Radionuclide Production, Radiopharmaceuticals, and Internal Dosimetry

16.1 RADIONUCLIDE PRODUCTION

Although many naturally occurring radioactive nuclides exist, all of those commonly administered to patients in nuclear medicine are artificially produced. Artificial radioactivity was discovered in 1934 by Irene Curie (daughter of Marie and Pierre Curie) and Frederic Joliot, who induced radioactivity in boron and aluminum targets by bombarding them with alpha (α) particles from polonium. Positrons continued to be emitted from the targets after the alpha source was removed. Today, more than 2,500 artificial radionuclides have been produced by a variety of methods. Most radionuclides used in nuclear medicine are produced by particle accelerators (e.g., cyclotrons), nuclear reactors, or radionuclide generators.

16.1.1 Cyclotron-Produced Radionuclides

Cyclotrons and other charged-particle accelerators produce radionuclides by bombarding stable nuclei with high-energy charged particles. Positively charged ions such as protons (H⁺), deuterons (²H⁺), and alpha particles (⁴He²⁺) as well as negatively charged hydrogen ions (H⁻) are commonly used to produce radionuclides used in medicine. Charged particles must be accelerated to high kinetic energies to overcome and penetrate the repulsive coulombic barrier of the target atoms’ nuclei.

In 1930, Cockcroft and Walton applied a clever scheme of cascading a series of transformers, each capable of stepping up the voltage by several hundred thousand volts. The large potentials generated were used to produce artificial radioactivity by accelerating particles to high energies and using them to bombard stable nuclei.

In Berkeley, California, in 1931, E.O. Lawrence capitalized on this development but added a unique dimension in his design of the cyclotron (Fig. 16-1). A cyclotron has a vacuum chamber between the poles of an electromagnet. Inside the vacuum chamber is a pair of hollow, semicircular electrodes, each shaped like the letter D and thus referred to as “dees.” The two dees are separated by a small gap. An alternating high voltage is applied between the two dees. When positive ions are injected into the center of the cyclotron, they are attracted to and accelerated toward the negatively charged dee. The static magnetic field constrains the ions to travel in a circular path, whereby the radius of the circle increases as the ions gain kinetic energy (Fig. 16-2). Halfway around the circle, the ions approach the gap between the dees; at this time, the polarity of the electrical field between the two dees is reversed, causing the ions to be accelerated toward the negative dee. This cycle is repeated again and again, with the particles accelerated each time they cross the gap, acquiring additional kinetic energy and sweeping out larger and larger circles. As the length of the path between successive accelerations increases, the speed of the particle also increases; hence,
the time interval between accelerations remains constant. The cyclic nature of these events led to the name “cyclotron.” The final kinetic energy achieved by the accelerated particles depends on the type of particle (e.g., protons or deuterons), the diameter of the dees, and the strength of the static magnetic field. Finally, as the ions reach the periphery of the dees, they are removed from their circular path by a negatively
charged deflector plate (if positive ions are accelerated) or electron stripping foil (if H⁻ ions are accelerated), emerge through the window, and strike the target. Depending on the design of the cyclotron, particle energies can range from a few million electron volts (MeV) to several hundred MeV.

▪ FIGURE 16-1 Schematic view of a cyclotron. Two “dees” (A and B) are separated by a small gap.

▪ FIGURE 16-2 A constant magnetic field imposes a force (F) on a moving charged particle that is perpendicular to the direction of the particle’s velocity (v). This causes an ion in a cyclotron to move in a circular path. The diameter of the circular path is proportional to the speed of the ion.

The accelerated ions collide with the target nuclei, causing nuclear reactions. An incident particle may leave the target nucleus after interacting, transferring some of its energy to it, or it may be completely absorbed. The specific reaction depends on the type and energy of the bombarding particle as well as the composition of the target. In either case, target nuclei are left in an excited state, and this excitation energy is disposed of through the emission of particulate (protons and neutrons) and electromagnetic (γ-rays) radiations. Gallium-67 (Ga-67) is an example of a cyclotron-produced radionuclide. The production reaction is written as follows:

where the target material is zinc-68 (Zn-68), the bombarding particle is a proton (p) accelerated to approximately 20 MeV, two neutrons (2n) are emitted, and Ga-67 is the product radionuclide. In some cases, the nuclear reaction produces a radionuclide that decays to the clinically useful radionuclide (see iodine-123 and thallium-201 production below). Most cyclotron-produced radionuclides are neutron poor and therefore decay by positron emission or electron capture. The production methods of several cyclotron-produced radionuclides important to nuclear medicine are shown below (EC = electron capture, T_½ = physical half-life).

Iodine-123 production:

Iodine-111 production:

¹⁰⁹Ag (α, 2n)¹¹¹In or ¹¹¹Cd (p, n)¹¹¹In or ¹¹²Cd (p, 2n)¹¹¹In.

Cobalt-57 production:

⁵⁶Fe (d, n)⁵⁷Co.

Thallium-201 production:

Industrial cyclotron facilities that produce large activities of clinically useful radionuclides are very expensive and require substantial cyclotron and radiochemistry support staff and facilities. Cyclotron-produced radionuclides are usually more expensive than those produced by other technologies.

Much smaller, specialized cyclotrons, installed in commercial radiopharmacies serving metropolitan areas or in hospitals, have been developed to produce positron-emitting radionuclides for positron emission tomography (PET) (Fig. 16-3). These cyclotrons operate at lower energies (10 to 30 MeV) than industrial cyclotrons and commonly accelerate H⁻ ions, which is a proton with two orbital electrons. In such a cyclotron, the beam is extracted by passing it through a carbon stripping foil, which removes the electrons thus creating an H⁺ ion (proton) beam.

▪ FIGURE 16-3 Commercial self-shielded cyclotron for radionuclide production capable of producing a 60 µA beam of protons accelerated to – 11 MeV is shown with the radiation shields closed (A). The unit is designed with a small footprint to fit into a relatively small room (24′ × 23′ × 14′ height). (B) Power supply and control cabinet. (C) Cyclotron assembly approximately 10,000 kg (22,000 lb). (D) Retractable radiation shielding (open) approximately 14,500 kg (32,000 lb) of borated concrete and polyethylene. Neutrons and γ radiation are an unavoidable by-product of the nuclear reactions that are used to produce the desired radioactive isotopes. Boron and polyethylene are added to the radiation shield to absorb neutrons. The shielding is designed so that radiation exposure rates are reduced to the point where technologists and other radiation workers can occupy the room while the accelerator is in operation (less than 20 µSv/h at 24 ft from the center of the cyclotron). (E) Cyclotron assembly open. Hydrogen gas line at the top of the cyclotron assembly provides the source of hydrogen ions to be accelerated. (F) One of four cyclotron dees. The acceleration potential is supplied by high frequency voltage. In this system, four dees provide eight accelerations per orbit, thus reducing acceleration path length and beam loss. (G) Beam shaping magnets act as powerful lenses to confine ions to the midplane. The dotted white arrow shows the beam path through one of the dees. The radiochemicals produced (in gas or liquid) are sent through tubing in a shielded channel running under the floor to the automated radiochemistry unit located in a shielded enclosure in a room next to the cyclotron. A typical production run from a cyclotron in a commercial radiopharmacy serving a metropolitan area will produce approximately 131 GBq (3.5 Ci) of F-18 during a 2 h irradiation. The radiopharmacy may have three to four production runs a day depending on the clinical demand in the area. (© Siemens Healthineers 2019. Used with permission.)

Because of the change in the polarity of the charge on each particle, the direction of the forces on the moving particles from the magnetic field is reversed and the beam is diverted out of the cyclotron and onto a target. These commercially available specialized medical cyclotrons have a number of advantages, including automated cyclotron operation and radiochemistry modules, allowing a technologist with proper training to operate the unit. Radiation shielding of cyclotrons is always an important consideration; however, the use of negative ions avoids the creation of unwanted radioactivity in the cyclotron housing and thus reduces the amount of radiation shielding necessary. These features substantially reduce the size and weight of the cyclotron facility allowing it to be placed within the hospital close to the PET
imaging facilities. Production methods of clinically useful positron-emitting radionuclides are listed below.

In the interests of design simplicity and cost, some medical cyclotrons accelerate only protons. These advantages may be offset for particular productions such as ¹⁵O when an expensive rare isotope ¹⁵N that requires proton bombardment must be used in place of the cheap and abundant ¹⁴N isotope that requires deuteron bombardment. The medical cyclotrons are usually located near the PET imaging system because of the short half-lives of the radionuclides produced. Fluorine-18 (F-18) is an exception to this generalization owing to its longer half-life (110 min). Regional production and distribution of ¹⁸F is thus an option for this commonly used PET radionuclide.

16.1.2 Nuclear Reactor-Produced Radionuclides

Nuclear reactors are another major source of clinically used radionuclides. Neutrons, being uncharged, have an advantage in that they can penetrate the nucleus without being accelerated to high energies. There are two principal methods by which radionuclides are produced in a reactor: nuclear fission and neutron activation.

Nuclear Fission

Fission is the splitting of an atomic nucleus into two smaller nuclei. Whereas some unstable nuclei fission spontaneously, others require the input of energy to overcome the nuclear binding forces. This energy is often provided by the absorption of neutrons. Neutrons can induce fission only in certain very heavy nuclei. Whereas highenergy neutrons can induce fission in several such nuclei, there are only three nuclei of reasonably long half-life that are fissionable by neutrons of all energies; these are called fissile nuclides.

The most widely used fissile nuclide is uranium-235 (U-235). Elemental uranium exists in nature primarily as U-238 (99.3%) with a small fraction of U-235 (0.7%). U-235 has a high fission cross section (i.e., high fission probability); therefore, its concentration is usually enriched (typically to 3% to 5%) to make the fuel used in nuclear reactors.

When a U-235 nucleus absorbs a neutron, the resulting nucleus (U-236) is in an extremely unstable excited energy state that usually promptly fissions into two smaller nuclei called fission fragments. The fission fragments separate with very high kinetic energies, with the simultaneous emission of γ radiation and the ejection of two to five neutrons per fission (Eq. 16-4).

The fission of uranium creates fission fragment nuclei having a wide range of mass numbers. More than 200 radionuclides with mass numbers between 70 and 160 are
produced by the fission process (Fig. 16-4). These fission products are neutron-rich and therefore almost all of them decay by beta-minus (β⁻) particle emission.

▪ FIGURE 16-4 Fission yield as a percentage of total fission products from uranium 235.

Nuclear Reactors and Chain Reactors

The energy released by the nuclear fission of a uranium atom is more than 200 MeV. Under the right conditions, this reaction can be perpetuated if the fission neutrons interact with other U-235 atoms, causing additional fissions and leading to a self-sustaining nuclear chain reaction (Fig. 16-5). The probability of fission with U-235 is greatly enhanced as neutrons slow down or thermalize. The neutrons emitted from fission are very energetic (called fast neutrons) and are slowed (moderated) to thermal energies (˜0.025 eV) as they scatter in water in the reactor core. Good moderators are low-Z materials that slow the neutrons without absorbing a significant fraction of them. Water is the most commonly used moderator, although other materials, such as graphite (used in the reactors at the Chernobyl plant in Ukraine) and heavy water (²H₂O), are also used.

Some neutrons are absorbed by non-fissionable material in the reactor, while others are moderated and absorbed by U-235 atoms and induce additional fissions. The ratio of the number of fissions in one generation to the number in the previous generation is called the multiplication factor. When the number of fissions per generation is constant, the multiplication factor is 1 and the reactor is said to be critical. When the multiplication factor is greater than 1, the rate of the chain reaction increases, at which time the reactor is said to be supercritical. If the multiplication factor is less than 1 (i.e., more neutrons being absorbed than produced), the reactor is said to be subcritical and the chain reaction will eventually cease.

This chain reaction process is analogous to a room whose floor is filled with mousetraps, each one having a ping-pong ball placed on the trap. Without any form of control, a self-sustaining chain reaction will be initiated when a single ping-pong ball is tossed into the room and springs one of the traps. The nuclear chain reaction is maintained at the desired level by limiting the number of available neutrons through the use of neutron-absorbing control rods (containing boron, cadmium, indium, or a mixture of these elements), which are placed in the reactor core between the fuel elements. Inserting the control rods deeper into the core absorbs more neutrons,
reducing the reactivity (i.e., causing the neutron fluence rate and power output to decrease with time). Removing the control rods has the opposite effect. If a nuclear reactor accident results in loss of the coolant, the fuel can overheat and melt (socalled meltdown). However, because of the design characteristics of the reactor and its fuel, an atomic explosion, like those from nuclear weapons, is impossible.

▪ FIGURE 16-5 Schematic of a nuclear chain reaction. The neutrons (shown as small blackened circles) are not drawn to scale with respect to the uranium atoms.

Figure 16-6 is a diagram of a typical radionuclide production reactor. The fuel is processed into rods of uranium-aluminum alloy approximately 6 cm in diameter and 2.5 m long. These fuel rods are encased in zirconium or aluminum, which have favorable neutron and heat transport properties. There may be as many as 1,000 fuel rods in the reactor, depending on the design and the neutron fluence rate requirements. Water circulates between the encased fuel rods in a closed loop, whereby the heat generated from the fission process is transferred to cooler water in the heat exchanger. The water in the reactor and heat exchanger are in separate closed loops that do not come into direct physical contact with the fuel. The heat transferred to the cooling water is released to the environment through cooling towers, evaporation ponds, or heat exchangers that transfer the heat to a large body of water. The cooled water is pumped back toward the fuel rods, where it is reheated and the process is repeated.

In commercial nuclear power electric generation stations, the heat generated from the fission process produces high-pressure steam that is directed through a steam turbine, which powers an electrical generator. The steam is then condensed to water by the condenser.

Nuclear reactor safety design principles dictate numerous barriers between the radioactivity in the core and the environment. For example, in commercial power reactors, the fuel is encased in metal fuel rods that are surrounded by water and enclosed in a sealed, pressurized, approximately 30-cm-thick steel
reactor vessel. These components, together with other highly radioactive reactor systems, are enclosed in a large steel-reinforced concrete shell (˜1 to 2 m thick), called the containment structure. In addition to serving as a moderator and coolant, the water in the reactor acts as a radiation shield, reducing the radiation levels adjacent to the reactor vessel. Specialized nuclear reactors are used to produce clinically useful radionuclides from fission products or neutron activation of stable target material.

▪ FIGURE 16-6 NRU Radionuclide research/production reactor. (Adapted from diagram provided courtesy of Atomic Energy of Canada and Chalk River Laboratories, Chalk River, Ontario.) Fuel rod assemblies and the fission process are illustrated to show some of the detail and the relationships associated with fission-produced radionuclides.

Fission-Produced Radionuclides

The fission products most often used in nuclear medicine are molybdenum-99 (Mo-99), iodine-131 (I-131), and xenon-133 (Xe-133). These products can be chemically separated from other fission products with essentially no stable isotopes (carrier) of
the radionuclide present. Thus, the concentration or specific activity (measured in MBq or Ci per gram) of these “carrier-free” fission-produced radionuclides is very high. High-specific-activity, carrier-free nuclides are preferred in radiopharmaceutical preparations to increase the labeling efficiency of the preparations and minimize the mass and volume of the injected material.

Neutron Activation-Produced Radionuclides

Neutrons produced by the fission of uranium in a nuclear reactor can be used to create radionuclides by bombarding stable target material placed in the reactor. Ports exist in the reactor core between the fuel elements where samples to be irradiated are inserted. This process, called neutron activation, involves the capture of neutrons by stable nuclei, which results in the production of radioactive nuclei. The most common neutron capture reaction for thermal (slow) neutrons is the (n,γ) reaction, in which the capture of the neutron by a nucleus is immediately followed by the emission of a γ-ray. Other thermal neutron capture reactions include the (n,p) and (n,α) reactions, in which the neutron capture is followed by the emission of a proton or an alpha particle, respectively. However, because thermal neutrons can induce these reactions only in a few, low-atomic-mass target nuclides, most neutron activation uses the (n,γ) reaction. Almost all radionuclides produced by neutron activation decay by beta-minus particle emission. Examples of radionuclides produced by neutron activation useful to nuclear medicine are listed below.

A radionuclide produced by an (n,γ) reaction is an isotope of the target element. As such, its chemistry is identical to that of the target material, making chemical separation techniques useless. Furthermore, no matter how long the target material is irradiated by neutrons, only a small fraction of the target atoms will undergo neutron capture and become activated. Therefore, the material removed from the reactor will not be carrier-free because it will always contain stable isotopes of the radionuclide. In addition, impurities in the target material will cause the production of other radionuclides. The presence of carrier in the mixture limits the ability to concentrate the radionuclide of interest and therefore lowers the specific activity. For this reason, many of the clinically used radionuclides that could be produced by neutron activation (e.g., ¹³¹I, ⁹⁹Mo) are instead produced by nuclear fission to maximize specific activity. An exception to the limitations of neutron activation is the production of ¹²⁵I, in which neutron activation of the target material, ¹²⁴Xe, produces a radioisotope, ¹²⁵Xe, that decays to form the desired radioisotope (Eq. 16-6). In this case, the product radioisotope can be chemically or physically separated from the target material. Various characteristics of radionuclide production are compared in Table 16-1.

16.1.3 Radionuclide Generators

Since the mid-1960s, technetium-99m (Tc-99m) has been the most important radionuclide used in nuclear medicine for a wide variety of radiopharmaceutical applications. However, its relatively short half-life (6 h) makes it impractical to store even
a weekly supply. This supply problem is overcome by obtaining the parent Mo-99, which has a longer half-life (67 h) and continually produces Tc-99m. The Tc-99m is collected periodically in sufficient quantities for clinical operations. A system for holding the parent in such a way that the daughter can be easily separated for clinical use is called a radionuclide generator.

TABLE 16-1 COMPARISON OF RADIONUCLIDE PRODUCTION METHODS

CHARACTERISTIC	PRODUCTION METHOD
CHARACTERISTIC	*Linear Accelerator/Cyclotron*	*Nuclear Reactor (Fission)*	*Nuclear Reactor (Neutron Activation)*	*Radionuclide Generator*
Bombarding particle	Proton, alpha	Neutron	Neutron	Production by decay of parent
Product	Neutron poor	Neutron excess	Neutron excess	Neutron poor or excess
Typical decay pathway	Positron emission, electron capture	Beta-minus	Beta-minus	Several modes
Typically carrier free	Yes	Yes	No	Yes
High specific activity	Yes	Yes	No	Yes
Relative cost	High	Low	Low	Low (^99mTC) High (⁸²Rb)
Radionuclides for nuclear medicine applications	¹¹C, ¹³N, ¹⁵O, ¹⁸F, ⁵⁷Co, ⁶⁷Ga, ⁶⁸Ge, ¹¹¹In, ¹²³I, ²⁰¹Tl	⁹⁹Mo, ¹³¹I, ¹³³Xe	³²P, ⁵¹Cr, ⁸⁹Sr, ¹²⁵I, ¹⁵³Sm	⁶⁸Ga, ^81mKr, ⁸²Rb, ⁹⁰Y, ^99mTc

Molybdenum-99/Technetium-99m Radionuclide Generator

In a molybdenum-99/technetium-99m radionuclide generator, Mo-99 (produced by nuclear fission of U-235 to yield a high-specific-activity, carrier-free parent) is loaded, in the form of ammonium molybdenate (NH4⁺)(MoO₄⁻), onto a porous column containing 5 to 10 g of an alumina (Al₂O₃) resin. The ammonium molybdenite becomes attached to the surface of the alumina molecules (a process called adsorption). The porous nature of the alumina provides a large surface area for the adsorption of the parent.

As with all radionuclide generators, the chemical properties of the parent and daughter are different. In the Mo-99/Tc-99m or “moly” generator, the Tc-99m is much less tightly bound than the Mo-99. The daughter is removed (eluted) by the flow of isotonic (normal, 0.9%) saline (the “eluant”) through the column. When the saline solution is passed through the column, the chloride ions easily exchange with the TcO₄⁻ (but not the MoO₄⁻) ions, producing sodium pertechnetate, Na⁺(^99mTcO₄⁻). Technetium-99m pertechnetate (^99mTcO₄⁻) is produced in a sterile, pyrogen-free form with high specific activity and a pH (˜5.5) that is ideally suited for radiopharmaceutical preparations.

Commercially moly generators have a large reservoir of oxygenated saline (the eluant) connected by tubing to one end of the column and a vacuum extraction vial to the other. On insertion of the vacuum collection vial (contained in a shielded elution tool), saline is drawn through the column and the eluate is collected during elution, which takes about 1 to 2 min. Figure 16-7 is a picture and cross-sectional diagram of a moly generator together with an insert that shows details of the generator column. Sterility is achieved by a Millipore filter connected to the end of the column, by the

use of a bacteriostatic agent in the eluant, or by autoclave sterilization of the loaded column by the manufacturer.

▪ FIGURE 16-7 A. Picture of a “wet” molybdenum 99/technetium 99m generator in the process of being eluted (left). A spent generator that is no longer radioactive was used in order to minimize dose. For picture clarity, the shielding normally surrounding the generator is not shown [as illustrated in the accompanying diagram (right)]. However, correct radiation safety principles (discussed further in Chapter 21) are shown including the use of disposable gloves, finger ring and body dosimeters, and disposable plastic backed absorbent paper on the bench top to minimize the spread of any contamination. An explosion diagram depicting the generator components, and auxiliary radiation shielding is shown on the right. B. A cross-sectional diagram of the generator interior and column detail. Consult the text for additional information on the elution process. (Adapted from photo and diagrams provided courtesy of Curium US LLC, St. Louis, MO.)

Moly generators are typically delivered with approximately 37 to 740 GBq (1 to 20 Ci) of Mo-99, depending on the workload of the department. The larger activity generators are typically used by commercial radiopharmacies supplying radiopharmaceuticals to multiple nuclear medicine departments. The generators are shielded by the manufacture with lead, tungsten, or, in the case of higher activity generators, depleted uranium. Additional shielding is typically placed around the generator to reduce the exposure of staff during elution. The activity of the daughter at the time of elution depends on the following:

The activity of the parent
The rate of formation of the daughter, which is equal to the rate of decay of the parent (i.e., A₀e^-λ_pt)
The decay rate of the daughter
The time since the last elution
The elution efficiency (typically 80% to 90%)

Transient Equilibrium

Between elutions, the daughter (Tc-99m) builds up or “grows in” as the parent (Mo-99) continues to decay. After approximately 23 h, the Tc-99m activity reaches a maximum, at which time the production rate and the decay rate are equal and the parent and daughter are said to be in transient equilibrium. Once transient equilibrium has been achieved, the daughter activity decreases, with an apparent half-life equal to the half-life of the parent. Transient equilibrium occurs when the half-life of the parent is greater than that of the daughter by a factor of approximately 10. In the general case of transient equilibrium, the daughter activity will exceed the parent at equilibrium. If all of the (Mo-99) decayed to Tc-99m, the Tc-99m activity would slightly exceed (˜10% higher) that of the parent at equilibrium. However, approximately 12% of Mo-99 decays directly to Tc-99 without first producing Tc-99m, Figure 16-8.
Therefore, at equilibrium, the Tc-99m activity will be only approximately 97% (1.1 × 0.88) that of the parent (Mo-99) activity.

▪ FIGURE 16-8 Simplified decay scheme of Mo-99. Mo-99 decays to Tc-99m approximately 88% of the time. Thus is due to the β₈ transition directly to Tc-99m (˜82.2%) along with several other beta transitions to excited states that emit γ-rays (principally the β₄ γ₁₀ and β₇ γ₄) to yield Tc-99m. The balance (12%) of Mo-99 decays occurs by other beta transitions to excited states that ultimately yield Tc-99 bypassing the metastable form of Tc (Tc-99m).

Moly generators (sometimes called “cows”) are usually delivered weekly and eluted (a process referred to as “milking the cow”) each morning, allowing maximal yield of the daughter. The elution process is approximately 90% efficient. This fact, together with the limitations on Tc-99m yield in the Mo-99 decay scheme, results in a maximum elution yield of approximately 85% of the Mo-99 activity at the time of elution. Therefore, a typical elution on Monday morning from a moly generator with 55.5 GBq (1.5 Ci) of Mo-99 yields approximately 47.2 GBq (1.28 Ci) of Tc-99m in 10 mL of normal saline (a common elution volume). By Friday morning of that same week, the same generator would be capable of delivering only about 17.2 GBq (0.47 Ci). The moly generator can be eluted more frequently than every 23 h; however, the Tc-99m yield will be less. Approximately half of the maximal yield will be available 6 h after the last elution. Figure 16-9 shows a typical time-activity curve for a moly generator.

Secular Equilibrium

Although the moly generator is by far the most widely used in nuclear medicine, other generator systems produce clinically useful radionuclides. When the halflife of the parent is much longer than that of the daughter (i.e., more than about 100 times longer), secular equilibrium occurs after approximately five to six half-lives of the daughter. In secular equilibrium, the activity of the parent and the daughter are the same if all of the parent atoms decay directly to the daughter. Once secular equilibrium is achieved, the daughter will have an apparent half-life equal to that of
the parent. The strontium-82/rubidium-82 (Sr-82/Rb-82) generator, with parent and daughter half-lives of 25.5 d and 75 s, respectively, reach secular equilibrium within approximately 7.5 min after elution. Figure 16-10 shows a time-activity curve demonstrating secular equilibrium. The characteristics of radionuclide generator systems are compared in Table 16-2.

▪ FIGURE 16-9 Time-activity curve of a molybdenum 99/technetium 99m radionuclide generator system demonstrating the ingrowth of Tc-99m and subsequent elution. The time to maximum Tc-99m activity, approximately 23 h, assumes there is no residual Tc-99m from a previous elution of the column. Typical elution efficiency is approximately 85% (˜15% residual Tc-99m), thus time to maximum Tc-99m activity following the first elution is approximately 21 h. Approximately 50% of the maximum Tc-99m activity is obtained in 6 h. The maximum Tc-99m activity in the eluate is typically 80% to 90% of Mo-99 activity.

▪ FIGURE 16-10 Time-activity curve demonstrating secular equilibrium.

Quality Control

The users of moly generators are required to perform molybdenum and alumina breakthrough tests. Mo-99 contamination in the Tc-99m eluate is called molybdenum breakthrough. Mo-99 is an undesirable contaminant because its long half-life and highly energetic betas increase the radiation dose to the patient without providing any clinical information. The high-energy γ-rays (˜740 and 780 keV) are very penetrating and cannot be efficiently detected by scintillation cameras. The U.S. Pharmacopeia (USP) and the U.S. Nuclear Regulatory Commission (NRC) limit the Mo-99 contamination to 0.15 µCi of Mo-99 per mCi of Tc-99m or (0.15 kBq/MBq) at the time of administration. Contaminant limits are specified in 10CFR35.204 and include
those for the Rb-82 generators: 0.02 µCi of Sr-85 per mCi of Rb-82 or (0.02 kBq/MBq). The Mo-99 contamination is evaluated by placing the Tc-99m eluate in a thick (˜6 mm) lead container (provided by the dose calibrator manufacturer), which is placed in the dose calibrator. The high-energy photons of Mo-99 can be detected, whereas virtually all of the Tc-99m 140-keV photons are attenuated by the lead container. Eluates from moly generators rarely exceed permissible Mo-99 contamination limits. The quality control procedures to evaluate breakthrough of radionuclidic contaminates in the eluates from Mo-99/Tc-99m and Sr-82/Rb-82 generators are discussed further in Chapter 17 in the context of dose calibrator operations and quality control. It is also possible (although rare) for some of the alumina from the column to contaminate the Tc-99m eluate. Alumina interferes with the preparation of some radiopharmaceuticals (especially sulfur colloid and Tc-99m-labeled red blood cell preparations). The USP limits the amount of alumina to no more than 10 mg alumina per mL of Tc-99m eluate. Commercially available paper test strips and test standards are used to assay for alumina concentrations.

TABLE 16-2 CLINICALLY USED RADIONUCLIDE GENERATOR SYSTEMS IN NUCLEAR MEDICINE

PARENT	DECAY MODE AND (HALF-LIFE)	DAUGHTER	TIME OF MAXIMAL INGROWTH (EQUILIBRIUM)	DECAY MODE AND (HALF-LIFE)	DECAY PRODUCT
⁶⁸Ge	EC (271 d)	⁶⁸Ga	˜6.5 h (S)	β⁺ EC (68 min)	⁶⁸Zn (stable)
⁹⁰Sr	β⁻ (28.8 y)	⁹⁰Y	˜1 mo (S)	β⁻ (2.67 d)	⁹⁰Zr (stable)
⁸¹Rb	β⁺ EC (4.6 h)	^81mKr	˜80 s (S)	IT (13.5 s)	⁸¹Kr^a
⁸²Sr	EC (25.5 d)	⁸²Rb	˜7.5 min (S)	β⁺ (75 s)	⁸²Kr (stable)
⁹⁹Mo	β⁻ (67 h)	^99mTc	˜24 h (T)	IT (6 h)	⁹⁹Tc^a
Note: Decay modes: EC, electron capture; β⁺, positron emission; β⁻, beta-minus; IT, isometric transition (i.e., γ-ray emission). Radionuclide equilibrium: T, transients; S, secular.
^aThese nuclides have half-lives greater than 10⁵ years and for medical applications can be considered to be essentially stable.

16.2 RADIOPHARMACEUTICALS

16.2.1 Characteristics, Applications, Quality Control, and Regulatory Issues in Medical Imaging

The vast majority of radiopharmaceuticals in nuclear medicine today use Tc-99m as the radionuclide. Most Tc-99m radiopharmaceuticals are easily prepared by aseptically injecting a known quantity of Tc-99m pertechnetate into a sterile vial containing the lyophilized (freeze-dried) pharmaceutical. The radiopharmaceutical complex is, in most cases, formed instantaneously and can be used for multiple doses over a period of several hours. Radiopharmaceuticals can be prepared in this fashion (called “kits”) as needed in the nuclear medicine department, or they may be delivered to the department by a centralized commercial radiopharmacy that serves several hospitals in the area. Although most Tc-99m radiopharmaceuticals can be prepared rapidly and easily at room temperature, several products (e.g., Tc-99m sulfur colloid), require multiple steps such as boiling the Tc-99m reagent complex for several minutes. In almost all cases, however, the procedures are simple and the labeling efficiencies are very high (typically greater than 95%).

Other radionuclides common to diagnostic nuclear medicine imaging include ¹²³I, ⁶⁷Ga, ¹¹¹In, ¹³³Xe, and ²⁰¹Tl. Positron-emitting radionuclides are used for PET. F-18, as fluorodeoxyglucose (FDG), is used in approximately 85% of all clinical PET applications. Rubidium-82 (⁸²Rb) is used to assess myocardial perfusion using PET/CT imaging systems, in place of Tl-201 and Tc-99m based myocardial perfusion agents that are imaged using scintillation cameras. A wide variety of other positron-emitting radionuclides are currently being evaluated for their clinical utility, including carbon-11 (¹¹C), nitrogen-13 (¹³N), oxygen-15 (¹⁵O), and gallium-68 (⁶⁸Ga). The physical characteristics, most common modes of production, decay characteristics, and primary imaging photons (where applicable) of the radionuclides used in nuclear medicine are summarized in Table 16-3.

16.2.2 Ideal Diagnostic Radiopharmaceuticals

Although there are no truly “ideal” diagnostic radiopharmaceuticals, it is helpful to think of currently used agents in light of a set of ideal characteristics for radiopharmaceuticals applied to medical imaging of disease or evaluating the progress of prescribed therapy.

TABLE 16-3 PHYSICAL CHARACTERISTICS OF CLINICALLY USED RADIONUCLIDES

RADIONUCLIDE	METHOD OF PRODUCTION	MODE OF DECAY (%)	PRINCIPAL PHOTONS keV (% ABUNDANCE)	PHYSICAL HALF-LIFE	COMMENTS
Radionuclides Used in Diagnostic Nuclear Medicine Imaging (Planar Imaging and SPECT)
Chromium-51 (⁵¹Cr)	Neutron activation	EC (100)	320 (9)	27.8 d	Used for in vivo red cell mass determinations (not used for imaging; samples counted in a Nal(TI) well counter).
Cobalt-57 (⁵⁷Co)	Cyclotron produced	EC (100)	122 (86) 136 (11)	271 d	Principally used as a uniform flood field source for scintillation camera quality control.
Gallium-67 (⁶⁷Ga)	Cyclotron produced	EC (100)	93 (40) 184 (20) 300 (17) 393 (4)	78 h	Typically use the 93, 184, and 300 keV photons for imaging.
Indium-111 (¹¹¹In)	Cyclotron produced	EC (100)	171 (90) 245 (94)	2.8 d	Typically used when the kinetics require imaging more than 24 h after injection. Both photons are used in imaging.
Iodine-123 (¹²³I)	Cyclotron produced	EC (100)	159 (83)	13.2 h	Has replaced ¹³¹I for diagnostic imaging to reduce patient radiation dose.
Iodine 125 (¹²⁵I)	Neutron activation	EC (100)	35 (6) 27 (39) XR 28 (76) XR 31 (20) XR	60.2 d	Typically used as ¹²⁵I albumin for in vivo blood/plasma volume determinations (not used for imaging; samples counted in a Nal(TI) well counter).
Krypton-81m (^81mKr)	Generator product	IT (100)	190 (67) 181 (6) 740 (12)	13 s	This ultrashort-lived generator-produced radionuclide is a gas and is used to perform serial lung ventilation studies with very little radiation exposure to patient or staff. The expense and short T_1/2 of the parent (⁸¹Rb, 4.6 h) limits its use.
Molybdenum-99 (⁹⁹Mo)	Nuclear fission (²³⁵U)	β⁻ (100)	740 (12) 780 (4)	67 h	Parent material for Mo/Tc generator. Not used directly as a radiopharmaceutical; 740- and 780-keV photons used to identify “moly breakthrough.”
Technetium-99m (^99mTc)	Generator product	IT (100)	140 (88)	6.02 h	This radionuclide is used in radiopharmaceuticals that account for >70% of all nuclear medicine imaging studies.
Xenon-133 (¹³³Xe)	Nuclear fission (²³⁵U)	β⁻ (100)	81 (37)	5.3 d	¹³³Xe is a heavier-than-air gas. Low abundance and low energy of photon reduces image resolution.
Thallium-201 (²⁰¹TI)	Cyclotron produced	EC (100)	69-80 (94) XR	73.1 h	The majority of clinically used photons are low-energy x-rays (69-80 keV) from mercury 201 (²⁰¹Hg), the daughter of ²⁰¹TI. Although these photons are in high abundance (94%), their low energy results in significant patient attenuation. This issue is of particular concern in female patients in whom breast artifacts may appear in myocardial imaging.
Radionuclides Used in Diagnostic Nuclear Medicine Imaging (PET)
Carbon-11 (¹¹C)	Cyclotron produced	β⁺ (99.8)	511 AR (200)	20.4 min	Carbon-11 production: 14N (p,α) ¹¹C Short half-life requires onsite cyclotron for imaging. Primarily clinical research applications.
Fluorine-18 (¹⁸F)	Cyclotron produced	β⁺ (97) EC (3)	511 AR (193)	110 min	This radionuclides accounts for more than 70%-80% of all clinical PET studies; typically formulated as FDG Cyclotron produced via ¹⁸O (p,n)¹⁸F reaction.
Nitrogen-13 (¹³N)	Cyclotron produced	β⁺ (99.8)	511 AR (200)	10 min	Cyclotron produced via ¹⁶O (p,α)¹³N reaction. Short half-life requires on-site cyclotron for imaging. Primarily clinical research applications.
Oxygen-15 (¹⁵O)	Cyclotron produced	β⁺ (99.9)	511 AR (200)	122 s	Cyclotron produced via ¹⁴N(d,n)¹⁵O or ¹⁵N(p,n)¹⁵O. Short half-life requires on-site cyclotron for imaging. Primarily clinical research applications.
Gallium-68 (⁶⁸Ga)	Generator product	β⁺ (89) EC (11)	511 AR (184)	68 min	Ga-68 is a generator decay product of Ge-68, which is linear accelerator produced via a ⁶⁹Ga(p,2n)⁶⁸Ge reaction.
Rubidium-82 (⁸²Rb)	Generator product	β⁺ (95) EC (15)	511 AR (190) 776 (13)	78 s	Rb-82 is a generator decay product of Sr-82, which in turn is cyclotron produced via the 85Rb(p,4n)82Sr reaction. The half-life of Sr-82 is 25 d (or 600 h). Sr-82 is thus in secular equilibrium with Rb-82 within ˜8 min after elution.
Radionuclides Used in Radiopharmaceutical Therapy (Alpha-Emitters)
Astatine-211 (²¹¹At)	Cyclotron produced	α (42) EC (5)	²¹¹Po x-rays ²¹¹Po γ-rays	7.2 h	²¹¹Po (α), ²⁰⁷Bi (β⁻), and ²⁰⁷Pb (stable)
Lead-212 (²¹²Pb)	Generator product	β⁻ (100)	239 (43)	10.6 h	²¹²Pb is the parent of ²¹²Bi (see below)
Bismuth-212 (²¹²Bi)	Generator product or decay of ²³²Th	α (36) β⁻ (64)	67-91 220-257	1.0 h	²⁰⁸Tl (β⁻), ²¹²Po (α), ²⁰⁸Pb (stable)
Bismuth-213 (²¹³Bi)	Generator product	α (2) β⁻ (98)	440 (26)	45.6 m	²⁰⁹Tl (β⁻), ²¹³Po (α), ²⁰⁹Pb (β⁻), ²⁰⁹Bi (stable)
Radium-223 (²²³Ra)	Neutron activation	α (100)	154, 270, 351, 405 (progeny)	11.4 d	²¹⁹Rn (α), ²¹⁴Po (β), ²¹¹Pb (β⁻), ²¹¹Bi (α/β⁻), ²⁰⁷Tl (β⁻), ²¹¹Po(α), ²⁰⁷Pb (stable)
Actinium-225 (²²⁵Ac)	Cyclotron produced or decay of ²³³U	α (100)	440 (26) of ²¹³Bi	10.0 d	²¹¹Fr (α), ²¹⁷At (α), and followed by ²¹³Bi (see above)
Thorium-227 (²²⁷Th)	Neutron activation of decay of ²³⁵U	α (100)	12.3, 15.2, 236	18.7 d	²²⁷Th is the parent of ²²³Ra (see above)
Radionuclides Used in Radiopharmaceutical Therapy (Beta-Emitters)
Phosphorus-32 (³²P)	Neutron activation	β⁻ (100)	None	14.3 d	Prepared as either sodium phosphate for treatment of myeloproliferative disorders such as polycythemia vera and thrombocytosis or colloidal chromic phosphate for intracavitary therapy of malignant ascites, malignant pleural effusions, malignant pericardial effusions, and malignant brain cysts.
Strontium-89 (⁸⁹Sr)	Neutron activation	β⁻ (100)	Bremsstrahlung x-rays	50.5 d	As strontium chloride for pain relief from metastatic bone lesions.
Yittrium-90 (⁹⁰Y)	Generator product daughter of ⁹⁰Sr	β⁻ (100)	Bremsstrahlung x-rays (0.0032) AR	2.7 d	The radionuclide is bound to microspheres (glass or resin) for intrahepatic arterial delivery of the Y-90 microspheres for the treatment of unresectable metastatic liver tumors ⁹⁰Y-DOTATOC and ⁹⁰Y-DOTATATE are therapy radiopharmaceutical for treatment of neuroendocrine tumors that express somatostatin receptors. ⁹⁰Y-ibritumomab is used for therapy of CD20+ relapsed or refractory, low-grade or follicular B-cell non-Hodgkin’s lymphoma.
Iodine-124 (¹²⁴I)	Cyclotron produced		511 (46) AR	4.2 d	I-124 is an alternative to 131 for treatment of differentiated thyroid cancer, which further allows PET imaging of its biodistribution.
Iodine-131 (¹³¹I)	Neutron activation or nuclear fission (²³⁵U)	β⁻ (100)	80 (2.6) 284 (6) 364 (82) 637 (7) 732 (1.8)	8.0 d	Used for treatment of hyperthyroidism and thyroid cancer: 364-keV photon used for imaging. Resolution and detection efficiency are poor due to high energy of photons. High patient dose, mostly from β-particles. Used as the therapy radionuclide for ¹³¹I-MIBG (metaiodobenzylguanidine) for the treatment of neuroblastoma in children and young adults. Prior to 2014, ¹³¹I-tositumomab was used for therapy of CD20+ relapsed or refractory, low-grade or follicular B-cell non-Hodgkin’s lymphoma.
Samarium-153 (¹⁵³Sm)	Neutron activation	β⁻ (100)	69 (4.8) 103 (30) 635 (32)	46.3 h	As ¹⁵³Sm ethylene diamine tetra methylene phosphonic acid (EDTMP) used for pain relief from metastatic bone lesions. Advantage compared to ⁸⁹Sr is that the ¹⁵³Sm distribution can be imaged.
Lutetium-177 (¹⁷⁷Lu)	Neutron activation	β⁻ (100)	133 (6) 208 (11)	6.7 d	¹⁷⁷Lu-DOTATATE is a therapy radiopharmaceutical for treatment of neuroendocrine tumors that express somatostatin receptors.
Rhenium-186 (¹⁸⁶Re)	Neutron activation	β⁻ (92.5) EC (7.5)	137 (9)	3.7 d	¹⁸⁶Re-HEDP has been used as an alternative to ¹⁵³Sm-EDTMP for pain relief from metastatic bone lesions. The radionuclide has also be used in radiopharmaceuticals for prostate, breast, colon, lung, and skin cancer therapy.
Rhenium-188 (¹⁸⁸Re)	Neutron activation	β- (100)	155 (15) 478 (1) 633 (1)	17 h	¹⁸⁸Re-HEDP has been used as an alternative to ¹⁵³Sm-EDTMP for pain relief from metastatic bone lesions. The radionuclide has also be used in radiopharmaceuticals for prostate, breast, colon, lung, and skin cancer therapy.
Radionuclides Used in Radiopharmaceutical Therapy (Auger Electron-Emitters)
Palladium-103 (¹⁰³Pd)	Neutron activation	EC (100)	20-27 (7) XR	17.0 d	Pd-103 is been traditionally used as a therapy radionuclide for brachytherapy seeds in the treatment of prostate and cervical cancer. The radionuclide is also a potential therapy agent as labeled to albumin microspheres (AMS).
Indium-111 (¹¹¹In)	Cyclotron produced	EC (100)	171 (90) 245 (94)	2.8 d	High-administered activity ¹¹¹In-octreotide therapy has been used for patients with disseminated neuroendocrine tumors (NET) with high somatostatin receptor (SSR)
Tin-117m (¹¹⁷Sn)	Neutron activation	IT (100)	156 (2) 159 (86)	13.6 d	Sn-117m diethylenetriaminepentaacetic acid (^117mSn DTPA) is a radiopharmaceutical agent for the palliation of pain from bony metastases.
Iodine-123 (¹²³I)	Cyclotron produced	EC (100)	159 (83)	13.2 h	I-123 is used in pretherapy scans of patients with thyroid cancer to provide information on the amount of thyroid remnant, sometimes indicating the need for two-step I-131 ablation. It may also detect unsuspected local lymph node involvement or distant metastases, indicating the requirement for a higher I-131 dose after thyroidectomy.
Iodine-125 (¹²⁵I)	Neutron activation	EC (100)	35 (6) 27 (39) XR 28 (76) XR 31 (20) XR	60.2 d	Use as ¹²⁵I lotrex liquid brachytherapy source in Proxima GliaSite radiation therapy system for treatment of recurrent gliomas and metastatic brain tumors.
Platinum-193m (^193mPt)	Cyclotron produced	IT (100)	135 (0.11) 65-79 (14)	4.3 d	Potential radionuclide for chemoradiotherapy in which the chemotherapy drug cisplatin forms DNA-platinum adducts that are targeted to cancer cells. Cisplatin and Auger-electron radiation has demonstrated synergism in their cell killing effects.
Platinum-195m (^195mPt)	Neutron activation	IT (100)	31 (2) 99 (11) 130 (3)	4.2 d	Potential radionuclide for chemoradiotherapy in which the chemotherapy drug cisplatin forms DNA-platinum adducts that are targeted to cancer cells. Cisplatin and Auger-electron radiation has demonstrated synergism in their cell killing effects.
Note: α, alpha decay, β⁻, beta-minus decay; β⁺, beta-plus (positron) decay; AR, annihilation radiation; EC, electron capture; IT, isomeric transition (i.e., γ-ray emission), XR, x-ray.