Development of ²²⁵Ac Radiopharmaceuticals: TRIUMF Perspectives and Experiences

Abstract

Background:

The development of radiopharmaceuticals containing 225Ac for targeted alpha therapy is an active area of academic and commercial research worldwide.

Objectives:

Despite promising results from recent clinical trials, 225Ac-radiopharmaceutical development still faces significant challenges that must be overcome to realize the widespread clinical use of 225Ac. Some of these challenges include the limited availability of the isotope, the challenging chemistry required to isolate 225Ac from any co-produced isotopes, and the need for stable targeting systems with high radio-labeling yields.

Results:

Here we provide a review of available literature pertaining to these challenges in the 225Ac-radiopharmaceutical field and also provide insight into how performed and planned efforts at TRIUMF - Canada’s particle accelerator centre - aim to address these issues

1. INTRODUCTION

Targeted radionuclide therapy using alpha-emitting isotopes combined with disease-specific targeting vectors (antibodies or peptides) has the potential to treat metastatic disease by delivering a source of cytotoxic radiation directly to targeted cells [1-8]. Given specific targeting, the short range and high cytotoxicity of alpha particles result in the destruction of nearby diseased cells with limited harm to healthy tissues [9]. Due to this potential, the development of alpha-emitting radiopharmaceuticals for Targeted Alpha Therapy (TAT) is an active area of academic and commercial research worldwide. Though one alpha-emitting radiopharmaceutical, Xofigo (²²³RaCl₂), is approved for clinical use, the clinical approval of an alpha-emitting isotope combined with a disease-targeting biomolecule has yet to occur and these radiopharmaceuticals remain in the development stages. Several candidate isotopes for TAT are currently under clinical and preclinical evaluation, including ¹⁴⁹Tb, ²¹¹At, ²¹²Bi, ²¹²Pb, ²¹³Bi, ²²³Ra, ²²⁴Ra, ²²⁷Th, and ²²⁵Ac. ²²⁵Ac is one promising candidate isotope for TAT due to its 9.9 day half-life - suitable for targeting with antibodies - and the emission of four alpha particles in its decay chain (it decays to stable ²⁰⁹Bi via four alpha- and two beta-decays - Fig. 1). ²²⁵Ac can also be used as a generator of ²¹³Bi (t_1/2 = 45.6 min), itself a promising TAT isotope.

Fig. (1).

Decay schematic showing the decay and production pathways for ²²⁵Ac. Gamma emissions useful for quantification of ²²⁵Ac are shown in red. (The color version of the figure is available in the electronic copy of the article).

While several clinical trials have demonstrated the potential of ²²⁵Ac or ²¹³Bi radiopharmaceuticals to treat advanced cancers [10-13], the development of these drugs faces many challenges that have slowed progress, including: 1) the limited supply of ²²⁵Ac - currently only 63 GBq (1.7 Ci) globally annually - which could treat fewer than 1000 patients per year; 2) the need for adequate chemical purification processes required to separate ²²⁵Ac from co-produced stable or unstable isotopes during production; and 3) the need for stable targeting systems with high radiolabeling yields and appropriate pharmacodynamics, the development of which is hindered by a limited understanding of actinium chemistry and complicated by the need to also successfully retain ²²⁵Ac progeny isotopes at the target site despite recoiling daughter nuclei and changing chemistry as the decay chain progresses [14].

Each section of this review addresses one of these challenges. For each, we summarize the current literature including the current standard methods and any proposed alternatives, and also discuss how performed, planned, and potential activities at TRIUMF attempt to address these challenges facing the development of ²²⁵Ac-radiopharmaceu-
ticals.

2. PRODUCTION OF ²²⁵Ac

2.1. Existing ²²⁵Ac Supplies

Current sources of ²²⁵Ac are primarily derived from the build-up of ²²⁹Th through the decay of ²³³U stockpiles (see Fig. (1) of the ²³³U decay chain). The majority of this ²³³U (t_1/2 = 1.6 × 10⁵ y) was produced between 1954 and 1970 via neutron irradiation of ²³²Th while being investigated for its use in nuclear weapons and reactors that were never fully deployed [15]. Between 1995 and 2005, ²²⁹Th (t_1/2 = 7340 y) generated from ²³³U decay was extracted from stockpiles stored at Oak Ridge National Laboratory (ORNL, Oak Ridge, TN). This ²²⁹Th now exists in two sources: one at ORNL (∼5.55 GBq (150 mCi), or ∼704 mg) [16] and another (1.7 (46 mCi), or 215 mg) [17] transferred to the Institute for Transuranium Elements (ITU, Karlsruhe, Germany). A third ²²⁹Th source (5.55 GBq (150 mCi), 704 mg) obtained from Russia ²³³U stockpiles exists at the Leipunskii Institute for Physics and Power Engineering (IPPE, Obninsk, Russia) [18]. These three sources serve as generators of ²²⁵Ac and its parent ²²⁵Ra (t_1/2 = 14.9 d) and act as the major ²²⁵Ac sources worldwide, producing approximately 26.6 GBq (720 mCi) (ORNL) [19] and 13.1 GBq (350 mCi) (ITU) [17] of ²²⁵Ac annually. While the IPPE source contains as much ²²⁹Th as the ORNL source, reported values indicate ²²⁵Ac production from this source is sporadic [18, 20]. Overall, the accepted global annual production from ²²⁹Th is 63 GBq (1.7 Ci) [21-25].

While a key advantage of this production method is an ²²⁵Ac product free of other actinium isotopes, 63 GBq (1.7 Ci) is insufficient to meet the current global demand for researchers and the development of new agents and will be even more inadequate should any ²²⁵Ac therapies become deployed clinically [19]. From research into fundamental ²²⁵Ac chemistry [9] to the most promising clinical trials [10], the development of ²²⁵Ac radiopharmaceuticals is slowed by the small supply and resulting high cost that makes ²²⁵Ac inaccessible to many researchers.

2.2. Leveraging Unique Facilities to Supply ²²⁵Ac Research

Due to the high cost of ²²⁵Ac, ²²⁵Ac-radiopharmaceutical development at TRIUMF currently relies on internal production using the Isotope Separator and Accelerator (ISAC) Facility [26]. Commissioned in 2000, ISAC produces beams of rare isotopes for experiments primarily studying nuclear structure and nuclear astrophysics. Irradiation of uranium or thorium targets with protons at 480 MeV results in the production of a number of isotopes that are extracted into a heterogeneous ion beam. Isotope Separation On-Line (ISOL) mass-separates these isotopes to produce a homogeneous isobaric ion beam [27]. Isolation of mass 225 produces an ion beam containing ²²⁵Ra and ²²⁵Ac that is directed onto an aluminum target in which the isotopes are deposited at a depth of 20 nm (as determined by SRIM [28]). Etching of the aluminum post-implantation followed by separation of ²²⁵Ra and ²²⁵Ac on a solid phase extraction (DGA) resin provides both a primary ²²⁵Ac fraction for immediate studies, as well as a number of subsequent ²²⁵Ac batches isolated through the decay of ²²⁵Ra. When eluted at the optimal time (every 17.5 days) this generator produces ²²⁵Ac with an activity equal to 44.4% of the ²²⁵Ra activity present at the previous elution1. Since 2015, ²²⁵Ac production at ISAC has enabled radiolabeling and preclinical studies at TRIUMF. ISOL methods have also been applied at ISOLDE (CERN, Geneva) to produce ²²⁵Ac for radiopharmaceutical development [29].

Though ISOL provides isotopically pure ²²⁵Ac sources, yields remain insignificant compared to quantities available from ²²⁹Th generators. At TRIUMF, the maximum measured ISAC beam intensities of 1.3 × 10⁸ ions/s for ²²⁵Ac and 1.6 × 10⁸ ions/s for ²²⁵Ra could theoretically produce up to 370 MBq (10 mCi) of ²²⁵Ac per month², ISAC does not operate as a dedicated medical isotope production facility and ²²⁵Ac production occurs within the overarching context of the laboratory’s research program. Total ²²⁵Ac production for 2016 was only 44.4 MBq (1.2 mCi).

In order to better accommodate medical isotope research, the European Organization for Nuclear Research (CERN) has launched the Medical Isotopes Collected from ISOLDE facility (MEDICIS), which plans to produce mass-separated isotope beams from offline targets starting in 2017 and with uranium targets for ²²⁵Ac production available in late 2018 [29, 30]. As a dedicated medical isotope production facility, MEDICIS suggests the capacity to produce up to about 112 MBq (3 mCi) of ²²⁵Ac per month³.

2.3. The Need for New ²²⁵Ac Production Methods

Estimates of current demand for ²²⁵Ac are less than 185 GBq (5 Ci) per year, however, this is likely significantly tempered by both supply constraints and cost. While predicting future demand is difficult, it can be estimated to grow by about 200 to 400 GBq per year (about 5 to 10 Ci per year) for each ²²⁵Ac-based therapy that is approved for clinical use4. Should efforts to develop ²¹³Bi-radiopharmaceuticals also increase, ²²⁵Ac demands will be even higher.

While facilities like ISAC and MEDICIS can facilitate radiopharmaceutical development by providing access to medical isotopes that are otherwise challenging to obtain, their application will remain limited to medical isotopes in the development stages. Even with potential proposed upgrades to the MEDICIS facility that could enable monthly production of up to 1.7 GBq (45 mCi) of ²²⁵Ac [29], and potential upgrades to ISAC that could increase yields by a factor of ∼10005, these facilities are not expected to meet existing ²²⁵Ac demand, 
are not commercially viable production sources, and could not

supply enough ²²⁵Ac to support the widespread use of a clinically approved therapy. These facilities should instead be viewed as valuable research enablers whose utility comes from their ability to provide quick access to a range of high-purity medical isotopes so that the feasibility of a given isotope’s applications can be explored before having to build dedicated large-scale, isotope-specific production infrastructure.

While harnessing untapped ²²⁹Th supplies has the potential to more significantly impact ²²⁵Ac availability, in 2005 the U.S. Congress ordered the Department of Energy (DOE) to cease extraction of ²²⁹Th from ²³³U stockpiles and to instead begin down-blending (dilution with ²³⁸U to a non-weaponizable ²³³U concentration) and permanent disposal of the two tonnes of stockpiled ²³³U [15]. Petitions to recover ²²⁹Th before ²³³U disposition have been denied [20], and completion of these efforts is scheduled for 2018 [15]. From the high- and intermediate-purity ²³³U sources within the inventory [32], this represents the loss of 32.6 g (∼260 GBq or 7 Ci) of ²²⁹Th or a potential 1.5 TBq (40.5 Ci) of annual ²²⁵Ac production. Other estimates suggest this is a loss of 37 g (∼8 Ci) of ²²⁹Th [16] and suggest a loss of 2.2 TBq (60 Ci) of annual ²²⁵Ac production [20]. Without new ²²⁹Th sources, ²²⁵Ac production from current DOE ²²⁹Th generators could increase by only 20% if the current elution schedule is optimized for ²²⁵Ac production instead of unit cost [19]. While quantities of additional ²²⁹Th-containing ²³³U sources may exist in other countries (ex. Russia), quantities - to the best of our knowledge - are unknown.

Without the existence of significant additional ²²⁹Th sources, the use of ²²⁵Ac or ²¹³Bi in multiple approved therapies will require the development of new ²²⁵Ac production methods. The remainder of this section aims to present a comprehensive list of alternative ²²⁵Ac production options. The potential of each method to meet projected ²²⁵Ac demand and the practical challenges associated with each method will be discussed. Possible production methods proposed in the literature are summarized in Table 1, while Table 2 summarizes other nuclear reactions capable of producing ²²⁵Ac but that are considered impractical at this time. When not derived from original sources, details of calculations are provided in footnotes.

Table 1.

Summary of current and potential future ²²⁵Ac production methods. Production values for current sources list current production levels, while values for potential sources list estimates of maximum possible production at sample of existing and operational facilities that have dedicated stations for large-scale medical isotope production. Details of calculations or references to cited values can be found in the text. Values listed for ²²⁶Ra targets assume a target mass of 1 g.

	Production Method	Facility	Capabilities	Monthly 225Ac Production [GBq (Ci)]
Current Sources	229Th generator	ORNL ITU IPPE	0.704 g (150 mCi) of 229Th 0.215 g (46 mCi) of 229Th 0.704 g (150 mCi) of 229Th	2.2 (0.06) 1.1 (0.03) 2.2 (0.06)
Potential	232Th(p, x)225Ac	TRIUMF	500 MeV, 120 µA	11266.5 (304.05)
		BNL	200 MeV, 173 µA	2675.84 (72.32)
		INR Arronax	160 MeV, 120 µA 70 MeV, 2×375 µA	1002.0 (27.08) 462.1 (12.49)
		LANL	100 MeV, 250 µA	444.0 (12.00)
		iThemba LABS	66 MeV, 250 µA	127.7 (3.45)
Future	226Ra(p, 2n)225Ac	20 MeV, 500 µA cyclotron 15 MeV, 500 µA cyclotron		3983.1 (107.65) 1157.4 (31.28)
Sources
	ISOL	TRIUMF (existing) TRIUMF (potential upgrades)		0.37 (0.01) 190.6 (5.15)
	226Ra(γ, n)225Ra	medical linac ALTO	18 MeV, 26 µA 50 MeV, 10 µA	48.1 (1.3) 55.5 (1.5)
	226Ra(n, 2n)225Ra	fast breeder reactor		∼37 (1)

Table 2.

Possible yet impractical methods for ²²⁵Ac production.

Production Method	Comments
226Ra(p, pn)225Ra	Yields insignificant compared to 226Ra(p, n) 225Ac production (105× less according to FLUKA simulation)
232Th(p,4n)229Pa	Low cross-section
natU(p, x)225Ac	Produces ∼ 10× less 225Ac and 225Ra compared to thorium spallation, creates fissile 239Pu and 235U, can handle less beam current than thorium spallation targets
232Th(n, γ)233U	Would take decades for 229Th to build up
230Th(γ, n)229Th	230Th not available in sufficient quantities
Reactor production of 229Th	Potential target materials 228Ac, 228Ra, 228Th, and 230Th not available in sufficient quantities. Production yields from 226Ra irradiation (110 MBq/month/g, or 3 mCi/month/g) too low considering cost and difficulty of 226Ra source production.

2.4. Potential for ²²⁵Ac Production in Nuclear Reactors

While the majority of medical isotopes today are sourced from nuclear reactors [33], the potential for reactor-based ²²⁵Ac production is limited. The parent isotope, ²²⁵Ra, can be produced in reactors via the ²²⁶Ra(n, 2n) ²²⁵Ra reaction, however, this reaction requires an intense source of high (>6.4 MeV) neutrons found only near the tail end of a typical breeder reactor neutron energy spectrum. Given that significantly more lower energy neutrons would be present, these irradiations would be dominated by the co-production of ²²⁷Ac, a long-lived (t_1/2 = 21.8 y) and highly toxic actinium isotope, the presence of which in significant quantities may prevent the clinical approval of a pharmaceutical. To the best of our knowledge, this method has not been investigated experimentally or thoroughly modelled. However, rough estimates suggest this method could produce MBq to GBq (µCi to mCi) amounts of ²²⁵Ac per month per gram of ²²⁶Ra target material at a single reactor facility6.

The potential to increase ²²⁹Th stocks using reactors has also been investigated [36, 37]. Results suggest the irradiation of ²²⁶Ra targets at a single reactor could produce 100 MBq (2.7 mCi) of ²²⁹Th per month per gram of ²²⁶Ra target material [36]. Other results have suggested this value may be closer to 59 MBq (1.6 mCi) [37]. While larger specific yields were seen when using ²²⁸Ra and ²²⁷Ac target materials (352 and 600 MBq (9.5 and 16.2 mCi) of ²²⁹Th per month per gram, respectively), these isotopes cannot be supplied in sufficient quantities [36]. The ²²⁸Th(n,γ)²²⁹Th reaction is impractical for the same reason. For example, 2.5 thousand tonnes of natural thorium would have to be processed to produce a single gram of ²²⁸Ra (t_1/2 = 5.8 y), which could potentially be used to slowly generate ²²⁸Th (t_1/2 = 1.9 y).

Given the challenges and costs associated with safely handling large ²²⁶Ra sources (see Section 2.7) and the resulting low ²²⁵Ra or ²²⁹Th yields, reactor production of sufficient ²²⁵Ac quantities would likely be prohibitively expensive.

2.5. Potential for ²²⁵Ac Production Using Electron Accelerators

The use of the ²²⁶Ra(γ, n)²²⁵Ra reaction for ²²⁵Ac production has been experimentally explored [38] and modelled [39]. These works have explored irradiating old radium needles on electron linear accelerators (linacs) found in most modern cancer centres. These linacs typically use electron beams incident on tungsten targets to produce bremsstrahlung x-rays for external beam radiation therapy. Experimental results measured the production of 2.44 MBq (66 µCi) of ²²⁵Ac after a single-hour irradiation by 18 MV 
x-rays of a 20 mg of ²²⁶Ra source located 12.5 cm from the tungsten target and with an incident electron beam of 26 µA average current [38]. This scales to a potential 48 GBq (1.3 Ci) of ²²⁵Ac per month for a 1 g ²²⁶Ra source, which could be potentially increased by irradiation parameter optimization. This method has the advantage of producing ²²⁵Ac without contamination from other actinium isotopes. While co-production of ²²⁴Ra occurs for photons above 12 MeV, this should not impact the desired ²²⁵Ra/²²⁵Ac generator as ²²⁴Ra (t_1/2 = 3.7 d) decays to inert ²²⁰Rn and does not result in the production of any Ac isotopes.

While many medical linacs capable of this ²²⁵Ac production method may exist, these facilities are used for patient care and, to our knowledge, none are currently equipped with the infrastructure required for safe large-scale isotope production and processing. Again, the 48 GBq (1.3 Ci) of ²²⁵Ac per month would likely be cost prohibitive given the challenges (see Section 2.7) associated with a 1 g ²²⁶Ra target. It has been estimated that linac-based ²²⁵Ac production could be increased by up to a factor of 16, although accompanied by an increase in target mass [40].

Given this low yield, sufficient ²²⁵Ac production via the ²²⁶Ra(γ, n)²²⁵Ra reaction would require the use of a facility with significantly higher electron beam current. Though none are dedicated medical isotope production facilities, some such facilities exist for which ²²⁵Ac production values determined by scaling experimental medical linac irradiation results by electron beam current can be found in Table 1. In addition, TRIUMF’s planned Advanced Rare IsotopE Laboratory (ARIEL) facility will use a 35 MeV, 10 mA electron beam to produce intense high-energy x-rays for radioisotope production by photofission [26]. While the ARIEL electron target is intended for operation as an ISOL facility for fundamental research - not a medical isotope production facility - scaling experimental results for ²²⁵Ra production on medical linacs to account for the higher current and different irradiation geometry suggests ARIEL could theoretically produce up to 74 TBq (2000 Ci) of ²²⁵Ac per month from a 1 g ²²⁶Ra target. However, how an isotope production target could survive a 100 kW beam is another unsolved problem - current designs for ARIEL ISOL targets only consider 50 kW. Other lower current electron accelerators, such as the existing 50 MeV, 10 µA ALTO electron accelerator (Orsay, France) [41], could theoretically produce up to 56 GBq (1.5 Ci) per month.

2.6. Potential for ²²⁵Ac Production using Low Energy Proton Accelerators

The promising use of the ²²⁶Ra(p,2n)²²⁵Ac reaction to produce ²²⁵Ac on low-energy proton accelerators was first demonstrated in 2005 by Apostolidis et al. [42]. This reaction has a high (710 mb) cross-section peak at 16.8 MeV and could thus be performed on the many low energy cyclotrons already in use worldwide for medical isotope production. An estimated >550 of these cyclotrons have an energy over 16 MeV, some of which operate at up to 500 µA [43]. Another advantage of this approach is that it would not coproduce ²²⁷Ac. While the (p, n) reaction is expected to produce some ²²⁶Ac (t_1/2 = 29.4 h), measurements of co-production of ²²⁶Ac have not been reported from experiments found in the literature [42]. A simple FLUKA [44,45] simulation approximating the Apostolidis et al. experiment suggests an ²²⁶Ac activity at end of bombardment (EOB) equal to ∼11% the ²²⁵Ac activity. However, unlike with ²²⁷Ac contaminants, the ratio of expected ²²⁶Ac to ²²⁵Ac activity would decrease over time due to the differences in half-lives. The co-production of ²²⁵Ra via the ²²⁶Ra(p,pn)²²⁵Ra reaction is expected to be negligible at the optimal energies required for direct ²²⁵Ac production [46].

Given the high cross-section, large-scale production of ²²⁵Ac via the (p,2n) reaction would be capable of meeting long-term demand for ²²⁵Ac with only a single production site. Combining available cross-section data [42] with stopping power for ²²⁶Ra [28] suggests a single 20 MeV, 500 µA proton beam incident on a ²²⁶Ra target (∼1 g) could produce a theoretical maximum of 4 TBq (108 Ci) of ²²⁵Ac per month7.

The use of low energy proton irradiation of ²³²Th to produce ²²⁹Pa, which decays to ²²⁹Th, has also been explored [47]. The peak measured cross-section for this ²³²Th(p, 4n)²²⁹Pa reaction is 162 mb for a proton energy of 29.8 MeV. However, yields for this production method are low, with a potential to produce only 7.4 MBq (0.2 mCi) of ²²⁹Th per month on a 50 MeV, 500 µA cyclotron.

2.7. Challenges Associated with ²²⁶Ra Targets

All alternative ²²⁵Ac production methods discussed so far have involved the use of ²²⁶Ra as a target material. Cost-effective isotope production requires the use of stable or naturally occurring and long-lived target materials, and as the closest such isotope to ²²⁵Ac, ²²⁶Ra is one of only a few options - spallation reactions on naturally occurring ²³²Th and ²³⁸U being the only other possibilities. Other potential target materials such as ²³⁰Th, ²²⁸Th, ²²⁸Ac, and ²²⁸Ra are not available in large enough quantities to be of practical use.

Despite its potential, the use of ²²⁶Ra targets poses significant challenges due to the availability of the isotope and safety hazards that complicate the target manufacturing, irradiation, processing, and recycling. Part of the ²³⁸U decay chain, ²²⁶Ra ultimately decays to stable ²⁰⁶Pb and is typically found in equilibrium with most isotopes in its decay chain. ²²⁶Ra was the first radioactive isotope discovered and was produced in large quantities from the 1920s for use in a number of medical and industrial applications until production stopped in 1960 [48]. Due to its high radiotoxicity, reactivity with water and air, and decay to the noble alpha-emitting ²²²Rn gas, ²²⁶Ra sources typically contained radium salts encapsulated in platinum [49]. The internal production of helium from the five alpha decays in the ²²⁶Ra chain caused most of these sources to rupture, after which ²²²Rn gases are released and ²²⁶Ra salts can leak out. Even when sealed, the high energy gamma rays present from ²²⁶Ra progeny present external radiation hazards, with a dose rate of 
8.1 mSv/h at 1 m from a 37 GBq (1 Ci, 1 g) ²²⁶Ra source.

While the use of ²²⁶Ra sources declined after the health effects of radiation exposure became known and safer reactor-based isotopes became available, many ²²⁶Ra sources remained in storage - primarily in hospitals - for decades. The hazards associated with the presence of ²²⁶Ra sources lead many governments to push for the elimination of ²²⁶Ra inventories and in 1996 the International Atomic Energy Agency (IAEA) established guidelines for the disposal of ²²⁶Ra sources in long-term geological repositories [49]. This limits the availability of large ²²⁶Ra quantities, with the IAEA estimating only a few kilograms of ²²⁶Ra exist among these sources worldwide [49]. Typical medical sources contained <100 mg of ²²⁶Ra, with some industrial sources containing up to 1000 mg quantities. For this reason, calculations in Table 1 assume 1 g as a reasonable upper limit on the size of potential ²²⁶Ra targets.

New ²²⁶Ra sources could be extracted from the waste of current uranium mining operations. Approximately 50 thousand tonnes of uranium ore is mined each year [50], from which ²²⁶Ra is separated and disposed of as waste. With the potential to extract 257 mg of ²²⁶Ra from each tonne of U₃O₈ [39], this amounts to about 12.85 kg of ²²⁶Ra waste per year.

Whether obtained using old sources or through uranium mine tailings, the manufacturing of ²²⁶Ra targets for ²²⁵Ac production would require infrastructure beyond what is typically used to make medical isotope production targets. ²²⁶Ra regulatory and safety issues - specifically those associated with ²²²Rn - would also require additional infrastructure during the target irradiation, processing, and the necessary recycling of irradiated ²²⁶Ra material. While ²²⁶Ra targets have the greatest potential for ²²⁵Ac production per gram of target material, difficulties and costs associated with these targets is a significant disadvantage of the ²²⁶Ra(p,2n)²²⁵Ac and ²²⁶Ra(γ, n)²²⁵Ra production methods.

2.8. Potential for ²²⁵Ac Production via High-Energy Proton Spallation of Thorium

An alternative ²²⁵Ac production method that avoids the use of ²²⁶Ra targets involves the irradiation of natural thorium targets with protons above 70 MeV. This ²³²Th(p, x)²²⁵Ac reaction produces ²²⁵Ac through a number of reaction pathways, though the total cross-section peaks ∼40 times lower than for ²²⁶Ra(p, 2n) ²²⁵Ac production (for details on cross-sections, the reader is referred to: [51]). Thorium metal is the preferred chemical form for post-irradiation processing of the target, and isolation of MBq to GBq (µCi to mCi) quantities of ²²⁵Ac from irradiated thorium metal has been demonstrated by both American and Russian research groups at Brookhaven National Laboratory (BNL, Brookhaven, NY), Los Alamos National Lab (LANL, Los Alamos, NM), and the Institute for Nuclear Research of the Russian Academy of Sciences (INR) [20, 22, 23, 51-58]. Unlike ²²⁶Ra (3.7×10¹⁰ Bq/g, or 1 Ci/g), ²³²Th (4.1×10³ Bq/g, 110 nCi/g) is not prohibitively radioactive, poses fewer radiological hazards and is readily available as a target material. Tens of kilograms are known to exist in stockpiles within a number of countries, and more thorium metal is able to be produced in bulk quantities from thorium oxide or thorium nitrate, hundreds of tonnes of which are produced annually worldwide as a by-product of rare-earth mining [50]. This availability means recycling of irradiated ²³²Th target material may not be necessary. Another advantage of this method is that facilities already exist with demonstrated ability to perform target fabrication, irradiation, and processing. Examples of accelerator facilities capable of producing large amounts of ²²⁵Ac via proton spallation are listed in Table 1.

While spallation of naturally occurring ²³⁸U will also produce ²²⁵Ac, ²³²Th irradiation is preferred for a number of reasons. The ²³⁸U(p,x)²²⁵Ac reaction cross-section is ∼10 times lower (as modelled using FLUKA and GEANT4), and due to the higher density and lower melting point of uranium, thorium targets could more safely handle the higher heat load induced by higher beam currents. The co-production of fissile ²³⁹Pu and ²³⁵U is also avoided by ²³²Th irradiation.

The spallation of thorium produces a number of isotopes other than ²²⁵Ac. While this may provide an opportunity for recovery of other useful isotopes, it also complicates target processing by requiring the separation of dozens of elements. An overview of the different methods for isolating ²²⁵Ac from irradiated thorium is described in Section 3.2.

Concerns exist in the field that the amount of ²²⁷Ac coproduced from thorium spallation will prevent its use as a method for clinical-grade ²²⁵Ac production. An ²²⁷Ac to ²²⁵Ac activity ratio of 0.1-0.2% is typically found in irradiated targets at end of bombardment. However, potential exists for current target processing methods to be modified to produce ²²⁵Ac quantities that are free of ²²⁷Ac by isolation of an radium-actinium generator [59]. Most methods already isolate radium from the irradiated thorium matrix, and if this is done days after EOB, only ²²⁸Ra, ²²⁶Ra, ²²⁵Ra, ²²⁴Ra, and ²²³Ra will be present because of the length of their half-lives (t_1/2 = 5.7 y, 1600 y, 14.9 d, 3.6 d, and 11.4 d, respectively). Of these, ²²⁸Ra and ²²⁵Ra beta-decay to actinium isotopes, while the others alpha-decay to radon isotopes. Use of this mixture as a radium-actinium generator will produce ²²⁵Ac free of ²²⁷Ac. While ²²⁸Ac (t_1/2 = 6.2 h) will be present after ²²⁵Ra/²²⁵Ac separation, the ratio of ²²⁸Ac to ²²⁵Ac activity (∼0.88% at the time of optimal ²²⁵Ac elution [59]) will decrease with time. After sufficient ²²⁸Ac decay, ²²⁵Ac could then be removed from the ²²⁸Th (t_1/2 = 1.9 y) produced by ²²⁸Ac decay to obtain a final ²²⁵Ac product with significantly reduced radioactive impurities when compared to the directly produced ²²⁵Ac fraction. While the total ²²⁵Ac yield from this method will be reduced by a factor of about 10, it does not prevent the use of the directly produced, ²²⁷Ac-containing batch of ²²⁵Ac from being used for research or for use in ²²⁵Ac/²¹³Bi generators.

Only a few existing accelerators can produce proton beams with a current and energy sufficient for large-scale ²²⁵Ac production. A list of some of these facilities is given in Table 1 along with estimates of the maximum amount of ²²⁵Ac each could produce per month. These values only include directly produced ²²⁵Ac and exclude ²²⁵Ac produced from ²²⁵Ra generators that would increase production for each by roughly 10% to 20%. Without knowing details of each institutions’s target irradiation facilities, all are compared based on their maximal yield estimates that assume a target station capable of handling a thorium target thick enough to completely stop the proton beam (the same assumption was made for ²²⁶Ra (p,2n)²²⁵Ac maximal yields)8. As a result, practical yields will be lower. For example, while TRIUMF’s 500 MeV, 120 µA beam could theoretically produce 11.2 TBq (304 Ci) of ²²⁵Ac per month, 3 TBq (82 Ci) of monthly ²²⁵Ac production is a more practical limit given the existing target station’s size and cooling capacity. Similarly, practical estimates for production at BNL and LANL are 165 GBq (4.5 Ci) per month [58].

2.9. TRIUMF Perspective on ²²⁵Ac Production

Given the costs and challenges associated with ²²⁶Ra targets, the existing facilities already capable of thorium target spallation, and the large number of successful thorium target irradiations described in the literature, we believe that the development of ²²⁵Ac production via proton spallation of thorium is the fastest way to reliably meet the current global demand for ²²⁵Ac and support the widespread clinical use of any future therapies.

To this end, TRIUMF is working towards the development of routine ²²⁵Ac production via the irradiation of thorium metal targets on its primary beamline (BL1A), which delivers up to 500 MeV, 120 µA protons to the currently under-utilized legacy 500 MeV Isotope Production Facility (IPF). Since IPF is located directly before the BL1A beam dump, IPF routinely receives >100 µA of beam. Progress so far has included the low-level (2 µAh) irradiation of a prototype thorium oxide target in December 2016. The irradiation of thorium metal targets followed by processing on-site to isolate 370 MBq (10 mCi) of ²²⁵Ac is anticipated for late 2017, followed by 3700 MBq (100 mCi) production in 2018.

3. RADIOCHEMICAL ISOLATION OF ²²⁵Ac

The most established production method is the ²²⁹Th generator, which currently provides the main source of ²²⁵Ac and ²²⁵Ac/²¹³Bi generators used in preclinical or clinical trials. In addition, radiochemical procedures for several alternative strategies for ²²⁵Ac production from thorium and radium targets are also discussed. Details on radiochemical aspects of ²²⁵Ac/²¹³Bi generators are beyond the scope of this review, and the reader is referred to a review by Mogenstern et al. for additional insight into this area [24].

3.1. Isolation of ²²⁵Ac from ²²⁹Th Generators

Separation of ²²⁵Ac and ²²⁵Ra from ²²⁹Th (Fig. 1) is routinely performed at Oak Ridge National Laboratory (ORNL, TN, USA) [16], Institute for Transuranium Elements (JRC-ITU, Karlsruhe, Germany) [17, 61], and Leipunskii Institute for Physics and Power Engineering (IPPE, Obninsk, Russia) [62]. At ORNL, ²²⁹Th (5.6 BGq, 150 mCi) stock is divided into several batches and separation occurs every three weeks. A four-step chemical recovery procedure is used, including the combination of anion and cation exchange columns in nitric and hydrochloric media [16]. An anion exchange resin is first used to remove the bulk thorium mass. A 5×60 cm² (1.2 L) column filled with MP1 (250±50 mesh) resin, allows the sorption of up to 30 g of thorium in 8 M HNO₃, while ²²⁵Ac and ²²⁵Ra pass through the column. This process is repeated using a second MP1 column (2×10 cm², 30 mL) in order to separate residual thorium quantities. The fraction containing Ac and Ra is evaporated to dryness and re-dissolved in 10 M HCl and then loaded onto the next anion exchange (MP1) resin, which allows purification of actinium and radium from iron and uranium micro-impurities. Final separation of ²²⁵Ac from radium is performed on two cation exchange (AG50x4) columns in nitric acid media. Average recovery yield of ²²⁵Ac is 80% over one campaign with radionuclidic purity of over 99%. The use of isolated ²²⁵Ra as an additional ²²⁵Ac source occurs by storing ²²⁵Ra on a cation exchange resin for ²²⁵Ac accumulation and subsequent separation.

At ITU, a similar strategy using anion exchange column for ²²⁵Ac and ²²⁵Ra separation from ²²⁹Th is employed [17]. Due to the smaller ²²⁹Th (1.7 GBq, 46 mCi) stock compared to ORNL, the production of clinically relevant ²²⁵Ac quantities requires separation once every nine weeks after maximum ²²⁵Ac buildup is achieved. Separation from thorium is performed on two anion exchange (AG1X8) columns (0.5 L and 80 mL) in 8 M HNO₃, where ²²⁵Ac and ²²⁵Ra first pass through without sorption and ²²⁹Th is later eluted from the resin in 0.05 M HNO₃. The ²²⁹Th fraction is evaporated to near dryness and re-dissolved in 8 M HNO₃ and mixed with anion exchange resin for storage until the next elution. The Ra and Ac fraction is subject to further purification from residual thorium via solid phase extraction chromatography, using three columns filed with diamyl, amylphosphonate (UTEVA) resin. Separation of Ra from Ac is then performed inoctyl(phenyl)-N,N-diisobutylcarbamoyl-methylphosphine oxide (RE-resin) resin in nitric acid. The reported recovery yield for ²²⁵Ac is higher than 95% with a de-contamination factor from thorium of about 10². The same group also reports alternative separation of ²²⁵Ra and ²²⁵Ac via N,N,N’,N”-tetrakis-2-ethylhexyldiglycolamide (DGA) based resin in nitric acid [61].

At IPPE, separation of ²²⁵Ac and ²²⁵Ra from ²²⁹Th stock (5.6 BGq, 150 mCi) occurs on an anion exchange column in nitric acid followed by Ra/Ac separation on a cation exchange column in diluted nitric acid [18]. The final ²²⁵Ac fraction is additionally purified with a combination of cation and anion exchange columns. A portion of the stock of ²²⁹Th is loaded on an anion exchange column (Dowex 1X8, 0.5 L) in 8 M HNO₃. As previously described, Ac and Ra pass through the column, while ²²⁹Th is retained. The column is additionally washed with 1 L of 8 M HNO₃ for complete Ac/Ra elution. ²²⁹Th is then stripped with 2.2 L of 0.05 M HNO₃ and converted to 8 M HNO₃ (250-300 mL) for future use. The eluted Ac and Ra fractions are evaporated to dryness and re-dissolved in 0.5 M HNO₃ and passed through a cation exchange column (Dowex 50X8, 6 mL) on which ²²⁵Ac is retained while ²²⁵Ra passes through. The column is further washed with 1.5 M HNO₃ to remove residual ²²⁵Ra isotopes. Actinium is desorbed with 8 M HNO₃ and converted to 10 M HCl for further purification. Finer purification of the ²²⁵Ac solution is performed using anion and cation exchange columns in hydrochloric and nitric acids, respectively. The final actinium product is eluted in 10-12 mL of 8 M HNO₃. ²²⁵Ac separation yield varies from 85-95% with ^{229, 228}Th impurities below 10^-5 of ²²⁵Ac by activity.

Additionally, several alternative methods were previously reported for isolation of research quantities of ²²⁵Ac and ²²⁵Ra from ²²⁹Th [63,64]. One purifies ²²⁵Ac from ²²⁹Th by chelation with citric acid at various pH on a cation exchange column [63]. The final separation scheme includes sorption of Ac, Ra, and Th on a cation exchange (Aminex A-5, 4×40 mm²) column after loading with 0.25 M citrate (pH <1). Th was further eluted with ammonia citrate (pH 2.3) and Ac and Ra were eluted with ammonia citrate pH 4.5 and 4 M HNO₃, respectively. Another separation strategy based on the combination of cation and anion exchange chromatography with mixed HNO₃-methanol solution was also tested [64]. This strategy provides a tandem system for consequent separation of ²²⁵Ac from ²²⁹Th followed by loading of the ²²⁵Ac fraction on a generator for ²¹³Bi.

3.2. ²²⁵Ac Isolation from Thorium Targets

As described in Section 2.8, the irradiation of thorium metal with protons above 100 MeV can produce large quantities of ²²⁵Ac. The main challenges associated with the chemical isolation of ²²⁵Ac from these targets are a limited number of facilities capable of handling targets containing large ²²⁵Ac quantities (>10 GBq, or >0.3 Ci) and the challenging chemistry required for separating actinium from multiple grams of thorium and several hundreds of co-produced fission products. As discussed in Section 2.8, another drawback is the co-production of other actinium isotopes, most of which have relatively short half-lives (≤30 hours). One exception is the longest-lived Ac isotope, ²²⁷Ac (t_1/2 = 21.8 y), co-produced at a rate of 0.1-0.2% when compared to ²²⁵Ac activity. While the production of ²²⁵Ac without ²²⁷Ac is possible via the isolation of a ²²⁵Ra/²²⁵Ac generator (see Section 2.8), this further complicates the required chemical processing.

Strategies similar to those described in Section 3.1 that use anion exchange columns to retain thorium while eluting Ac and Ra can also be applied to the processing of thorium targets. However, this method requires a large volume of resin to remove the many grams of thorium used as target material - compared to the mg quantities used in ²²⁹Th generators - as well as additional purification steps to remove other spallation products. While several strategies were developed and tested for separation of actinium and radium from thorium [54, 65], none of these provide actinium with the minimal impurities required for radiolabeling of biomolecules and clinical application.

More recently, two alternative separation methods were developed and tested for isolation of ²²⁵Ac suitable for radiopharmaceutical application. Both methods are practically suitable for masses of thorium of up to 100 grams - larger ²³²Th targets will require larger volumes of extraction and chelation agents. The first, described by Aliev et al. [23], uses a three-step procedure that includes liquid-liquid and solid phase extraction chromatography [23, 57, 66]. First thorium metal target was dissolved in nitric acid with a small portion of HF. Further, two extractions with Di-(2- ethylhexyl)phosphoric acid (HDEHP) were performed, where the bulk thorium mass was extracted into in the organic phase while actinium, radium and most spallation products were retained in aqueous phase. Following phase separation, the aqueous phase is passed through N,N,N’,N”-tetrakis-2-ethylhexyldiglycolamide (DGA) column for separation of actinium and lathanides from fission products and radium in nitric acid media. The Ac and lanthanide fraction is then processed using TRU resin in 3 M HNO₃. Reported recovery yields for actinium are higher than 85% and the process was adopted to a hot cell environment for remote manipulation [23]. A modified strategy has also been applied to co-extract ²²³Ra along with ²²⁵Ac [67].

The second method was the result of an multi-institutional collaboration within the US DOE that included Los Alamos, Oak Ridge and Brookhaven National Laboratories [22, 68, 69]. This two-step procedure first removes the bulk thorium mass by chelating thorium and spallation products in 1 M oxalic acid at pH 2. At these conditions, all cations with charge 4+ and higher form negatively charged complexes, while cations with lower charge form positively charged complexes. Therefore, positively charged actinium and radium complexes are retained on the cation exchange resin while the bulk mass of thorium and majority of the fission products pass through the column without absorption. A similar strategy had also been previously used for ²²⁹Th/²²⁵Ac generators [63]. Ac and Ra are then eluted in 6 M HNO₃ and directly loaded onto a solid phase extraction chromatography column (N,N,N’,N”- tetrakis-2-ethylhexyldiglycolamide (DGA branched). At loading conditions, 2+ charged cations (e.g. Ra and Ba) pass through the column without sorption and actinium and lanthanides are retained. Actinium can be separated from lanthanides by specifically eluting with 10 M HNO₃ from the DGA column, while lanthanides are retained under such conditions. Reported recovery yield of actinium is higher than 98% with high radionuclidic and radiochemical purity suitable for radiolabeling or generator application.

This strategy has been further improved by additional steps for the co-extraction of other valuable medical radionuclides - including ^{225, 224, 223}Ra, ²³⁰Pa, ¹⁰³Ru and ¹¹¹Ag - without disturbing the actinium purification process [70].

3.3. ²²⁵Ac Isolation from ²²⁶Ra Targets

Irradiation of radium with low energy protons and photons also results in ²²⁵Ac production (see Sections 2.5 and 2.6). From a radiochemical separation standpoint, ²²⁵Ac needs to be isolated from macroscopic amounts of radium isotopes and their daughters. Previously published work suggests the use of a two-step procedure for ²²⁵Ac purification [42]. Irradiated ²²⁶Ra targets are dissolved in 0.01 M HCl, followed by purification of Ac from the bulk Ra mass (plus small Po and Pb quantities) by sorption of ²²⁵Ac on di(2-ethylhexyl) orthophosphoric acid (HDEHP) resin (LN-resin) in 0.01 M HCl. The column is washed with 0.1 M HCl and actinium was eluted with 2 M HCl. The second purification step uses 4,4’(5’)-di-t-butylcyclohexano 18-crown-6 (crown ether) resin (Sr-resin) for removal of residual amounts of radium and its daughters. The actinium fraction was passed through the Sr-resin column in 2 M HCl. No radiochemical yields or specific values for radiochemical purities have been reported for this method reported, though radiochemical purity is stated to be similar to ²²⁵Ac derived from ²²⁹Th generators.

3.4. Status at TRIUMF

At TRIUMF, several production routes for ²²⁵Ac are currently in use and under consideration. The collection of ²²⁵Ra and ²²⁵Ac ion beams at ISAC (see Section 2.2), is followed by a radiochemical separation consisting simply of one DGA column in nitric acid. However, chemical purity of ²²⁵Ac is still under evaluation due to the possible stable impurities originating from the implantation target. As mentioned in Section 2.9, ²²⁵Ac production via irradiation of sealed thorium metal targets is also planned at TRIUMF’s 500 MeV Isotope Production Facility, located on beamline 1A. While isolation of actinium and radium isotopes can be performed with one of the previously described strategies, any potential future full utilization of the high proton beam energy will require irradiation of large targets (>100 g). Therefore, alternative separation strategies are under evaluation. The upcoming ARIEL facility [26, 71] will also provide additional opportunity for irradiation of thorium targets with protons (480 MeV, 10-100 µA) as well irradiation of radium targets with photons. Design of radium targets and radiochemical separation of actinium from bulk radium are currently under consideration by our Life Sciences group.

4. TARGETING DESEASE with 225AC

4.1. Chemistry of 225Ac and Radiolabeling Challenges

The lack of any stable actinium isotopes has restrained the advancement of actinium chemistry, and as a consequence the chemistry of Ac(III) is virtually unknown [72, 73]. Only very recently have some studies begun to elucidate the fundamental coordination chemistry of this highly radioactive element [74, 75]. Actinium isotopes are typically +3 ions with a documented ionic radius of 112 pm (CN 6) [76]; its large size is likely suited to large polydentate chelators of high denticities, since most commonly used chelates for Ac(III) range between 8-12 coordinate [72]. Actinium is similar to other actinides and rare earth elements, and can undergo hydrolysis in solution in the absence of a chelating agent to form [Ac(OH)_3−x]^x- ; the sub-picomolar concentrations of ²²⁵Ac will cause the hydroxide species in turn to form radiocolloids that bind to surfaces such as reaction vessels [77].

The emission of multiple alpha-particles in the ²²⁵Ac decay chain (Fig. 1) makes ²²⁵Ac a particular effective isotope to kill cancer cells, yet also makes the directed delivery of the nuclide and its decay daughters a challenge. Due to the conservation of momentum, the emission of an energetic alpha particle (energies shown in Fig. (1)) imparts a recoil energy to the daughter nucleus often >100 keV, 1000 times larger than the binding energy for any chemical bond [14]. This results in release of the daughter nuclide from the original delivery vector (Fig. 2). The subsequent redistribution of the alpha- emitting daughter nuclides in vivo can cause substantial harm to untargeted healthy tissues and reduce the therapeutic effect. Consequently, renal toxicity induced by ²¹³Bi is considered to be a major constraint to the application of ²²⁵Ac in a large number of clinical trials [78]. There are three main strategies for limiting the toxicity of recoil daughters in the literature: fast uptake and internalization of the alpha emitters in the target tissue, encapsulation of the nuclide in a nanoparticle, or local administration of radioactivity directly into the target site via injection [14]. Herein, a literature review of the first two strategies is included.

Fig. (2).

Depiction of the recoil effect associated with α-decay of ²²⁵Ac. Daughter isotope ²²¹Fr is released from the chelating agent due to the 100 keV recoil energy associated with the alpha emission of ²²⁵Ac. ²²¹Fr and its decay daughters are consequently free to migrate within in the body.

4.2. Chelating Agents for ²²⁵Ac(III)

The discovery of a chelating agent that binds Ac(III) with sufficient stability and that also controls the release of its daughter nuclides remains a challenge. Moreover, limited global availability of ²²⁵Ac and the absence of a stable surrogate nuclide has limited the study of this isotope to a handful of institutions around the world that have secured a reliable ²²⁵Ac supply. The majority of initial ²²⁵Ac chelation studies have focused on screening a variety of commercially available polydentate macrocyclic or acyclic ligands for their ability to bind ²²⁵Ac and form stable complexes in vitro or in vivo. Despite the unique coordination preferences of the large +3 actinide, very few studies investigating new ligands specifically designed to coordinate Ac(III) can be found throughout the literature. A brief summary of ligands tested with ²²⁵Ac can be found in Table 3.

Table 3.

Tested ²²⁵Ac chelates coupled with targeting vectors for in vitro or in vivo application and their radiolabeling yields (RCY) and stabilities. Ligands rating system: red = not suitable for use; orange = adequate labeling, but likely unstable complex, or needs improvement; green = sufficient labeling and complex stability in vitro and in vivo, recommended for use.

Chelate (and Corresponding Tested Bifunctional Analogues)	Donor Set (CN#)	Grade	Radiolabeling Conditions & RCY	In vitro or In vivo Stability	Ref.
DOTA 1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid	N4O4 CN = 8	Green- orange	0.02 M ligand, NH4Ac pH 6, 37 ◦C, 2 h, RCY = 99%	In vitro human stability, 37 ◦C remained 90% intact after 10 days.	[79]
DOTPA 1,4,7,10-tetraazacyclododecane-1,4,7,10-tetrapropionic acid	N4O4 CN = 8	Red	0.02 M ligand, NH4Ac pH 6, 37 ◦C, 2 h, RCY = 0%	n/a	[79]
DOTMP 1,4,7,10-tetraazacyclododecane-1,4,7,10-tetramethylene-phosphinic acid	N4O4 CN = 8	Red	0.02 M ligand, NH4Ac pH 6, 37 ◦C, 2 h, RCY = 78%	In vitro human serum stability - rapid decom- plexation	[79]
TETPA 1,4,8,11-tetraazacyclotetradecane-1,4,8,11-tetrapropionic acid	N4O4, CN = 8	Red	0.02 M ligand, NH4Ac pH 6, 37 ◦C, 2 h, RCY = 0%	n/a	[79]
DTPA diethylenetriaminepentaacetic acid	N3O5, CN = 8	Red	0.02 M ligand, NH4Ac pH 6, 37 ◦C, 2 h, RCY = 0%	n/a	[79]

Early studies by Davis et al. Deal et al. and McDevitt 
et al. screened a library of ligands for their ability to coordinate ²²⁵Ac and tested the resultant complex’s in vitro or in vivo stability [77, 79, 80]. In the study presented by Davis et al., ligands EDTA (ethylenediaminetetraacetic acid, N₂O₄), CHX-A”-DTPA (cyclohexyl-diethylenetriaminepen-
taacetic acid, N₃O₅), and PEPA (1,4,7,10,13-pentaazacyc-
lopentadecane-N,N’,N”,N”’,N””-pentaacetic acid, N₅O₅) were radiolabeled with ²²⁵Ac with radiochemical yields (RCY) of 80-90%. Biodistribution profiles over the course of 8 days for each of the purified ²²⁵Ac-complexes were assessed by injecting 92 kBq (2.5 µCi) of each complex, and compared to the ²²⁵Ac-acetate biodistribution as a control. Since uncomplexed ²²⁵Ac accumulates predominantly in the liver with small amounts in the bone, kidney, and heart, high liver uptake of an ²²⁵Ac-chelate indicates an unstable complex in vivo. CHX-A”-DTPA and PEPA reduced liver uptake of the ²²⁵Ac-complex by more than 5.5 times compared to ²²⁵Ac-acetate, and although their data suggests ²²⁵Ac-CHX-A”-DTPA to be the most effective tested chelator complex with regard to its in vivo stability, improvements can still be made to further reduce non-target tissue accumulation [77]. As such, CHX-A”-DTPA provides inadequate chelation of Ac(III). Another important finding of the initial in vivo study was that the maximum tolerated dose of ²²⁵Ac-CHX-A”-DTPA was less than 185 kBq (5 µCi), since at doses of 185 kBq (5 µCi) or higher, severe tissue damage was observed as early as 1 hr post-injection (p.i.), which ultimately lead to death causing 100% mortality by day 8 p.i. [77].

In search of a better ²²⁵Ac chelator, Deal et al. developed a novel dodecadentate (coordination number, CN = 12) chelator with an extended macrocyclic ring to accommodate the large Ac³⁺ ion, HEHA (1,4,7,10,13,16-hexaazacyclohexadecane-N,N’,N”,N”’,N””,N””’-hexaacetic acid, N₆O₆), and compared its in vivo stability to ²²⁵Ac labeled EDTA, CHX-A”-DTPA, DOTA, and PEPA [80]. ²²⁵Ac-HEHA demonstrated the highest complex stability evidenced by the low uptake of the complex in all tissues; essentially all radioactivity was excreted within 1 hour. Despite this perceived in vivo stability, the authors suggested the predicted -3 charge at physiological pH of the ²²⁵Ac-HEHA complex may be mediating the fast excretion of the complex, giving it the appearance of stability since the radiometal ion doesn’t have time to dissociate within the time frame of excretion [80]. Given these initial promising results, efforts towards the preparation of a bifunctional HEHA analogue were undertaken [81]. A C-functionalized isothiocyanate HEHA derivative was successfully prepared and conjugated to three monoclonal antibodies (mAbs): BL-3, T101, and CC49. One-step labeling of ²²⁵Ac to HEHA-mAb incubated for 30 min at 37 °C, pH 7.0 produced moderate to high RCYs of 60-85% with specific activities of 7.4-14.8 MBq/mg (200-400 µCi/mg) (for ligand:mAb ratios > 1.0); these bioconjugates were sufficient for animal therapy studies. In vitro serum stability revealed the ²²⁵Ac-HEHA-BL-3 mAb conjugate to be only 50% stable in serum after 24 h, suggesting that the HEHA system may very well not be a suitable chelator for sequestering ²²⁵Ac [81].

McDevitt et al. screened the ²²⁵Ac radiolabeling efficiency and in vitro stability of several polydentate chelators including DTPA (diethylenetriaminepentaacetic acid, N₃O₅), DOTA (1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid, N₄O₄), TETA (1,4,8,11-tetraazacyclotetradecane-1,4,8,11-tetraacetic acid, N₄O₄), DOTPA (1,4,7,10-tetraazacyclododecane-1,4,7,10-tetrapropionic acid, N₄O₄), TETPA (1,4,8,11- tetraazacyclotetradecane-1,4,8,11-tetrapropionic acid, N₄O₄), and DOTMP (1,4,7,10- tetraazacyclododecane-1,4,7,10-tetramethylene-phosphinic acid, N₄O₄) [79]. Of the six ligands tested, only DOTA and DOTMP showed any complexation of ²²⁵Ac after 2 h at 
37 °C with RCYs of >99 and 78%, respectively. Subsequent in vitro stability assays in serum suggested that the ²²⁵Ac-DOTA complex was robust, remaining >90% intact after 10 days, while the ²²⁵Ac-DOTMP complex rapidly dissociated.

The initial promising in vitro stability of ²²⁵Ac-DOTA motivated the authors to prepare conjugates of DOTA with antibodies HuM195 (anti-CD33), mJ591 and huJ591 (anti-PSMA), B4 (anti-CD19), and 3F8 (anti-GD2). The elevated temperatures required to achieve adequate labeling yield of Ac-DOTA were not amenable to antibody conjugates since such reaction conditions would denature proteins and result in loss of function. Consequently, a two-step labeling process was employed which required ²²⁵Ac radiolabeling of the bifunctional DOTA-NCS ligand first, followed by mAb conjugation (pH 8.7, 37 °C for 52 min). Despite low overall radiochemical yields of only 9.8 ± 4.5%, reasonable specific activity (4.1 ± 2.6 GBq/g, or 0.11 ± 0.07 Ci/g) was achieved which would allow for preclinical therapeutic studies. Low yields were attributed to the first ²²⁵Ac labeling step of DOTA-NCS which required heating and, consequently, degradation of the isothiocyanate linker resulting in poor mAb conjugation in the following step.

Attempts to increase the 2-step labeling yields of ²²⁵Ac-DOTA-mAb conjugates via the modification of a more robust DOTA-linker chelate system have yielded some improvements [82]. Antczak et al. established a 2-step labeling protocol for thiol-based DOTA-linkers which provided a marked improvement compared to the DOTA- NCS 2-step method, with chelation yields of 95-99% and labeling yield up to 40%. In the same study, the authors incorporated an enzymatically cleavable linker which could minimize the toxicity associated with long-circulating mAbs to normal tissues by allowing the release of a small molecular weight radiometal-chelate complex from the mAb to promote fast clearance of the therapeutic nuclide. However, constructs resulted in high accumulation of ²²⁵Ac in the liver in small animal models, indicative of ²²⁵Ac release from the chelate [82].

Perhaps the most noteworthy improvement to ²²⁵Ac radiolabeling to date was presented by Maguire et al. [83], which offered for the first time an efficient 1-step radiolabeling procedure for ²²⁵Ac-DOTA-antibody constructs. Typical radiolabeling proceeded in 2 M tetramethyl ammonium acetate buffer (pH 7.5) with the addition of L- ascorbic acid as radioprotectant to the addition of DOTA-antibody construct and ²²⁵Ac³⁺ with a typical final reaction pH of 5.8. Heating to 37 °C for 2 hours allowed a 10-fold increase in radiochemical yield (80%) compared to previous 2-step methods (6-12%), and resulted in the preparation of bioconjugates with up to 30-fold higher specific activities (120 GBq/g compared to 3.7-14.8 GBq/g) [83]. The highest specific activity achieved was equivalent to 1 actinium for every 25 antibodies. These results will likely have great implications in preclinical and clinical uses of ²²⁵Ac labeled antibodies, since the ease of synthesis can more easily be translated in a clinical setting and the significant reduction of ²²⁵Ac losses during labeling can save the cost of this rare and valuable isotope.

4.3. Bioconjugates for ²²⁵Ac Labelling

Despite often requiring elevated temperatures and extended reaction times, promising stability in serum and efficient labeling of ²²⁵Ac-DOTA complexes has cemented the commercial chelator and its bifunctional analogues as the most exploited chelator for ²²⁵Ac thus far. With the relatively limited scope of chelates tested with ²²⁵Ac to date (see Table 3), and very few examples of specifically tailored Ac-chelates in the literature, it seems apparent there is much room for improvement in the field of Ac-chelation, particularly as more powerful spectroscopic and computational techniques evolve that will continue to help elucidate the unique chemical differences between Ac³⁺ and other +3 actinides and lanthanides [74]. Nonetheless, ²²⁵Ac labeled DOTA-small molecule [10], peptide [84-86], and antibody [78, 87, 88] conjugates have been tested in a variety of in vitro and in vivo preclinical studies, and a select few have seen clinical successes. Specifically, the ²²⁵Ac-labeled humanized anti-CD33 (HuM195) mAb is in clinical trials for the treatment of patients with advanced myeloma [89]. Most recently a brief communication by Kratochwil et al. [10] reported remarkable treatment success in two patients with metastasized castration-resistant rostate cancer (mCRPC) who had extensive pretreatments and showed resistance to other therapies including beta-emitting radiopharmaceuticals. The small molecule ²²⁵Ac-PSMA-617 (Fig. 3) was administered bi-monthly at doses of 100 kBq/kg (2.7 µCi/kg); patient 2 saw complete remission after 3 treatment cycles.

Fig. (3).

DOTA-urea conjugate PSMA-617 used for targeting of prostate-specific membrane antigen.

4.4. ²²⁵Ac Labelled Nanoparticles

To circumvent the inevitable loss of ²²⁵Ac daughters after alpha decay from an actinium-chelate complex, researchers have sought to encapsulate the highly potent alpha-emitter into a nanoparticle structure. It is hypothesized that the ²²⁵Ac³⁺ ion and its decay daughters can be retained within the cavity of the nanoparticle structure, while the alpha particles are released and able to deposit their therapeutic dose at the intended target site. However, the use of nanoparticles as a platform to affix radionuclides or other biomolecular targeting vectors comes with several limitations. The biodistribution of nanoparticles is dominated by their large size and ability to take advantage of the enhanced permeability and retention (EPR) effect of cancer cells, where “leaky” vessels of poorly vascularized tumours allow for the uptake and retention of large macromolecules [72]. Moreover, the relatively large particles are often primarily excreted through the hepatic pathway which can cause unwanted high liver uptake. The accumulation of a highly toxic alpha-emitter in the liver may cause damage to the organ. Much of the available literature describing ²²⁵Ac-labeled nanoparticles provides in vitro data only [90-95]. Nonetheless, a brief overview of some strategies to prepare ²²⁵Ac radiolabeled nanoparticles is found below.

Work by Matson et al. [94] investigated the encapsulation of ²²⁵Ac 3+ ions in single-walled carbon nanotubes (SWNTs) by co-encapsulation of Gd³⁺ in an ion cluster. Although the Gd³⁺ ions remained inside the SWNTs, continual leakage of the ²²⁵Ac³⁺ ions was seen when challenged with serum. McLaughlin et al. [90] employed a multilayered nanoparticle structure which can contain the recoiling daughters of the in vivo alpha generator at the centre cavity, while coupling the outer layer to antibodies and without preventing the release of emitted alpha-particles. The shells included a radiation resistant lanthanide phosphate crystal doped with ²²⁵Ac and layered with a magnetic GdPO₄ layer, plus a gold outer shell for the attachment of targeting vectors. Polymer vesicles (polymersomes) composed of poly(butadiene-b-ethylene oxide) have also been used to encapsulate ²²⁵Ac [92]. Preliminary in vitro studies in cells showed that smaller particles were absorbed by the cells and gathered around the cell nucleus. However, experiments and simulation indicated that larger polymerases are needed to attain satisfactory retention of recoiling daughters [92]. PEGylated liposomes loaded with ²²⁵Ac and labeled with mouse antihuman PSMA J951 antibody or with the A10 PSMA aptamer were tested in vitro for their targeting, internalization, and cytotoxicity on a prostate cancer cell line [91, 95]. These studies demonstrated anti- PSMA targeted liposome loaded with ²²⁵Ac can selectively bind, become internalized, and kill PSMA-expressing cells. Similarly, a ²²⁵Ac-loaded lipid-based nanocarrier was labeled with a PSMA targeting antibody or small molecule urea-based agent, and the targeting selectivity and cytotoxicity were compared to the radiolabeled antibody on its own [95]. It was found that the loaded lipid vesicles improved the killing efficacy 3-fold compared to the same levels of activity per cell when delivered by the PSMA-targeting antibody.

4.5. Assessing the Biodistribution of the ²²⁵Ac Decay Chain

While this discussion so far has been limited to the biodistribution of ²²⁵Ac itself, assessment of the biodistribution of each alpha-emission in the decay chain is necessary when evaluating the performance of ²²⁵Ac-radiopharmaceuticals. As previously mentioned in Section 4.1, the retention or redistribution of ²²¹Fr, ²¹⁷At, and ²¹³Bi at the target site impacts the efficacy and toxicity of the radiopharmaceutical. While the half-life of ²¹⁷At is short enough that its biodistribution can be assumed to be effectively identical to ²²¹Fr, the short ²²¹Fr half-life makes accurately determining its biodistribution - and also independently determining the biodistribution of its ²¹³Bi granddaughter - a challenge using conventional ex vivo counting methods. Speedy harvesting and counting of organs is essential, since while successive measurements of the same ex vivo tissue samples over time can be used to estimate the amount of ²²¹Fr or ²¹³Bi present at the time of sacrifice, the uncertainty in these estimates increases the longer after sacrifice the first measurements are made [14].

Imaging-based methods can also be used to assess the biodistribution of the radionuclides in vivo, and quantitative SPECT imaging of ²²⁵Ac progeny isotopes has been demonstrated on small-animal SPECT/CT systems for ²¹³Bi alone [96], and for both ²²¹Fr and ²¹³Bi simultaneously, via their 218 keV and 440 keV gamma lines, respectively [97]. Unfortunately, quantitative imaging of the high energy ²¹³Bi photopeak (440 keV) requires the use of a high energy collimator not available on most imaging systems. However, qualitative SPECT imaging of ²¹³Bi has been performed clinically, as has qualitative ²²¹Fr SPECT in preclinical settings [11, 12, 90, 98, 99]. The use of Cerenkov imaging has also been demonstrated in vivo for the ²²⁵Ac decay chain [86], though this imaging modality is incapable of quantitative biodistribution measurements and cannot distinguish between individual ²²⁵Ac decay chain components.

While quantitative SPECT imaging of ²²¹Fr and ²¹³Bi with sub-millimeter spatial resolution has the potential to assess the retention of ²²⁵Ac progeny within the tumour and determine uptake within whole organs [97], the short range (<100 µm) of alpha particles mean that information regarding the sub-organ biodistribution - a level of detail not provided by current in vivo imaging modalities - is necessary for alpha-particle dosimetry [100, 101]. While ex vivo imaging using alpha-cameras can determine ²²⁵Ac biodistributions with spatial resolutions sufficient for dosimetry [102-106], alpha particle dosimetry itself faces additional challenges that currently limit the translation of preclinical dosimetric data to biological outcomes in the clinic [100, 101], a discussion of which is beyond the scope of this review.

4.6. Progress at TRIUMF

By leveraging our existing infrastructure TRIUMF has established ∼37 MBq (1 mCi) annual production of ²²⁵Ac via our ISOL facility (Section 2.2). This ²²⁵Ac has enabled TRIUMF to conduct a variety of preclinical radiolabeling, complex stability, and imaging studies [97]. In particular, the apparent lack of chelating agents available to complex ²²⁵Ac under conditions commensurate for “kit” type preparation of radiopharmaceuticals has motivated our researchers to search for alternate ligands which can quantitatively sequester the radioactive metal under mild, room temperature conditions with low ligand concentrations while forming thermodynamically stable and kinetically inert radiometal-chelate complexes. To this end, we are currently screening a wide variety of macrocyclic and acyclic chelates for their potential as ligands in ²²⁵Ac radiopharmaceutical development. Ligand candidates which display promising ²²⁵Ac radiolabeling efficiencies and high in vitro stability will be conjugated to selected antibody or peptide targeting vectors, and small animal in vivo studies can be conducted.

Conclusion

With some recent and remarkable clinical results in the treatment of late-stage cancers using beta- and alpha-emitting radiopharmaceuticals, the community has experienced a surge in interest in certain therapeutic isotopes. Supply challenges, especially for isotopes such as ²²⁵Ac, have limited the development of existing and emerging targeted radiotherapeutics, and pose a significant challenge in the discovery of new agents. That said, there exists a substantial amount of untapped production potential across a number of facilities.

Discussed within are the different ²²⁵Ac production routes which span the spectrum of neutron (reactor), electron (gamma) and proton-induced reactions. At this time, the most promising short-term approach to greater availability of ²²⁵Ac would be to irradiate thorium metal with high energy (>70 MeV) protons. By enable existing facilities with access to protons in this energy range to produce ²²⁵Ac, it will allow for a focus on addressing unanswered questions on a specific activity, radiochemical and/or radionuclidic purity. There exists the possibility of exploiting the co-production of ²²⁵Ra to assemble and distribute ²²⁵Ra/²²⁵Ac generators - a route that could help reduce the amount of ²²⁷Ac in the ²²⁵Ac used for radiopharmaceutical preparation. ²²⁵Ac produced during the irradiation itself could be used for manufacturing ²²⁵Ac/²¹³Bi generator. Beyond this, safety challenges will continue in the handling and processing of irradiated Th targets given their extensive radionuclide content, which can be mitigated by leveraging the experience and repurposing equipment from facilities designed to handle large quantities of radioactive material from reactor fuel and waste processing.

Given the widespread availability of lower energy (16-24 MeV) medical cyclotrons, there also exists the tantalizing possibility of large-scale, decentralized production across the fleet of the hundreds of medical cyclotrons in operation across the globe today. The key challenge, in this case, would be more on the procurement of sufficient quantities of ²²⁶Ra, not to mention the safety implications of a failed irradiation in a hospital-based setting.

Adequate radiopharmaceuticals will also require the optimization of chelate design, with several advances having been recently reported. Enhanced supply will allow for the study and identification of compounds with the most suited pharmacokinetic profiles for therapy with minimized bystander organ dose or side effects. With clinical interest increasing, the most efficient way forward would be to enable increased production quantities by building or enabling existing facilities to produce isotopes such as ²²⁵Ac at the bulk-scale. By using these facilities to supply sufficient and routine quantities of ²²⁵Ac to advance existing clinical trials for compounds in clinical development - and should the early results seen to date represent results anticipated from a larger patient pool - there will undoubtedly be an increase in demand for more alpha-therapeutics toward more indications.

Consent for Publication

Not applicable.

Chelate (and Corresponding Tested Bifunctional Analogues)	Donor Set (CN#)	Grade	Radiolabeling Conditions & RCY	In vitro or In vivo Stability	Ref.
PEPA 1,4,7,10,13-pentaazacyclopentadecane-N,N’,N”,N”’,N””-pentaacetic acid	N5O5, CN = 10	Red	0.01 M ligand, NH4OAc pH 5.8, 40 ◦C, 30 min, RCY = 80%	inadequate stability in vivo	[80]
HEHA 1,4,7,10,13,16-hexaazacyclohexadecane-N,N’,N”,N”’,N””,N””’-hexaacetic acid	N6O6, CN = 12	Orange	M ligand, NH4OAc pH 5.8, 40 ◦C, 30 min, RCY = >95%, or >98% after 2 h	Rapid whole body clearance in vivo, facilitated by -3 charge. Low liver uptake suggests stability over short time in body	[80, 81]
CHX-A”-DTPA	N3O5, CN = 8	Red	0.01 M ligand, NH4OAc pH 5.8, 40 ◦C, 30 min, RCY = >95%	In vivo decom- plexation indicated by high liver uptake	[77, 80]
EDTA ethylenediaminetetraacetic acid	N2O4, CN = 6	Red	0.01 M ligand, NH4OAc pH 5, 40 ◦C, 30 min, RCY = 80-90%	In vivo decom- plexation indicated by high liver uptake	[77]

Conflict of Interest

The authors declare no conflict of interest, financial or otherwise.