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National Research Council (US) Committee on Medical Isotope Production Without Highly Enriched Uranium. Medical Isotope Production without Highly Enriched Uranium. Washington (DC): National Academies Press (US); 2009.

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Medical Isotope Production without Highly Enriched Uranium.

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2Molybdenum-99/Technetium-99m Production and Use

The 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 determined 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:


The decay product of Mo-99, technetium-99m1 (Tc-99m), is used in about two-thirds2 of all diagnostic medical isotope procedures in the United States.


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.


Other medical isotopes such as iodine-131 (I-131) and xenon-133 (Xe-133) are by-products of the Mo-99 production process and will be sufficiently available if Mo-99 is available.


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.3

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.

FIGURE 2.5. Fission yield for thermal neutron fission of U-235.


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.org/sgnucdat/c1.htm.

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.


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 chemically 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 instruments such as gamma cameras. The data collected by the camera are analyzed 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 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).

FIGURE 2.1. (a) External view of a technetium generator produced by the Australian Nuclear Science and Technology Organisation (ANSTO).


(a) External view of a technetium generator produced by the Australian Nuclear Science and Technology Organisation (ANSTO). SOURCE: Courtesy of ANSTO. (b) Schematic diagram showing the internal structure of a typical technetium generator.

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 replaced4 with a fresh generator (Figure 2.2).

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.


Plot of typical Mo-99 and Tc-99m activity on a logarithmic scale versus time for multiple elution of a technetium generator.

There are numerous Tc-99m kits 5 for producing radiopharmaceuticals to examine the brain, kidney, heart, bone, liver, and lung. Table 2.1 provides a selected list of Tc-99m labeled radiopharmaceuticals in use today. 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 diagnostic imaging procedures.

TABLE 2.1. Selected Examples of Tc-99m Kits for Nuclear Medicine Diagnostic Imaging.


Selected Examples of Tc-99m Kits for Nuclear Medicine Diagnostic Imaging.

FIGURE 2.3. (a) Image acquired from a Tc-99m cerebral blood flow brain scan of a person with Alzheimer’s disease.


(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 flow due to the disease. SOURCE: Courtesy of Satoshi Minoshima, University of Washington. (b) Images (more...)

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 scheduled and highly time dependent. An interruption at any point in the production, transport, or delivery of Mo-99 or technetium generators can have substantial impacts on patient care, as discussed in Chapter 4.


There are two primary approaches for producing the medical isotope Mo-99, as described in Appendix D: fission of U-235, which produces 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 described 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 quantities of Mo-99. In fact, as will be discussed in Chapter 3, the International Atomic Energy Agency has Coordinated Research Projects that are partly focused on production by this method. Additionally, Japan recently announced that it will produce Mo-99 using neutron activation to provide a stable domestic supply.6

This chapter focuses on the production of Mo-99 by neutron irradiation 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:


Fabrication of uranium targets,


Irradiation of targets in a nuclear reactor,


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 equipment 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 designed to satisfy several requirements: First, it must be properly sized to fit into the irradiation position inside the reactor.7 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 overheating8 (which could result in target failure) during irradiation. 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.

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 compositions include uranium metal, uranium oxides, and alloys of uranium, nearly always with aluminum. Metallic targets are typically encapsulated in aluminum 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.9 Sometimes an intermediate 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.

FIGURE 2.4. CNEA’s high-density LEU-aluminum dispersion targets.


CNEA’s high-density LEU-aluminum dispersion targets. These targets have been used since 2002 to produce Mo-99 in Argentina. The target is approximately 15 cm in length. SOURCE: Courtesy of Pablo Cristini, CNEA, Argentina.

TABLE 2.2. Uranium-Bearing Targets for Mo-99 Production.


Uranium-Bearing Targets for Mo-99 Production.

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 (Figure 2.5).

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 radioactive 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 processing 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.

FIGURE 2.6. (a) Hot cells in use at CNEA for processing of LEU targets to recover Mo-99.


(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.

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 in the hot cell is constructed using commercially available components or components that are easily fabricated in machine or glass-blowing shops.

FIGURE 2.7. (a) View into a hot cell at CNEA showing the target processing equipment.


(a) View into a hot cell at CNEA showing the target processing equipment. (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.

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 aluminum. 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, typically 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 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 combined 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. Worldwide, 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 downblend 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 producers. Moreover, the Argentine organization CNEA has demonstrated that the alkaline process can be used with LEU targets, and work is underway (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 disadvantages.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, relative 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 molybdenum 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.



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.


For example Russian English Venture in Isotope Supply Services (REVISS) sells Russian-produced isotopes.


The technetium generator is replaced after about a week because it loses its elution efficiency 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.


This requirement is reactor specific, because the locations and sizes of the irradiation positions 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.


The target has a “sandwich” structure: The metal cladding is the “bread” and the uranium-bearing material is the “meat.”


A thermal neutron is a low-energy neutron of about 0.025 electron volts at room temperature. This energy is typical for neutrons in light-water (i.e., ordinary water) reactors.


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.


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).


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 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).


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.


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 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).


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.


In some processes ion exchange resins have been substituted for the alumina column for this separation.


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.


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.

Copyright 2009 by the National Academy of Sciences. All rights reserved.
Bookshelf ID: NBK215133


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