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At least 271 records · Page 15

An engineering perspective on evaluating mechanisms governing ductility in pure molybdenum

The United States lacks a stable domestic supply of 99 mTc, a critical medical imaging isotope generated from 99 Mo. Accelerator-based production using 100 Mo targets (aMo) introduces mechanical concerns due to the ductile-to-brittle transition temperature inherent to refractory metals like Mo. This study evaluates the tensile behavior of aMo targets from 25 to 1000 °C, compared to powder-metallurgy-processed natural Mo (PMo) and cast-and-rolled Mo (RMo), both as-received and after 5 ppm O 2 exposure in flowing He. RMo showed superior ductility (7.3% at 25 °C, 48% at 1000 °C) and strength, attributed to its fine, elongated grains and high geometrically necessary dislocation density. PMo exhibited variable ductility (up to 33%), while aMo remained brittle, with a maximum elongation of 10.4% at 600 °C. EBSD analysis revealed weak texture in PMo and aMo, but high defect density in aMo limited dislocation mobility. This work links processing, microstructure, and deformation mechanisms to guide fabrication of ductile refractory targets.

Hyer, Holden C. [Oak Ridge National Laboratory (OR↗

Molybdenum Recovery from Filters Used in Large-scale Dissolution of Sintered Mo-disks

99 Mo production with linear accelerators can be achieved via the bremsstrahlung photonuclear reaction 100 Mo(γ,n) 99 Mo or the neutron capture reaction 98 Mo(n,γ) 99 Mo. For commercial producers, maximum recovery of enriched 98 Mo and 100 Mo target material is critical for sustaining an economic production cycle. During the peroxide dissolution of Mo metal disks and subsequent conversion to K 2 MoO 4 , several grams of Mo can be lost during filtration from solution when several hundred grams of sintered Mo disks are processed. This investigation shows that 5–8 g Mo is routinely retained on the filter units, but it can be almost fully recovered using aqueous washes. Washing can be done immediately and incorporated into the dissolution procedure, or it can occur several months after the initial filtration process to decrease processing time.

07 ISOTOPE AND RADIATION SOURCES↗

Temperature Effect over Gas Bubble Evolution in Uranium-10 wt.% Molybdenum alloy Irradiated by Swift Xe Ions

Results from post-irradiation examination (PIE) of U-10Mo binary metallic fuel irradiated by 84 MeV Xe ions at two different temperatures (450°C & 275°C) have been evaluated and reported. The main goal of this study was to understand gas bubble formation (due to accumulation of Xe gas) and corresponding size evolution dependent upon net Xe deposition at two different temperatures, for the U-10Mo fuel samples. From the post ion irradiation examination analysis, it can be concluded that at 450°C the Xe gas bubbles seems to coarsen more easily compared to 275°C. The results of ion irradiation testing can provide initial data that may be beneficial for fuel system down-selections, and for the development and validation of fuel performance models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents AMP-PAN and CST R91290-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions, but CST should be investigated at a higher pH. Limited removal of transuranic species and almost no removal of uranium was observed in this testing. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in the range of 0.2 to 5 mg Cs per gram of absorbent appears to be achievable.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Small Punch Testing of Molybdenum-99 Targets

Northstar Medical Radioisotopes is developing an accelerator-based method to produce 99 Mo, which is a parent isotope of the commonly used 99m Tc medical isotope. The Mo targets being designed for the accelerator will be produced from enriched 100 Mo, also known as aMo . Pressed and sintered powder feedstock is used to fabricate aMo targets, producing 29 mm disk-shaped targets. The targets are subjected to 1–6 days in line of an electron beam with subsequent dissolution of the disk to retain the 99 Mo, which decays to 99m Tc at radio-pharmacies. The press and sinter method is advantageous because the inherent porosity produced by this method enables increased surface area and therefore increased flow of dissolution media, decreasing the dissolution time and reducing the need for a highly acidic media. Although porosity aids in dissolution, it reduces the mechanical strength and ductility. Targets require good mechanical integrity when subjected to the conditions in the accelerator. Therefore, Northstar is seeking methods to rapidly test disk samples after fabrication to assure mechanical performance metrics are achieved. This report details the design and testing of a small punch test (SPT) that accommodates the 29 mm disk. Initial data were used to relate the SPT data to tensile properties, such as the yield strength (YS), ultimate tensile strength (UTS), and total elongation to failure. Although the SPT has been established as a somewhat reliable method for testing metallic materials, few studies have applied the SPT to refractory materials such as Mo. Based on tensile testing performed at Oak Ridge National Laboratory on different Mo samples, correlation between the Mo tensile and SPT properties could be performed, establishing a standard calibration that could be applied to other Mo samples. To test the efficacy of the SPT with Mo, multiple different disk batches were fabricated under different conditions (e.g., pressure, lubricant) with commercially available pure Mo powder. Generally, only a UTS could be well defined because the press and sinter disks failed under brittle fracture, making it difficult to determine the YS and elongation. Compared with disks fabricated with aMo powder, the aMo samples underperformed their pure nat Mo counterparts. This report summarizes the current status of the SPT, but further evaluation is needed before it can be applied as a reliable quality assurance tool.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production: Part II The Effect of Sodium Carbonate Adjustment to pH 9 on Strontium Removal

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents – AMP-PAN and CST R9120-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions. CST was shown to be useful at removing both Cs and Sr after adjustment of the acidic sulfate solution with sodium carbonate to pH 9. Limited removal of transuranic species and almost no removal of uranium was observed for CST R9120B absorbent. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in excess of 0.7 mg Cs per gram of CST absorbent while loading 0.14 mg Sr per gram of CST absorbent appears to be achievable from pH 9 carbonate solution.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Completion of IER 305: Molybdenum Sleeve Experiments and Preparations for Performing IER 441: Epithermal Tantalum Experiments [Slides]

This presentation includes a look into Sandia critical experiments including the 7uPCX, BUCCX, and assembly design. This presentation touches on the completion of IER 305 with CED-3b, CED-4a, and CED-4b. Finally, there are preparations to perform IER 441 including new hardware, critical configurations, and next steps.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Control of iodine and tellurium for use in domestic molybdenum-99 production efforts

Three initial proof-of-concept experiments were performed to test the removal of I and Te from solutions containing Mo for potential abatement during chemical processing to purify 99 Mo. All three proof-of-concept experiments produced results indicating that further optimization could provide methods for I and/or Te abatement from multiple matrices. Decamp and Happel (2013) describe the use of CL resin from Eichrom Technologies loaded with Ag + combined with Amberlite XAD-4 in the chromatographic separation of irradiated fission products in 1 M HNO 3 to separate iodine species where iodide and iodate react with Ag + and I2 adheres to the Amberlite resin. The method worked well under processing conditions in a medical isotope facility although many separation methods use higher concentrations of HNO 3 which could adversely impact the Ag + loading. The experiment performed here used a roughly 2 mL column of CL resin loaded with Ag + and Amberlite XAD-4 to separate µg quantities of I, Te, and Mo in 3 M HNO 3 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Isotopic Evolution of the Inner Solar System Inferred from Molybdenum Isotopes in Meteorites

The fundamentally different isotopic compositions of non-carbonaceous (NC) and carbonaceous (CC) meteorites reveal the presence of two distinct reservoirs in the solar protoplanetary disk that were likely separated by Jupiter. However, the extent of material exchange between these reservoirs, and how this affected the composition of the inner disk, are not known. Here we show that NC meteorites display broadly correlated isotopic variations for Mo, Ti, Cr, and Ni, indicating the addition of isotopically distinct material to the inner disk. The added material resembles bulk CC meteorites and Ca–Al-rich inclusions in terms of its enrichment in neutron-rich isotopes, but unlike the latter materials is also enriched in s-process nuclides. The comparison of the isotopic composition of NC meteorites with the accretion ages of their parent bodies reveals that the isotopic variations within the inner disk do not reflect a continuous compositional change through the addition of CC dust, indicating an efficient separation of the NC and CC reservoirs and limited exchange of material between the inner and outer disk. Instead, the isotopic variations among NC meteorites more likely record a rapidly changing composition of the disk during infall from the Sun's parental molecular cloud, where each planetesimal locks the instant composition of the disk when it forms. Here, a corollary of this model is that late-formed planetesimals in the inner disk predominantly accreted from secondary dust that was produced by collisions among pre-existing NC planetesimals.

58 GEOSCIENCES↗