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Title: Waste Management Strategies for Production of Mo-99

Abstract

Production of Mo-99 for medical isotope use is being investigated using dissolved low enriched uranium (LEU) fissioned using an accelerator driven process. With the production and separation of Mo-99, a low level waste stream will be generated. Since the production facility is a commercial endeavor, waste disposition paths normally available for federally generated radioactive waste may not be available. Disposal sites for commercially generated low level waste are available, and consideration to the waste acceptance criteria (WAC) of the disposal site should be integral in flowsheet development for the Mo-99 production. Pending implementation of the “Uranium Lease and Take-Back Program for Irradiation for Production of Molybdenum-99 for Medical Use” as directed by the American Medical Isotopes Production Act of 2012, there are limited options for disposing of the waste generated by the production of Mo-99 using an accelerator. The commission of a trade study to assist in the determination of the most favorable balance of production throughput and waste management should be undertaken. The use of a waste broker during initial operations of a facility has several benefits that can offset the cost associated with using a subcontractor. As the facility matures, the development of in-house capabilities can be expandedmore » to incrementally reduce the dependence on a subcontractor.« less

Authors:
 [1];  [1]
  1. Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)
Publication Date:
Research Org.:
Savannah River Site (SRS), Aiken, SC (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
1342717
Report Number(s):
SRNL-STI-2017-00036
TRN: US1701900
DOE Contract Number:
DE-AC09-08SR22470
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES; LOW-LEVEL RADIOACTIVE WASTES; MOLYBDENUM 99; ENRICHED URANIUM; WASTE MANAGEMENT; MOLYBDENUM; ISOTOPE PRODUCTION; Mo-99 waste treatment

Citation Formats

Cozzi, A., and Johnson, F. Waste Management Strategies for Production of Mo-99. United States: N. p., 2017. Web. doi:10.2172/1342717.
Cozzi, A., & Johnson, F. Waste Management Strategies for Production of Mo-99. United States. doi:10.2172/1342717.
Cozzi, A., and Johnson, F. Tue . "Waste Management Strategies for Production of Mo-99". United States. doi:10.2172/1342717. https://www.osti.gov/servlets/purl/1342717.
@article{osti_1342717,
title = {Waste Management Strategies for Production of Mo-99},
author = {Cozzi, A. and Johnson, F.},
abstractNote = {Production of Mo-99 for medical isotope use is being investigated using dissolved low enriched uranium (LEU) fissioned using an accelerator driven process. With the production and separation of Mo-99, a low level waste stream will be generated. Since the production facility is a commercial endeavor, waste disposition paths normally available for federally generated radioactive waste may not be available. Disposal sites for commercially generated low level waste are available, and consideration to the waste acceptance criteria (WAC) of the disposal site should be integral in flowsheet development for the Mo-99 production. Pending implementation of the “Uranium Lease and Take-Back Program for Irradiation for Production of Molybdenum-99 for Medical Use” as directed by the American Medical Isotopes Production Act of 2012, there are limited options for disposing of the waste generated by the production of Mo-99 using an accelerator. The commission of a trade study to assist in the determination of the most favorable balance of production throughput and waste management should be undertaken. The use of a waste broker during initial operations of a facility has several benefits that can offset the cost associated with using a subcontractor. As the facility matures, the development of in-house capabilities can be expanded to incrementally reduce the dependence on a subcontractor.},
doi = {10.2172/1342717},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Tue Jan 31 00:00:00 EST 2017},
month = {Tue Jan 31 00:00:00 EST 2017}
}

Technical Report:

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  • One of the missions of the Reduced Enrichment for Research and Test Reactors (RERTR) program (and now the National Nuclear Security Administrations Material Management and Minimization program) is to facilitate the use of low enriched uranium (LEU) targets for 99Mo production. The conversion from highly enriched uranium (HEU) to LEU targets will require five to six times more uranium to produce an equivalent amount of 99Mo. The work discussed here addresses the technical challenges encountered in the treatment of uranyl nitrate hexahydrate (UNH)/nitric acid solutions remaining after the dissolution of LEU targets. Specifically, the focus of this work is themore » calcination of the uranium waste from 99Mo production using LEU foil targets and the Modified Cintichem Process. Work with our calciner system showed that high furnace temperature, a large vent tube, and a mechanical shield are beneficial for calciner operation. One- and two-step direct calcination processes were evaluated. The high-temperature one-step process led to contamination of the calciner system. The two-step direct calcination process operated stably and resulted in a relatively large amount of material in the calciner cup. Chemically assisted calcination using peroxide was rejected for further work due to the difficulty in handling the products. Chemically assisted calcination using formic acid was rejected due to unstable operation. Chemically assisted calcination using oxalic acid was recommended, although a better understanding of its chemistry is needed. Overall, this work showed that the two-step direct calcination and the in-cup oxalic acid processes are the best approaches for the treatment of the UNH/nitric acid waste solutions remaining from dissolution of LEU targets for 99Mo production.« less
  • The goal of the Reduced Enrichment for Research and Test Reactors Program is to limit the use of high-enriched uranium (HEU) in research and test reactors by substituting low-enriched uranium (LEU) wherever possible. The work reported here documents our work to develop the calcining technologies and processes that will be needed for 99Mo production using LEU foil targets and the Modified Cintichem Process. The primary concern with the conversion to LEU from HEU targets is that it would result in a five- to six-fold increase in the total uranium. This increase results in more liquid waste from the process. Wemore » have been working to minimize the increase in liquid waste and to minimize the impact of any change in liquid waste. Direct calcination of uranium-rich nitric acid solutions generates NO 2 gas and UO 3 solid. We have proposed two processes for treating the liquid waste from a Modified Cintichem Process with a LEU foil. One is an optimized direct calcination process that is similar to the process currently in use. The other is a uranyl oxalate precipitation process. The specific goal of the work reported here was to characterize and compare the chemical reactions that occur during these two processes. In particular, the amounts and compositions of the gaseous and solid products were of interest. A series of experiments was carried out to show the effects of temperature and the redox potential of the reaction atmosphere. The primary products of the direct calcination process were mixtures of U 3O 8 and UO 3 solids and NO 2 gas. The primary products of the oxalate precipitation process were mixtures of U 3O 8 and UO 2 solid and CO 2 gas. Higher temperature and a reducing atmosphere tended to favor quadrivalent over hexavalent uranium in the solid product. These data will help producers to decide between the two processes. In addition, the data can be used to design appropriate off-gas systems for pilot and production facilities.« less
  • Technetium-99m is a widely used radiopharmaceutical. Its parent, Mo-99, is produced worldwide to supply this important isotope. One means to produce Mo-99 is by bombarding a Mo-100 target with an electron beam from a linear accelerator; the γ/n reaction on Mo-100 produces Mo-99. After dissolving Mo-100 enriched disks in hydrogen peroxide, the solution is converted to potassium molybdate (0.2 g-Mo/mL) in 5 M KOH. After milking the Tc-99m in the TechneGen generator over a period of 7-10 days, the molybdenum solution needs to be treated to recover valuable Mo-100 for production of sintered Mo disks. However, during the production ofmore » Mo-99 by (γ, n) reaction on the Mo-100 target, several byproducts are formed. Therefore, recycling Mo will require the conversion of K 2MoO 4 in 5 M KOH solution to MoO 3 powder, and purification from other metals present in the Mo solution. The starting Mo-100 enriched material contains less than 20 mg of potassium in 1 kg of molybdenum (<20 ppm). However, after dissolving the irradiated Mo-100 target in hydrogen peroxide and converting it to K 2MoO 4 in 5 M KOH (0.2 g-Mo/mL), the solution contains about 1.8 kg of potassium per kilogram of molybdenum. The most challenging separation for this recovery step is purifying molybdenum from potassium. One requirement to facilitate the acceptance of the recycled material by the U.S. Food and Drug Administration (FDA) is that the impurities in the recycled material need to be at or below the levels present in the starting material. Therefore, the amount of potassium (K) in purified MoO 3 powder should be below 20 ppm; this will require a decontamination factor for removal of K to be ~1 × 10 5. Such a low K-contamination level will also prevent the production of large amounts of K-42 during irradiation of Mo-100. Based on economic concerns (due to the significant cost of enriched Mo-100) recycling Mo requires the conversion of K 2MoO 4 in a 5 M KOH solution to MoO 3 powder with high Mo recovery yields (>98%).« less
  • A series of four one-day irradiations was conducted with 100Mo-enriched disk targets. After irradiation, the enriched disks were removed from the target and dissolved. The resulting solution was processed using a NorthStar RadioGenix™ 99mTc generator either at Argonne National Laboratory or at the NorthStar Medical Radioisotopes facility. Runs on the RadioGenix system produced inconsistent analytical results for 99mTc in the Tc/Mo solution. These inconsistencies were attributed to the impurities in the solution or improper column packing. During the irradiations, the performance of the optic transitional radiation (OTR) and infrared cameras was tested in high radiation field. The OTR cameras survivedmore » all irradiations, while the IR cameras failed every time. The addition of X-ray and neutron shielding improved camera survivability and decreased the number of upsets.« less
  • A six-and-a-half day irradiation of enriched Mo-100 target disks was performed by Argonne’s electron linac. This report describes the irradiation conditions and the means used to process the targets for shipment to NorthStar Medical Isotopes, LLC, for feed to their RadioGenixTM technetium generator.