Radionuclide and Radiopharmaceutical Production

Most of the naturally occurring radionuclides are very long-lived (e.g., ⁴⁰ K, T _1/2 ~ 10 ⁹ years), represent very heavy elements (e.g., uranium and radium) that are unimportant in metabolic or physiologic processes, or both. Some of the first applications of radioactivity for medical tracer studies in the 1920s and 1930s made use of natural radionuclides; however, because of their generally unfavorable characteristics indicated here, they have found virtually no use in medical diagnosis since that time. The radionuclides used in modern nuclear medicine all are of the manufactured or “artificial” variety. They are made by bombarding nuclei of stable atoms with subnuclear particles (such as neutrons and protons) so as to cause nuclear reactions that convert a stable nucleus into an unstable (radioactive) one. This chapter describes the methods used to produce radionuclides for nuclear medicine as well as some considerations in the labeling of biologically relevant compounds to form radiopharmaceuticals.

a

Reactor-Produced Radionuclides

1 Reactor Principles

Nuclear reactors have for many years provided large quantities of radionuclides for nuclear medicine. Because of their long and continuing importance for this application, a brief description of their basic principles is presented.

The “core” of a nuclear reactor contains a quantity of fissionable material, typically natural uranium ( ²³⁵ U and ²³⁸ U) enriched in ²³⁵ U content. Uranium-235 undergoes spontaneous nuclear fission ( T _1/2 ~ 7 × 10 ⁸ years), splitting into two lighter nuclear fragments and emitting two or three fission neutrons in the process (see Chapter 3 , Section I). Spontaneous fission of ²³⁵ U is not a significant source of neutrons or energy in of itself; however, the fission neutrons emitted stimulate additional fission events when they bombard ²³⁵ U and ²³⁸ U nuclei. The most important reaction is

{}^{235}U + n \to {}^{236}{U*}

${}^{235}U + n \to {}^{236}{U*}$

The ²³⁶ U ^* nucleus is highly unstable and promptly undergoes nuclear fission, releasing additional fission neutrons. In the nuclear reactor, the objective is to have the fission neutrons emitted in each spontaneous or stimulated fission event stimulate, on the average, one additional fission event. This establishes a controlled, self-sustaining nuclear chain reaction.

Figure 5-1 is a schematic representation of a nuclear reactor core. “Fuel cells” containing fissionable material (e.g., uranium) are surrounded by a moderator material. The purpose of the moderator is to slow down the rather energetic fission neutrons. Slow neutrons (also called thermal neutrons ) are more efficient initiators of additional fission events. Commonly used moderators are “heavy water” [containing deuterium (D ₂ O)] and graphite. Control rods are positioned to either expose or shield the fuel cells from one another. The control rods contain materials that are strong neutron absorbers but that do not themselves undergo nuclear fission (e.g., cadmium or boron). The fuel cells and control rods are positioned carefully so as to establish the critical conditions for a controlled chain reaction. If the control rods were removed (or incorrectly positioned), conditions would exist wherein each fission event would stimulate more than one additional nuclear fission. This could lead to a runaway reaction and to a possible “meltdown” of the reactor core. (This sequence occurs in a very rapid time scale in nuclear explosives. Fortunately, the critical conditions of a nuclear explosion cannot be achieved in a nuclear reactor.) Insertion of additional control rods results in excess absorption of neutrons and terminates the chain reaction. This procedure is used to shut down the reactor.

FIGURE 5-1, Schematic representation of a nuclear reactor.

Each nuclear fission event results in the release of a substantial amount of energy (200-300 MeV per fission fragment), most of which is dissipated ultimately as thermal energy. This energy can be used as a thermal power source in reactors. Some radionuclides are produced directly in the fission process and can be subsequently extracted by chemical separation from the fission fragments.

A second method for producing radionuclides uses the large neutron flux in the reactor to activate samples situated around the reactor core. Pneumatic lines are used for the insertion and removal of samples. The method of choice largely depends on yield of the desired radionuclide, whether suitable sample materials are available for neutron activation, the desired specific activity, and cost considerations.

2 Fission Fragments

The fission process that takes place in a reactor can lead to useful quantities of medically important radionuclides such as ⁹⁹ Mo, the parent material in the ^99m Tc generator (see Section C ). As described earlier, ²³⁶ U ^* promptly decays by splitting into two fragments. A typical fission reaction ( Fig. 5-2A ) is

{}_{92}^{235}U + n \to {}_{92}^{236}{U*} \to {}_{56}^{144}B a + {}_{36}^{89}K r + 3 n

${}_{92}^{235}U + n \to {}_{92}^{236}{U*} \to {}_{56}^{144}B a + {}_{36}^{89}K r + 3 n$

FIGURE 5-2, A, Example of production of fission fragments produced when neutrons interact with 236 U*. B, Mass distribution of fragments following fission of 236 U*.

More than 100 nuclides representing 20 different elements are found among the fission products of ²³⁶ U*. The mass distribution of the fission fragments is shown in Figure 5-2B . It can be seen that fission of ²³⁶ U* generally leads to one fragment with a mass number in the range of 85 to 105 and the other fragment with a mass number in the range of 130 to 150. It also is apparent that fission rarely results in fragments with nearly equal masses.

The fission products always have an excess of neutrons and hence undergo further radioactive decay by β ^– emission, until a stable nuclide is reached. If one of the radioactive intermediates has a sufficiently long half-life, it can be extracted from the fission products and used as a medical radionuclide. For example,

{}_{39}^{99}Y \overset{β^{-} (1.5 s)}{⟶} {}_{40}^{99}Z r \overset{β^{-} (21 s)}{⟶} {}_{41}^{99}N b \overset{β^{-} (15 s)}{⟶} {}_{42}^{99}M o

${}_{39}^{99}Y \overset{β^{-} (1.5 s)}{⟶} {}_{40}^{99}Z r \overset{β^{-} (21 s)}{⟶} {}_{41}^{99}N b \overset{β^{-} (15 s)}{⟶} {}_{42}^{99}M o$

The half-life of ⁹⁹ Mo is 65.9 hours, which is sufficiently long to allow it to be chemically separated from other fission fragments. Molybdenum-99 plays an important role in nuclear medicine as the parent radionuclide in the ⁹⁹ Mo- ^99m Tc generator (see Section C ). Technetium-99m is the most common radionuclide used in clinical nuclear medicine procedures today. Fission has also been used to produce ¹³¹ I and ¹³³ Xe for nuclear medicine studies.

Radionuclides produced by the fission process have the following general characteristics:

1
Fission products always have an excess of neutrons, because N/Z is substantially higher for ²³⁵ U than it is for nuclei falling in the mass range of the fission fragments, even after the fission products have expelled a few neutrons (see Fig. 2-9 ). These radionuclides therefore tend to decay by β ^– emission.
2
Fission products may be carrier free (no stable isotope of the element of interest is produced), and therefore radionuclides can be produced with high specific activity by chemical separation. (Sometimes other isotopes of the element of interest are also produced in the fission fragments. For example, high-specific-activity ¹³¹ I cannot be produced through fission because of significant contamination from ¹²⁷ I and ¹²⁹ I.)
3
The lack of specificity of the fission process is a drawback that results in a relatively low yield of the radionuclide of interest among a large amount of other radionuclides.

3 Neutron Activation

Neutrons carry no net electrical charge. Thus they are neither attracted nor repelled by atomic nuclei. When neutrons (e.g., from a nuclear reactor core) strike a target, some of the neutrons are “captured” by nuclei of the target atoms. A target nucleus may be converted into a radioactive product nucleus as a result. Such an event is called neutron activation. Two types of reactions commonly occur.

In an ( n, γ) reaction a target nucleus, , captures a neutron and is converted into a product nucleus, , which is formed in an excited state. The product nucleus immediately undergoes de-excitation to its ground state by emitting a prompt γ ray. The reaction is represented schematically as

{}_{Z}^{A}X (n, γ) {}_{Z}^{A + 1}X

${}_{Z}^{A}X (n, γ) {}_{Z}^{A + 1}X$

The target and product nuclei of this reaction represent different isotopes of the same chemical element.

A second type of reaction is the ( n,p ) reaction. In this case, the target nucleus captures a neutron and promptly ejects a proton. This reaction is represented as

{}_{Z}^{A}X (n, p) {}_{Z - 1}^{A}Y

${}_{Z}^{A}X (n, p) {}_{Z - 1}^{A}Y$

Note that the target and product nuclei for an (n,p) reaction do not represent the same chemical element.

In these examples, the products ( or ) usually are radioactive species. The quantity of radioactivity that is produced by neutron activation depends on a number of factors, including the intensity of the neutron flux and the neutron energies. This is discussed in detail in Section D . Production methods for biomedically important radionuclides produced by neutron activation are summarized in Table 5-1 .

TABLE 5-1

NEUTRON-ACTIVATED RADIONUCLIDES OF IMPORTANCE IN BIOLOGY AND MEDICINE

Radionuclide	Decay Mode	Production Reaction	Natural Abundance of Target Isotope (%) ^*	σ _c (b) ^†
¹⁴ C	β ^–	¹⁴ N(n,p) ¹⁴ C	99.6	1.81
²⁴ Na	(β ^– ,γ)	²³ Na(n,γ) ²⁴ Na	100	0.53
³² P	β ^–	³¹ P(n,γ) ³² P	100	0.19
		³² S(n,p) ³² P	95.0	0.1
³⁵ S	β ^–	³⁵ Cl(n,p) ³⁵ S	75.8	0.4
⁴² K	(β ^– ,γ)	⁴¹ K(n,γ) ⁴² K	6.7	1.2
⁵¹ Cr	(EC,γ)	⁵⁰ Cr(n,γ) ⁵¹ Cr	4.3	17
⁵⁹ Fe	(β ^– ,γ)	⁵⁸ Fe(n,γ) ⁵⁹ Fe	0.3	1.1
⁷⁵ Se	(EC,γ)	⁷⁴ Se(n, γ) ⁷⁵ Se	0.9	30
¹²⁵ I	(EC,γ)		0.1	110
¹³¹ I	(β ^– ,γ)		33.8	0.24

EC, Electron capture.

* Values from Browne E, Firestone RB: Table of Radioactive Isotopes . New York, 1986, John Wiley.

† Thermal neutron capture cross-section, in barns (b) (see “ Activation Cross-Sections ”). Values from Wang Y: Handbook of Radioactive Nuclides, Cleveland, Chemical Rubber Company, 1969.

Radionuclides produced by neutron activation have the following general characteristics:

1
Because neutrons are added to the nucleus, the products of neutron activation generally lie above the line of stability (see Fig. 2-9 ). Therefore they tend to decay by β ^– emission.
2
The most common production mode is by the (n,γ) reaction, and the products of this reaction are not carrier free because they are the same chemical element as the bombarded target material. It is possible to produce carrier-free products in a reactor by using the (n,p) reaction (e.g., ³² P from ³² S) or by activating a short-lived intermediate product, such as ¹³¹ I from ¹³¹ Te using the reaction

${}^{130}T e (n, γ) {}^{131}T e \overset{β^{-}}{⟶} {}^{131}I$ ${}^{130}T e (n, γ) {}^{131}T e \overset{β^{-}}{⟶} {}^{131}I$
3
Even in intense neutron fluxes, only a very small fraction of the target nuclei actually are activated, typically 1 : 10 ⁶ to 10 ⁹ (see Section D ). Thus an (n,γ) product may have very low specific activity because of the overwhelming presence of a large amount of unactivated stable carrier (target material).

There are a few examples of the production of electron capture (EC) decay or β ⁺ -emitting radionuclides with a nuclear reactor, for example, ⁵¹ Cr by (n,γ) activation of ⁵⁰ Cr. They may also be produced by using more complicated production techniques. An example is the production of ¹⁸ F (β ⁺ , T _1/2 = 110 min). The target material is lithium carbonate (Li ₂ CO ₃ ). The first step is the reaction

{}^{6}L i (n, γ) {}^{7}L i

${}^{6}L i (n, γ) {}^{7}L i$

Lithium-7 is very unstable and promptly disintegrates:

{}_{3}^{7}L i \to {}_{2}^{4}H e + {}_{1}^{3}H + energy

${}_{3}^{7}L i \to {}_{2}^{4}H e + {}_{1}^{3}H + energy$

Some of the energetic recoiling tritium nuclei ( ) bombard stable ¹⁶ O nuclei, causing the reaction

{}_{8}^{16}O ({}_{1}^{3}H, n) {}_{9}^{18}F

${}_{8}^{16}O ({}_{1}^{3}H, n) {}_{9}^{18}F$

Useful quantities of ¹⁸ F can be produced in this way. One problem is removal from the product (by chemical means) of the rather substantial quantity of radioactive tritium that is formed in the reaction. More satisfactory methods for producing ¹⁸ F involve the use of charged particle accelerators, as discussed in Section B .

b

Accelerator-Produced Radionuclides

1 Charged-Particle Accelerators

Charged-particle accelerators are used to accelerate electrically charged particles, such as protons, deuterons ( nuclei), and α particles ( nuclei), to very high energies. When directed onto a target material, these particles may cause nuclear reactions that result in the formation of radionuclides in a manner similar to neutron activation in a reactor. A major difference is that the particles must have very high energies, typically 10-20 MeV, to penetrate the repulsive coulomb forces surrounding the nucleus.

Two types of nuclear reactions are commonly used to produce radionuclides using a charged-particle accelerator. In a ( p,n ) reaction , the target nucleus captures a proton and promptly releases a neutron. This reaction is represented as

{}_{Z}^{A}X (p, n) {}_{Z + 1}^{A}Y

${}_{Z}^{A}X (p, n) {}_{Z + 1}^{A}Y$

This reaction can be considered the inverse of the (n,p) reaction that uses neutrons as the bombarding particle and was discussed in Section A.3 .

A second common reaction is the ( d,n ) reaction in which the accelerated particle is a deuteron (d). The target nucleus captures a deuteron from the beam and immediately releases a neutron. This reaction is represented as

{}_{Z}^{A}X (d, n) {}_{Z + 1}^{A + 1}Y

${}_{Z}^{A}X (d, n) {}_{Z + 1}^{A + 1}Y$

and results in a change of both the element (atomic number) and the mass number. In some cases, more than one neutron may be promptly released from the target nucleus after the bombarding particle has been captured. For example, a (p,2n) reaction involves the release of two neutrons following proton capture and a (d,3n) reaction involves the release of three neutrons following deuteron capture. Some accelerators also use alpha-particles to bombard a target and produce radionuclides. Indium-111 can be produced in this way using the reaction ¹⁰⁹ Ag(α,2n) ¹¹¹ In.

Van de Graaff accelerators, linear accelerators, cyclotrons, and variations of cyclotrons have been used to accelerate charged particles. The cyclotron is the most widely used form of particle accelerator for production of medically important radionuclides. Many larger institutions have their own compact biomedical cyclotrons for onsite production of the shorter-lived, positron-emitting radionuclides. The principles and design of cyclotrons dedicated to production of radionuclides for nuclear medicine are described briefly.

2 Cyclotron Principles

A cyclotron consists of a pair of hollow, semicircular metal electrodes (called dees because of their shape), positioned between the poles of a large electromagnet ( Fig. 5-3 ). The dees are separated from one another by a narrow gap. Near the center of the dees is an ion source, S, (typically an electrical arc device in a gas) that is used to generate the charged particles. All these components are contained in a vacuum tank at ~10 ^–3 Pa(~10 ^–8 atm).

FIGURE 5-3, Schematic representation of a positive ion cyclotron: top ( left ) and side ( right ) views. The accelerating voltage is applied by a high-frequency oscillator to the two “dees.” S is a source of positive ions.

During operation, particles are generated in bursts by the ion source, and a high-frequency alternating current (AC) voltage generated by a high-frequency oscillator (typically 30 kV, 25-30 MHz) is applied across the dees. The particles are injected into the gap and immediately are accelerated toward one of the dees by the electrical field generated by the applied AC voltage. Inside the dee there is no electrical field, but because the particles are in a magnetic field, they follow a curved, circular path around to the opposite side of the dee. The AC voltage frequency is such that the particles arrive at the gap just as the voltage across the dees reaches its maximum value (30 kV) in the opposite direction. The particles are accelerated across the gap, gaining about 30 keV of energy in the process, and then continue on a circular path within the opposite dee.

Each time the particles cross the gap they gain energy, so the orbital radius continuously increases and the particles follow an outwardly spiraling path. The increasing speed of the particles exactly compensates for the increasing distance traveled per half orbit, and they continue to arrive back at the gap exactly in phase with the AC voltage. This condition applies so long as the charge-to-mass ratio of the accelerated particles remains constant. Because of their large relativistic mass increase, even at relatively low energies (~100 keV), it is not practical to accelerate electrons in a cyclotron. Protons can be accelerated to 20-30 MeV, and heavier particles can be accelerated to even higher energies (in proportion to their rest mass), before relativistic mass changes become limiting. *

* Even at low energies, protons, deuterons, and α particles gain some mass when accelerated in a cyclotron. Magnetic “field shaping” is used in the cyclotron to compensate for this effect.

Higher particle energies can be achieved in a variation of the cyclotron called the synchrocyclotron or synchrotron, in which the AC voltage frequency changes as the particles spiral outward and gain energy. These machines are used in high-energy nuclear physics research.

The energy of particles accelerated in a cyclotron is given by

E (MeV) \approx 4.8 \times 10^{- 3} {(H \times R \times Z)}^{2} / A

$E (MeV) \approx 4.8 \times 10^{- 3} {(H \times R \times Z)}^{2} / A$

in which H is the magnetic field strength in tesla, R is the radius of the particle orbit in centimeters, and Z and A are the atomic number (charge) and mass number of the accelerated particles, respectively. The energies that can be achieved are limited by the magnetic field strength and the dee size. In a typical biomedical cyclotron with magnetic field strength of 1.5 tesla and a dee diameter of 76 cm, protons (Z = 1, A = 1) and α particles (Z = 2, A = 4) can be accelerated to approximately 15 MeV and deuterons (Z = 1, A = 2) to approximately 8 MeV.

When the particles reach the maximum orbital radius allowed within the cyclotron dees, they may be directed onto a target placed directly in the orbiting beam path (internal beam irradiation). More commonly, the beam is extracted from the cyclotron and directed onto an external target (external-beam radiation). Typical beam currents at the target are in the range of 50-100 µA. For cyclotrons using positively charged particles (positive ion cyclotron), the beam is electrostatically deflected by a negatively charged plate and directed to the target ( Fig. 5-3 ). Unfortunately electrostatic deflectors are relatively inefficient, as much as 30% of the beam current being lost during extraction. This “lost” beam activates the internal parts of the cyclotron, thus making servicing and maintenance of the cyclotron difficult.

In a negative-ion cyclotron, negatively charged ions (e.g. H ^– , a proton plus two electrons) are generated and then accelerated in the same manner as the positive ions in a positive-ion cyclotron (but in the opposite direction because of the different polarity). When the negatively charged ions reach the outermost orbit within the dee electrodes, they are passed through a thin (5-25 µm) carbon foil, which strips off the electrons and converts the charge on the particle from negative to positive. The interaction of the magnetic beam with this positive ion bends its direction of motion outward and onto the target ( Fig. 5-4 ). The negative-ion cyclotron has a beam extraction efficiency close to 100% and can therefore be described as a “cold” machine that requires minimal levels of shielding. Furthermore, two beams can be extracted simultaneously by positioning a carbon-stripping foil part way into the path of the beam, such that only a portion of the beam is extracted to a target. The remainder of the beam is allowed to continue to orbit and then is extracted with a second stripping foil onto a different target ( Fig. 5-4 ). This allows two different radionuclides to be prepared simultaneously. One disadvantage of negative-ion cyclotrons is the requirement for a much higher vacuum (typically 10 ^–5 Pa compared with 10 ^–3 Pa for positive ion machines) because of the unstable nature of the H ^– ion, the most commonly used particle in negative ion cyclotrons.

FIGURE 5-4, Left, Schematic representation of a negative-ion cyclotron. The carbon stripping foils remove two electrons from negative hydrogen (H – ) ions, converting them into protons (p + ) that bend in the opposite direction in the applied magnetic field. Right, The first stripping foil intersects only part of the beam, allowing two beams to be extracted simultaneously.

3 Cyclotron-Produced Radionuclides

Cyclotrons are used to produce a variety of radionuclides for nuclear medicine, some of which are listed in Table 5-2 . General characteristics of cyclotron-produced radionuclides include the following:

1
Positive charge is added to the nucleus in most activation processes. Therefore, the products lie below the line of stability (see Fig. 2-9 ) and tend to decay by EC or β ⁺ emission.
2
Addition of positive charge to the nucleus changes its atomic number. Therefore cyclotron-activation products usually are carrier free.
3
Cyclotrons generally produce smaller quantities of radioactivity than are obtained from nuclear reactors. In part this results from generally smaller activation cross-sections for charged particles as compared with neutron irradiation (see Section D ) and in part from lower beam intensities obtained in cyclotrons as compared with nuclear reactors.

TABLE 5-2

SOME CYCLOTRON-PRODUCED RADIONUCLIDES USED IN NUCLEAR MEDICINE

Product	Decay Mode	Common Production Reaction	Natural Abundance of Target Isotope * (%)	Energy Threshold (MeV) ^†
¹¹ C	β ⁺ , EC	¹⁴ N(p,α) ¹¹ C	99.6	3.1
¹¹ C	β ⁺ , EC	¹⁰ B(d,n) ¹¹ C	19.9	0
¹³ N	β ⁺	¹⁶ O(p,α) ¹³ N	99.8	5.5
¹³ N	β ⁺	¹² C(d,n) ¹³ N	98.9	0.35
¹⁵ O	β ⁺	¹⁴ N(d,n) ¹⁵ O	99.6	0
¹⁵ O	β ⁺	¹⁵ N(p,n) ¹⁵ O	0.37	—
¹⁸ F	β ⁺ , EC	¹⁸ O(p,n) ¹⁸ F	0.20	2.57
¹⁸ F	β ⁺ , EC	²⁰ Ne(d,α) ¹⁸ F	90.5	0
⁶⁷ Ga	(EC,γ)	⁶⁸ Zn(p,2n) ⁶⁷ Ga	18.8	5.96
¹¹¹ In	(EC,γ)	¹⁰⁹ Ag(α,2n) ¹¹¹ In	48.2	—
¹¹¹ In	(EC,γ)	¹¹¹ Cd(p,n) ¹¹¹ In	12.8	—
¹²³ I	(EC,γ)	¹²² Te(d,n) ¹²³ I	2.6	—
¹²³ I	(EC,γ)	¹²⁴ Te(p,3n) ¹²³ I	4.8	—
²⁰¹ Tl	(EC,γ)	²⁰¹ Hg(d,2n) ²⁰¹ Tl	13.2	—

EC, electron capture.

* Values from Browne E, Firestone RB: Table of Radioactive Isotopes . New York, 1986, John Wiley.

† Values from Helus F, Colombetti LG: Radionuclides Production , Vols I, II. Boca Raton, 1983, CRC Press.

Cyclotron products are attractive for nuclear medicine imaging studies because of the high photon/particle emission ratios that are obtained in β ⁺ and EC decay. Of special interest are the short-lived positron emitters ¹¹ C ( T _1/2 = 20.4 min), ¹³ N ( T _1/2 = 9.97 min), and ¹⁵ O ( T _1/2 = 2.03 min). These radionuclides represent elements that are important constituents of all biologic substances, and they can be used to label a wide variety of biologically relevant tracers. Because of their very short lifetimes, these positron-emitting radionuclides must be prepared on site with a dedicated biomedical cyclotron. The high cost of owning and operating such machines has impeded their widespread use. Nevertheless, because of the importance of several positron emitter–labeled radiopharmaceuticals, there are now many hundreds of cyclotrons worldwide producing short-lived positron-emitting isotopes for nuclear medicine imaging studies. A typical biomedical cyclotron is shown in Figure 5-5 .

FIGURE 5-5, Photograph of a negative-ion biomedical cyclotron. Left, Cyclotron within concrete shield. Right, The cyclotron itself.

Fluorine-18 ( T _1/2 = 110 min) is another important positron-emitting radionuclide. One of its main applications is in the labeling of a glucose analog, ¹⁸ F-fluorodeoxyglucose (FDG), which provides a measure of the metabolic rate for glucose in the cells of the body. The longer half-life of the ¹⁸ F label allows FDG to be produced in regional distribution centers and shipped to hospitals tens or even hundreds of miles away. FDG is the most widely used positron-emitting radiopharmaceutical with a wide range of clinical applications in the heart, and brain and especially in cancer imaging. (See Chapter 18 , Section F.)

c

Radionuclide Generators

A radionuclide generator consists of a parent-daughter radionuclide pair contained in an apparatus that permits separation and extraction of the daughter from the parent. The daughter product activity is replenished continuously by decay of the parent and may be extracted repeatedly.

Table 5-3 lists some radionuclide generators of interest to nuclear medicine. They are an important source of metastable radionuclides. The most important generator is the ⁹⁹ Mo- ^99m Tc system, because of the widespread use of ^99m Tc for radionuclide imaging. Technetium-99m emits γ rays (140 keV) that are very favorable for use with a gamma camera ( Chapter 13 ). It has a reasonable half-life (6 hours), delivers a relatively low radiation dose per emitted γ ray ( Chapter 22 ), and can be used to label a wide variety of imaging agents. More than 1850 TBq (50,000 Ci) of ⁹⁹ Mo per week are required to meet the worldwide requirements for nuclear medicine procedures.

TABLE 5-3

SOME RADIONUCLIDE GENERATORS USED IN NUCLEAR MEDICINE

Daughter *	Decay Mode	T _1/2	Parent	T _1/2
⁶² Cu	β ⁺ ,EC	9.7 min	⁶² Zn	9.3 hr
⁶⁸ Ga	β ⁺ ,EC	68 min	⁶⁸ Ge	271 d
⁸² Rb	β ⁺ ,EC	1.3 min	⁸² Sr	25 d
^87m Sr	IT	2.8 hr	⁸⁷ Y	80 hr
^99m Tc	IT	6 hr	⁹⁹ Mo	66 hr
^113m In	IT	100 min	¹¹³ Sn	120 d

EC, electron capture; IT, isomeric transition.

* Generator product.

A ⁹⁹ Mo- ^99m Tc generator is shown in Figure 5-6 . The parent ⁹⁹ Mo activity in the form of molybdate ion, is bound to an alumina (Al ₂ O ₃ ) column. The daughter ^99m Tc activity, produced in the form of (pertechnetate), is not as strongly bound to alumina and is eluted from the column with 5 to 25 mL of normal saline. Technetium-99m activity builds up again after an elution and maximum activity is available about 24 hours later ( Equation 4-28 ); however, usable quantities are available 3 to 6 hours later. Commercially prepared generators are sterilized, well shielded, and largely automated in operation. Typically they are used for approximately 1 week and then discarded because of natural decay of the ⁹⁹ Mo parent.

FIGURE 5-6, Cut-away view of a 99 Mo- 99m Tc generator.

Decay of the ⁹⁹ Mo- ^99m Tc parent-daughter pair is an example of transient equilibrium (see Chapter 4 , Section G.3). Equation 4-25 and Figure 4-8 describe the buildup and decay of activity for such a pair. Under idealized conditions, and a branching ratio of 0.876, the ratio of ^99m Tc / ⁹⁹ Mo activity in a generator in a state of transient equilibrium (see Equation 4-27 ) would be approximately 0.96, and the time to maximum activity following an elution ( Equation 4-28 ) would be approximately 23 hours.

However, these equations do not accurately predict the amount of ^99m Tc actually obtained in individual elutions, because most generators do not yield 100% of the available activity. Typical generator elution efficiencies are 80% to 90%, depending on the size and type of generator, volume of eluant, and so on. Furthermore, the efficiency can vary from one elution to the next. In practice, efficiency variations of ±10% or more can occur in successive elutions of the same generator. These may be caused by chemical changes in the column, including some that are caused by the intense radiation levels. Failure to keep a “dry” column in a dry state also can substantially degrade elution efficiency. These issues, as well as other complexities of ⁹⁹ Mo- ^99m Tc generators, are discussed in detail in references .

If 90% of the ^99m Tc activity in a generator is removed during an elution, the activity obtained would be 10% less than predicted from Equation 4-25 and Figure 4-8 . Furthermore, the 10% residual ^99m Tc activity left in the generator becomes “ A _d (0)” in Equation 4-25 for the next elution interval. This activity provides a “jump start” for regrowth of ^99m Tc in the generator, thereby shortening the time to maximum activity after an elution from that predicted by Equation 4-28 .

Figure 5-7A , shows the available ^99m Tc activity, relative to parent ⁹⁹ Mo activity, for a generator that is eluted with 90% efficiency at 24-hour intervals, starting at t = 0 hours. Under these conditions, the activity obtained is approximately 77% of the parent ⁹⁹ Mo activity in the generator at the time of elution, and the time to maximum activity after an elution is shortened to approximately 21 hours.

FIGURE 5-7, Orange lines: 99 Mo activity in a generator, normalized to 1.0 at t = 0. Blue lines: 99m Tc activity available for elution, assuming 90% elution efficiency. A , Generator eluted at regular 24-hour intervals. B, Generator eluted at irregular intervals. 99m Tc activities also are expressed relative to the 99 Mo activity in the generator, and assume consistent 90% elution efficiency from one elution to the next.

If a generator is eluted at irregular intervals, the situation becomes more complicated, because the residual ^99m Tc activity left in the generator varies from one elution to the next. In this situation, the ^99m Tc activity in generator can be predicted using Equation 4-25 , using the ideal versus actual yield to estimate the amount of residual ^99m Tc for A _d (0) for the next elution interval. Figure 5-7B , shows the results of such a calculation for elutions at 0, 24, 30, 48, and 96 hours, each done with 90% elution efficiency.

In a practical environment, it is useful to keep records comparing generator yields to those predicted from the idealized equations. This can be helpful for identifying “low-yield” generators, as well as possible problems that may develop in an individual generator. A simplified equation that can be used to predict yields for elutions performed at regular 24-hour or other similarly “long” intervals is

Y_{2} = \frac{Y_{1} \times (e^{- λ_{p} Δ t_{2}} - e^{- λ_{d} Δ t_{2}})}{[1 - e^{- (λ_{d} - λ_{p}) Δ t_{1}}]}

$Y_{2} = \frac{Y_{1} \times (e^{- λ_{p} Δ t_{2}} - e^{- λ_{d} Δ t_{2}})}{[1 - e^{- (λ_{d} - λ_{p}) Δ t_{1}}]}$

Here, Y ₂ is the predicted yield of an elution (in units of activity), Y ₁ is the actual yield of the immediately preceding elution, λ t ₂ is the time since that elution, λ t ₁ is the elution interval between that elution and the one immediately preceding it (i.e., prior to the elution yielding Y ₁ ), and λ _p and λ _d are the decay constants of ⁹⁹ Mo (~0.0105 hr ^-1 ) and ^99m Tc (~0.115 hr ^-1 ), respectively. This equation assumes that the elution efficiency is constant from one elution to the next and that there is insignificant carryover of residual ^99m Tc activity in the column at the time of the next elution. The latter condition is reasonably satisfied for 24-hr or similarly long elution intervals that allow for virtually complete decay of any ^99m Tc left over from previous elutions.

Molybdenum-99 activity is obtained by separation from fission fragments produced in a target containing uranium or by (n,γ) activation of stable molybdenum (23.8% ⁹⁸ Mo). The former, sometimes called fission moly, has significantly higher specific activity and is the production method of choice for large quantities. The reaction by which it is produced sometimes is called an (n,f) reaction, indicating neutron irradiation causing fission. The production of ⁹⁹ Mo is described in detail in reference .

Fission moly is produced by inserting a target (typically shaped as a pin, cylinder, or plate) containing natural uranium, enriched with ²³⁵ U, via an access port into the reactor core. The target is encapsulated in aluminum or stainless steel. Fission neutrons from the reactor core induce fission reactions in the target, as shown in Equation 5-1 . Molybdenum-99 is one of the more abundant fission products (6.1% of fission products), but a wide variety of others are produced as well (see Fig. 5-2B ).

After a suitable period of irradiation (typically 5-7 days), the uranium target is removed, allowed to cool, and dissolved either using an acid or alkaline dissolution process. The ⁹⁹ Mo is then extracted by chemical means. Special care is required to assure that the many other radioactive fission products do not contaminate the desired ⁹⁹ Mo product. As well, a large fraction of the original ²³⁵ U remains in the solution and must be stored as long-term radioactive waste.

The amount of stable molybdenum produced by the (n,f) reaction is small, as compared with its concentration in a target used for neutron activation of ⁹⁸ Mo. Therefore the specific activity of ⁹⁹ Mo in “fission moly” is much higher, and can be loaded into generators containing much smaller quantities of the alumina column.

The volume of alumina required in a ⁹⁹ Mo- ^99m Tc generator is determined essentially by the amount of stable ⁹⁹ Mo carrier that is present. Therefore “fission moly” generators require much smaller volumes of alumina per unit of ⁹⁹ Mo activity. They can be eluted with very small volumes of normal saline (∼5 mL), which is useful in some dynamic imaging studies requiring bolus injections of very small volumes of high activity (740 MBq, 20 mCi) of ^99m Tc.

One problem with ^99m Tc generators is ⁹⁹ Mo “breakthrough,” that is, partial elution of the ⁹⁹ Mo parent along with ^99m Tc from the generator. From the standpoint of patient radiation safety, the amount of ⁹⁹ Mo should be kept to a minimum. Maximum amounts, according to Nuclear Regulatory Commission regulations, are 0.15 Bq ⁹⁹ Mo per kBq ^99m Tc (0.15 µCi ⁹⁹ Mo per mCi ^99m Tc). It is possible to assay ⁹⁹ Mo activity in the presence of much larger ^99m Tc activity using NaI(Tl) counting systems by surrounding the sample with approximately 3 mm of lead, which is an efficient absorber of the 140 keV γ rays of ^99m Tc but relatively transparent to the 740-780-keV γ rays of ⁹⁹ Mo. Thus small quantities of ⁹⁹ Mo can be detected in the presence of much larger amounts of ^99m Tc. Some dose calibrators are provided with a lead-lined container called a “moly shield” specifically for this purpose. Other radioactive contaminants also are occasionally found in ⁹⁹ Mo- ^99m Tc generator eluate.

A second major concern is breakthrough of aluminum ion, which interferes with labeling processes and also can cause clumping of red blood cells and possible microemboli. Maximum permissible levels are 10 µg aluminum per mL of ^99m Tc solution. Chemical test kits are available from generator manufacturers to test for the presence of aluminum ion.

You're Reading a Preview

Become a Clinical Tree membership for Full access and enjoy Unlimited articles

Become membership

If you are a member. Log in here

a

Reactor-Produced Radionuclides

1

Reactor Principles

2

Fission Fragments

3

Neutron Activation

b

Accelerator-Produced Radionuclides

1

Charged-Particle Accelerators

2

Cyclotron Principles

3

Cyclotron-Produced Radionuclides

c

Radionuclide Generators

You're Reading a Preview

Related Posts

Convolution

The Fourier Transform

Effective Dose Equivalent (mSv/MBq) and Radiation Absorbed Dose Estimates (mGy/MBq) to Adult Subjects from Selected Internally Administered Radiopharmaceuticals

Mass Attenuation Coefficients for Water, NaI(Tl), Bi 4 Ge 3 O 12 , Cd 0.8 Zn 0.2 Te, and Lead