MELCOR for High Temperature Gas-Cooled and Fluoride High Temperature Reactor Modeling.
Abstract not provided.
Beeny, Bradley↗
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Methodology for Assessing Stainless Steel Compatibility in Fluoride Salts at 550°-850°C
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Pint, Bruce↗
Lithium yttrium ytterbium fluoride nanocrystals for laser cooling applications
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Material Science↗
Investigation of Misuse Potential and Detection for Homogeneous Fluoride Fast Molten Salt Reactors Based on Neutronics Signatures
The International Atomic Energy Agency (IAEA) has obligations to apply safeguards to nuclear material and facilities within States subject to IAEA safeguards agreements. As a result of these obligations, safeguards will need to be applied to Generation IV reactors, including liquid fueled molten salt reactors (MSRs), if they are deployed in States with safeguards agreements. Recently, the IAEA has noted that a growing number of Member States are showing interest in MSR technology. This has been accompanied by an increase in short term development and deployment activities relating to MSRs. To ensure MSRs can be safeguarded effectively and efficiently upon possible future deployment, safeguards approaches and techniques need to be investigated now.
DeGuire, Thomas↗
A laboratory-scale process for producing dilithium beryllium tetrafluoride (FLiBe) with dissolved uranium tetrafluoride
Flibe Energy, Incorporated (FEI)'s conceptual Lithium Fluoride Thorium Reactor (LFTR) incorporates a chemical processing facility aimed at recovering uranium and other valuable volatile radionuclides while managing harmful radionuclides from the used fuel. The fuel utilized in this reactor is a combination of dilithium beryllium tetrafluoride (Li 2 BeF 4 or FLiBe) and uranium tetrafluoride (UF 4 ), (FLiBe/U). FEI's plan involves extracting the uranium and other valuable volatile fluoride-forming radionuclides using nitrogen trifluoride (NF 3 ). To facilitate laboratory-scale testing of uranium extraction using NF 3 and address the toxicity and physical hazards associated with beryllium and beryllium fluoride (BeF 2 ), we used a two-step process to prepare the simulated fuel salt. The first step entailed thermally decomposing ammonium beryllium tetrafluoride [(NH 4 ) 2 BeF 4 ] (ABeF) through a nominal 3-step process, combined with appropriate amounts of lithium fluoride (LiF) and UF 4 , resulting in the formation of beryllium fluoride (BeF 2 ). In the second step, the mixture was repeatedly melted and frozen at the melting point of FLiBe to prepare the eutectic FLiBe with dissolved UF 4 . Although the concept appears straightforward, the production of FLiBe/U involved various challenges. These challenges included transporting the gaseous decomposition products of ABeF, hydrogen fluoride (HF) and ammonia (NH 3 ), while preventing the formation of ammonium fluoride (NH 4 F). Additionally, it was necessary to control the reaction between the higher-than-anticipated water content in the commercial ABeF with NH 3 , HF, and the condensed NH 4 F, protect UF 4 from forming an unknown black compound, select suitable structural materials to mitigate fluoride corrosion, address the risks associated with beryllium toxicity through equipment design and operational protocols, and monitor process conditions. This article provides an account of the thermal decomposition chemistry observed in the commercial ABeF, describes the FLiBe/U production apparatus, describes the experiences and process refinements developed to prepare FLiBe/U, and presents our characterizations of prepared FLiBe/U.
Ammonium beryllium fluoride thermal decomposition↗
Sulfur Hexafluoride Treatment of Used Nuclear Fuel to Enhance Separations
Reactive Gas Recycling (RGR) technology development has been initiated at Savannah River National Laboratory (SRNL), with a stretch-goal to develop a fully dry recycling technology for Used Nuclear Fuel (UNF). This approach is attractive due to the potential of targeted gas-phase treatment steps to reduce footprint and secondary waste volumes associated with separations relying primarily on traditional technologies, so long as the fluorinators employed in the reaction are recycled for use in the reactors or are optimized for conversion of fluorinator reactant. The developed fluorination via SF 6 , similar to the case for other fluorinators such as NF 3 , can be used to address multiple fuel forms and downstream cycles including continued processing for LWR via fluorination or incorporation into a aqueous process (e.g. modified FLUOREX) or for subsequent pyro treatment to be used in advanced gas reactor designs such metal- or gas-cooled reactors. This report details the most recent experimental results on the reaction of SF 6 with various fission product surrogate materials in the form of oxides and metals, including uranium oxides using a high-temperature DTA apparatus capable of temperatures in excess of 1000°C. The experimental results indicate that the majority of the fission products form stable solid fluorides and sulfides, while a subset of the fission products form volatile fluorides such as molybdenum fluoride and niobium fluoride, as predicted thermodynamically. Additional kinetic analysis has been performed on additional fission products. A key result is the verification that SF 6 requires high temperatures for direct fluorination and subsequent volatilization of uranium oxides to UF 6 , and thus is well positioned as a head-end treatment for other separations technologies, such as the volatilization of uranium oxide by NF 3 as reported by colleagues at PNNL, advanced pyrochemical separations or traditional full recycle approaches. Based on current results of the research at SRNL on SF 6 fluoride volatility for UNF separations, SF 6 treatment renders all anticipated volatile fluorides studied to be volatile, and all non-volatile fluorides studied to be non-volatile, with the notable exception of uranium oxides. This offers an excellent opportunity to use this as a head-end separations treatment process because: 1. SF 6 can be used to remove volatile fluorides from a UNF matrix while leaving behind uranium oxides. Therefore an agent such as NF 3 should be able to very cleanly separate a pure UF 6 stream, leaving compounds in the bottoms such as PuF 4 , SrF 2 and CsF after the UNF matrix has been pre-treated with SF 6 . 2. Due to the fact that the uranium oxide is not separated in the volatilization step upon direct contact with SF 6 at moderately high temperatures (≤ 1000°C), this fluoride approach may be well-suited for head-end processing for Gen IV reactor designs where the LWR is treated as a fuel stock, and it is not desired to separate the uranium from plutonium, but it is desired to separate many of the volatile fission products. 3. It is likely that removal of the volatile fission products from the uranium oxide should simplify both traditional and next generation pyroprocessing techniques. 4. SF 6 treatment to remove volatile fission products, with or without treatment with additional fluorinators, could be used to simplify the separations of traditional aqueous processes in similar fashion to the FLUOREX process. Further research should be conducted to determine the separations efficiency of a combined SF 6 /NF 3 separations approach which could be used as a stand-alone separations technology or a head-end process.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling
The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsin had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re-evaluate thermophysical properties of flibe and flinak. Pacific Northwest National Laboratories has focused on evaluating the fluorinating gas nitrogen trifluoride as a potential salt purification agent. Work there was performed on removing hydroxides and oxides from flinak salt under controlled conditions. Lastly, the University of California Berkeley has spent considerable time designing and simulating reactor components with fluoride salts at high temperatures. Despite the hurdles presented by the innate chemical hazards, considerable progress has been made. The stage has been set to perform new research on salt chemical control which could advance the fluoride salt cooled reactor concept towards commercialization. What were previously thought of as chemical undesirable, but nuclear certified, alloys have been shown to be theoretically compatible with fluoride salts at high temperatures. This preliminary report has been prepared to communicate the construction of the basic infrastructure required for flibe, as well as suggest original research to performed at the University of Wisconsin. Simultaneously, the contents of this report can serve as a detailed, but introductory guide to allow anyone to learn the fundamentals of chemistry, engineering, and safety required to work with flibe salt.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Electrokinetic comparison of CaF 2 and AgCl from a MDS perspective
CaF 2 has limited electrokinetic sensitivity to fluoride ions in solution and the positive CaF 2 surface charge is not reversed at high fluoride ion concentration. Similar insensitivity of surface charge to high fluoride ion concentration has been observed for other alkali and alkaline earth fluoride salts. On the contrary, chloride ions act as potential determining ions for AgCl and reverse the surface charge, as expected. FTIR-IRS results indicate that the AgCl/water interface has a water structure-breaking tendency, whereas the CaF 2 /water interface has a structure-making tendency. Here, it is speculated that the strong hydrogen bonding between fluoride ions and water prevents the accommodation of excess fluoride ions at surface lattice sites of fluorite, and therefore explains the low sensitivity of the fluorite surface charge sign and magnitude to fluoride ion concentration. However, this speculation has not been examined at the molecular level and is evaluated by molecular dynamics simulations (MDS) in the research presented in this paper. Higher hydrogen bond density at the CaF 2 mineral/water interface confirms that the water structure has stronger hydrogen bonding at both the CaF 2 mineral surface and with the F - ion in solution. In this way, the interaction of the F - ion with the CaF 2 surface is inhibited.
36 MATERIALS SCIENCE↗
Kinetic Parameters and Diffusivity of Uranium in FLiNaK and ClLiK
Anodic stripping voltammetry (ASV) and cyclic voltammetry (CV) measurements at 773, 823, and 873 K were made of uranium trifluoride (UF 3 ) in lithium fluoride-sodium fluoride-potassium fluoride eutectic (FLiNaK) and uranium trichloride (UCl 3 ) in lithium chloride-potassium chloride eutectic (ClLiK). ASV data were used to estimate the charge transfer coefficients, exchange current densities, and activation energies of the uranium reactions. Charge transfer coefficients of both salt systems were within the range of 0.17 to 0.38. Exchange current densities in the fluoride and chloride salts were estimated within the range of 0.060 to 0.12 A cm -2 . Activation energy of uranium exchange current was 16.0 kJ mol -1 in the fluoride salt and 28.9 kJ mol -1 in the chloride salt. Kinetics of charge transfer were found to be faster in FLiNaK. Analyses of the CV data suggest the electrochemical system was diffusion controlled and irreversible in the chloride and fluoride salt mixtures. Diffusion coefficients of uranium in the range of temperatures were on the order of 10-5 cm 2 s -1 in both systems. Greater values of diffusivity in ClLiK are attributed to its lower density compared with FLiNaK. Activation energy of uranium diffusion in the fluoride and chloride salt mixtures were 53.4 and 89.4 kJ mol -1 , respectively.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Mathematical modeling of positron emission tomography (PET) data to assess radiofluoride transport in living plants following petiolar administration
Background: Ion transport is a fundamental physiological process that can be studied non-invasively in living plants with radiotracer imaging methods. Fluoride is a known phytotoxic pollutant and understanding its transport in plants after leaf absorption is of interest to those in agricultural areas near industrial sources of airborne fluoride. Here we report the novel use of a commercial, high-resolution, animal positron emission tomography (PET) scanner to trace a bolus of [¹⁸F]fluoride administered via bisected petioles of Brassica oleracea , an established model species, to simulate whole plant uptake of atmospheric fluoride. This methodology allows for the first time mathematical compartmental modeling of fluoride transport in the living plant. Radiotracer kinetics in the stem were described with a single-parameter free- and trapped-compartment model and mean arrival times at different stem positions were calculated from the free-compartment time-activity curves. Results: After initiation of administration at the bisected leaf stalk, [¹⁸F] radioactivity climbed for approximately 10 minutes followed by rapid washout from the stem and equilibration within leaves. Kinetic modeling of transport in the stem yielded a trapping rate of 1.5 +/- 0.3%/min (mean +/- s.d., n = 3), velocity of 2.2 +/- 1.1 cm/min, and trapping fraction of 0.8 +/- 0.5%/cm. Conclusion: Quantitative assessment of physiologically meaningful transport parameters of fluoride in living plants is possible using standard positron emission tomography in combination with petiolar radiotracer administration. Movement of free fluoride was observed to be consistent with bulk flow in xylem, namely a rapid and linear change in position with respect to time. Trapping, likely in the apoplast, was observed. Future applications of the methods described here include studies of transport of other ions and molecules of interest in plant physiology.
54 ENVIRONMENTAL SCIENCES↗
Apparatus and methods for purifying lead
Disclosed is an exemplary method of purifying lead which includes the steps of placing lead and a fluoride salt blend in a container; forming a first fluid of molten lead at a first temperature; forming a second fluid of the molten fluoride salt blend at a second temperature higher than the first temperature; mixing the first fluid and the second fluid together; separating the two fluids; solidifying the molten fluoride salt blend at a temperature above a melting point of the lead; and removing the molten lead from the container. In certain exemplary methods the molten lead is removed from the container by decanting. In still other exemplary methods the molten salt blend is a Lewis base fluoride eutectic salt blend, and in yet other exemplary methods the molten salt blend contains sodium fluoride, lithium fluoride, and potassium fluoride.
36 MATERIALS SCIENCE↗
Electrochemical Behavior of Lanthanum in Molten LiF-NaF-KF 'Flinak' Eutectic Salt
In order to develop the Fluoride High- Temperature Reactor, an emerging Gen-IV reactor concept, a primary coolant cleanup system must be developed to allow online reprocessing of the primary molten fluoride salt. This research is focused on the determination of electrochemical properties of rare earths (REs) in molten fluoride salt related to the FHR. The FHR utilizes two fluoride salts. These fluoride salts are LiF-NaF-KF (46.5-11.5-42 mol%), melting point 454 deg. C, also known as 'Flinak', and LiFBeF{sub 2} (33-67 mol%), melting point 459 deg. C, also known as 'Flibe'. Flibe is selected as the primary coolant salt due to its low neutron cross-section (Li{sup 7}-enriched), superior heat transfer properties, and low melting point. However, due to the beryllium content in Flibe, it is highly toxic to humans and Flinak is preferred for the secondary and tertiary coolant salts. The work presented here is specifically focused on the behavior of lanthanum trifluoride (LaF{sub 3}) in molten Flinak. Furthermore, updated data and analysis regarding the electrochemical behavior of dysprosium trifluoride (DyF{sub 3}) is also presented. Previous work related to this topic was presented at the 2016 ANS winter conference. The goal of this research is to develop a contaminant cleanup system for the primary loop of the FHR.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
On the Use of Thermal NF3 as the Fluorination and Oxidation Agent in Treatment of Used Nuclear Fuels
This paper presents results of our investigation on the use of nitrogen trifluoride as the fluorination or fluorination/oxidation agent for use in a process for separating valuable constituents from used nuclear fuels by employing the volatility of many transition metal and actinide fluorides. Nitrogen trifluoride is less chemically and reactively hazardous than the hazardous and aggressive fluorinating agents used to prepare uranium hexafluoride and considered for fluoride volatility based nuclear fuels reprocessing. In addition, nitrogen trifluoride’s less aggressive character may be used to separate the volatile fluorides from used fuel and from themselves based on the fluorination reaction’s temperature sensitivity (thermal tunability) rather than relying on differences in sublimation/boiling temperature and sorbents. Our thermodynamic calculations found that nitrogen trifluoride has the potential to produce volatile fission product and actinide fluorides from candidate oxides and metals. Our simultaneous thermogravimetric and differential thermal analyses found that the oxides of lanthanum, cerium, rhodium, and plutonium fluorinated but did not form volatile fluorides and that depending on temperature volatile fluorides formed from the oxides of niobium, molybdenum, ruthenium, tellurium, uranium, and neptunium. We also demonstrated near-quantitative removal of uranium from plutonium in a mixed oxide.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Removal of Surface Carbonate from Lithium-Ion Battery Cathode Materials via Vapor-Phase Fluorination
Herein, ultrathin metal-fluoride barrier coatings on the surface of lithium-ion battery (LIB) cathodes can improve cycling stability and prevent corrosion by acidic byproducts in the electrolyte. Atomic layer deposition (ALD) is an effective method to deposit ultrathin metal fluoride coatings on LIB cathodes. Although numerous studies have demonstrated the benefit of ALD metal fluoride coatings to LIB performance, comparatively few works have examined the effect of individual ALD precursors on the cathode surface. This paper uses X-ray photoelectron spectroscopy (XPS) measurements to elucidate the surface chemical changes on LIB cathode material surfaces upon exposure to the ALD metal fluoride precursor, hydrogen fluorine pyridine (HFPy). We found a decrease in surface carbonate and an increase in surface fluoride after HFPy exposure suggesting the conversion of lithium carbonate (Li 2 CO 3 ) to lithium fluoride (LiF). This conversion is desirable given that Li 2 CO 3 degrades LIB performance, whereas LiF provides an excellent physio-chemical barrier against chemical attack during cycling. Scanning transmission electron microscopy, X-ray energy dispersive spectroscopy, and XPS measurements following HFPy exposure to Li 2 CO 3 powder revealed the formation of a conformal LiF shell around the Li 2 CO 3 particles. Finally, we confirmed the complete conversion of similar to 7 nm ALD Li 2 CO 3 films on silicon from HFPy exposure using XPS and spectroscopic ellipsometry. The elimination of problematic Li 2 CO 3 from LIB cathode surfaces and conversion into a protective LiF coating via a single precursor vapor treatment may provide a cost-effective method for enhancing LIB performance.
25 ENERGY STORAGE↗
Characterization of the kinetics of NF 3 -fluorination of NpO 2
The exploitation of selected actinide and fission product fluoride volatilities has long been considered as a potentially attractive compact method for recycling used nuclear fuels to avoid generating the large volumes of radioactive waste arising from aqueous reprocessing . The most developed process uses the aggressive and hazardous fluorinating agents hydrogen fluoride (HF) and/or molecular fluorine (F 2 ) at high temperatures to volatilize the greatest fraction of the used nuclear fuel into a single gas stream. The volatilized fluorides are subsequently separated using a series of fractionation and condensation columns to recover the valuable fuel constituents and fission products. In pursuit of a safer and less complicated approach, we investigated an alternative fluoride volatility-based process using the less hazardous fluorinating agent nitrogen trifluoride (NF 3 ) and leveraging its less aggressive nature to selectively evolve fission product and actinide fluorides from the solid phase based on their reaction temperatures into a single recycle stream. In this approach, successive isothermal treatments using NF 3 will first evolve the more thermally susceptible used nuclear fuel constituents leaving the other constituents in the residual solids until subsequent isothermal temperature treatments cause these others to volatilize. During investigation of this process, individual neat used fuel components were treated with isothermal NF 3 in an attempt to characterize the kinetics of each fluorination reaction to provide input into the design of a new volatile fluoride separations approach. In these directed investigations, complex behavior was observed between NF 3 and certain solid reactants such as the actinide oxides of uranium, plutonium, and neptunium. Given the similar thermal reaction susceptibilities of neptunium oxide (NpO 2 ) and uranium dioxide (UO 2 ) and the importance of Np and U, we initially focused our efforts on determining the reaction kinetic parameters for NpO 2 . Characterizing the NF3 fluorination of NpO 2 using established models for gas-solid reactions [16] proved unsuccessful so we developed a series of successive fundamental reaction mechanisms to characterize the observed successive fluorination reactions leading to production of the volatile neptunium hexafluoride (NpF 6 ).
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
FY24 Task 5: Leachate Disposition
Directly feeding sludge solids to the high-level waste (HLW) Waste Treatment Plant represents an alternative flowsheet seeking to initiate sludge processing as soon as possible. Key processing functions previously captured during baseline pretreatment operations include leaching and washing prior to solids concentration. These operations should be considered in the potential direct feed flowsheets to maximize waste feed loading, minimize HLW volume, and mitigate corrosion challenges associated with vitrification of high phosphate and fluoride concentrations. Additionally, single-shell tank (SST) retrievals and waste transfers to double-shell tanks (DSTs) in a direct feed flowsheet would likely also benefit from some level of leaching, washing, and solids concentration in order to reduce DST space and mission duration. These operations could occur in a new facility or potentially in available DSTs. If washing and leaching are utilized, an effective disposition pathway for the wash water and leachate solutions are needed. Three target species that benefit significantly from leaching and washing are phosphate, fluoride and aluminum. Phosphate (PO 4 3- ) and fluoride (F - ) can contribute substantially to the amount of carrier fluid needed for dissolution, and the resulting volume of liquid generated. Disposition of this retrieval solution should be evaluated in order to prevent crystallization of these anions throughout system processing. Since there is a high probability that any retrieval solutions will be at or near their PO 4 3- and F - solubility limits, evaporation or blending with a high Na supernate (>3.5 M) is not recommended for the wash water streams without a method to remove precipitants prior to solution disposal. Additionally, aluminum present in the southeast quadrant of the Hanford site represents roughly 60% of the waste solids in the initial processing tanks. These aluminum solids are in the form of gibbsite (Al(OH) 3 ) and can pose significant challenges for processing due to the fast-settling times and high solids loading associated with these materials. Easily remediated by caustic addition to the solids, these wash solutions could be processed through crystalline silicotitanate (CST) ion exchange columns to prepare the supernate solutions for disposition. The current target for feed conditions to the Low Activity Waste (LAW) melter are waste streams that contain nominally 5-6 M Na. Fractions within the tanks contain upwards of 0.2 M phosphate and fluoride in solution at 3.5 M Na. Concentrating these solutions above 5 M Na would result in an exceedance of the solubility limits, and potential for uncontrolled precipitation of the phosphate and fluoride crystal material. The resulting crystalline salt material is typically sodium fluoride phosphate, also referred to as natrophosphate (Na 7 FPO 4 ·19H 2 O). Salt phases are of importance in tank waste due to their chemical reactivity, which can result in precipitation, dissolution, or transformation, impacting any downstream processes (Bolling et al. 2020, Russell, Snow, and Peterson 2010). Salt generation and precipitation could pose challenges by causing system plugging and melter corrosion if left in the supernate stream, or limit sodium molarity of the supernate that would be accepted without incident in waste operations. To understand the impact of this salt generation, the crystallization of natrophosphate in multiple simulant feed matrices was studied to understand the implications of various tank waste supernate chemistries. Three matrices were examined: high PO 4 3- /low F - , low PO 4 3- /high F - , and an average matrix. Subsequent testing was performed with the average matrix with the inclusion of CsNO 3 , and a final run with the average matrix including CsNO 3 and a 137 Cs spike for tracer purposes.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Chemico-therapeutic approach to prevention of dental caries
The program of chemical preventive dentistry is based primarily upon the development of a procedure for stabilizing stannous fluoride in solution by forcing it into glycerin. New topical fluoride treatment concentrates, fluoride containing gels and prophylaxis pastes, as well as a completely stable stannous fluoride dentifrice are made possible by the development of a rather complicated heat application method to force stannous fluoride into solution in glycerin. That the stannous fluoride is clinically effective in such a preparation is demonstrated briefly on orthodontic patients.
Shannon, I. L.↗
Effects of Disinfectants in Water on Mir- and Earth-Grown Wheat
Iodine and silver fluoride are used to purify water onboard U. S. Shuttles and the Russian Space Station, Mir, respectively. In 1995, iodine-treated water, which ranged from 1.0-4.0 mg x kg(exp -1) with a mean of 2.9 mg x kg(exp -1), was applied to Super Dwarf wheat (Triticum aestivum L.) plants when Mir water (grey or tech grade) became scarce. The potential phytotoxicity of iodine on Super Dwarf wheat is an unknown. Since use of iodine-treated water was not part of the experiment, we sought to determine whether it accounted for the subsequent poor wheat seedling growth and floral development onboard the Mir. Super Dwarf wheat seeds were imbibed in iodine or silver fluoride concentrations of 0.0, 1.0, 2.0, 4.0, 8.0 or 16.0 mg x kg(exp -1) for 96 h at 4 C. Five seeds were then planted per 13.3 cm x 13.3 cm pots containing a granular clinoptilolite (Cp) zeolite (1 -2 mm dia.) and placed in Percival(TM) growth chambers programmed for 20/15 C and 18/6 h d/n regime. Plants were irrigated with distilled water, and Iodine- or silver fluoride-treated distilled water. In separate experiments, seeds were treated as above and germination and early seedling growth were determined by examining seedling responses to disinfectants in rolled paper towels. Silver fluoride had very little effect on wheat seed germination. By contrast, iodine reduced germination at all treatment levels. Seedlings exposed to 1.0, 2.0, and 4.0 mg x kg(exp -1) of iodine or silver fluoride levels exhibited a slight stimulation in shoot and root growth. Both disinfectants at 8 and 16 mg x kg(exp -1) showed significantly (p is less than or equal to 0.01) reduced seedling shoot and root lengths and fresh biomasses compared to the control and lower disinfectant levels. The number of spikelets per spike, florets per spikelet, seeds per spike and seed weight were also significantly reduced at the 8 and 16 mg x kg(exp -1) compared to the control and lower levels of disinfectant. Based on these ground-based post-flight analyses, the levels of iodine- and/or silver fluoride-treated water used on Mir-grown plants onboard the Mir did not cause the poor growth and development of the wheat plants.
Campbell, William .F.↗