumor, sparing other tissues. Medical iodine for imaging or treatment is typically administered orally in the form a pill or liquid solution. A typical adult thyroid scan using iodine-123 involves having the patient swallow between one and four 0.1 millicuries pills, with the exact dose dependent on the patient’s weight. This results in a whole-body committed effective dose of 80 – 320 mrem, and a thyroid equivalent dose of 1443 – 5772 mrem. Note that the whole-body effective dose relates to the overall cancer risk, while the larger equivalent dose to the thyroid only indicates that most of this risk is the result of exposure to the thyroid. These doses are considered safe, although the procedure is not recommended for pregnant or breastfeeding women. In contrast, the quantities of iodine-131 used for treatment of hyperthyroidism and thyroid cancers are much higher. For treatment of hyperthyroidism, 4 – 10 millicuries are administered, while for thyroid cancer the administration can range from 50 – 150 millicuries of I-131. In addition to medical exposures, large populations were exposed to radioiodine as result of the atomic bombings of Hiroshima and Nagasaki in Japan, and the Chernobyl nuclear accident. These populations have been carefully followed for many years to assess the effect of their radiation exposures on cancer risk. As a result, a great deal is known about the cancer risks associated with radioiodine exposure. Because iodine is concentrated in the thyroid, the principal risk of exposure to radioiodine is thyroid cancer. Children have the highest risk of thyroid cancer after exposure to radioactive iodine. According to a large study of Japanese atomic bomb survivors, an effective dose of one Sievert (100,000 mrem) has been observed to increase the risk of thyroid cancer by a factor of 9.5 in children aged zero to nine years old, by a factor of 3 in children aged 10 to 19 years old, and by barely detectable amounts in adults. Another way of quantifying the risk from radioiodine exposure is from risk coefficients, which provide the risk per unit intake of radionuclides in terms of both morbidity (any cancer) and mortality (death). Both morbidity (risk of cancer) and mortality (death) risks are shown in the table below. Note that because thyroid cancer is almost never fatal, the morbidity coefficients are much larger than the mortality coefficients.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Ligand Desorption and Fragmentation in Oleate-Capped CdSe Nanocrystals under High-Intensity Photoexcitation

Semiconductor nanocrystals (NCs) offer prospective use as active optical elements in photovoltaics, light-emitting diodes, lasers, and photocatalysts due to their tunable optical absorption and emission properties, high stability, and scalable solution processing, as well as compatibility with additive manufacturing routes. Over the course of experiments, during device fabrication, or while in use commercially, these materials are often subjected to intense or prolonged electronic excitation and high carrier densities. The influence of such conditions on ligand integrity and binding remains underexplored. Here, in this study, we expose CdSe NCs to laser excitation and monitor changes in oleate that is covalently attached to the NC surface using nuclear magnetic resonance as a function of time and laser intensity. Higher photon doses cause increased rates of ligand loss from the particles, with upward of 50% total ligand desorption measured for the longest, most intense excitation. Surprisingly, for a range of excitation intensities, fragmentation of the oleate is detected and occurs concomitantly with formation of aldehydes, terminal alkenes, H 2 , and water. After illumination, NC size, shape, and bandgap remain constant although low-energy absorption features (Urbach tails) develop in some samples, indicating formation of substantial trap states. The observed reaction chemistry, which here occurs with low photon to chemical conversion efficiency, suggests that ligand reactivity may require examination for improved NC dispersion stability but can also be manipulated to yield desired photocatalytically accessed chemical species.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Irradiation-induced swelling of pure chromium with 5 MeV Fe ions in the temperature range 450–650 °C

A surface coating using pure chromium has been proposed to increase the accident tolerance of Zircaloy cladding in pressurized water reactors. However, there is not much irradiation experience with Cr alloys and especially pure Cr. In the present study, pure chromium was irradiated with 5 MeV Fe ions to 50 peak dpa (displacements per atom) at temperatures of 450, 500, 550, 600 and 650 °C. Then irradiation at the peak swelling temperature of 550 °C was conducted to 50, 100, 150 peak dpa. Swelling at 50 dpa was observed over the entire temperature range studied, 450–650 °C, but appeared to be decreasing strongly at the temperature boundaries of the experiment. After an initial transient of rapid swelling, chromium was observed to swell at a rate of ~0.03–0.04%/dpa (up to 120 local dpa), which is much lower than pure Fe at 0.2%/dpa. This low swelling rate was found to be relatively insensitive to dpa rate, which varied by a factor of ~2 over the depth of data collection. Here, swelling was observed to begin quickly with an incubation period less than 10 dpa. Self-organization in the form of void ordering was observed to be developing at 50 dpa, becoming better defined with increasing dose. The void alignment direction is determined to be the $\langle$111$\rangle$ axial direction.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Mechanical Property Measurements of HFIR-irradiated FM and ODS Alloys

This report provides a summary of the post-irradiation examination of the tensile properties of three ODS ferritic alloys, 14YWT-SM13, 9YWTV-PM2 and OFRAC, and two HT9 ferritic/martensitic steels, one with low N content and the other with high N content. The neutron irradiations were conducted in the flux trap of the High Flux Isotope Reactor as part of the A dditive M anufacturing, O xide Dispersion Strengthened Alloys and W rought Steels (AMOW) irradiation campaign. Sub-sized tensile specimens fabricated from each material were exposed to nominal target temperatures and neutron doses of 300°C, 385°C, and 525°C and 8 dpa, 16 dpa, and 50 dpa, respectively.

14YWT-SM13, 9YWTV-PM2, and OFRAC, and two HT9 ferr↗
DOE OSTI2020

Key Technical Issues for Greater-Than-Class-C (GTCC) Waste Disposal - 20164

Greater-than-Class C (GTCC) waste is low-level waste (LLW) that exceeds the Class C concentrations tabulated in Title 10, Code of Federal Regulations (CFR), Part 61. Disposal of GTCC waste in the near surface (i.e. upper 30 m of the earth's surface) is prohibited in the United States. Though GTCC waste disposal is generally prohibited, the Commission can approve disposal according to 10 CFR 61.55(a)(2)(iv) on a case-by-case basis. The waste classification tables, developed in the early 1980's, were based on model projections of dose to an inadvertent intruder in agricultural, construction, and discovery exposure scenarios. The assumptions and parameters used were documented in a series of public reports. Though different disposal facility designs and site conditions were considered, the waste classification tables were based on a disposal facility design located in a specific environment. The tables were based on shallow (i.e., top few meters) trench facility designs and did not consider deeper facilities. To determine the suitability of GTCC waste disposal in the near-surface, site-specific analyses must account for differences between GTCC waste and Class A, B, and C LLW (hereafter, traditional LLW). GTCC waste can have concentrations of radionuclides that are much higher than traditional LLW. Because of these higher concentrations, processes that are typically not significant for traditional LLW may be significant with respect to disposal of GTCC waste. These processes include, but are not limited to, heat generation, criticality, and radiolysis. The form of the waste as well as the barriers to release of the waste (e.g. waste package) could be substantially different than they are for traditional LLW. These barriers need to be considered when assessing the impacts of accidents during receipt and placement and in evaluating long-term performance. Some GTCC radioactivity is either embedded in stainless steel or contained in stainless steel barriers. Stainless steel can have very low corrosion rates under a variety of environmental conditions. Finally, and possibly most importantly, GTCC waste would likely need to be disposed deeper than traditional LLW to reduce the probability of disturbance. The waste classification table values of traditional LLW are based on an inadvertent intruder excavating into the waste and bringing some of the material to the land surface. If waste is deeper than approximately 5 m, the excavation scenario becomes very unlikely. Therefore, other intruder scenarios, such as drilling exposure scenarios, need to be evaluated. This paper summarizes key technical issues for the disposal of GTCC waste. Our previous study shows that certain GTCC waste may be suitable for near surface disposal whereas others may not. This study may support ongoing technical analyses assessing potential disposal of GTCC in a near surface disposal facility. (authors)

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

Dispersion calibration for the National Ignition Facility electron–positron–proton spectrometers for intense laser matter interactions

Electron–positron pairs, produced in intense laser–solid interactions, are diagnosed using magnetic spectrometers with image plates, such as the National Ignition Facility Electron–Positron–Proton Spectrometers (EPPSs). Although modeling can help infer the quantitative value, the accuracy of the models needs to be verified to ensure measurement quality. The dispersion of low-energy electrons and positrons may be affected by fringe magnetic fields near the entrance of the EPPS. We have calibrated the EPPS with six electron beams from a Siemens Oncor linear accelerator (linac) ranging in energy from 2.7 MeV to 15.2 MeV as they enter the spectrometer. A Geant4 Tool for Particle Simulation Monte Carlo simulation was set up to match depth dose curves and lateral profiles measured in water at 100 cm source–surface distance. An accurate relationship was established between the bending magnet current setting and the energy of the electron beam at the exit window. The simulations and measurements were used to determine the energy distributions of the six electron beams at the EPPS slit. Furthermore, analysis of the scanned image plates together with the determined energy distribution arriving in the spectrometer provides improved dispersion curves for the EPPS.

47 OTHER INSTRUMENTATION↗
DOE OSTI2020

Low Count and Background Radionuclides Analysis - 20488

The US Department of Energy (DOE) is often faced with the need to evaluate radionuclides at low concentrations. When site sample data are likely to be close to threshold activity concentrations of interest, then the means by which the radiochemical analysis is performed and reported is critical. This situation can occur when differentiating from zero (presence/absence) for radionuclides that do not occur naturally, close comparison with environmental background for naturally occurring radionuclides, close comparison with a risk- or dose-based threshold concentrations of interest, or even comparisons across studies. There are several analytical issues that are of concern, but the two that appear to cause incorrect decisions to be made most often involve establishing detection limits and subtracting ambient background conditions in the laboratory. These issues are not critical when radionuclide activity concentrations are large relative to thresholds of concern, but they seem to be poorly understood when it matters. When the comparisons are important and are likely to be close to a threshold of interest, then the general contract with the analytical laboratories needs to be changed so that the right or appropriate data are obtained. The concern is that important decisions are made incorrectly more often as greater scrutiny is placed on DoE's radionuclide cleanup or monitoring decisions by the public and other stakeholders. Examples are presented of problems that have been observed for different projects, both within and outside the realm of DOE and NRC remediation and radioactive waste disposal problems, and solutions are offered that should lead to better data from which important decisions need to be made. The first example is from Los Alamos National Laboratory (LANL) and involves radionuclide concentrations in soil and rock beneath LANL's Material Disposal Area (MDA) G. An initial review of the data led to a conclusion that americium and plutonium are a long way present beneath MDA G. A more thorough review of the data that accounted properly for ambient background and the detection limits that had been established led to the opposite conclusion. Another example is from the Nevada National Security Site where tritium results from one of the wells were unexpectedly high. Proper understanding and analysis of ambient background led to the conclusion that the increased concentrations were not so obvious, and that a different contract with the analytical laboratory was needed to provide more appropriate data to support a better determination. Other examples are used from regulatory review of projects in Nevada, where background levels and secular equilibrium for naturally occurring radionuclides are not established correctly because of analytical issues. The same basic issues have also been found to create difficulties analyzing historical data from the West Valley Demonstration Project. There is evidence in the data that the apparent lack of secular equilibrium where it is expected to exist is related to ambient background subtraction or other analytical issues. A final example is presented for analysis of Tc-99 in samples of depleted uranium. In this case, two different studies that were performed only three months apart provide quite different results. The US Environmental Protection Agency (EPA) established the data quality objectives (DQO) process in the mid-1980's to establish decision performance criteria for data collection. EPA guidance (EPA G-4, for example) clearly distinguishes between DQOs and measurement performance objectives (MQOs) that should be addressed for laboratory analysis of samples. The language of DQOs and MQOs has become confused over time it seems, and the subsequent effects seem to include a lack of attention to decision performance and a routine approach to measurement quality. In order to better address radionuclide sample analysis when the concentrations are close to thresholds of concern, which might be zero for some radionuclides, background for others, and risk-based thresholds for yet others, it is important that routine laboratory analysis methods are adjusted, and that the project team and the laboratory work closely together to ensure that the data meets the MQO requirements of laboratory analysis and reporting of results, and that the MQOs effectively support project-specific DQOs. This basic approach will be applied in Los Alamos in the coming year to the collection of moisture data from underneath MDA T that will be analyzed for americium and neptunium isotopes. Proper understanding of the radiochemistry methods and reporting, and of appropriate statistical methods is critical to the success of such projects, ensuring that the right decisions are made. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Non-Fouling, Low Cost Electrolytic Coagulation & Disinfection for Treating Flowback and Produced Water for Reuse

Executive Summary This Final Report is composed of two major sections. The first section presents experimental results obtained in the laboratory during Budget Period 1. The second section presents results from the field test conducted during Budget Period 2. Laboratory Results This research investigated a novel electrochemical process for producing a ferric iron coagulant for use in treating flowback and produced water from hydraulic fracturing and oil production operations. The treatment system improves the effectiveness and lowers the cost of coagulation processes using Fe3+ as the coagulant. The electrolytic coagulant generation (ECG) system uses an electrochemical cell to produce acid and base from oilfield brine solutions. The acid is used to dissolve scrap iron to provide a Fe3+ coagulating agent. The base is used to neutralize the treated water. Compared to conventional electrocoagulation (EC), the main advantage of the ECG system is an order of magnitude lower cost for the source of iron. The second advantage over conventional EC is that it can deliver Fe3+ doses greater than 1 mM, since it is not limited by the amount of dissolved oxygen in the water required to oxidize ferrous to ferric iron. The capital costs for conventional EC and the ECG system are similar, but the operational costs for the ECG system are an order of magnitude lower than conventional EC. The combined costs for iron and electrical energy for treating 1 m3 of FPW with 1 mM Fe3+ is estimated to be $0.87 for conventional EC, and $0.087 for the ECG system. The estimated all-in cost for treating FPW with a 2 mM Fe3+ dose is $0.73/m3 ($0.12/bbl). Field Test Results The field test was conducted at the Paul Foster Central Tank Battery (CTB) in Lea County, New Mexico from November 10, 2022 through December 15, 2022. The feed water to the system was produced water from the Tatanka 1H formation. After approximately two weeks of testing, the initial batch of produced water had been treated and no untreated produced water was available. Thus, after this time, the feed water to the system consisted of previously treated water (i.e., recycled water). The recycled water had nearly all colloidal particles removed, and had a much lower alkalinity due to precipitation of carbonate minerals during the first pass through the system. Although the recycled water was not an ideal test solution due to its low particulate concentrations, its lower alkalinity did allow us to identify the main problem with the treatment system. The main problem with the treatment system was caused by the high alkalinity of the initial feed water (5.4 meq/L) that consumed a significant fraction of the electrochemically generated acid. This resulted in pH values exiting the iron contact tank that were too high to dissolve enough iron to effectively treat the produced water. Tests performed with recycled water with lower alkalinity did not have this issue, and dissolved iron concentrations greater than 20 mM could be achieved. One consistent observation was that effluent water from the iron contact tank was always free of particulates, even when fed with circumneutral solutions. This suggests that there is no need to dissolve high concentrations of iron if all the water to be treated is passed through the scrap iron canister. In this case, dissolved O2 and hypochlorous acid can promote sufficient iron corrosion to provide an effective coagulating agent – even in neutral pH water. This solves the problem resulting from highly alkaline produced water. Modifications to the design of the treatment system were made based on the field test results. These modifications will add minimal additional cost, and were tested in bench-scale laboratory experiments. In short-term testing, the modified treatment process was able to remove colloidal FeS particulates to levels below detection. Long-term, steady state testing will be required to determine whether the modified process is suitable for commercial treatment systems.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2025

Consequence analyses of sabotage-induced radiological releases in sodium-cooled fast microreactors

Analysis of three sodium-cooled fast microreactors (SFMs) with thermal powers of 10, 30, and 50 MWt showed that smaller reactors result in lower radiological consequences during a postulated sabotage-induced event because of their reduced core inventory. All SFMs used U-10Zr metal fuel enriched to 15 wt% high-assay low-enriched uranium and operated until their respective effective multiplication factor (k eff ) reduced to less than 1 or until the end of their operational lifespan. Sabotage scenarios were simulated at this point, when the fuel inventory within the core contains the highest-level of radioactivity. Radionuclide core inventories were calculated using the SCALE code at shutdown and 3 days post-shutdown. Dose consequence analyses were performed for three sabotage scenarios using the RASCAL tool. As microreactor developers plan for minimal on-site or complete off-site emergency response, it remains essential to evaluate their physical protection needs and potential hazards, including assessing postulated sabotage-induced events that could become more relevant. SFM licensees should identify a credible worst-case, major accident, estimate release source terms, and perform dose consequence analyses to evaluate site-specific physical protection measures. In conclusion, this recommendation supports a risk-informed, performance-based approach, aligning with applicable regulatory requirements, i.e., 10 CFR Parts 100 and 53 rulemaking in the United States.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2022

On the Effect of Indium Chloride Dose on the Recrystallization of Cu(In,Ga)Se 2 Thin Films and associated Devices

Cu(In,Ga)Se 2 thin films deposited by a single-stage co-evaporation process at 350 ºC on molybdenum coated soda lime glass substrate were annealed post-deposition in InCl 3 vapor. The amount of InCl 3 and Se was varied. Here, the annealing treatment was done at 450 ºC for 30 minutes. Increase in grain size was observed after the treatment in all cases by X-ray diffraction. Device performance was low, but improved slightly after KCN etching.

14 SOLAR ENERGY↗
DOE OSTI2022

Radiolytic evaluation of select sulfur chlorides (S 2 Cl 2 and SOCl 2 ) for advanced low temperature chlorination of zirconium-based used nuclear fuel cladding

New sulfur chloride compounds—sulfur monochloride (S 2 Cl 2 ) and thionyl chloride (SOCl 2 )—have been proposed as alternative chlorinating agents for the chemical decladding of zirconium-based used nuclear fuel cladding materials. However, the radiation stability of these compounds has not been thoroughly evaluated, despite envisioned process conditions involving intense, multi-component radiation fields. Knowledge of the radiation chemistry of these sulfur chlorides is essential for the development and optimization of alternative chlorination technologies. To this end, we present preliminary findings on the gamma radiolysis of neat S 2 Cl 2 and SOCl 2 at ambient temperature up to ~ 34 MGy. In this study, Raman spectroscopy and density functional theory calculations were used to identify the Raman-active degradation products from S 2 Cl 2 and SOCl 2 radiolysis. Both sulfur chloride compounds exhibited significant radiation resistance, with respect to changes in the Raman signatures of the parent compounds and the ingrowth of Raman-active degradation products. For S 2 Cl 2 , molecular chlorine (Cl 2 ) was directly identified as the predominant degradation product, which was completely consumed at higher absorbed doses (> 9 MGy). Similarly, the main degradation product from SOCl 2 radiolysis was also Cl 2 , although in this system it continued to grow in with dose over the entire dose range. Further, an “S=O” containing degradation product(s) was also identified as a compliment to Cl 2 . Overall, the perceived radiation resistance of these sulfur chloride compounds makes them suitable for inclusion in a used nuclear fuel chemical decladding process, especially as radiolytically formed Cl 2 can be used to reform the parent compounds, thereby increasing the longevity of the solvents used.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Identifying the gene responsible for non‐photochemical quenching reversal in Phaeodactylum tricornutum

SUMMARY Algae such as diatoms and haptophytes have distinct photosynthetic pigments from plants, including a novel set of carotenoids. This includes a primary xanthophyll cycle comprised of diadinoxanthin and its de‐epoxidation product diatoxanthin that enables the switch between light harvesting and non‐photochemical quenching (NPQ)‐mediated dissipation of light energy. The enzyme responsible for the reversal of this cycle was previously unknown. Here, we identified zeaxanthin epoxidase 3 (ZEP3) from Phaeodactylum tricornutum as the candidate diatoxanthin epoxidase. Knocking out the ZEP3 gene caused a loss of rapidly reversible NPQ following saturating light exposure. This correlated with the maintenance of high concentrations of diatoxanthin during recovery in low light. Xanthophyll cycling and NPQ relaxation were restored via complementation of the wild‐type ZEP3 gene. The zep3 knockout strains showed reduced photosynthetic rates at higher light fluxes and reduced specific growth rate in variable light regimes, likely due to the mutant strains becoming locked in a light energy dissipation state. We were able to toggle the level of NPQ capacity in a time and dose dependent manner by placing the ZEP3 gene under the control of a β‐estradiol inducible promoter. Identification of this gene provides a deeper understanding of the diversification of photosynthetic control in algae compared to plants and suggests a potential target to improve the productivity of industrial‐scale cultures.

Ware, Maxwell A.↗
DOE OSTI2024

Improving to the neutron fluence rate monitor measurement system at the Advanced Test Reactor [Poster]

The existing fluence monitor wire scanning system at the Advanced Test Reactor (ATR) was designed and installed for use in the Engineering Test Reactor (ETR) when it began operation in 1958. The wire scanner was operated in ETR for over 20 years until ATR began operation, when it was moved to the ATR west canal area in 1971 and subsequently moved to the west canal in 2006 where it presently resides. With a continued service life of 65 years the system is well beyond the typical design life of 20 years for these types of systems. The need to update the data acquisition and control system was identified, and the benefits of replacing the existing sodium iodide (NaI) detector with an electronically cooled high-purity germanium (HPGe) detector are discussed. The wirescanner system in the ATR canal is utilized after every reactor cycle by the ATR Radiation Measurements Laboratory (RML) to assess the activation of cobalt and nickel dosimeter wires during the cycle. These wires become activated through exposure to thermal and fast neutrons respectively during the irradiation cycle and are highly radioactive upon shutdown. It is for this reason that the wirescanner is used in the ATR canal rather than transporting the dosimeters to another facility. A scoping study was performed to develop a base-line design to ensure that existing capabilities could be replaced with a new system. The new hardware will enable automated measuring of several flux monitor holders without necessitating the removal of the flux wires. In this way, flux wire measurements will be performed with minimal dose to the technicians and will not be limited by canal operations as is presently the case. The new control and acquisition software will be based on commercially available and supported systems that have a wide user-base to provide long-term stability. An electronically cooled HPGe detector will be used to provide high-resolution gamma-ray measurements, an improvement from the low-resolution sodium-iodide detector that is presently deployed. The electronic cooler eliminates the need for liquid nitrogen to cool the detector head. A new collimator has been designed to house the new detector and allow for sufficient counting rates.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗