Medical Isotope Production Without Highly Enriched Uranium (2009)

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2 Molybdenum-99/Technetium-99m Production and Use T he congressional mandate for this study calls for an examination of the production of medical isotopes to include “molybdenum 99, iodine 131, xenon 133, and other radioactive materials used to produce radiopharmaceuticals for diagnostic and therapeutic procedures or for research and development.” However, the authoring committee deter­ mined that for the purposes of addressing the statement of task for this study (Sidebar 1.2), it is sufficient to focus on the production of the medical isotope molybdenum-99 (Mo-99). This is so because: 1. The decay product of Mo-99, technetium-99m (Tc-99m), is used in about two-thirds of all diagnostic medical isotope procedures in the United States. 2. Between 95 and 98 percent of Mo-99 is currently being produced using highly enriched uranium (HEU) targets (NNSA and ANSTO, 2007), which was the major concern of Congress when it mandated this study. 3. Other medical isotopes such as iodine-131 (I-131) and xenon-133  The symbol “m” denotes that the isotope is metastable. The nucleus of a metastable isotope has an elevated energy state and, in the case of Tc-99m, releases this energy by emitting a gamma ray. The decay process is referred to as isomeric transition.  Higher percentages of procedures utilizing Tc-99m are estimated by some other sources. For example, NNSA and ANSTO (2007) estimated that about 70 percent of all procedures utilize Tc-99m. Some of the industry presenters at the committee’s information-gathering meetings estimated that 80–85 percent of all procedures utilize Tc-99m. 16 

MOLYBDENUM-99/TECHNETIUM-99m PRODUCTION AND USE 17 (Xe-133) are by-products of the Mo-99 production process and will be suf- ficiently available if Mo-99 is available. 4. These other medical isotopes are not being recovered for sale by all major Mo-99 producers because they can be more cheaply produced and purchased from other sources. Point 3 deserves additional elaboration. The fission of uranium-235 (U-235) produces a spectrum of fission products (see Figure 2.5) including Mo-99, I-131, and Xe-133. These fission products are produced in the same proportions to each other whether HEU or low enriched uranium (LEU) targets are used. All of these isotopes can be recovered when the targets are processed to obtain Mo-99. The primary purpose of this chapter is to provide a brief overview of the production and use of Mo-99 in nuclear medicine and is intended primarily for nonexpert readers. Knowledgeable readers may wish to skip directly to Chapter 3. MOLYBDENUM-99 USE IN NUCLEAR MEDICINE The decay product of Mo-99, Tc-99m, is the workhorse isotope in nuclear medicine for diagnostic imaging. Tc-99m is used for the detection of disease and for the study of organ structure and function. Tc-99m is especially useful for nuclear medicine procedures because it can be chemi- cally incorporated into small molecule ligands and proteins that concentrate in specific organs or tissues when injected into the body. The isotope has a half-life of about 6 hours and emits 140 keV photons when it decays to Tc-99, a radioactive isotope with about a 214,000-year half-life. This photon energy is ideally suited for efficient detection by scintillation instru- ments such as gamma cameras. The data collected by the camera are ana- lyzed to produce detailed structural and functional images. A recent report of the National Research Council and Institute of Medicine (NAS and IOM, 2007) provides a description of the imaging process. As will be described in more detail in the following section, Tc-99m is currently produced through a multistep process that begins with the neutron irradiation of fissile U-235 contained in HEU (see Sidebar 1.1) or LEU targets in a nuclear reactor. This irradiation causes U-235 to fission and produces Mo-99 and many other fission products, including I-131 and Xe-133. Following irradiation, the targets are chemically processed to separate Mo-99 from other fission products. If desired, these other fission products can be recovered separately. The separated Mo-99, which is con-   For example Russian English Venture in Isotope Supply Services (REVISS) sells Russian- produced isotopes. 

18 MEDICAL ISOTOPE PRODUCTION WITHOUT HIGHLY ENRICHED URANIUM tained in a solution, is then adsorbed onto an alumina (Al2O3) column that is contained in cylinders that are about the diameter of a large pencil. The columns are shipped to radiopharmacies and hospitals in radiation-shielded cartridges known as technetium generators (Figure 2.1). The Mo-99 in the generators decays with about a 66-hour half-life to Tc-99m. The Tc-99m is typically recovered by passing a saline solution through the alumina column in the generator, a process known as eluting the generator. The saline removes the Tc-99m but leaves the Mo-99 in place. A technetium generator can be eluted several times a day for about a week before it needs to be replaced with a fresh generator (Figure 2.2). There are numerous Tc-99m kits for producing radiopharmaceuti- cals to examine the brain, kidney, heart, bone, liver, and lung. Table 2.1 provides a selected list of Tc-99m labeled radiopharmaceuticals in use to- day. The list is not intended to be exhaustive but to illustrate the range of diseases and conditions where Tc-99m based diagnostic imaging is useful. Figure 2.3 provides examples of images that can be obtained from diag­ nostic imaging procedures. Because of its relatively short half-life (66 hours), Mo-99 cannot be stockpiled for use. It must be made on a weekly or more frequent basis to ensure continuous availability. The processes for producing Mo-99 and technetium generators and delivering them to customers are tightly sched- uled and highly time dependent. An interruption at any point in the produc- tion, transport, or delivery of Mo-99 or technetium generators can have substantial impacts on patient care, as discussed in Chapter 4. Mo-99 PRODUCTION PROCESS There are two primary approaches for producing the medical iso- tope Mo-99, as described in Appendix D: fission of U-235, which pro- duces Mo-99 and other medically important isotopes such as I-131 and Xe-133, and neutron capture by Mo-98 to produce Mo-99. For the reasons d ­ escribed in Appendix D, the committee dismissed neutron capture as a v­iable process for producing Mo-99 in the quantities needed to meet U.S. or global demand for Mo-99. None of the four global producers of Mo-99 (Chapter 1) use the neutron capture method to produce Mo-99 because of its inefficiencies. However, this process can be used to make smaller   The technetium generator is replaced after about a week because it loses its elution effi- ciency and also because the Tc-99m can become contaminated with Mo-99 from the column. The latter process is referred to as Mo-99 breakthrough. After it is replaced, the old generator may continue to be used for research that does not involve human subjects.   Kits are composed of all of the required chemicals (e.g., the pharmaceutical agent, chelating compound, and saline solution) for formulating the radiopharmaceutical to which Tc-99m is added. 

MOLYBDENUM-99/TECHNETIUM-99m PRODUCTION AND USE 19 A B Evacuated Vial Saline Vial 2-1a Vent Filter One-Way Filter 99 Mo Column Lead Shield FIGURE 2.1  (a) External view of a technetium generator produced by the Austra- Figure 2-1b.eps lian Nuclear Science and Technology Organisation (ANSTO). SOURCE: Courtesy of ANSTO. (b) Schematic diagram showing the internal structure of a typical tech- netium generator. 

20 MEDICAL ISOTOPE PRODUCTION WITHOUT HIGHLY ENRICHED URANIUM 1,000 99 Mo Activity (mCi) 100 99m Tc 10 Elute Generator 0 0 24 48 72 96 120 Time in Hours FIGURE 2.2  Plot of typical Mo-99 and Tc-99m activity on a logarithmic scale versus time for multiple elution of a technetium generator. Figure 2-2.eps TABLE 2.1  Selected Examples of Tc-99m Kits for Nuclear Medicine Diagnostic Imaging Kit Name Imaging Procedure Technetium Tc-99m Medronate (MDP) Bone Scan Technetium Tc-99m Albumin Aggregated (MAA) Lung Perfusion Technetium Tc-99m Pentetate (DTPA) Kidney Scan and Function Technetium Tc-99m Sulfur Colloid Liver Scan Sentinel Lymph Node Localization Technetium Tc-99m Sestamibi Cardiac Perfusion Technetium Tc-99m Exametazime Brain Perfusion Technetium Tc-99m Mebrofenin Gall Bladder Function Technetium Tc-99m Etidronate Bone Scan Technetium Tc-99m Disofenin Gall Bladder Function Technetium Tc-99m Succimer (DMSA) Kidney Scan and Function Technetium Tc-99m Tetrofosmin Cardiac Perfusion Technetium Tc-99m Bicisate Brain Perfusion Technetium Tc-99m Red Blood Cell Blood Pool Imaging Technetium Tc-99m Sodium Pertechnetate Thyroid, Salivary Gland, Meckel’s Scan Technetium Tc-99m Lidofenin Gall Bladder Function Technetium Tc-99m Mertiatide (MAG3) Kidney Scan and Function Technetium Tc-99m Oxidronate (HDP) Bone Scan NOTE: MAA = methacrylic acid, MDP = methylene diphosphonate, DTPA = diethylene triamine p ­ entaacetic acid, DMSA = dimercaptosuccinic acid, MAG3 = mercapto acetyl triglycine, HDP = ­hydroxymethylene diphosphonate. SOURCE: Extracted from the Food and Drug Administration approved pharmaceutical list, 2008. 

MOLYBDENUM-99/TECHNETIUM-99m PRODUCTION AND USE 21 A B FIGURE 2.3 (a) Image acquired from a Tc-99m cerebral blood flow brain scan of a person with Alzheimer’s disease. The arrows indicate areas of diminished blood 2.3b flow due to the disease. SOURCE: Courtesy of Satoshi Minoshima, University of Washington. (b) Images acquired from a cardiac perfusion SPECT study at stress and rest using a Tc-99m radiotracer. The images on the top row are taken during stress, and the images at rest are shown on the bottom. The arrows indicate areas of decreased perfusion, visualized by the darker colors in the image. SOURCE: Reprinted with permission from Elsevier from Rispler et al., 2007. 

22 MEDICAL ISOTOPE PRODUCTION WITHOUT HIGHLY ENRICHED URANIUM quantities of Mo-99. In fact, as will be discussed in Chapter 3, the Interna- tional Atomic Energy Agency has Coordinated Research Projects that are partly focused on production by this method. Additionally, Japan recently a ­ nnounced that it will produce Mo-99 using neutron activation to provide a stable domestic supply. This chapter focuses on the production of Mo-99 by neutron irradia- tion of targets containing highly enriched uranium-235 (HEU) in a nuclear reactor. This section provides an overview of this production method and is organized in terms of the following three processes: 1. Fabrication of uranium targets, 2. Irradiation of targets in a nuclear reactor, 3. Dissolution of the uranium target and recovery and purification of Mo-99. These three processes apply whether Mo-99 is produced from HEU or LEU targets. The equipment used to produce Mo-99 is small: The process equip- ment used to dissolve the targets and recover Mo-99 and (if desired) other isotopes is “bench scale” compared to most industrial chemical processing applications. In fact, this process equipment has a footprint similar to that of a large dining room table. Of course, this processing equipment must be operated inside large and heavily radiation-shielded facilities because the irradiated targets that contain Mo-99 are highly radioactive. Fabrication of Uranium Targets The target used for Mo-99 production is a material containing ­uranium- 235 that is designed to be irradiated in a nuclear reactor. The target is d ­ esigned to satisfy several requirements: First, it must be properly sized to fit into the irradiation position inside the reactor. Second, it must contain a sufficient amount of U-235 to produce the required amount of Mo-99 when it is irradiated. Third, it must have good heat transfer properties to prevent overheating (which could result in target failure) during irradia- tion. Fourth, the target must provide a barrier to the release of radioactive products, especially fission gases, during and after irradiation. Fifth, the target materials must be compatible with the chemical processing steps that will be used to recover and purify Mo-99 after the target is irradiated.   http://www.jaif.or.jp/english/aij/member/2008/2008-11-26b.pdf.   This requirement is reactor specific, because the locations and sizes of the irradiation posi- tions depend on the particular design of the reactor.   This heat is the by-product of nuclear reactions in the target that result from neutron bombardment. 

MOLYBDENUM-99/TECHNETIUM-99m PRODUCTION AND USE 23 FIGURE 2.4  CNEA’s high-density LEU-aluminum dispersion targets. These targets have been used since 2002 to produce Mo-99 in Argentina. The target is approxi- mately 15 cm in length. SOURCE: Courtesy of Pablo Cristini, CNEA, Argentina. To meet these criteria, targets are fabricated in a wide variety of shapes and compositions to meet the needs of individual Mo-99 producers. Targets may be shaped as plates (Figure 2.4), pins, or cylinders. Target composi- tions include uranium metal, uranium oxides, and alloys of uranium, nearly always with aluminum. Metallic targets are typically encapsulated in alumi- num or stainless steel to protect the chemically reactive uranium metal or alloy and to contain the fission products produced during irradiation. This encapsulation is referred to as the target cladding. Sometimes an interme- diate barrier material such as aluminum or nickel is used to separate the cladding from the U-235 target material. Table 2.2 summarizes the types of targets used or planned to be used in the future by different producers. Irradiation of Targets in a Nuclear Reactor Mo-99 is produced in the uranium-bearing targets by irradiating them with thermal neutrons.10 Some of the U-235 nuclei absorb these neutrons, which can cause them to fission. The fission of the U-235 nucleus produces two but sometimes three lower-mass nuclei referred to as fission fragments. Approximately 6 percent of these fission fragments are Mo-99 atoms (Fig- ure 2.5).   The target has a “sandwich” structure: The metal cladding is the “bread” and the uranium- bearing material is the “meat.” 10  thermal neutron is a low-energy neutron of about 0.025 electron volts at room tempera- A ture. This energy is typical for neutrons in light-water (i.e., ordinary water) reactors. 

24 MEDICAL ISOTOPE PRODUCTION WITHOUT HIGHLY ENRICHED URANIUM TABLE 2.2  Uranium-Bearing Targets for Mo-99 Production Target Geometry Target Material Target Usersa Plate Uranium aluminide/ Mallinckrodt, Institut National des aluminum-alloy dispersion Radioéléments, Nuclear Technology Products, CNEA, Australian Nuclear Science and Technology Organisation (ANSTO, OPAL reactor) Pin Uranium aluminum alloy in MDS-Nordion (National Research aluminum-cladding Universal reactor) Cylinder UO2 deposited on the inside Indonesian National Atomic Energy surface of a stainless-steel Agency (BATAN; current) closed cylinder BATAN (planned) Foil target MDS Nordion (Maple reactors)b Compacted UO2 powder aSee Chapter 3 for a discussion of these producers. bIn May 2008, AECL announced that it was discontinuing development work on the Maple reactors. SOURCE: Data from George Vandegrift, Argonne National Laboratory. 10 1 Fission Yield (%) 0.1 0.01 0.001 60 80 100 120 140 160 180 Atomic Mass Number FIGURE 2.5  Fission yield for thermal neutron fission of U-235. SOURCE: Data from Joint Evaluated Fission and Fusion File, Incident-neutron data, http:// www-nds.iaea.org/exfor/endf00.htm, October 2, 2006; see http://www-nds.iaea. 2.5 org/sgnucdat/c1.htm. 

MOLYBDENUM-99/TECHNETIUM-99m PRODUCTION AND USE 25 Nuclear reactors provide an efficient source of thermal neutrons for Mo-99 production. This is why all major Mo-99 producers irradiate their targets in nuclear reactors. The amount of Mo-99 produced in a target is a function of irradiation time, the thermal neutron fission cross section for U-235,11 the thermal neutron flux12 on the target, the mass of U-235 in the target, and the half-life of Mo-99. For typical reactor thermal neutron fluxes on the order of 1014 neutrons per square centimeter per second, irradia­tion times of about 5 to 7 days are required to achieve near-­maximum Mo-99 production in the targets. Beyond these irradiation times, the amount of Mo-99 produced in the targets approximately balances the amount of Mo-99 being lost to radio­ active decay, so further irradiation is not productive (see Sidebar 3.1). Even at maximum production, only about 3 percent of the U-235 in the target is typically consumed. The remaining U-235 along with the other fission products and target materials are treated as waste. Dissolution and Mo-99 Recovery Once the targets are removed from the reactor, they are cooled13 in water typically for half a day or less before being transported to the pro- cessing facility in shielded casks. Once at the processing facility, the targets are placed into hot cells (Figure 2.6) for chemical processing. Processing is carried out quickly to recover the Mo-99 to minimize further losses from radioactive decay. About 1 percent of the Mo-99 produced in the target is lost to radioactive decay every hour after irradiation. The apparatus in the hot cell used to process the targets and recover the Mo-99 (Figure 2.7) consists of a container for dissolving the targets, which is connected to tubing and columns for subsequent chemical separations to isolate Mo-99. The components can be easily replaced or reconfigured by a human operator using remote manipulators. The most expensive part of the separation facilities are the hot cells themselves. Hot cell facilities can cost tens of millions of dollars to construct.14 The separation apparatus 11  Fission cross section is usually expressed in barns, where 1 barn = 1 × 10–24 cm2. This cross section is related to the probability that the nuclei will capture a thermal neutron and cause fission. 12 Neutron flux is a measure of the intensity of neutron radiation. It is defined as the number of neutrons crossing a unit area of one square centimeter in one second (neutrons/cm2-s). 13 Cooling is a safety measure to prevent the target from being damaged because of high temperatures, to provide time for short-lived fission gases to decay, and to reduce overall radiation doses in the target processing system. 14 For example, Ralph Butler, director of the Missouri University Research Reactor (MURR), estimated that it could cost between $30 million and $40 million to construct a new hot cell facility for Mo-99 production at MURR. The facility would have two processing lines with three or four hot cells plus one additional common hot cell. This cost estimate was character- 

26 MEDICAL ISOTOPE PRODUCTION WITHOUT HIGHLY ENRICHED URANIUM in the hot cell is constructed using commercially available components or components that are easily fabricated in machine or glass-blowing shops. There are two general approaches for chemically processing targets to recover Mo-99: alkaline dissolution and acidic dissolution. The processes can be used on both HEU and LEU targets. Alkaline Dissolution Process Alkaline dissolution is generally used for targets that contain alumi- num. This process is used by all of the major isotope producers except MDS Nordion. A sodium hydroxide (NaOH) solution is used to dissolve the entire target, including the aluminum cladding and the uranium/aluminum alloy “meat” (see footnote 9). Dissolution produces a sodium aluminate (NaAlO2) solution containing sodium molybdate (Na2MoO4) along with small amounts of fission products and plutonium (Pu)15 and a solid oxide/ hydrated oxide residue. Hydrogen gas is evolved during dissolution. The solid residue contains uranium and most of the fission products except the alkali metals, iodine, fission gases, alkaline earths, and the elements that can act as either an acid or base such as molybdenum and aluminum. The short-lived fission gases (e.g., Xe-133) can be collected for sale or stored for decay, and I-131 can also be separated for sale if desired. The solution is recovered by filtering to remove suspended solids, typi- cally purified by ion exchange, and passed through a column of ­alumina16 that preferentially adsorbs the molybdate (MoO4–2) ion. Mo-99 recovery yield from the solution typically exceeds 85 to 90 percent. The sorbed ­molybdate is typically washed with a dilute ammonium hydroxide (NH4OH) solution and then removed from the column using a concentrated saline or ammonium hydroxide solution. Mo-99 is recovered as a highly pure product. Acid Dissolution and Molybdenum Separations Process Acid dissolution is generally used for uranium metal and uranium oxide targets. It is currently used by only one major producer, MDS Nordion. In contrast to the alkaline dissolution process, only the uranium metal or ­ oxide is processed; the uranium target meat is physically separated or ized as “just a guess” pending completion of a conceptual design study for the facility (Ralph Butler, written communication with study director Kevin Crowley, November 24, 2008). 15  Plutonium is produced by neutron capture of U-238 to produce U-239 which rapidly undergoes beta decay to form neptunium-239 (Np-239). Subsequently, Np-239 undergoes beta decay to form Pu-239. Plutonium may also be produced by successive neutron captures of U-235. 16  In some processes ion exchange resins have been substituted for the alumina column for this separation. 

MOLYBDENUM-99/TECHNETIUM-99m PRODUCTION AND USE 27 A B FIGURE 2.6 (a)  Hot cells in use at CNEA for processing of LEU targets to recover Mo-99. (b) Worker operating hot cell manipulators at MDS Nordion. SOURCE: Courtesy of CNEA and MDS Nordion, respectively. 

28 MEDICAL ISOTOPE PRODUCTION WITHOUT HIGHLY ENRICHED URANIUM A B 2-6a FIGURE 2.7 (a)  View into a hot cell at CNEA showing the target processing equip- ment. (b) View into a hot cell at MURR showing the new dissolver for the LEU metal foil targets. SOURCE: Courtesy of CNEA and the University of Missouri, respectively. 

MOLYBDENUM-99/TECHNETIUM-99m PRODUCTION AND USE 29 leached from the target cladding and then dissolved in nitric acid. A nitrate (NO3–) solution containing uranium, molybdenum, and all other fission products (except volatile gases such as iodine, Xe-133, krypton-85, and nitrogen oxides) is formed. Additional processing steps are required to recover pure molybdenum. Molybdenum can be separated from the nitrate solution by any of several separation processes. Typical separation processes include adsorption of the molybdenum on ion exchange resins and solvent extraction. Mo-99 recovery yields from these separation processes typically exceed 85 to 90 percent. The adsorbed or extracted molybdenum is washed with an appro­priate solution to remove residual fission products and uranium. The wash solution becomes waste. The adsorbed molybdenum is then removed from the separation medium using an appropriate solution and recovered as a highly pure Mo-99 product. Waste Management Waste management is similar for both the alkaline and acid dissolution processes. In the alkaline process, the sodium aluminate and dissolved or suspended fission products that pass through the alumina column are com- bined with the other fission product wastes and precipitated oxide residues. This waste is stored temporarily either as-is or put into a solid form (e.g., in cement). The waste stream from the acid dissolution process includes the separated cladding and liquid waste from the Mo-99 separation or extraction processes. This liquid waste can be stored in tanks or mixed with cement to immobilize it. Most of these process wastes are stored at producers’ sites or are transported to offsite storage facilities. As noted in Chapter 3, one producer (Nuclear Technology Products Radioisotopes in South Africa) is disposing of these wastes. Approximately 97 percent of the uranium originally present in the targets ends up in the process waste. Consequently, the accumulating waste from Mo-99 production contains substantial quantities of HEU. World- wide, tens of kilograms of this HEU waste are accumulating annually from Mo-99 production. This HEU could be recovered for reuse, but currently no producer has active plans to do so, presumably because it is less costly to purchase fresh HEU. Additionally, no Mo-99 producers currently down- blend their HEU waste (by mixing it with natural or depleted uranium) to convert it to LEU. Process Trade-offs Both the alkaline and acid dissolution processes have been proven to be effective through many years of use with HEU targets by the major isotope 

30 MEDICAL ISOTOPE PRODUCTION WITHOUT HIGHLY ENRICHED URANIUM producers. Moreover, the Argentine organization CNEA has demonstrated that the alkaline process can be used with LEU targets, and work is under­ way (see Chapter 7) to develop an improved acid dissolution process for LEU targets. As discussed elsewhere in this report (see Chapter 10), the committee sees no technical barriers to adapting either of these processes for LEU-based Mo-99 production. However, each of these processes has inherent advantages and disad- vantages.17 For example, alkaline processing produces very pure Mo-99, solid waste that is suitable for storage, and fission gases that can be readily isolated for sale or for storage to allow for decay. On the other hand, rela- tive to the acid process, alkaline processing produces larger volumes18 of processing solutions, it can require more time than the acidic process for target dissolution, and Mo-99 yields can be lower because some molybde- num may be incorporated into the solid residue. Additionally, hydrogen gas is produced in the alkaline process, which requires additional safety procedures. Acidic processing, in contrast, generally requires shorter processing times, produces smaller volumes of processing waste, and results in slightly higher Mo-99 yields. On the other hand, additional steps have to be carried out to separate the Mo-99 from the processing solutions, and there needs to be a separate process for handling the treatment of the nitrogen oxide gases given off from the process. These characteristics should only be viewed as generalities. All of the major producers have optimized their processing systems over many years to improve processing times, enhance recovery efficiencies, and minimize the production of liquid and solid waste. 17  A review of both alkaline and acid dissolution processes was provided by George ­Vandegrift (Argonne National Laboratory) during a presentation to the Committee in 2007. 18  The operative word here is “relative” because the liquid volumes are small (typically of the order of one or a few liters per processing batch) for either process.