Arsenic-77 (T _1/2 = 1.6 d) is a promising medium-range β⁻-emitting radionuclide (E _βmax = 0.7 MeV) for in vivo radionuclide therapy owing to its favorable nuclear decay characteristics (Neves et al. 2002), decaying by 100 % via electron emission to stable ⁷⁷Se. The emission of low-energy gamma photons [239 keV (1.6 %), 520 keV (0.5 %)] allows imaging for evaluation of in vivo radiopharmaceutical localization and pharmacokinetics (Jennewein et al. 2006). The 1.6-d half-life of ⁷⁷As is compatible with the pharmacokinetics of several targeting biomolecules, including monoclonal antibodies (MAbs) (Emran and Phillips 1991) and is amenable for radiochemical synthetic procedures which may require several hours. The chemistry for effective labeling of biologically relevant molecules with arsenic radioisotopes is slowly evolving (Leonard 1991; Emran 1984; Hosain et al. 1982; Jennewein et al. 2003a, b; Zhu et al. 1997; Mirzadeh and Lambrecht 1996). Arsenic forms stable covalent bonds with carbon and sulfur and can substitute phosphorus in certain compounds with minimal alteration of the biological activity of the parent molecules. In addition, the potential use of ⁷⁷As for endoradiotherapeutic radiopharmaceuticals has also been proposed (Maki and Murakami 1974).

Reactor production of ⁷⁷As involves the As reaction, where GeO₂ or Ge metal targets of natural abundance are irradiated to undergo (n,γ) activation to produce ⁷⁷Ge (Table 5.2), which subsequently decays by β⁻ emission (t _½ = 11.3 h) to yield ⁷⁷As (Jennewein et al. 2005; Maki and Murakami 1974). Natural Ge is a mixture of five isotopes, ⁷⁰Ge, ⁷²Ge, ⁷³Ge, ⁷⁴Ge, and ⁷⁶Ge, with relative abundances of 21.23 %, 27.66 %, 7.72 %, 35.94 %, and 7.45 %, respectively (Jennewein et al. 2006). Neutron irradiation of this isotopic mixture results produces a mixture containing ^71mGe, ⁷¹Ge,⁷⁰Ga,⁷²Ga,⁷³Ga, ⁷⁴Ga, ^75mGe,⁷⁵Ge, ⁷⁶Ga, ^77mGe,⁷⁷Ge, ^69mZn, ⁶⁹Zn,^71mZn,⁷¹Zn, and ⁷³Zn (Erdtmann 1976). With the exception of ⁷¹Ge (T _½ = 11.2 d), the other activated Ge radionuclides are short-lived. The production of radioactive Ga and Zn radioisotopes is negligible owing to the very low neutron cross sections. On the other hand, 24 % of the ^77mGe produced (t _½ = 54 s) rapidly decays to ⁷⁷Ge (t _½ = 11.3 h), while the remainder disintegrates directly to ⁷⁷As (t _½ = 38.7 h) by β⁻ decay, which finally decays to stable ⁷⁷Se. During target cooling, most of the short-lived radionuclides decay, which not only facilitates the in-growth of ⁷⁷As to maximum levels but also reduces the radiation dose to a considerable extent.

The production, purification, and application of ⁷⁷As has been an active area of research in nuclear medicine and radiochemistry communities and ⁷⁷As can be cyclotron-/accelerator-produced by the deuteron-induced reaction on enriched ⁷⁶Ge targets by the ⁷⁶Ge(d,n) ⁷⁷As reaction, but this route has not been studied in detail (Jennewein et al. 2005). For processing, ⁷⁷As can be separated from cyclotron- or reactor-irradiated germanium or germanium oxide targets by methods involving distillation, coprecipitation, solvent extraction, thin layer chromatography, etc. (Jennewein et al. 2005; Maki and Murakami 1974; Chattopadhayay et al. 2007; Bokhari et al. 2009; Mirzadeh and Lambrecht 1996). Four successful preparative-scale separations of ⁷⁷As from Ge for clinical applications have been reported in the literature (Jennewein et al. 2005; Chattopadhayay et al. 2007; Bokhari et al. 2009; Chakravarty et al. 2011). One method is based on the formation of soluble GeF₆ ⁻² in concentrated hydrofluoric acid and conversion of As to AsI₃ by the addition of KI, followed by separation using a polystyrene-based solid-phase extraction system (Jeenewien et al. 2005). The method reported by Chattopadhayay et al. (2007) involves precipitation of GeO₂ followed by centrifugation or filtration and subsequent purification using six cycles of solvent extraction (twice with CCl₄ and four times with benzene). In an alternative approach, the target is initially dissolved in aqua regia, followed by removal of excess acid by evaporation, reconstitution of the feed in basic media, and passage of the solution through a hydrous zirconium oxide (HZO) column (Bokhari et al. 2009). The Ge is quantitatively retained on HZO, while ⁷⁷As is present in the effluent. The method adapted by BARC (Chakravarty et al. 2011) consists of a two-step ion-exchange procedure using a polymer-embedded nanocrystalline titania sorbent. The first step involves removal of the bulk Ge from ⁷⁷As and the second step involves purification and concentration.

Copper-67 (T _1/2 = 2.6 d) is a medium-energy β⁻ emitter with an energy of 0.6 MeV_max (141 keV average), with strong gamma emissions at 91.266 (7 %), 93.31 (16 %), and 184.57 keV (48.7 %) (Junde et al. 2005). Because of the degree of soft tissue penetration, the medium-energy β⁻ particle permits efficient use of ⁶⁷Cu for effective irradiation and sterilization of cells in solid tumors up to 4 mm in diameter. The range of the ⁶⁷Cu beta particle in tissue is of the same order as a cell diameter (Qaim 2001). The chemistry of copper is suitable for labeling purposes and precludes the concentrate in sensitive body tissues such as the bone marrow. The 2.6-d half-life is suitable for imaging slow in vivo pharmacokinetics with agents such as monoclonal antibodies and other carrier molecules without appreciable decay loss (DeNardo et al. 1999). The presence of relatively low-energy gamma photos can be effectively detected with Anger cameras [(184 keV 948.7 %), 93.3 (16.1 %)] and is useful for imaging studies (Shen et al. 1996). In vivo, ⁶⁷Cu in the inorganic form has favorable short-term retention, and both ⁶⁷Cu and its stable ⁶⁷Zn decay product are nontoxic since Cu and Zn are essential trace nutrients (Blower et al. 1996). Copper is not a skeletal or organ seeker and ⁶⁷Cu has a biological half-life of the same order as its radiological half-life (Johnson et al. 1992; Linder and Hazegh-Azam 1996).

Unfortunately^{, 67}Cu cannot be effectively produced in a research reactor in sufficient activity levels to make this route practical. Reactor production of ⁶⁷Cu via the ⁶⁷Zn(n, p)⁶⁷Cu reaction has been reported by bombarding ⁶⁷Zn with neutrons in high-flux reactors having a high epithermal neutron component; however, the cross-section values are low (Mausner et al. 1998; Johnsen et al. 2015). Production of ⁶⁷Cu can be much more effectively carried out following using accelerator-based pathways (Katabuchi et al. 2008; Medvedev et al. 2012; Smith et al. 2012; Szelecsényi et al. 2009) and the cyclotron production of ⁶⁷Cu and different nuclear reactions are discussed and summarized in Table 6.​2 in Chap.​ 6.

Erbium-169 (T _1/2 = 9.4 d) emits beta particle ([E _β(max) = 351 keV (55 %), 342 (45 %)]) with an average soft tissue range of 0.3 mm and is used in targeted therapy as a soft β⁻-emitting radionuclide, primarily for synovectomy of small joints as described in Chap.​ 14. Production of ¹⁶⁹Er can be conveniently carried out through neutron capture of stable ¹⁶⁸Er by the ¹⁶⁸Er(n, γ)¹⁶⁹Er reaction. Naturally occurring erbium is composed of six stable isotopes, which include ¹⁶²Er (0.14 %), ¹⁶⁴Er (1.61 %),¹⁶⁶Er (33.6 %), ¹⁶⁷Er (22.95 %),¹⁶⁸Er (26.8 %), and ¹⁷⁰Er (14.9 %). Neutron irradiation of ^nat Er not only leads to the coproduction of ¹⁶⁵Er and ¹⁷¹Er as impurities but also produces low-SA ¹⁶⁹Er. In order to circumvent such drawbacks, isotopically enriched ¹⁶⁸Er targets are generally used for ¹⁶⁸Er production. Owing to the poor neutron capture cross section of this process (σ = 1.95 barn), however, the activation product obtained is in only low specific activity.

Reactor production and electrochemical purification of ¹⁶⁹Er for vivo therapeutic applications have been reported (Chakravarty et al. 2014). Despite having useful radiation emission characteristics, poor specific activity is thus a major impediment that restricts the use of ¹⁶⁹Er for radiosynovectomy. However, use of ¹⁶⁹Er for the labeling of tumor receptor or antigen-targeting vectors is precluded because of the requirement for much higher specific activities for these applications. ¹⁶⁹Er is the most preferred choice for RSV of digital joints such as metacarpophalangeal metatarsophalangeal and digital interphalangeal joints (Karavida and Notopoulos 2010). Erbium-169-labeled silicate/citrate colloid is widely used as in the treatment of refractory painful arthritis (Gumpel et al. 1979; Bouvier et al. 1983; Kahan et al. 2004). Low volumes containing 0.5–1 mCi (20–40 Mbq) with a maximum of 20 mCi (750 Mbq) per administration are used for this application.

Gold-198 (T _1/2 = 2.7 d) has been of interest for many years and decays with emission a β⁻ particle with a maximum energy of 0.96 MeV (99 %) suitable for therapeutic applications. However, the high abundance of high-energy gamma photons (412 keV) (95.6 %) can be a significant disadvantage. Because of the unique properties of gold nanoparticles, in recent years there has been interest in designing and developing ¹⁹⁸Au nanoparticles for tumor therapy applications (Chanda et al. 2010). The ¹⁹⁸Au is reactor-produced by direct (n,γ) activation by the ¹⁹⁷Au(n, γ)¹⁹⁸Au reaction which has a very high thermal neutron cross section (σ 26,500 barn). The target used is either gold foil or metal and the neutron-irradiated target is dissolved in aqua regia (3:1 HCl/HNO₃) and partially evaporated to yield HAuCl₄ in dilute HCl (normally 0.05 M). In order to insure the product is free from trace radionuclide impurities, the AuCl₄ ⁻ is extracted into organic solvents such as chloroform, dichloromethane, or ethyl acetate as its tetrabutyl ammonium salt, TBA (AuCl₄).

Gold-199 [T _1/2 = 3.14 d] also decays by emission of β⁻ particles (E _β(max) = 452 keV (6.5 %), 293 keV (72 %), 244 keV (21.5 %)) and gamma photons [(E _γ = 494 keV (97 %), 158 keV (36.9 %)]. Although the high gamma abundance is a disadvantage, interest in the potential use of ¹⁹⁹Au for therapeutic applications has persisted. Radiolabeling of gold clusters containing 11 gold atoms site selectively to monoclonal antibodies has been reported (Hainfeld 1987; Vaughan and Varley 1988; Hainfeld 1988; Hainfeld et al. 1990). Production of ¹⁹⁹Au is performed by an indirect method, involving neutron capture of ¹⁹⁸Pt followed by β⁻ decay of the ¹⁹⁹Pt (T _1/2 = 30.80 m) to generate ¹⁹⁹Au (Table 5.1). The reaction cross section is moderate with a value of 3.6 barn.

The use of natural platinum leads to the concomitant production of ¹⁹³Pt and ¹⁹⁷Pt which results in the production of inactive ¹⁹⁷Au, thereby decreasing the ¹⁹⁹Au SA. Use of enriched targets is thus necessary to avoid handling high activity levels arising from decay of other Pt radionuclides. The indirect method of production leads to the production of NCA ¹⁹⁹Au and the radiochemical separation procedure are well described in the literature (Kolsky and Mausner 1993; Bonardi et al. 2001; Das et al. 1999). Since the ¹⁹⁹Au is available NCA, this radionuclide is of interest for radiolabeling peptides and antibodies. However, stabilization of gold by chelation is a challenging task, due to the inertness of this metal.

Holmium-166 (T _1/2 = 26.83 h) emits β⁻ particles with a maximum energy of 1.85 MeV [E _β(max) = 1854 keV (50.0 %), 1774 keV (48.7 %)] and γ rays with energies of 80.6 (6.2 %) and 1379 keV (1.13 %). The 80 keV photon is within the optimal energy range for effective gamma camera imaging. The two β⁻ emissions have mean soft tissue penetration range of 4 mm and a maximum range of 8.7 mm in soft tissues (Marque et al. 2005; Nijsen et al. 1999). Holmium-166 is an excellent radionuclide suited for in vivo therapeutic applications (Dadachova et al. 1997; Hong et al. 2002, 2004; Chakraborty et al. 2001; Chakraborty et al. 2006b; Majali et al. 2001, 2002; Das et al. 2009a, b), although the relatively short half-life may often limit distribution of ¹⁶⁶Ho-labeled radiopharmaceuticals to within only short proximity to the production site. Although ¹⁶⁶Dy/¹⁶⁶Ho radionuclide generator prototypes have been described, these systems are not yet further developed or practical for routine clinical use (see Chap.​ 7). The most widely utilized ¹⁶⁶Ho radiopharmaceutical which has been described for therapy rather than palliation of bone cancer is ¹⁶⁶Ho-DOTMP (DOTMP = 1,4,7,10-tetraazacyclododecane-1,4,7,10-tetramethylene-phosphonic acid; Fig. 11.​12, Chap.​ 11), which targets multiple myeloma, a cancer of the plasma white blood cells, that is presented in the bone marrow of these patients (Breits et al. 2006; Rajendren et al. 2002). Utility of ¹⁶⁶Ho-EDTMP for bone pain palliation has also been explored (Appelbaum et al. 1992; Bahrami-Samani et al. 2010; Sohaib et al. 2011).

For the treatment of unresectable liver cancer (see Chap.​ 11), several ¹⁶⁶Ho-labeled agents are in various stages of evaluation and include the percutaneous administration of the ¹⁶⁶Ho–chitosan complex (PHI) (Rajendren et al. 2002), the ¹⁶⁶Ho–oxine–lipiodol complex (Das et al. 2009a, b) and ¹⁶⁶Ho–polylactic acid microspheres (Nijsen et al. 1999).

Holmium-166 can be produced by direct neutron irradiation of ¹⁶⁵Ho, which is 100 % abundant (Table 5.1) and from decay of reactor produced ¹⁶⁶Dy (Fig. 5.2). Since the thermal neutron activation cross section is 66 barn, relatively high activity levels of ¹⁶⁶Ho can be produced, but with only moderate specific activity, because only a very small portion (~0.31 %) of target atoms are actually converted to ¹⁶⁶Ho at saturation yields (Fig. 5.3).

Although the modest SA of the ¹⁶⁶Ho produced by direct (n,γ) activation is not useful for radiolabeling receptor-specific biomolecules, these ¹⁶⁶Ho SA values are sufficient for several applications where much higher radiopharmaceutical masses can be administered, which include bone marrow ablation, therapy of hepatocellular carcinoma, and radiosynovectomy. For targeting low populations of binding sites such as receptor binding, the use of NCA ¹⁶⁶Ho is required, and the availability of NCA ¹⁶⁶Ho from a ¹⁶⁶Dy/¹⁶⁶Ho generator would be expected to be of interest for these applications. Potential of ¹⁶⁶Dy/¹⁶⁶Ho in vivo generator system for therapeutic use have also been explored (Ferro-Flores et al. 2003; 2004). Production of ¹⁶⁶Ho through hot atom reactions has also been explored (Nassan et al. 2011; Zeisler and Weber 1998).

For such applications, NCA ¹⁶⁶Ho can be produced via an indirect route (Table 5.2) in which an enriched ¹⁶⁴Dy target undergoes two sequential neutron captures (2n, γ) to provide ¹⁶⁶Dy that subsequently decays by β⁻ emission to the desired ¹⁶⁶Ho product by the route. This method necessitates availability of efficient methods to perform chemical separation of microscopic levels of ¹⁶⁶Ho from macroscopic amounts of ¹⁶⁶Dy. The initial neutron capture cross-section values are effectively 2731 barn thermal and 932 barn epithermal to form ¹⁶⁵Dy with a half-life of 2.33 h. Although the half-life of this intermediate radionuclide is relatively short, ¹⁶⁵Dy has thermal and epithermal neutron capture cross sections of 3600 and 22,000 barn, respectively, and thus a significant percentage of ¹⁶⁵Dy atoms undergo second neutron capture reaction to form ¹⁶⁶Dy. One milligram of enriched ¹⁶⁴Dy irradiated over 155 h at a thermal flux of 4 × 10¹⁴ neutrons cm⁻² s⁻¹ thermal and an epithermal flux of 1.6 × 10¹³ neutrons cm⁻² s⁻¹ at the MURR has been shown to produce close to the theoretical yield of 1.2 Ci of ¹⁶⁶Dy (Ma et al. 1996). Similar production studies at the ORNL HFIR have demonstrated production of ¹⁶⁶Dy and separation of ¹⁶⁶Ho (Dadchova et al. 1994; 1995; Lahiri et al. 2004). Efforts for separation of NCA ¹⁶⁶Ho from ¹⁶⁶Dy are described in Chap.​ 6.

Iodine-131 (T _1/2 = 8.02 d) has been widely used for therapy for many years and has many advantages and has traditionally represented one of the most important and most widely used reactor-produced radionuclides (Ambade et al. 2015). Decay involves emission of medium-energy beta particles [E _β(max) = 606 keV (89.9 %), 333 (7.27)] and gamma rays [E _γ = 364 keV (81.2 %), 636 (7.27 %)]. The 8.02-day half-life is long enough for uptake and incorporation of radioiodine into thyroglobulin by the follicular thyroid cells, which has been the primary treatment of thyroid cancer. The medium-energy beta particle has a mean path length of about 1 mm in soft issue and is well suited for the treatment of small tumors such as follicular cell carcinoma. The emission of gamma radiation—although of high energy—facilitates imaging, which permits the uptake and localization of low activity levels of ¹³¹I to be assessed prior to administration of therapeutic doses. Ready availability of ¹³¹I in liquid form at affordable cost permits straightforward dispensing and oral administration. The high-energy gamma radiation of ¹³¹I can result in poor images and contributes significantly to the whole-body patient radiation burden without significantly increasing the radiation damage to the target tissue. This factor also adds to the radiation dose to staff and relatives, necessitating in most countries the admission and isolation of patients undergoing radioiodine therapy.

Iodide anions are localized in thyroid cells by active transport and directly incorporated into thyroglobulin from which the thyroid hormones are released. Iodine-131 for the treatment of thyroid disease is utilized for hyperthyroidism or well-differentiated thyroid carcinoma indications. There is also a large and growing body of evidence to support effective clinical use of ¹³¹I for management of well-differentiated thyroid cancers (Ambrosetti et al. 2009; Pitoia and Cavallo 2012; Pagano et al. 2004; Reynolds and Robbins 1997; Giovanella 2011). Apart from its utility in the management of thyroid disorder, ¹³¹I-MIBG (m-iodobenzylguanidine) is widely used for the treatment of neuroendocrine-originating tumors (Castellani et al. 2000; Bomanji et al. 2003; Gedik et al. 2008; Ezzidin et al. 2012), while ¹³¹I-Tositumomab (anti CD20 antibody, Bexaar®) is a registered radiopharmaceutical for the treatment of non-Hodgkin’s lymphoma (Tomblyn 2012; Lewington 2005; Dewaraja et al. 2010; Tsai et al. 2004; Kern 2000). The comparatively long half-life of ¹³¹I provides logistical advantage for the radiochemical production and processing/dispensing/shipment of ¹³¹I radiopharmaceuticals.

Production of ¹³¹I on a large scale can be performed either from ²³⁵U(n,f)¹³¹I (Siri and Mondino 2005; Nazari et al. 2001; Al-Janabi and Kadem 1990; Tabasi et al. 2005; Wojdowska 2010; Khalafi et al. 2005; Ahmad et al. 1982) or from the nuclear reactions (Shikata and Amano 1973; Payamara 2011; Chattopadhyay and Saha-Das 2010; El-Absy et al. 2010; Constan 1958; El-Azony et al. 2004; Al-Janabi and Kadem 1990; Elom Achoribo et al. 2012; Alanis and Navarrete 1999; Sorantin and Bildstein 1965). Although the ²³⁵U fission route provides HSA ¹³¹I, this production method is of limited use because of the requirements for availability of an elaborate complex radiochemical processing technology, sophisticated equipment, a reliably well-trained workforce, and management of high levels of radioactive waste. The production of fission ¹³¹I can be sustained only by countries pursuing fission ⁹⁹Mo production where ¹³¹I could be produced as a by-product. Due to the inherent complexities in the production of fission ¹³¹I, alternative production technologies using neutron activation of ¹³⁰Te is widely practiced. The natural abundances of tellurium isotopes and their corresponding thermal neutron (n, γ) cross-section values are summarized in Table 5.4.

Reactor production of ¹³¹I is generally via irradiation of ¹³⁰Te via the ¹³⁰Te(n, γ)¹³¹Te (t _½ = 25 min) and ¹³⁰Te(n, γ)^131m Te (t _½ = 30 h) nuclear reactions, which have thermal-neutron capture cross sections of 0.2 and 0.04 barn, respectively, and the three decay branches are shown in Figs. 5.4 and 5.5.

Use of natural tellurium target also results in the production of ¹²⁷I and ¹²⁹I as depicted in Table 5.5.

¹²⁷I is inactive, whereas ¹²⁹I is long-lived with a half-life of 1.57 × 10⁷ years. The activity levels of ¹²⁹I formed are very low, due to the long half-life; however, the number of atoms of iodine (¹²⁷I and ¹²⁹I) at the end of bombardment (EOB) constitutes about 80 % of the total iodine atoms. Thus the isotopic abundance of ¹³¹I will not exceed 20 % at EOB and continues to reduce with progressive decay of ¹³¹I. Separation of ¹³¹I from irradiated Te is usually carried out following dry and wet chemical distillation method following the reported procedures (Ambade et al. 2015; Ahmad et al. 1982; Beyer and Pimentel-Gonzales 2000). However, dry distillation is preferred due to the significantly lower levels of radioactive liquid waste which are generated, the more rapid processing, and the ability to obtain high radioactive concentrations of the final ¹³¹I product.

Lutetium-177 (T _1/2 = 6.65 d) represents therapeutic radioisotopes of rapidly increasing importance which decays with emission of low- to medium-energy β⁻ emission [E _β(max) = 497 keV (78.6 %), 385 keV (9.1 %)] and photon emissions [E _γ = 208 keV (11.0 %), 113 keV (6.4 %)] which permit gamma camera imaging and which can be used as a 3+ lanthanide for facile radiolabeling chemistry. Lutetium-177 is also a very attractive theranostic radionuclide which can be used in diagnostic/therapeutic pairing and exhibits a soft tissue penetration range of several millimeters. The use of ¹⁷⁷Lu-labeled peptides and for other applications is described in Chap.​ 10. The 6.65-day half-life is long enough for preparation of targeted biomolecules which have short or long biological half-lives and allows the products to be distributed for use at clinics distant to the production site. The gamma rays emitted by ¹⁷⁷Lu permit imaging of the uptake and biodistribution of the agent both before and during therapy administration. The principal applications of ¹⁷⁷Lu currently include treatment of neuroendocrine tumors (Kam et al. 2012; Gulenchyn et al. 2012; Esser et al. 2006), especially for bone pain palliation of metastatic cancer (Bodei et al. 2011; Delpassand et al. 2014; Kwekkeboom et al. 2008; Swärd et al. 2010; van Essen et al. 2010; Zaknun et al. 2013; Chakraborty et al. 2008a, c; Mathe et al. 2010; Liu et al. 2012; Das et al. 2002; Chakraborty et al. 2008a, b, c; Bryan et al. 2009), radiation synovectomy (Chakraborty et al. 2006a, b, c; Abbasi et al. 2011), and non-Hodgkin’s lymphoma (Goldenberg and Sharkey 2006; Michel et al. 2005), and treatment of ovarian (Alvarez et al. 1997) and liver cancer (Chakraborty et al. 2008a, b, c) have also been studied.

Two alternative production routes are available for reactor production of ¹⁷⁷Lu (Fig. 5.6) which include the “direct” neutron activation of highly enriched ¹⁷⁶Lu targets by the ¹⁷⁶Lu (n, γ)¹⁷⁷Lu reaction, which can provide relatively high SA ¹⁷⁷Lu in medium- to high-flux research reactors due to the high thermal neutron cross section (Chakraborty et al. 2014; Dash et al. 2015a, b). The “indirect” route is based on neutron irradiation of enriched ytterbium targets by the ¹⁷⁶Yb(n, γ)¹⁷⁷Yb → β⁻ route to NCA ¹⁷⁷Lu. The radiochemical processing from the “direct” route involves simple dissolution of the irradiated targets, whereas purification of HSA ¹⁷⁷Lu from the “indirect” route involves more elaborate radiochemical processing to separate the microscopic levels of NCA ¹⁷⁷Lu from macroscopic ytterbium targets and other radionuclides.

In contrast to production of many radionuclides by the (n, γ) route (see Table 5.1), the direct neutron activation method of ¹⁷⁷Lu production is attractive because of the high thermal neutron capture cross-section value for ¹⁷⁶Lu (σ = 2090 b). In fact, this is the highest cross section encountered among all the therapeutic radionuclides of current interest reactor-produced by the (n, γ) route. It is possible to produce ¹⁷⁷Lu of specific activity of 740–1110 GBq/mg (20–30 Ci/mg) in medium-flux reactors using isotopically enriched ¹⁷⁶Lu in which only 20 − 30 % of the ¹⁷⁶Lu atoms are being converted to ¹⁷⁷Lu. The specific activity of ¹⁷⁷Lu can be augmented to 1850–2405 GBq/mg (50–65 Ci/mg) by irradiation in high-flux reactors such as the high-flux isotope reactor (HFIR) at Oak Ridge National Laboratory (ORNL) and at the SM3 reactor in Dimitrovgrad, Russian Federation. Experimental setup for radiochemical processing of irradiated lutetium target by the (n, γ) route is shown in Fig. 5.7.

The direct (n, γ) ¹⁷⁷Lu production route also leads to the co-production of low levels of the long-lived ^177mLu impurity which has a half-life of 160 days. Although the presence of low activity levels of the ^177mLu isomer is not expected to increase the patient radiation burden (Breeman et al. 2003), the presence of this long-lived radiocontaminate in radioactive waste may pose an issue in some countries (Bakker et al. 2006). For this reason, a careful optimization of the time of irradiation may be essential to obtain the highest SA and to also minimize ^177mLu contamination levels (Chakraborty et al. 2014; Dash et al. 2015a, b; Pillai et al. 2003). While the use of isotopically enriched ¹⁷⁶Lu is a prerequisite for providing¹⁷⁷Lu of adequate SA by the “direct” route amenable for radiolabeling peptides and antibodies, use of natural Lu targets is adequate to obtain ¹⁷⁷Lu of moderate SA adequate for preparation of several radiopharmaceuticals where high SA is not required for therapeutic application such as bone pain palliation, synovectomy, etc. Use of natural targets (see Table 5.6) does not pose any challenges with respect to radionuclidic impurity.

During irradiation of ^natLu, some of the ¹⁷⁵Lu atoms are transmuted to ¹⁷⁶Lu atoms during neutron irradiation which are subsequently transformed to ¹⁷⁶Lu, which enhances the contribution of ¹⁷⁷Lu due to double-neutron activation cross section (Fig. 5.8). The contribution of ¹⁷⁵Lu to ¹⁷⁷Lu production is higher in high-flux reactors since the ¹⁷⁷Lu formed from ¹⁷⁵Lu will be proportional to the square of the neutron flux (σ²). The tremendous prospects associated with the (n, γ) method of ¹⁷⁷Lu production have thus witnessed extensive activity directed toward the establishment of production strategies using a wide spectrum of reactors of varying neutron flux (Mikolajczak et al. 2004; Nir-El 2004; Dvorakova et al. 2008; Zhernosekov et al. 2008; Chinol et al. 2010).

In order to obtain NCA ¹⁷⁷Lu, several investigators have described radiochemical separation of NCA ¹⁷⁷Lu following neutron irradiation of macroscopic ytterbium targets (Mirzadeh et al. 2004; So et al. 2008; Morcos et al. 2008; Horwitz et al. 2005; Hashimoto et al. 2003; Lebedev et al. 2000; Lahiri et al. 1998; Bilewics et al. 2009; Kumric et al. 2006; Chakraborty et al. 2008b; Knapp et al. 2005; Dash et al. 2015a, b). In this method, an isotopically enriched ¹⁷⁶Yb target undergoes (n, γ) transmutation to produce ¹⁷⁷Yb which subsequently decays by β⁻-emission (T _½ = 1.9 h) to yield ¹⁷⁷Lu by the process. The prospects associated with this route of ¹⁷⁷Lu production along with the challenge associated with the separation of chemically similar neighboring lanthanides have thus led to a considerable research and innovative detection strategies. At least four successful preparative-scale separations of ¹⁷⁷Lu from Yb for clinical applications have been reported in the literature (Lebedev et al. 2000; Bilewicz et al. 2009; Mirzadeh et al. 2004; Chakravarty et al. 2010). The method reported by Lebedev et al. is based on the selective extraction of Yb by sodium amalgam from Cl⁻/CH₃COO⁻ electrolytes taking advantage of the higher solubility of metallic Yb than Lu in mercury. This rather complex and time-consuming process involves eight cementation cycles. The separated ¹⁷⁷Lu is reported to contain 10 μg Yb(III) from 200 mg of neutron-irradiated Yb₂O₃ and must be subjected to an ion exchange process for purification. Subsequently, Bilewicz et al. (2009) have reported a method based on the reduction of Yb(III) to Yb(II) with sodium amalgam followed by selective precipitation of Yb(II) as the sulfate. The ¹⁷⁷Lu solution obtained after the precipitation step contained 1 mg Yb(III) from 50 mg of neutron-irradiated Yb₂O₃ and was hence subjected to purification by an ion exchange process. The method described by Knapp and co-workers consists of a chromatographic extraction method containing LN resin comprised di(2-ethylhexyl)orthophosphoric acid (HDEHP) in which both Lu and Yb can be loaded. The column was eluted with 2 M HCl and the initial elution ¹⁷⁰Tm detected prior to Yb elution was presumed to be formed by neutron activation of the low levels of stable ¹⁶⁹Tm impurity present in the enriched ¹⁷⁶Yb target material. After the removal of ¹⁷⁰Tm and Yb, ¹⁷⁷Lu was then eluted with 6 M HCl. The specific activity of ¹⁷⁷Lu obtained by this method was estimated to be 3.7 TBq (100 Ci)/mg (i.e., 91 % of the 110 Ci/mg theoretical value).

The EXC technique was further exploited Horwitz and co-workers (Horwitz et al. 2005) and was found to be successful for the separation of NCA ¹⁷⁷Lu from a 300 mg irradiated ytterbium target. The whole separation process can be broadly divided in to three steps: (1) front-end target removal step, (2) primary separation step, and (3) secondary separation step. While the goals of each separation step differ, it basically consists of separation of Yb and Lu using the LN2 resin followed by concentration and acid adjustment of the Lu-rich eluate using Amberchrom® CG-71 resin. In another independent study, a multi-column solid-phase extraction (SPE) chromatography technique using di-(2-ethylhexyl)orthophosphoric acid (HDEHP)-impregnated OASIS-HLB sorbent-based SPE resins (OASIS-HDEHP) was used (Le et al. 2008a, b). The method was successful for the isolation of several hundred mCi of NCA ¹⁷⁷Lu using 50 mg Yb target irradiated in a medium neutron flux nuclear reactor (ϕ = 5.10¹³ n/cm²/sec).

An alternative method (Fig. 5.9) reported by Chakravarty et al. (Chakravarty et al. 2010) consists of an electro-amalgamation approach in which a two-cycle electrolysis procedure was adopted for separation of 18.5 GBq (500 mCi) of ¹⁷⁷Lu from ^169-177Yb/¹⁷⁷Lu mixture in lithium citrate medium. In the first step, the bulk of Yb target is removed, and in the second step, tracer quantities of Yb present in the ¹⁷⁷Lu are removed to obtain NCA grade ¹⁷⁷Lu.

Phosphorous-32 (T _1/2 = 14.26 d) is the first radionuclide used in therapeutic applications more than 50 years ago for the palliation of skeletal pain from bone metastases. Phosporus-32 is a pure β⁻ emitter with a maximum energy of 1.71 MeV and the mean and the maximum particle range in soft tissue are 3 and 8 mm, respectively. The traditional clinical application of ³²P has been of the treatment of polycythemia vera and leukemia (Najean and Rain 1997; Tefferi and Silverstein 1998; Meuret et al. 1975; Najean et al. 1996; Tefferi 2012). Phosphorus-32 was first produced in the cyclotron at the University of California in Berkeley, by E. Lawrence in 1936 by irradiating red phosphorus by the ³¹P(d,p)³²P [σ = 0.18 × 10⁻²⁷cm²] nuclear reaction (Lawrence and Cooksey 1936). Subsequently, when the first nuclear reactor (Graphite Reactor) became available at the Oak Ridge National Laboratory, production shifted to the reactor route, since ³²P can be readily produced by either the ³²S(n,p)³²P or ³¹P(n, γ)³²P nuclear reactions. Only relatively low SA ³²P can be produced by the neutron irradiation of ³¹P (100 % abundance, σ = 0.172 barn) by the ³¹P(n, γ)³²P route but could be used for the preparation of formulations for bone pain palliation (Vimalnath et al. 2013; 2014a, b). This production route requires only very simple chemical treatment after neutron irradiation, but the low SA restricts the useful therapeutic applications of ³²P produced by this method.

For these reasons, ³²P obtained by separation from the sulfur target irradiated in a fast neutron flux by the ³²S(n,p)³²P reaction (σ = 0.068b barn for fast neutrons) is currently used for the majority of research and clinical therapeutic applications. The ³²S(n,p)³²P nuclear reaction involves ejection of a charged particle which must overcome the Coulomb barrier and this reaction is thus favored with fast neutrons. However, the cross section of the ³²S(n,p)³²P reaction is only 0.068 barn for fast neutrons, and therefore, several hundred grams of sulfur are required for irradiation in order to provide a few hundred mCi of the ³²P product. Even though the cross-section values for the ³²S(n,p)³²P nuclear reaction are much lower than for the ^3lP(n, γ)³²P route, the irradiation of ³²S targets is still of interest and utility because of the production of higher-SA ³²P. The isotopic composition of ^natS consists of ³²S (95.02 %), ³³S (0.75 %), ³⁴S (4.21 %), and ³⁶S (0.02 %). It is imperative to consider all the radionuclides produced both by thermal and fast neutrons when sulfur undergoes reactor irradiation; Table 5.7 summarizes the nuclear reactions which occur when natural sulfur undergoes irradiation with thermal and fast neutrons in a nuclear reactor.

From an analysis of the possible radionuclides formed by neutron irradiation of natural sulfur, only ³⁵S (T _1/2 = 87.2 d) and ³³P (T _1/2 = 25.3 d) are of concern and could be radiochemical impurities of the final ³²P product (T _1/2 = 14.28 d). The ³⁵S levels contaminating the desired ³²P product are of concern, since they will be carried through the chemical separation process. The contribution of the longer-lived ³³P is a point of concern since separation from ³²P is not possible by traditional methods and it is not feasible to reduce the ³³P levels by cooling because of the longer half-life than ³²P. It has been experimentally observed that the ³²P and ³³P activity ratio produced in the same irradiation conditions is about 7000, because of the much lower isotopic abundance and low fast neutron cross section of the ³³S target as well as longer half-life of ³³P.

The bulk separation of only a few micrograms of ³²P from the bulk of ³²S must be conducted, and the ³²P SA obtained can be reduced by introduction of inactive phosphorous impurities during radiochemical processing. Due to the low neutron capture reaction cross section, significant large target irradiation volumes are required in the reactor for long periods of time in order to produce ³²P. The separation and purification of ³²P from irradiated S is usually carried out following wet chemical extraction and dry distillation methods. The wet extraction method consists of dissolution of irradiated powdered sulfur in boiling water in the presence of strong and weak acids to extract the phosphorus nuclide from the sulfur target using 2-octanol (Samsahl 1958; Razbash et al. 1991). One of the major impediments of this wet chemical extraction method is the dependence of extraction yield on the target sulfur particle size which is significantly decreased when the target is melted or solidified due to the exothermal heat generated during neutron irradiation. Additionally, the use of mineral acid is a major deterrent since its use introduces impurities and leaves solid waste behind which not only complicates the extraction process but also requires additional purification steps to achieve the required high chemical and radiochemical purity. The requirement of multistage processes results in poor recovery yields. Because of these drawbacks, the wet chemical extraction method is not often used for ³²P production.

The most widely used methods for the separation of ³²P from neutron-irradiated sulfur thus involve distillation methods, for instance, at 500 ° C in a nitrogen atmosphere in order to preclude the possibility of fire. Vacuum distillation of sulfur at 180–200 °C, which is lower than the sulfur ignition point, is conducted under a pressure of 1–10 mmHg (Gharemano et al. 1983; Yeh 1962). The distillation method has the advantage that high-purity products can be obtained since no reagents are added during the separation of ³²P from neutron-irradiated sulfur. The distillation method usually consists of a vacuum system, gas-feeding apparatus, and distillation assembly (Arrol 1953), in which pressure and temperature are optimally controlled (Gharemano et al. 1983; Yeh 1962; Arrol 1953; Alanis and Navarrete 2007; Karelin et al. 2000). Experimental setup for the radiochemical processing of irradiated elemental sulfur target is shown in Fig. 5.10.

Praseodymium-143 (T _1/2 = 13.57 d) is a long-lived medium-energy pure β⁻-emitting [E _β(max) = 934 keV (100 %)] radionuclide useful for therapy. Although reactor production of ¹⁴³Pr can be performed by the ¹⁴¹Pr(n, γ)¹⁴²Pr(n, γ)¹⁴³Pr double-neutron capture process, from this method the activated product consists of a ¹⁴²Pr and ¹⁴³Pr mixture, the proportion of which will depend on the irradiation duration and length of the post-irradiation decay period. For this reason, the indirect reactor production route is preferred. A target cooling period of a few days is required to insure decay of the short-lived ¹⁴³Ce (T _½ = 33.04 h), which decays to ¹⁴³Pr. This production route provides NCA ¹⁴³Pr with high radionuclidic purity. Naturally occurring Ce is composed of ¹³⁶Ce (0.185 %), ¹³⁸Ce (0.251 %), ¹⁴⁰Ce (88.45 %), and ¹⁴²Ce (11.114 %). Neutron activation of natural CeO₂ would therefore lead to the formation of ^l43Ce along with ^l39Ce, ¹⁴¹Ce, and ¹⁴³Ce as radionuclidic impurities. In order to reduce the radioactive burden from the irradiated target, use of enriched ¹⁴²Ce target is mandatory. The reaction cross section has a moderate value of 0.95 barn, which, coupled with the long half-life, not only necessitates long irradiation periods but also requires a high neutron flux irradiation to produce sufficient activity levels of ¹⁴³Pr (Vimalnath et al. 2005).