Chemical Decladding of Used Nuclear Fuel fact sheet
Fact sheet about chemical decladding of used nuclear fuel.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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Chemical Decladding of Nuclear Fuel
High level up date for chemical decladding projects
42 - ENGINEERING↗
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↗
Demonstration of In Situ Down Blending During ZIRCEX Process
The Material Recovery Pilot Plant has successfully demonstrated in situ down-blending of HEU fuel to support INL’s commitment to HALEU production technologies. Depleted uranium pellets, along with an HEU fuel piece were placed in the hydrochlorinator and subsequently declad and oxidized. UOx product was recovered, analyzed, and demonstrated to have reduced enrichment within the targeted range of 20 – 30 % U-235. This first attempt at in situ down-blending will be followed by continued process refinement over the remaining fuel operations at MRPP.
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Evaluation of Ammonia Discharge into PUREX Crib 216-A-37-1 and Nitrogen Species Fate in the Subsurface
Based on published data, this report identifies potential fate and transport mechanisms for the ammonium hydroxide condensate that was discharged to 216-A-37-1 Crib from Hanford fuel rod decladding before PUREX dissolution and separation operations. Crib 216-A-37-1 received 377 million liters of alkaline condensate discharged from March 1977 through April 1989. The following major processes were identified as controlling the ammonia (NH 3 ) fate and transport: 1. Liquid migration rates through the vadose zone 2. NH 3 microbial oxidation to nitrite and nitrate 3. Ammonia species [NH4 + , NH 3 (aq)] sorption to sediments 4. NH3 precipitation reactions after pH buffering occurs in sediments The travel time through the vadose zone was estimated to be 2.5 to 9 months during periods of high subsurface discharge rates. These travel times are based on tritium migration through the vadose zone, a co-contaminant also discharged to the 216-A-37-1 crib.
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Off-Gas Design and Testing from Advanced Chlorination Process
Chlorination of Zr alloy cladding with sulfur-chloride reagents is being considered as a means of separating the main metal from alloying elements, imbedded U, actinides, and fission and activation products to purify the Zr. This would allow the Zr to be disposed of as low-level nuclear waste or possibly be recycled. A multistage process was developed that involves several steps and waste streams. Recycling of sulfur-chloride reagents S 2 Cl 2 and SOCl 2 is planned to minimize the waste stream. The generation of off-gas and entrainment of volatiles is anticipated to be minimized. However, given the complex nature of spent nuclear fuel (SNF), some volatile components will be generated, including the solvents themselves; Cl 2 used for the regeneration of S 2 Cl 2 ; volatile fission product chlorides, such as Sb and 3 H; and chlorinated alloying components, such as Sn, Nb, Fe, and Mo. This report discusses the chemical phenomena that might give rise to volatile species and their planned capture. The volatility of the transition metals will depend on their oxidation state and the compounds that are formed, such as Zr tritide. Loading radionuclides in the process off-gas will depend on the inputs to the process. For instance, processing separated Zr to reduce waste will have a much lower loading of radionuclides than if alternative chlorination were used for decladding SNF. Much of the volatile process streams will be part of the solvent recycling, which may involve online scrubbing. Generally, a caustic scrubber should be included in the off-gas to remove acidic vapors (e.g., Cl 2 , SOCl 2 ). The decontamination procedures for removing these species could involve pretreatment (i.e., heating) options or reduction by contact with H 2 . This report discusses these options along with a framework for testing these alternatives in future experiments. Technologies for off-gas capture will leverage those developed in the Materials Recovery and Waste Forms program.
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Sister Rod Destructive Examinations (FY23): Appendix J: Leaching of High Burnup Used Nuclear Fuel in Deionized Water
The leaching experiment aims to understand the trends in the radiolysis-enhanced dissolution of HBU SNF when exposed to water (e.g., in-reactor or in-pool cladding failures). Specimens from a baseline M5-clad rod and a heat-treated M5 rod were cut from the fractured CIRFT specimens and placed in 100 mL deionized water for a period of 128 days. Both radial and axial sections were cut to provide different surface areas of fuel in contact with the leachate. During the four-month exposure period, aliquot samples of the leachate were analyzed using gamma spectroscopy and inductively coupled plasma - mass spectrometry (ICP-MS). The analysis quantified the amount of fuel leached into the solutions and provided individual isotopic release fractions (of 30+ isotopes) which were compared as a function of time and surface area of the fuel exposed. Consistent with existing literature, the leaching followed a trend in which isotopes of certain elements such as Cs and Mo were among the first species in the matrix to dissolve, and with the highest release rates. This was followed by a gradual matrix dissolution consisting of uranium and other actinides and a slower-than-matrix release from some isotopes, including Ru and Rh. It was also observed that the circumferential samples having less exposed fuel surface area (to begin with) leached more than the axial samples for a majority of the isotopes during the timespan of the study. Previous experiments were conducted by decladding or exposing the pellet-clad surface completely, but the samples used in this study retained the cladding. One possible explanation for the higher leaching rate of the circumferential samples is that the pellet-clad interface, which has a greater density of grain boundaries and defects, may be the most vulnerable area to leaching of fuel in the presence of water.
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Characterization of Nitrate, Nitrite, Ammonia, and Tritium in D0220 Cores
This report describes characterization of nitrogen species (nitrate, nitrite, ammonia) in aqueous and solid phases in cores taken in borehole D0220 (well 299-E25-245) under crib 216-A-37-1 at the Hanford Site in 2022 to evaluate (a) the types of nitrogen species currently in the vadose zone and (b) the migration of nitrogen species in the vadose zone. Ammonia and tritium from PUREX decladding condensate were sporadically discharged to the crib from 1977 to 1989. During discharge operations, the estimated travel time through the vadose zone to groundwater was 2.5 to 9 months. By 2003, a characterization borehole (C4106) showed residual pore water with elevated tritium and nitrate in the first 100 ft of the vadose zone, which is likely from the crib. However, there was significantly more nitrate present at the shallowest depth (15 to 22 ft), which may indicate a different source for nitrate, such as adsorbed ammonia slowly being oxidized or nitrogen species precipitates slowly dissolving. In this study, more extensive nitrogen species characterization was done on D0220 cores (drilled in 2022) at 40- and 262-ft depths, which included (a) aqueous nitrate, nitrite, and tritium; (b) adsorbed ammonia; (c) nitrogen in carbonates (or other minerals dissolved in acidic acid); (d) nitrogen in iron oxides (or other minerals dissolved in oxalic acid); and (e) nitrogen in hard-to-extract minerals (minerals dissolved in nitric acid). High pore water nitrate (226 to 331 mg/L) at 40-ft depth measured in D0220 (2022) compared to 60 mg/L at 40-ft depth in C4106 (2003) may indicate nitrate is migrating deeper. Tritium concentrations (pore water 132,000 to 148,000 pCi/L) measured in D0220 at 40-ft depth in 2022 were considerably higher than in C4106 at 40-ft depth (160 pCi/L). Additional nitrogen species mass was present in adsorbed and precipitated phases in D0220 cores. Low adsorbed ammonia was measured at 40.1- and 261.7-ft depths. Low concentrations of carbonate associated nitrogen and iron oxide-associated nitrogen were present at 40.1 and 40.6 ft depths. Nitrogen species in solid phase extractions indicate nitrogen precipitates or aqueous nitrate or ammonia trapped in sediment microfractures that are coated by precipitates. Overall, the nitrogen species and tritium characterization at two depths in D0220 showed that additional nitrogen species were present in the vadose zone. In addition, elevated pore water nitrate in D0220 at 40 ft depth from 2022 may indicate vertical migration compared to C4106 nitrate profile from 2003. Analysis of additional depths in D0220 and spatial variability of the nitrate plume along the length of the crib (from surface electrical resistivity) would be extremely useful for this comparison.
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FY24 Task 4: Studies of Phosphate and Fluoride Solubility for Dissolution of High Phosphate Tank Waste
Knowledge gaps have been identified in phosphate solubility in almost every single- and multi-component system, and particularly for aluminate in phosphate/hydroxide, where uncertainties in predicted values of aluminum solubility are greater than 50%. Solubility data of multicomponent, aqueous electrolytes containing sodium hydroxide, sodium fluoride, sodium phosphate, sodium nitrite, sodium nitrate, and dissolved gibbsite are also sparse. For example, for solutions of (i) sodium nitrate and sodium phosphate, data are only available at 30 and 50 °C, and for (ii) dissolved gibbsite in sodium hydroxide and sodium phosphate, only two data points are available at 20 and 40 °C. There is no data available on mixtures of sodium hydroxide, sodium phosphate, sodium fluoride, and dissolved gibbsite. These knowledge gaps were identified in a technical review of waste solubility data and the impact of dilution on solution stabilities and will be addressed in this work to predict aluminum hydroxide and sodium phosphate solubility in multicomponent electrolytes upon dilution, and upon variation of temperature. Results will provide the technical basis to develop accurate models for (i) gibbsite solubility and mass transfer of aluminum between solid and liquid forms following the sluicing of sludge and saltcake with water; and (ii) further blending of these suspensions with bismuth from bismuth phosphate waste, and zirconium and uranium left from the fuel decladding. This work will be essential to developing a disposition path for retrieval solutions from bismuth phosphate wastes. This effort would also support sludge washing to further reduce phosphate concentration if that process were to be added back to the flowsheet in the future.
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Impact of Molten Gallium on the Microstructure and Corrosion Behavior of Aluminum and Uranium-Aluminum Alloys for Used Nuclear Fuel Reprocessing
Test reactors around the world utilize highly enriched uranium fuel to achieve high neutron fluxes for materials testing. Once spent, the remaining uranium is a valuable resource for subsequent fuel fabrication. However, some of these test reactor cores consist of curved plate-type fuel elements, fabricated using aluminum alloy 6061 (AA6061) cladding to encapsulate a uranium-aluminum alloy (UAlx) fuel matrix. These assemblies require non-standard reprocessing approaches for uranium recovery, as aluminum dissolves readily in acidic solutions, generating large volumes of waste and complicating downstream chemical separations. In this work, we investigate a novel chemical decladding strategy based on the interaction between AA6061/UAlx and molten gallium (Ga). Ga is known to induce severe degradation of aluminum metal through liquid metal embrittlement (LME), even at relatively low Ga concentrations. By penetrating the aluminum crystal lattice, Ga disrupts grain cohesion and facilitates fracture or dissolution of the aluminum matrix. Thermodynamic analysis of the Al–Ga binary phase diagram suggests that Ga may offer a viable pathway to selectively weaken or dissolve the AA6061 cladding, and potentially the aluminum component of the UAlx fuel matrix within. To this end, parametric experiments were performed at 50 °C and 100 °C across a range of Al–Ga atomic fractions. At lower Al fractions, the AA6061 was completely molten after 2 hours of exposure to the Ga metal. In contrast, samples with higher Al fractions (0.9 Al, 0.1 Ga) contained residual solids after 2 hours, which were characterized by microstructural examination using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). These Al-Ga compositions were also evaluated using FactSage thermodynamic modeling to further elucidate the relationship between phase diagram behavior and LME.
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