Engineering topics

Rudisill, Tracy S.

Publications and source records attributed to Rudisill, Tracy S..

13 records

DOE OSTI2024

Purification of U from U-10Mo scrap generated during the fabrication of high performance research reactor fuel

A low enriched U-Mo alloy fuel is under development to replace highly enriched U fuels currently used in United States high performance research reactors. The alloy casting and fuel fabrication processes will generate scrap streams containing low enriched U (LEU) which must be recovered. Solvent extraction processes were designed using the Argonne Model for Universal Solvent Extraction (AMUSE) to purify solutions containing 20 and 50 g/L U. The feed for the solvent extraction processes was prepared from solutions generated from the dissolution of U-10Mo-Zr foils and U-10Mo-Zr-Al mini-plates. The U purification processes were demonstrated using two, 16-stage banks of miniature mixer-settlers. The solvent extraction experiments demonstrated that all design objectives for the U purification processes could be met. The U recovery in the product stream for each flowsheet was ≥99.9%. The flowsheet demonstrations also showed that the purity of the U Product will meet the requirements of the ASTM International C1462-21 specification for LEU metal enriched to less than 20% 235 U. In conclusion, the AMUSE modeling for both flowsheet demonstrations was validated by comparing predicted and measured concentrations of U, Mo, and Zr in the exit streams and stage samples at steady-state conditions in the mixer-settlers.

modified PUREX process↗

DOE OSTI2024

Evaluation of the Potential for Precipitation of Solids during Storage of Non-Aluminum SNF Solutions

Non-aluminum clad spent nuclear fuels (NASNF) stored in the L-Area basin will be dissolved in H-Canyon using the 6.3D electrolytic dissolver. The solutions will be stored in either the hot or warm canyon until the preparation of a sludge batch for the Defense Waste Processing Facility. Spent nuclear fuel solutions could be stored for 1-2 years before transfer to the H-Area Tank Farm depending on the interval between sludge batches. The solution level in the storage tanks will be maintained; therefore, precipitation of solids due to evaporation is not an issue. However, the precipitation of solids from completely dissolved SNF due to solution instabilities has been observed during intermediate storage of solutions generating hydrated oxides.The presence of fissile material in these solids is generally associated with zirconium molybdate, which is known to act as a host lattice for Pu and can carry the actinides upon precipitation. The formation of zirconium molybdate solids which carry fissile material is a potential concern for the storage of NASNF solutions. To address this concern, the Savannah River National Laboratory performed a literature review to identify knowledge gaps which may require experimental work to determine if the formation of solids is a concern during storage of these solutions. Based on the literature review, the precipitation of zirconium molybdate solids from the Campaign 1 NASNF solutions during intermediatestorage is expected. This conclusion is supported by the identification of zirconium molybdate solids found on the H-Canyon 6.1D Dissolver MK-12 insert spacer. The formation of the zirconium molybdate solids is attributed to hydrolysis and radiolytic processes in the nitric acid solution. As the molybdate solids form, U and Pu can substitute for Zr in the crystal lattice resulting in co-precipitation. Generally, the Pu substitutes directly into the crystal lattice during precipitation while the U associated with the molybdate solids more likely absorbs from the solution. The U in the NASNF solutions is present as uranyl nitrate, a 2+ cation which will not substitute as easily into the molybdate crystal lattice for the Zr 4+ ion.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Metallurgical Analysis of the High Flux Isotope Reactor (HFIR) Carrier Lifting Bails (Rev.1)

The dissolution rates of the aluminum alloys in the High Flux Isotope Reactor (HFIR) element carriers and the Material Test Reactor (MTR) L-bundles in the H-Canyon facility have been identified as the possible cause of extended dissolutions that result in significant time and financial expenditures. A study, carried out by Savannah River National Laboratory (SRNL) to determine relationships between the dissolution rates and the metallurgical properties of the aluminum alloy materials of construction of the HFIR carriers and the L-bundles, considered the dissolution rates of aluminum alloy (AA) series 1100, 6061, and 6063. The study determined that the aluminum alloy compositions played a principal role in the dissolution rate of the carrier/bundle components. Higher dissolution rates were correlated with lower concentrations of the minor element additions in the alloys and with specific element concentrations. Aluminum alloys 1100 and 6063 were found to have similar dissolution rates that were approximately two orders of magnitude (100X) greater than those of AA6061. Based on the results of the dissolution behavior study, a Technical Assistance Request (TAR) was first issued to determine if the replacement of AA6061-T6 with AA6063-T6 is feasible for the HFIR carrier lifting bails. A Technical Task Request was then issued to consider AA6063-T5 as well as other alloys to improve possible supply chain issues. The metallurgical properties of the L-bundle (specifically the end caps) were not evaluated in this report because L-Bundle drawings already allow for the use of AA6063-T6 in all structural components. The HFIR carriers are composed of thin-walled components with significant surface areas that allow for relatively quick overall dissolution times. Conversely, the carrier lifting bails and the supporting constituents are composed of solid bars and thick plate regions with relatively small surface areas that experience longer overall dissolution times. While the MTR L-bundle design includes allowances for the materials of construction to be either AA6061-T6 or AA6063-T6, the HFIR carriers are specified to be constructed fully with AA6061-T6 alloy. This report analyzes the recommendations of the dissolution behavior study to replace the materials of construction of the HFIR carrier lifting bails. The analysis considers the operational requirements of the lifting bail and its supporting structures. To decrease dissolution times, the analysis considers direct replacement of the material as well as reductions in the thicknesses of the components to decrease the mass of the elements. Material reductions are considered on options for using either AA6061 and/or AA6063. The calculations are based on specifications from the American Society of Mechanical Engineer (ASME) and The Aluminum Association, Inc. design codes. The analysis finds that direct replacement of the lifting bail material of construction with AA6063-T6, and AA6063-T5 as well as reductions in the dimensions of the lifting bail components are acceptable. Note that this study considers the structural suitability of the alloys. It does not consider their dissolution rates in the dissolvers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

H-Canyon Flowsheet for the Neutralization of Dissolved Fast Critical Assembly (FCA) Fuel

The H-Canyon facility will be dissolving unirradiated stainless-steel clad Fast Critical Assembly (FCA) fuel in HNO3 and neutralizing the resulting solution with 50 wt% NaOH with no recovery operations prior to transfer to the Concentration, Storage, and Transfer Facility (CSTF). There are two types of FCA fuel, including a Pu-Al metal alloy and a mixed U and Pu oxide. H-Canyon anticipates dissolving 16 batches of the metal fuel and two batches of the oxide fuel. Potassium fluoride will be added to the dissolver solution at 0.05 M to promote dissolution. Gadolinium will be added after the dissolution as a thermal neutron poison for criticality control. H-Canyon requested that SRNL evaluate Gd:239Pu equivalent (239PuEq) ratios of 1:1 and 10:1. The neutralization process precipitates metals including actinides, Gd, and cladding components resulting in a slurry. Caustic neutralization is a routine H-Canyon operation, but the FCA dissolver solution will be unique relative to solutions that are typically processed due to the stainless-steel components, high HNO3 concentration of ~8.5 M, and an initial Pu concentration of up to 4 g/L. Previous neutralization studies have been performed for Pu containing solutions but at less than half the initial Pu concentration and much lower initial HNO3 concentrations. Experimental neutralizations targeted a final free hydroxide (OH-) concentration of 0.6 M as this was anticipated to be the final endpoint, but H-Canyon now expects the endpoint to be 1.2 M OH-. The purpose of this study was to characterize the distribution of the actinide, Gd, and cladding components between the precipitate and supernate, determine if the slurry will back up in the header during the transfer to the CSTF, and determine if solids will settle in the pipeline during the transfer.

Mills, Matthew S.↗

DOE OSTI2023

Analysis of H-Canyon Process Tanks in Preparation of Consolidation and Blending for HALEU Fuels

High-Assay Low- Enriched Uranium (HALEU) fuels are being developed to support the replacement of Highly Enriched Uranium (HEU) fuels used in U.S. High-Performance Research Reactors (USHPRR) as well as advanced nuclear power reactor designs. The projected demand for HALEU far exceeds the supply and studies are underway to assess various options to partially mitigate the potential short supply. The H Canyon facility at the Savannah River Site (SRS) Low Enriched Uranium (LEU) containing 4.95% U-235 from the reprocessing of highly enriched foreign and domestic research reactor fuel for the Tennessee Valley Authority’s (TVA) commercial power reactor market for several decades. The production of LEU at the H-Canyon facility can be readily transitioned to produce 19.75% HALEU solutions from the current separated inventory of purified HEU solutions in H-Canyon storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Direct Extraction of Lanthanide Oxides and Nitrates in Tributyl Phosphate

This work investigated the dissolution rate of lanthanide oxides and nitrates in a 30 vol % TBP-n-paraffin solvent (pre-equilibrated with various nitric acid concentrations) using visible (Vis) spectroscopy over time. Some dissolution mechanics were observed, such as an aqueous layer forming, considerably longer dissolution times for the heavier lanthanides (hours) vs. the lighter lanthanides (minutes), and the impacts of mixing lanthanide oxides or nitrates during dissolution. Neodymium, samarium, holmium, and erbium were selected due to their unique spectroscopic signatures and to represent the lighter (neodymium and samarium) and heavier (holmium and erbium) lanthanides. Even though europium does not have a strong absorbance in the range studied, europium was used in some instances to also represent the lighter lanthanides. Cerium oxide was used to representant dissolution of tetravalent lanthanides.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Demonstration of the Blending of the H-Canyon 2nd U Cycle with Natural U to Produce a HALEU Solution

Continued growth in the demand for high assay low enriched uranium (HALEU) is projected to support the conversion of US high performance research reactors from high to low enrichment and the development and deployment of a new generation of advanced reactors. Current options to meet the growing US need for HALEU include (1) blend-down of excess of highly enriched uranium (HEU) from the nuclear weapons programs, (2) blend-down of HEU metal from other stockpiles, or (3) recycle of irradiated HEU recovered from spent nuclear fuel (SNF). The recycle of HEU from SNF has been performed at the Savannah River Site (SRS) H-Canyon facility for nearly two decades to produce 4.95 wt % U-235 for sale and subsequent use in fuel for Tennessee Valley Authority (TVA) reactors. The HALEU produced in H-Canyon will not exceed the U-236 concentration limit in the ASTM C1462 standard for U metal enriched to more than 15% and less than 20% U-235, which applies to advanced reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

The Potential Interactions of Novec 1230 and Its Thermal Degradation Products with Actinide Metals and Oxides

The environmentally friendly and mainly chemically inert fire suppressant, Novec 1230, is being increasingly used worldwide. However, no published study has considered using Novec 1230 with fires involving radioactive material including actinides; here, therefore, this research note focuses on the possible interactions between Novec 1230 and its thermal degradation products (TDPs) with some actinide compounds (e.g., actinide metal and oxides) commonly present in a radiological facility in the event of a fire. Previous studies of halogenated fire suppressants used in actinide metal fires indicate a possibility for Novec 1230 or its TDPs to chemically interact with actinide compounds at elevated temperatures; however, these reactions are highly unlikely to cause either runaway exothermic reactions or gaseous actinide release.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

Catalytic Effects of Silver in Iodine Reactors for Dissolved Used Nuclear Fuel

The dissolution of used nuclear fuel generates a variety of off-gasses including flammable hydrogen and other species that are a concern for environmental release. The H-Canyon facility at the Savannah River Site is currently dissolving aluminum-clad research reactor fuel from material test reactors and the High Flux Isotope Reactor (HFIR) using a mercury-catalyzed nitric acid flowsheet. Savannah River National Laboratory recently developed and deployed a Raman spectrometer to monitor the off-gas stream from the dissolution process. Results from these measurements indicated a lack of the expected hydrogen, nitrous oxide, and nitric oxide in the off-gas stream. It was proposed that the silver on the silver nitrate–coated berl saddles present in the reactors for iodine capture were acting as a catalytic hydrogen recombiner. Nitric oxide is readily oxidized to nitrogen dioxide under normal conditions, but it was unclear what happened to the nitrous oxide. A laboratory-scale iodine reactor was assembled and filled with silver nitrate–coated berl saddles to help ascertain the fate of nitrous oxide and hydrogen. Testing with this laboratory-scale reactor observed the recombination of hydrogen when a simulated dissolver off-gas was passed through the reactor containing silver nitrate–coated berl saddles at the approximate temperatures seen in H-Canyon. However, the nitrous oxide concentration was unchanged, suggesting a more complex process occurring within the off-gas stream before it reaches the iodine reactors at H-Canyon.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Preliminary Result from the Dissolution of Neodymium and Erbium Oxide in a Tributyl Phosphate Solvent

This work demonstrated that visible absorbance spectroscopy can track the dissolution of neodymium and erbium oxide in an organic solution containing tributyl phosphate and nitrates. The formation of an aqueous phase was unexpected but an important phenomenon to consider when developing a metal oxide dissolution in organic solution flowsheet. Since the metal oxide appeared to completely dissolve into either the organic or newly formed aqueous phase, dissolution (extraction) rates should be easier to determine.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Dissolution Flowsheet for Skull Oxide Generated during U-Mo Alloy Casting for High Performance Research Reactor Fuel

The Savannah River National Laboratory was requested to develop dissolution flowsheets for high assay low enriched U scrap generated during the fabrication of high performance research reactor fuel. The scrap streams include U-10Mo and U-10Mo-Zr foils, rejected Al-clad fuel plates, and skull oxide from casting molds. Flowsheets for the dissolution of the U-10Mo-Zr foils and Al-clad U-10Mo-Zr mini-plates received from BWX Technologies, Inc were developed and demonstrated in the first phase of this project. In the second phase (this work), the skull oxide from a U-10Mo casting mold was obtained from the Y-12 National Security Complex (Y-12) and small-scale experiments were performed to demonstrate dissolution flowsheets.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Solvent Recovery and Management at the Savannah River Site H-Canyon Facility

NIOWAVE, Inc. is a domestic supplier of medical and industrial isotopes from uranium and radium. The Savannah River National laboratory (SRNL) is currently providing support to NIOWAVE, which plans to deploy a superconducting electron linear accelerator (LINAC) to fission uranium for Mo-99 production without the need for a nuclear reactor or HEU. The uranium from the Mo-99 production targets will be purified using a modified PUREX (Plutonium Uranium Reduction Extraction) solvent extraction process to recover the uranium in the product stream. The uranium will then be precipitated as an oxalate which is calcined to U 3 O 8 to fabricate pellets for new Mo-99 targets. In previous support provided to NIOWAVE, the SRNL demonstrated a solvent washing process to remove degradation products from the tributyl phosphate (TBP) solvent used in the modified PUREX process under development for uranium recovery. To supplement this technology demonstration, NIOWAVE requested the SRNL to provide summary information on the solvent recovery and management activities which are used at the Savannah River Site (SRS) H-Canyon facility. An existing reference document for the reprocessing of irradiated HEU fuels at the SRS was used as the primary reference for the solvent management activities; although, other reference documents were used to provide supplementary information. The information provided includes a brief summary of the solvent degradation issues which have been observed in the H-Canyon solvent extraction cycles and resulting process safety concerns. The solvent recovery processes for the three cycles of solvent extraction used in the H-Canyon were subsequently described including the process equipment which consists of the continuous and batch solvent washers, pumps, and tanks. A final section is provided on the monitoring and analysis of solvent quality based on the previous work performed at the SRNL for NIOWAVE and past research and development activities performed to support the solvent extraction processes in both the SRS F-Canyon and H-Canyon facilities.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2021

Evaluation of Niowave's Proposed Solvent Washing Approach

Niowave, Inc., is a domestic supplier of medical and industrial isotopes from uranium (U) and radium (Ra). The company has recently entered into a cooperative agreement with the U.S. Department of Energy’s National Nuclear Security Administration (NNSA) and plans to deploy a superconducting electron accelerator (LINAC) to fission U for molybdenum-99 ( 99 Mo) production without the need for a nuclear reactor or highly enriched uranium (HEU). NNSA provided funding to the Savannah River National Laboratory (SRNL) to support Niowave in this effort. SRNL evaluated the application of the solvent washing process for Niowave.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗