Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “dissolving”

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 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Influence of dissolved salts and impurities on seawater on heat transfer and fluid flow through fuel channels and debris bed

This project aims to better understand this mechanism and its effect on heat transfer. Specifically, experimental research was conducted to investigate the long term coolability issues in scaled test setups. Advanced diagnostic and measurement techniques were investigated and incorporated into an annular test facility and a debris bed test facility. The diagnostic techniques investigated included: radioactively tagged salt solution, Frisch collar radiation detector array, neutron and x-ray imaging systems, and a Rayleigh backscatter-based fiber-optic temperature system. Models were checked against the in-situ experimental data and recommendations were made for modeling purposes.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

ASC Poster - Concentration of Xe from Simulated Dissolver Off-gas Streams Utilizing a Solid Sorbent, AgZ-PAN

Idaho National Laboratory (INL) has developed and tested engineered sorbents to separate and capture volatile fission products such iodine (I), xenon (Xe), and krypton (Kr) from off-gas streams. As noble gases, Xe and Kr can be difficult to capture and separate. Historically, cryogenic distillation has been used to execute the separation, but this method can pose significant hazards. INL has successfully developed two sorbents for Xe and Kr capture, silver mordenite polyacrylonitrile (AgZ-PAN) and hydrogen mordenite polyacrylonitrile (HZ-PAN). Adsorption studies to date successfully separated and captured Xe from carrier gas streams, but those studies focused primarily on initial separation and capture. In addition, the ability to concentrate Xe through multiple adsorption and desorption cycles has not been studied and therefore is the focus on this presentation. Concentrating Xe during desorption is ideal to produce Xe that can be used for commercial and research applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

DISSOLUTION OF PLUTONIUM METAL IN 8-10 M NITRIC ACID

The H-Canyon facility will be used to dissolve Pu metal for subsequent purification and conversion to plutonium dioxide (PuO 2 ) using Phase II of HB-Line. To support the new mission, the development of a Pu metal dissolution flowsheet which utilizes concentrated (8-10 M) nitric acid (HNO 3 ) solutions containing potassium fluoride (KF) is required. Dissolution of Pu metal in concentrated HNO 3 is desired to eliminate the need to adjust the solution acidity prior to purification by anion exchange. The preferred flowsheet would use 8-10 M HNO 3 , 0.015-0.07 M KF, and 0.5-1.0 g/L Gd to dissolve the Pu up to 6.75 g/L. An alternate flowsheet would use 8-10 M HNO 3 , 0.05-0.2 M KF, and 1-2 g/L B to dissolve the Pu. The targeted average Pu metal dissolution rate is 20 mg/min-cm 2 , which is sufficient to dissolve a “standard” 2250-g Pu metal button in 24 h. Plutonium metal dissolution rate measurements showed that if Gd is used as the nuclear poison, the optimum dissolution conditions occur in 10 M HNO 3 , 0.04-0.05 M KF, and 0.5-1.0 g/L Gd at 112 to 116 °C (boiling). These conditions will result in an estimated Pu metal dissolution rate of ~11-15 mg/min-cm 2 and will result in dissolution times of 36-48 h for standard buttons. The recommended minimum and maximum KF concentrations are 0.03 M and 0.07 M, respectively. The data also indicate that lower KF concentrations would yield dissolution rates for B comparable to those observed with Gd at the same HNO 3 concentration and dissolution temperature. To confirm that the optimal conditions identified by the dissolution rate measurements can be used to dissolve Pu metal up to 6.75 g/L in the presence of representative concentrations of Fe and Gd or B, a series of experiments was performed to demonstrate the flowsheets. In three of the five experiments, the offgas generation rate during the dissolution was measured and samples were analyzed for hydrogen gas (H 2 ). The use of 10 M HNO 3 containing 0.03-0.05 M KF, 0.5-1.0 g/L Gd, and 1.9 g/L Fe resulted in complete dissolution of the metal in 2.0-3.5 h. When B was used as the neutron poison, 10 M HNO 3 solutions containing 0.05-0.1 M KF, 1.9 g/L Fe, and 1 g/L B resulted in complete dissolution of the metal in 0.75-2.0 h. Dissolution rates estimated using data from the flowsheets demonstrations agreed reasonably well with the measured rates; although, a discrepancy was observed in the Gd system. The presence of 1 g/L Gd or B in the dissolving solution had about the same effect on the dissolution rate. The predominant Pu valence in the dissolving solution was Pu(IV). The concentration of Pu(VI) was evaluated by UV-visible spectroscopy and was estimated to be significantly less than 1 wt %. The offgas generation rates and H 2 concentrations measured in the offgas from experiments performed using 10 M HNO 3 containing 0.05 M KF, 1.9 g/L Fe and either 1 g/L Gd or B were approximately the same. These data support the conclusion that the presence of either 1 g/L Gd or B had the same general effect on the dissolution rate. The calculated offgas generation during the dissolutions was 0.6 mol offgas/mol of Pu. The H 2 concentration measured in the offgas from the dissolution using Gd as the neutron poison was approximately 0.5 vol %. In the B system, the H 2 ranged from nominally 0.8 to 1 vol % which is about the same as measured in the Gd system within the uncertainty of the analysis. The offgas generation rate for the dissolution performed using 10 M HNO 3 containing 0.03 M KF, 0.5 g/L Gd, and 1.9 g/L Fe was approximately a factor of two less than produced in the other dissolutions; however, the concentration of H 2 measured in the offgas was higher. The adjusted concentration ranged from 2.7 to 8.8 vol % as the dissolution proceeded. Higher concentrations of H 2 occur when the Pu dissolution proceeds by a metal/acid reaction rather than nitrate oxidation. The higher H 2 concentration could be attributed to the reduced activity of the fluoride due to complexation with Pu as the dissolution progressed. Dissolution of Pu metal at 20 °C in 10 M HNO 3 containing 0.05 M KF showed that the Pu metal dissolves slowly without any visible gas generation. As the Pu metal dissolves, it forms a more-dense Pu-bearing solution which sank to the bottom of the dissolution vessel. The dissolved Pu did not form a boundary layer around the sample and failed to distribute homogeneously due to minimal (thermally-induced) mixing. This indicates that in the H-Canyon dissolver insert, the Pu will diffuse out of the insert into the bulk dissolver solution where it will disperse. At 35 °C, the Pu metal dissolved without visible gas generation. However, due to thermal currents caused by maintaining the solution at 35 °C, the dissolved Pu distributed evenly throughout the dissolver solution. It did not form a boundary layer around the sample.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

AFS-2 FLOWSHEET MODIFICATIONS TO ADDRESS THE INGROWTH OF PU(VI) DURING METAL DISSOLUTION

In support of the Alternate Feed Stock Two (AFS-2) PuO{sub 2} production campaign, Savannah River National Laboratory (SRNL) conducted a series of experiments concluding that dissolving Pu metal at 95°C using a 6–10 M HNO{sub 3} solution containing 0.05–0.2 M KF and 0–2 g/L B could reduce the oxidation of Pu(IV) to Pu(VI) as compared to dissolving Pu metal under the same conditions but at or near the boiling temperature. This flowsheet was demonstrated by conducting Pu metal dissolutions at 95°C to ensure that PuO{sub 2} solids were not formed during the dissolution. These dissolution parameters can be used for dissolving both Aqueous Polishing (AP) and MOX Process (MP) specification materials. Preceding the studies reported herein, two batches of Pu metal were dissolved in the H-Canyon 6.1D dissolver to prepare feed solution for the AFS-2 PuO{sub 2} production campaign. While in storage, UV-visible spectra obtained from an at-line spectrophotometer indicated the presence of Pu(VI). Analysis of the solutions also showed the presence of Fe, Ni, and Cr. Oxidation of Pu(IV) produced during metal dissolution to Pu(VI) is a concern for anion exchange purification. Anion exchange requires Pu in the +4 oxidation state for formation of the anionic plutonium(IV) hexanitrato complex which absorbs onto the resin. The presence of Pu(VI) in the anion feed solution would require a valence adjustment step to prevent losses. In addition, the presence of Cr(VI) would result in absorption of chromate ion onto the resin and could limit the purification of Pu from Cr which may challenge the purity specification of the final PuO{sub 2} product. Initial experiments were performed to quantify the rate of oxidation of Pu(IV) to Pu(VI) (presumed to be facilitated by Cr(VI)) as functions of the HNO{sub 3} concentration and temperature in simulated dissolution solutions containing Cr, Fe, and Ni. In these simulated Pu dissolutions studies, lowering the temperature from near boiling to 95 °C reduced the oxidation rate of Pu(IV) to Pu(VI). For 8.1 M HNO{sub 3} simulated dissolution solutions, at near boiling conditions >35% Pu(VI) was present in 50 h while at 95 °C <10% Pu(VI) was present at 50 h. At near boiling temperatures, eliminating the presence of Cr and varying the HNO{sub 3} concentration in the range of 7–8.5 M had little effect on the rate of conversion of Pu(IV) to Pu(VI). HNO{sub 3} oxidation of Pu(IV) to Pu(VI) in a pure solution has been reported previously. Based on simulated dissolution experiments, this study concluded that dissolving Pu metal at 95°C using a 6 to 10 M HNO{sub 3} solution 0.05–0.2 M KF and 0–2 g/L B could reduce the rate of oxidation of Pu(IV) to Pu(VI) as compared to near boiling conditions. To demonstrate this flowsheet, two small-scale experiments were performed dissolving Pu metal up to 6.75 g/L. No Pu-containing residues were observed in the solutions after cooling. Using Pu metal dissolution rates measured during the experiments and a correlation developed by Holcomb, the time required to completely dissolve a batch of Pu metal in an H-Canyon dissolver using this flowsheet was estimated to require nearly 5 days (120 h). This value is reasonably consistent with an estimate based on the Batch 2 and 3 dissolution times in the 6.1D dissolver and Pu metal dissolution rates measured in this study and by Rudisill et al. Data from the present and previous studies show that the Pu metal dissolution rate decreases by a factor of approximately two when the temperature decreased from boiling (112 to 116°C) to 95°C. Therefore, the time required to dissolve a batch of Pu metal in an H-Canyon dissolver at 95°C would likely double (from 36 to 54 h) and require 72 to 108 h depending on the surface area of the Pu metal. Based on the experimental studies, a Pu metal dissolution flowsheet utilizing 6–10 M HNO{sub 3} containing 0.05–0.2 M KF (with 0–2 g/L B) at 95°C is recommended to reduce the oxidation of Pu(IV) to Pu(VI) as compared to near boiling conditions. The time required to completely dissolve a batch of Pu metal will increase, however, by approximately a factor of two as compared to initial dissolutions at near boiling (assuming the KF concentration is maintained at nominally 0.1 M). By lowering the temperature to 95°C under otherwise the same operating parameters as previous dissolutions, the Pu(VI) concentration should not exceed 15% after a 120 h heating cycle. Increasing the HNO{sub 3} concentration and lowering Pu concentration are expected to further limit the amount of Pu(VI) formed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Use of Stable Mercury Isotopes to Assess Mercury and Methylmercury Transformation and Transport across Critical Interfaces from the Molecular to the Watershed Scale (Final Report)

This project titled “Use of Stable Mercury Isotopes to Assess Mercury and Methylmercury Transformation and Transport across Critical Interfaces from the Molecular to the Watershed Scale” represents a collaborative effort between the University of Michigan (Jason Demers, PI) and Oak Ridge National Laboratory (Scott Brooks, co-I). Much has been learned about mercury (Hg) cycling in stream ecosystems, and East Fork Poplar Creek (EFPC) in particular, through decades of previous research. Nevertheless, some of the most fundamental questions regarding the sources of bioavailable Hg and its transformation to toxic methylmercury (MeHg) have remained unanswered. These fundamental questions include: (1) what are the sources and biogeochemical processes that lead to the input of dissolved Hg to stream water across critical subsurface interfaces within stream ecosystems, and EFPC in particular? and (2) what are the sources and biogeochemical processes that control the production and fate of bioaccumulative MeHg within stream ecosystems, and in EFPC in particular? To address these fundamental questions, our project aimed to couple laboratory experiments and field observations, both utilizing natural abundance Hg stable isotope techniques, to identify the processes responsible for generating mobile, bioavailable dissolved Hg from recalcitrant legacy sources within critical subsurface zones (e.g., streambed hyporheic zone, riparian floodplain subsurface). We used the isotopic signature of this bioavailable dissolved Hg to track its mobilization across these critical interfaces in order to link diffuse subsurface sources of dissolved Hg with increases in surface water dissolved Hg flux measured at the watershed scale. Additionally, our research aimed to determine the isotopic composition of MeHg within these same critical subsurface zones. We directly assessed the isotopic composition of MeHg within biota in order to gain insight into which subsurface sources of inorganic Hg and toxic MeHg are available for bioaccumulation within the EFPC ecosystem. Net fluxes of dissolved Hg along the flow path of EFPC were shown to vary spatially and temporally. In the Upper EFPC, within the Y12 boundary, stream water flux of dissolved Hg consistently decreased between the outfall and the downstream boundary of Y12 (57% ± 29%, 1SD). Within the Upper EFPC, an assessment of Hg isotopic composition suggested that losses were strongly reaction-driven, although isotopic diagnostics did not conform to any known processes. Downstream of Y12, in the upper reach of the Lower EFPC, dissolved Hg fluxes tended to increase during the dormant season (net gain of 11-120%), and decrease during the growing season (net loss of 23% +/- 18%, 1SD). In the downstream-most reach of Lower EFPC, dissolved Hg fluxes increased by 12-108% in 9 out of 10 monthly assessments. Overall, diffuse fluxes from the non-Y12 watershed accounted for 34% (+/- 17%, 1SD) of all dissolved Hg exported during base flow. Within Lower EFPC, an assessment of Hg isotopic composition was consistent with the contribution of diffuse Hg inputs from high-concentration hotspots within riparian floodplains and streambed hyporheic pore water. To investigate remobilization of recalcitrant Hg from legacy sediment sources, we developed procedures that coupled isotopic analysis with sequential extractions of streambed sediment. We found that the proportion of weakly-bound Hg within EFPC streambed sediment was relatively small, but could still account for a large proportion of the annual flux of dissolved Hg from EFPC. These sequential extractions also showed that this weakly-bound Hg fraction could be replenished from the much larger fraction of recalcitrant Hg in sediment. The isotopic composition of these weakly-bound and remobilized recalcitrant Hg fractions within the sediment was consistent with high-concentration dissolved Hg hotspots within hyporheic pore water. Thus, this research provided novel evidence that legacy mercury sources within streambed sediment could provide an ongoing contribution of dissolved Hg to surface waters. Finally, we developed new methods for the direct determination of the MeHg isotopic composition of organisms, which allowed a more direct evaluation of inorganic Hg and MeHg sources accumulating in the food web. We found that fish and aquatic invertebrates in both EFPC and a regional background site obtained inorganic Hg and MeHg from multiple isotopically distinct sources, including sediment, suspended particulates, and periphyton. Photodemethylation was found to be an important reaction influencing MeHg dynamics at both sites. However, the balance of microbial methylation and demethylation processes differed between the two streams, with fractionation resulting from methylation and demethylation processes being relatively in balance within the regional background site, whereas microbial methylation appeared to be dominant over microbial demethylation within the EFPC ecosystem. Broadly, the application of Hg isotopic analysis in this study led to numerous novel insights regarding the biogeochemical cycling of Hg in stream ecosystems. This research project promoted the development of two new approaches, including the coupling of sequential extractions with Hg isotopic analysis to assess remobilization of recalcitrant Hg within sediments, and a new method for the isotopic analysis of MeHg isolated from environmental samples. Both of these efforts represent advances in capacity for the field of mercury isotopic analysis and environmental assessment. Overall, this study demonstrates that the application of Hg stable isotope techniques continues to provide new insights into the biogeochemical cycling of Hg in complex aquatic environments, both within the EFPC and beyond.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Dissolution Flowsheet for High Flux Isotope Reactor Fuel

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3 O 8 ) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2 . The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H 2 ) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy, and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H 2 and other permanent gases in the dissolution offgas, allowing the development of H 2 generation rate profiles. The H 2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the calculated lower flammability limit (LFL) for H 2 at a given Hg concentration. Complete dissolution of the Al 1100 and Al 6061 T6 alloys up to a final Al concentration of 2 M was obtained using a 7 M HNO 3 solution containing a 0.002 M Hg catalyst. However, following the dissolutions, solids were observed in the solution. The solids were amorphous, but likely originated from the Si present in the alloys. No crystalline materials, such as Al(NO 3 ) 3 were observed. During the course of the dissolution experiments, it was determined that delaying the addition of Hg once the HNO 3 solution reached the boiling point can reduce the total offgas and H 2 generation rates. The delay in starting the Hg addition is not necessary for HFIR fuel dissolution, but could be useful in other research reactor dissolution campaigns. The potential to generate flammable concentrations of H 2 in the offgas during a HFIR fuel dissolution was evaluated using the experimental data. The predicted H 2 concentration in the dissolver offgas stream was compared with 60% of the calculated H 2 LFL at 200 °C using several prototypical experiments. The calculations showed that a full HFIR core can be dissolved using nominally 0.002 M Hg to catalyze the dissolution. The margin between the predicted H 2 concentration and the calculated LFL was greater when the solution was allowed to boil for 45 min prior to initiating the Hg addition. When the Hg was increased to 0.004 M, the predicted H 2 concentration exceeded the calculated LFL early in the dissolution. The dissolution experiments also demonstrated that additional Hg (beyond the initial 0.002 M) could be added as the Al concentration increases. The ability to add more Hg during a HFIR fuel dissolution could be beneficial if slow dissolution rates are observed at high Al concentrations. Experimental data were used to demonstrate that the predicted H 2 concentration in a dissolver was below 60% of the calculated LFL at 200 °C when 0.004 M Hg was used to catalyze the dissolution if the Al concentration is conservatively greater than 0.5 M. Data also show that the Hg concentration during a HFIR fuel dissolution can be increased from 0.002 to 0.008 M at an Al concentration of 1.3 M.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Development of Additively Manufactured Complex Tools for Autoclave Cure Composites

IACMI Project 4.9, Tooling for Composites with Washout Features Produced by Additive Manufacturing, assembled a team including the industry lead, Ability Composites, NREL and Colorado State University (CSU). Ability Composites had originally expressed interest in alternate methods of producing tooling for composite parts. In follow-up discussions, it became clear that one of the bigger tooling challenges revolved around small production volume composite parts that were tooled on washout material due to the complex geometry. To build an understanding of the potential, both from a technology and a cost perspective, for replacing conventional washout tooling with 3D printed thermoplastic tooling, a number of commercially available dissolvable FDM printing materials were evaluated, leading to tooling representative of commercial articles of interest to Ability Composites. Ultimately, Ability Composites was able to directly compare autoclave processed prepreg composite parts produced on conventional washout tooling to composite parts molded on 3D printed dissolvable tooling produced at CSU. Small, laboratory test specimens were developed to investigate the structural performance of the candidate materials under autoclave processing conditions, which were nominally 121 °C (250 °F) and 345 kPa (50 psi). In addition, several internal structural configurations (infills) were evaluated under autoclave conditions using model materials. The results of these tests indicated that two materials, Stratasys ST 130 and Infinite Materials Solutions Aquasys 180 (AQ 180), were the best candidates, given the specified autoclave processing conditions. ST-130 was slightly more robust than AQ-180; however, the AQ-180 was carried forward as it was dissolvable in water, not requiring the basic solution needed to dissolve ST-130. Based on the preliminary material and 3D printed structures evaluations, larger tools with a truncated square pyramid geometry were created to produce prepreg composite test articles for 3D printed dissolvable tool evaluation under standard autoclave fabrication conditions. Two tools were manufactured using ST-130 and one tool using traditional ceramic washout tooling media. The tools were evaluated for geometric fidelity and surface roughness changes before and after carbon fiber/epoxy prepreg composites were manufactured on the tooling. The autoclave processing did not impact the geometry significantly and was completed at 121 °C and 345 kPa, indicating satisfactory tool performance. The results from surface roughness testing of both the resulting composite and the associated tooling indicated that an adequate surface resulted without the need for a surface sealing step, as was required for the conventional washout tooling. Based on results of the truncated pyramid tests as a basis, ST-130, AQ-120 and AQ-180 materials were carried forward to the tool geometry of interest to Ability Composites. These hollow rectangular bent ducts, which were complex in nature and not extractable after cure, were used to understand the impacts of tool material and thickness. One ST-130 tool was produced as a partially solid part, with an enclosed 40% dense infill region to reduce weight and material use. This was the same approach evaluated in the truncated pyramid portion of the study. This tool was to be envelope vacuum bagged and directly compared to a monolithic tool of conventional washout material. The traditional monolithic ceramic tool was manufactured by Ability composites using CNC-based subtractive methods. An additional five dissolvable polymer tools, manufactured from ST-130, AQ-120, and AQ-180, using a hollow design were 3D printed and used to produce carbon fiber/epoxy composite evaluation articles. These hollow dissolvable tools were expected to be less influenced by the autoclave conditions as the wall was solid and vacuum bagging was inside and outside the tool. This alternative geometry was also evaluated as an option in techno-economic modeling. Print times were reduced from in excess of 3 days to under 30 hours, while surface quality and and tool integrity were substantially improved in the transition from the partially solid tool to the hollow tooling concept. Ability Composites produced autoclave-cured prepreg ducts on each of these tools. The autoclave conditions utilized were more severe than those of the initial trials, reaching temperatures of 160 °C and a pressure of 414 kPa. Under these conditions, the partially solid 3D printed tool with skin and 40% dense infill crushed significantly; however, the thicker ST-130 hollow tool showed good promise, deforming only slightly. The thinner hollow tool walls were unsuccessful as were the other materials. Overall, the hollow tool manufacturing process saved significant amounts of time and material in manufacturing as compared to the solid ducts and produced composite surface quality improvements compared to the traditional washout tooling. The TEM was developed to allow direct comparisons between conventional washout tool manufacturing processes and those developed at CSU. It also allowed for two separate 3D printed tool geometries to be analyzed and compared. In this case, the square bent duct tool geometry was determined to be representative of common washout tools. This geometry was compared with a scaled-up version of it to assess differences in the two manufacturing processes based on tool size. The model was developed to make use of user input in the form of geometry details, process steps, manufacturing parameters, bulk material costs, capital equipment costs, and general costs to calculate overall labor, material, capital equipment, and energy costs per manufactured tool for the conventional and additive manufacturing processes for the two representative geometries. It was also able to estimate step-by-step process times for the manufacturing process and geometries. Based on significant input from Ability Composites and CSU from their knowledge gained from hands-on manufacturing of the 3D printed bent duct tool geometry, costs and process times were calculated for the two manufacturing processes. Results showed that the additive manufacturing techniques developed at CSU can substantially reduce the costs of tool manufacturing by reducing labor times and material usage. This is because additive manufacturing is a relatively hands-off process and allows for the tool design to be optimized to reduce material usage. The disadvantage, however, is that process times for additive manufacturing are significantly longer. The three-dimensional (3D) printing process is slow if tight tolerances are required, but the analysis did show that print times could be reduced with the hollow tool geometry. Also, further advances in additive manufacturing could expedite the process. Costs and process times for the tool washout process were calculated separately. They showed that costs are relatively insignificant when compared to the overall tool manufacturing processes, but with increases in tool size, costs for the conventional manufacturing approach are larger than for additive manufacturing. Again, the washout process for conventional tools is very hands-on, whereas for additively manufactured tools the print medium is dissolved in an automated detergent bath at the sacrifice of process time. The analysis showed that optimizing the additively manufactured tools may also reduce washout times. Overall, Project 4.9 demonstrated that commercially available dissolvable 3D printing materials exist that can be used to produce dissolvable tooling capable of surviving prepreg composites fabrication under autoclave conditions of 121 °C (250 °F) and 345 kPa (50 psi). An alternative hollow dissolvable tool design was developed which was structurally superior to the initial concept and was cost and time effective versus conventional washout tooling. The 3D printed sacrificial tool required no added surface sealing steps prior to composite part layup and cure, offering a significant advantage over the porous conventional washout tooling.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Dissolution of a Can Carrier Pin in Concentrated Nitric Acid

H=Canyon will be dissolving Fast Critical Assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA). The FCA fuel will be dissolved in the electrolytic dissolver in a solution that is a 50 wt.% nitric acid (HNO 3 ), 0.5 g/l gadolinium (Gd), and 0.05 M potassium fluoride (KF). The nitric acid concentration is expected to decrease during the batch process from 10.3 to 7 M. The temperature of the nitric acid solution may be as low as 15 °C. The fuel can will be placed in the dissolver basket insert utilizing a reusable charging device. The charging device is a coated stainless-steel rod with a clevis design. The charging device employs a linchpin to secure the FCA fuel can as it is being charged into the H-canyon dissolver. Prior to beginning the electrolytic dissolution process, the pin will be dissolved, the fuel can will remain in the dissolver basket insert and the charging device will be removed. Currently the time necessary for complete dissolution of the pin is unknown and thus the timing of the removal of the charging device cannot be planned. This process is to be performed remotely and therefore complete dissolution of the pin will not be able to be visually determined. The pin will be made of a material that dissolves in the concentrated nitric acid solution in the dissolver. The facility desired to know the time that the pin would be dissolved so that the charging device for the FCA can could be removed from the dissolver. In particular, the pin dissolution time as a function of the nitric acid concentration and the temperature was desired. The facility would like to ensure that the solution environment was such that the charging device could be removed within an operational shift (8-12 hours).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Dissolution Flowsheet for Non-Aluminum Spent Nuclear Fuel Campaign 1

As part of the Accelerated Basin De-inventory (ABD) program, H Canyon plans to dissolve non-aluminum spent nuclear fuel (NASNF) in the 6.3D electrolytic dissolver. NASNF Campaign 1 plans to electrolytically dissolve 68 bundles of fuel assemblies from the Carolinas-Virginia Tube Reactor (CVTR), Heavy Water Components Test Reactor (HWCTR), and Experimental Boiling Water Reactor (EBWR). The fuel assemblies are intact Zircaloy or stainless steel (SS) clad UO 2 rods, tubes, and plates. The H Canyon electrolytic dissolver previously dissolved a variety of UO 2 core fuel types in SS, Zircaloy, Nichrome, or Incoloy cladding from 1969 to 1980. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely dissolve NASNF Campaign 1 bundles in the H Canyon electrolytic dissolver. Bench-scale electrolytic dissolution tests were performed to demonstrate a flowsheet for NASNF Campaign 1 bundles. The outer bundles are composed of SS or Al alloy, Al 6061-T6, and contain intact Zircaloy or SS clad UO 2 fuel assemblies. The key objectives of these tests were to determine bounding dissolver chemistries and the sparge requirement to ensure H 2 concentration remain less than 60 vol % of the lower flammability limit (LFL) during dissolution. The impact of HNO 3 concentration and the addition of fluoride on the dissolution efficiency of Zircaloy, 304L SS, Al 6061-T6, and Inconel 625 were examined. While SS, Al, and Inconel 625 readily dissolve utilizing electrolytic dissolution, Zircaloy disintegrated anodically; the surface of Zircaloy oxidized and the oxide layer spalled off and settled at the bottom of the dissolver as an insoluble material. The black flakes were identified as ZrO 2 and 85% of the Zr processed was converted to black ZrO 2 flakes when Zr was anodically disintegrated in 9.5 M HNO 3 .

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Diffusional motion of redox centers in carbonate electrolytes

Ferrocene (Fc) and N-(ferrocenylmethyl)-N,N-dimethyl-N-ethylammonium bistrifluoromethyl-sulfonimide (Fc1N112-TFSI) were dissolved in carbonate solvents and self-diffusion coefficients (D) of solutes and solvents were measured by {sup 1}H and {sup 19}F pulsed field gradient nuclear magnetic resonance (NMR) spectroscopy. The organic solvents were propylene carbonate (PC), ethyl methyl carbonate (EMC), and a ternary mixture that also includes ethylene carbonate (EC). Results from NMR studies over the temperature range of 0–50 °C and for various concentrations (0.25–1.7 M) of Fc1N112-TFSI are compared to values of D simulated with classical molecular dynamics (MD). The measured self-diffusion coefficients gradually decreased as the Fc1N112-TFSI concentration increased in all solvents. Since TFSI{sup −} has fluoromethyl groups (CF{sub 3}), D{sub TFSI} could be measured separately and the values found are larger than those for D{sub Fc1N112} in all samples measured. The EC, PC, and EMC have the same D in the neat solvent mixture and when Fc is dissolved in EC/PC/EMC at a concentration of 0.2 M, probably due to the interactions between common carbonyl structures within EC, PC, and EMC. A difference in D (D{sub PC} < D{sub EC} < D{sub EMC}), and both a higher E{sub a} for translational motion and higher effective viscosity for PC in the mixture containing Fc1N112-TFSI reflect the interaction between PC and Fc1N112{sup +}, which is a relatively stronger interaction than that between Fc1N112{sup +} and other solvent species. In the EC/PC/EMC solution that is saturated with Fc1N112-TFSI, we find that D{sub PC} = D{sub EC} = D{sub EMC} and Fc1N112{sup +} and all components of the EC/PC/EMC solution have the same E{sub a} for translational motion, while the ratio D{sub EC/PC/EMC}/D{sub Fc1N112} is approximately 3. These results reflect the lack of available free volume for independent diffusion in the saturated solution. The Fc1N112{sup +} transference numbers lie around 0.4 and increase slightly as the temperature is increased in the PC and EMC solvents. The trends observed for D from simulations are in good agreement with experimental results and provide molecular level understanding of the solvation structure of Fc1N112-TFSI dissolved in EC/PC/EMC.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Dissolution flowsheet for high flux isotope reactor fuel

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3 O 8 ) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2 . The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H 2 ) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H 2 and other permanent gases in the dissolution offgas allowing the development of H 2 generation rate profiles. The H 2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the calculated lower flammability limit (LFL) for H 2 at a given Hg concentration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Evaluation of the Dissolution Behavior of L-Bundle End Caps and HFIR Fuel Carriers

The H-Canyon facility is currently using the 6.1D dissolver for the dissolution of Material Test Reactor (MTR) fuel and the 6.4D dissolver for the dissolution of High Flux Isotope Reactor (HFIR) fuel using mercury-catalyzed HNO₃ dissolution flowsheets. The processing strategy for both dissolvers involves the dissolution of multiple charges of fuel per batch. After the designated heating cycle, the dissolvers are opened, and the charging wells are probed to determine if the MTR or HFIR fuel has dissolved. If undissolved fuel fragments are beyond a certain height, the dissolver must be closed and heated for an additional amount of time to dissolve the remaining material. In recent MTR fuel dissolutions, “high probes” (i.e., excessive undissolved material) were frequently observed, which resulted in extended dissolution times. The suspected cause of the high probes was the incomplete dissolution of the L-Bundle End Cap, rather than the fuel or fuel bundles. The End Cap is hypothesized to be binding in the insert well and not dropping into the acid as the L-Bundle and fuel dissolve. Once the End Cap is dislodged by the probe and drops into the acid, the dissolution rate of the End Cap appears to be significantly reduced compared to the dissolution rate of the fuel and other parts of the L-Bundle. A similar issue has also been observed with the lifting bail on the HFIR fuel carriers. During HFIR fuel dissolutions, the lifting bail on the outer carrier has resulted in high probes due to incomplete dissolution. In one case, a partially dissolved bail was caught in one of the insert well holes which prevented the probe from going to the bottom of the well.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Combining RHF and HFIR Disposition Campaigns - Analysis, Opportunity, and Lessons Learned

Savannah River Nuclear Solutions has tasked its H Canyon facility to prepare for disposition of used Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) cores. HFIR cores consist of an inner and outer assembly of highly enriched uranium fuel plates. The two elements are annular, nest inside each other, and the fuel is a curved plate held between two collars. The H-Canyon facility has a dissolver insert specifically designed for HFIR cores and has performed this type of campaign before between 1976 and 1988. At H-Canyon, a dissolver is a large stainless steel pot loaded with a nitric acid solution into which uranium-alloy fuel is dissolved. The dissolver uses a device called an insert to feed the fuels to the acid solution in a controlled manner. The mission will be to dissolve a number of HFIR cores, purify the uranium, and down blend it to low enriched uranium suitable for commercial power reactors. HFIR has a sister core. A French high flux research reactor, called Reactor A Haut Flux (RHF), has nearly the same geometry as the outer element of HFIR. The annulus is only slightly smaller and the RHF active fuel height is {approx}97 cm while the HFIR active fuel height is only {approx}51 cm. SRNS is currently storing RHF cores (repatriated through circa 2003) awaiting a path for disposition. There is now an opportunity. RHF cores fit in the HFIR dissolver insert. Dispositioning them concurrent with HIFR avoids a future facility evolution of just to change dissolver inserts again. This work is the result of an initiative to propose appending the RHF cores to the anticipated HFIR campaign. Geometry and chemistry differences are evaluated and criticality safety assessed. The RHF cores were semi-explicitly modeled in various stages of dissolution. Normal and credible abnormal conditions were assessed. RHF was shown to be bounded by the dissolution of a complete HFIR core in both neutron multiplication and dissolver chemistry. Computational modeling was performed using KENOVI in the SCALE 6.1 code package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Trace Metal Uptake and Mercury Methylation by Sediments from a Stream in Tennessee

The formation and transport of methylmercury (MeHg), a neurotoxin, in aquatic environments is a global concern for human health as MeHg can bioaccumulate and biomagnify to high concentrations in aquatic food webs. MeHg is formed by conversion from inorganic mercury through microbial mediated methylation. Sulfate-reducing bacteria have been identified as the primary organisms responsible for MeHg production. Pure-culture studies suggest that low availability of cobalt and copper may inhibit mercury methylation, but whether such limitations occur in the environment is unclear. To explore the possible interaction between trace metal availability and mercury methylation, sediments from the East Fork Poplar Creek in Oak Ridge, Tennessee were sampled and then incubated in the presence and absence of added dissolved cobalt and copper. Three types of data are provided in this package. The first reports the uptake of dissolved cobalt and copper by these stream sediments on short time scales (24 hours) in the form of final dissolved and adsorbed concentrations. The second data component consists of a time series of dissolved concentrations and pH values for stream sediments incubated with artificial stream water containing different addition levels of dissolved cobalt or copper. The dissolved concentrations reported include total iron, manganese, sulfur, phosphorus, nickel, zinc and cobalt, dissolved concentrations of sulfate and orthophosphate, and the amount of cobalt or copper adsorbed by the sediment. The third data component reports data at 0 and 72 hours of incubation time for stream sediments to which cobalt was added. These data include concentrations of methylmercury with isotope labeling to enable determination of methylation and demethylation rates as well as dissolved concentrations of sulfate, phosphate, chloride, iron, cobalt, and organic carbon. All data are provided in text-based CSV format with header sections indicating the data contained in each file and the corresponding units. Note that "u" is used in place of Greek lower-case mu to indicate the micro prefix on units. A Table of Contents file (Data_package_TableofContents.txt) provides an index for the data contained in the individual files.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2011

Characterization of Tank 16H Annulus Samples Part II: Leaching Results

The closure of Tank 16H will require removal of material from the annulus of the tank. Samples from Tank 16H annulus were characterized and tested to provide information to evaluate various alternatives for removing the annulus waste. The analysis found all four annulus samples to be composed mainly of Si, Na, and Al and lesser amounts of other elements. The XRD data indicate quartz (SiO 2 ) and sodium aluminum nitrate silicate hydrate (Na 8 (Al 6 Si 6 O 24 )(NO 3 ) 2 .4H 2 O) as the predominant crystalline mineral phases in the samples. The XRD data also indicate the presence of crystalline sodium nitrate (NaNO 3 ), sodium nitrite (NaNO 2 ), gibbsite (Al(OH) 3 ), hydrated sodium bicarbonate (Na 3 H(CO 3 ) 2 .2H 2 O), and muscovite (KAl 2 (AlSi 3 O 10 )(OH) 2 ). Based on the weight of solids remaining at the end of the test, the water leaching test results indicate 20-35% of the solids dissolved after three contacts with an approximately 3:1 volume of water at 45°C. The chemical analysis of the leachates and the XRD results of the remaining solids indicate sodium salts of nitrate, nitrite, sulfate, and possibly carbonate/bicarbonate make up the majority of the dissolved material. The majority of these salts were dissolved in the first water contact and simply diluted with each subsequent water contact. The water leaching removed large amounts of the uranium in two of the samples and approximately 1/3 of the 99 Tc from all four samples. Most of the other radionuclides analyzed showed low solubility in the water leaching test. The oxalic acid leaching test result indicate approximately 34-47% of the solids in the four annulus samples will dissolve after three contacts with an approximately 3:1 volume of acid to solids at 45°C. The same sodium salts found in the water leaching test comprise the majority of dissolved material in the oxalic acid leaching test. However, the oxalic acid was somewhat more effective in dissolving radionuclides than the water leach. In contrast to the water leaching results, most constituents continued to dissolve during subsequent cycles of oxalic acid leaching. The somewhat higher dissolution found in the oxalic acid leaching test versus the water leaching test might be offset by the tendency of the oxalic acid solutions to take on a gel-like consistency. The filtered solids left behind after three oxalic acid contacts were sticky and formed large clumps after drying. These two observations could indicate potential processing difficulties with solutions and solids from oxalic acid leaching. The gel formation might be avoided by using larger volumes of the acid. Further testing would be recommended before using oxalic acid to dissolve the Tank 16H annulus waste to ensure no processing difficulties are encountered in the full scale process.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗