Recovery of ¹⁷⁷Lu from Irradiated HfO₂ Targets for Nuclear Medicine Purposes

Abstract

A new method of production of one of the most widely used isotopes in nuclear medicine, ¹⁷⁷Lu, with high chemical purity was developed; this method includes irradiation of the HfO₂ target with bremsstrahlung photons. The irradiated target was dissolved in HF and then diluted and placed onto a column filled with LN resin. Quantitative sorption of ¹⁷⁷Lu could be observed during this process. The column later was rinsed with the mixture of 0.1 M HF and 1 M HNO₃ and then 2 M HNO₃ to remove impurities. Quantitative desorption of ¹⁷⁷Lu was achieved by using 6 M HNO₃. The developed method of ¹⁷⁷Lu production ensures high purification of this isotope from macroquantities of hafnium and zirconium and radioactive impurities of carrier-free yttrium. The content of ^177mLu in ¹⁷⁷Lu in photonuclear production was determined. Due to high chemical and radionuclide purity, ¹⁷⁷Lu obtained by the developed method can be used in nuclear medicine.

1. Introduction

¹⁷⁷Lu is one of the most known and widely used therapeutic radioisotopes in nuclear medicine [1]. The first clinical application of ¹⁷⁷Lu happened in the 1960s; however, the breakthrough in using radiopharmaceuticals based on this radionuclide occurred later with the development of the ¹⁷⁷Lu-DOTATATE complex designed to treat neuroendocrine tumors [2,3,4]. Nowadays, the main method of ¹⁷⁷Lu production is the irradiation of ¹⁷⁶Lu or ¹⁷⁶Yb in nuclear reactors, and this production has its disadvantages. Thus, the operation of reactors leads to the accumulation of nuclear waste; irradiation of ¹⁷⁶Lu leads to ¹⁷⁷Lu with carrier; recovery of ¹⁷⁷Lu from neighbor lanthanide Yb is not a simple task [1]. As a result, other perspective methods of ¹⁷⁷Lu production have been investigated recently, the photonuclear one in particular. Today, this method is used for medical isotope production. For instance, light isotopes ¹¹C, ¹³N, ¹⁵O, and ¹⁸F as well as radiometals ⁴⁷Sc, ⁶⁷Cu, and ⁹⁹Mo/^99mTc generators are regularly obtained in sufficient quantities using electron accelerators [5]. Compared to the use of nuclear reactors, the photonuclear method has the following advantages: the compact sizes of electron accelerators that create an opportunity of placing one near the hospital and the relatively cheap cost of accelerators’ functioning. The main disadvantage is cross-sections of photonuclear reactions that are usually lower than ones in reactor production. Data about photonuclear production of ¹⁷⁷Lu from hafnium are limited [6,7,8].

One of the issues of ¹⁷⁷Lu production is the formation of long-lived isomer ^177mLu (T_1/2 = 160.4 d), the content of which should be minimized in ¹⁷⁷Lu-based radiopharmaceuticals (the activities ratio of ^177mLu and ¹⁷⁷Lu should not exceed 0.02 %). The method of ¹⁷⁷Lu recovery from irradiated HfO₂ using extraction chromatography was developed by us to determine the amount of formed ^177mLu in purified lutetium fractions [8]; any other techniques of recovery of trace amounts of lutetium from macroquantities of hafnium are absent. However, no gamma-ray peaks of ^177mLu were observed during a long registration of the gamma-ray spectrum of purified lutetium solution due to the registration of yttrium isotopes forming from zirconium impurities contained in the initial sample of HfO₂; their Compton plateau in the gamma-ray spectrum overlaps with ^177mLu peaks. Consequently, the development of methods of additional purification of recovered ¹⁷⁷Lu from forming during irradiation yttrium isotopes is a task of current interest.

The isotope ¹⁷⁷Lu can be produced either from hafnium with natural isotopic composition or from enriched ¹⁷⁸Hf, which is notably more expensive. It was mentioned that to produce ¹⁷⁷Lu with radionuclide purity sufficient for nuclear medicine, it is necessary to irradiate massive (a few grams) targets made of ¹⁷⁸Hf [8] and, as a result, such expensive target material should be reused after the recovery of ¹⁷⁷Lu.

The purpose of this work was to develop a promising method of ¹⁷⁷Lu production from HfO₂ using electron accelerator with subsequent purification from hafnium, zirconium, and especially yttrium, to develop a technique of irradiated HfO₂ regeneration, and also to determine the amount of ^177mLu produced during this irradiation.

2. Results and Discussion

In our work, a new method of ¹⁷⁷Lu production for nuclear medicine was developed: irradiation of HfO₂ with bremsstrahlung photons, its dissolution, and recovery of ¹⁷⁷Lu using extraction chromatography; the corresponding scheme is presented in Figure 1.

2.1. Development of Method of Recovery of ¹⁷⁷Lu from Macroquantities of Hafnium and Zirconium, Trace Amounts of Yttrium Isotopes

Method of ¹⁷⁷Lu recovery from irradiated HfO₂, developed previously by us, included the following steps: target material dissolution in HF_conc and dilution of the obtained solution fifteen times with 1 M HNO₃; sorption of lutetium on the column filled with LN resin (based on di(2-ethylhexyl)orthophosporic acid); rinsing the column with 1 M HNO₃ and 0.1 M HF mixture and then with 1 M HNO₃ for the remaining hafnium and for fluoride ion removals, accordingly; and, finally, rinsing with 6 M HNO₃ to desorb ¹⁷⁷Lu [8]. To further purify ¹⁷⁷Lu from yttrium, the step of column rinsing with 2.5 M HNO₃ was introduced before the desorption of ¹⁷⁷Lu; this solution was proved to be the optimal media for quantitative desorption of yttrium according to the conducted experiments (see Supplementary Section).

Table 1. Quantities of ¹⁷⁷Lu and other elements in eluate during different stages of ¹⁷⁷Lu recovery.

Stage	Content of Elements in Eluate, % of Initial
	Hf	Zr	Y	Lu
Sorption of 177Lu onto the column, rinsing with the mixture of 1 M HNO3 and 0.1 M HF	100	100	40	<2
Rinsing with 2.5 M HNO3	0	0	60	0
Desorption of 177Lu with 6 M HNO3	0	0	0	>98

contains data on the content of hafnium, zirconium, yttrium, and lutetium in eluates obtained during different stages of the developing technique; a chromatogram is presented in Figure 2.

It was established that hafnium and zirconium did not sorb onto the column during the sorption of lutetium; however, approximately 60% of yttrium was sorbed. Yttrium quantitatively desorbed during the column rinsing with 2.5 M HNO₃, while lutetium remained on the column. During the subsequent rinsing with 6 M HNO₃, ¹⁷⁷Lu quantitatively (no less than 98%) desorbed from the column.

Hafnium content in the obtained solution of lutetium was lower than the detection limit of ICP-MS. This led to the conclusion that the real content of hafnium in the ¹⁷⁷Lu solution during the recovery following this method is 1.2 × 10¹⁰ times lower compared to the content in the initial solution, which is five orders of magnitude higher than the result obtained by us earlier [8]. Zirconium content in the final product, according to ICP-MS, is 1.7 × 10⁶ times lower than in the initial solution. As for the purification of lutetium from yttrium, no peaks of yttrium isotopes were detected in a gamma-ray spectrum of lutetium solution aliquot during prolonged registration. Thus, according to the detection limit of ⁸⁸Y, yttrium content in the final solution was 10⁴ times lower than in the initial one.

2.2. Recovery of ¹⁷⁷Lu from Irradiated HfO₂, Determination of ^177mLu Content, and Regeneration of HfO₂

A target with a mass of 16 g was irradiated with bremsstrahlung photons with energy up to 55 MeV for 8 h; then, it was dissolved in HF, and the recovery was conducted according to the technique described above. It was established that the purification degree was achieved as outlined above, and the lutetium yield was 98.5 ± 0.5%.

The high purification level of lutetium from macroquantities of hafnium, zirconium, and trace amounts of yttrium, formed during the irradiation of zirconium achieved in our work, allowed us to detect ^177mLu peaks during prolonged registration of the gamma-ray spectrum of the obtained lutetium. Figure 3 presents the dependency of count rate of the 208 keV line, which is the most intense for both ¹⁷⁷Lu and ^177mLu, on time after the lutetium isotope’s recovery. It can be seen in Figure 1 that it is possible to determine the contribution of ^177mLu in the count rate of this line after ¹⁷⁷Lu decay, which allows us to precisely determine the radioactivity of ^177mLu after the irradiation. Thus, the ratio of the activity of ^177mLu to the activity of ¹⁷⁷Lu in the photonuclear production of ¹⁷⁷Lu was established to be (2.87 ± 0.07) × 10⁻⁵ (or 0.00287%).

Table 2. ^177mLu/¹⁷⁷Lu activity ratios in production of ¹⁷⁷Lu via different methods.

Nuclear Reactions Producing 177Lu	177mLu/177Lu Activity Ratio
178Hf(γ,p)177Lu (determined in this work)	0.00287%
178Hf(γ,p)177Lu (determined previously)	<0.013% [8]
176Lu(n,γ)177Lu	<0.02% [9]
176Yb(n,γ)177Yb → 177Lu	0 [10]
176Yb(d,p)177Yb → 177Lu + 176Yb(d,n)177Lu	<0.0045% [11]

allows us to compare the activity ratios ^177mLu/¹⁷⁷Lu in production by different methods, and it is clear that the ratio in case of the photonuclear method is minimal among direct production routes, and ¹⁷⁷Lu obtained by this method can be used in nuclear medicine.

According to X-ray diffraction (XRD), the spectra of commercial HfO₂ and the product of calcination of hafnium hydroxide obtained during the recovery of ¹⁷⁷Lu are identical, and values of interplanar distances coincide with the values for HfO₂ from the database. Thus, after heating, HfO₂ can be stored to decrease activity of hafnium isotopes if necessary, and can be reused for irradiation for ¹⁷⁷Lu production.

2.3. Comparison of Methods of Obtaining Carrier-Free ¹⁷⁷Lu

We demonstrated the possibility of producing and separating ¹⁷⁷Lu for nuclear medicine using electron accelerators. In conclusion, we present a comparison of this method and the production of ¹⁷⁷Lu without a carrier in a reactor and a cyclotron.

In the case of ¹⁷⁶Yb irradiation in the reactor, it is possible to produce 1.8 GBq of ¹⁷⁷Lu (therapeutic activity) by irradiating 5 mg of 97.6% ¹⁷⁶Yb₂O₃ for 10 days using flux of 1 × 10¹⁴ n∙cm⁻²∙s⁻¹ [12]. According to calculations based on experimental data, the same activity of ¹⁷⁷Lu can be obtained by irradiating a 100 µm plate of 100% ¹⁷⁶Yb with deuterons for about 2 h at a current of 0.1 mA [13]. According to our earlier theoretical calculations, 1.8 GBq of carrier-free ¹⁷⁷Lu can be produced in an electron accelerator by irradiating an enriched ¹⁷⁹HfO₂ target at a current of 0.1 mA [8]. However, it is important to note that the results of calculating the yields of photoproton reactions are usually underestimated from several times to several orders of magnitude. Thus, the determination of the experimental values of ¹⁷⁷Lu yields upon irradiation of enriched targets made of ¹⁷⁸Hf or ¹⁷⁹Hf is an urgent problem.

As for the separation of ¹⁷⁷Lu from irradiated Yb targets, the process takes a long time, and the loss of ¹⁷⁷Lu can reach 15% [12]. At the same time, in the present work, we demonstrated the possibility of rapid and quantitative recovery of ¹⁷⁷Lu from irradiated HfO₂. When ¹⁷⁷Lu is produced in a reactor, the long-lived ^177mLu isomer is completely absent [10]; when produced in an electron accelerator, isomer activity is 0.00287% of the activity of ¹⁷⁷Lu; and during cyclotron production, isomer activity does not exceed 0.0045% of ¹⁷⁷Lu one [11].

It is difficult to compare the cost of ¹⁷⁷Lu obtained by different methods for a number of reasons. The cost of production in the reactor is the lowest, but this method has disadvantages mentioned in the Introduction, including radioactive waste generation. Per unit of time, a higher ¹⁷⁷Lu activity is generated in the cyclotron than in an electron accelerator; however, the cost of the operation of the latter is lower. Finally, it is worth considering that the regeneration of HfO₂ targets is easy, as we demonstrated, while ¹⁷⁶Yb is usually not regenerated.

As a result, each of the described methods has its advantages and disadvantages, and each can be used to obtain ¹⁷⁷Lu for nuclear medicine purposes. In any case, it is currently possible to produce ¹⁷⁷Lu for preclinical studies in electron accelerators. Further development of the photonuclear method for obtaining ¹⁷⁷Lu consists of establishing the exact values of the yields of the desirable isotope after irradiation of different enriched hafnium targets.

3. Materials and Methods

3.1. Irradiation of HfO₂

^natHfO₂ with a weight of 16 g was placed in cylindrical polypropylene container with a volume of 5 mL; the remaining space in container was filled with cotton wool. The container was then irradiated for 8 h in RTM-55 microtron with maximum energy of electron beam being 55 MeV [14]. Tungsten plate of 2 mm thickness was used as an electron convertor; the usual value of average current was 100–200 nA for used accelerator. During radiochemical analysis of irradiated target isotopes ^177,178,179Lu, ^173,175Hf, ⁸⁹Zr, and ⁸⁸Y were found; the same isotopes were also observed in our work [8].

3.2. Target Dissolution, Recovery of ¹⁷⁷Lu, and Regeneration of HfO₂

Irradiated HfO₂ was dissolved in HF_conc by boiling for 1.5 h. Obtained solution was diluted 15 times with 1 M HNO₃, resulting in approximately 260 mL.

Four identical columns with volume of 3 mL and diameter of 0.6 cm each, filled with LN resin (100–150 mesh, Triskem Int, Bruz, France), were used in following recovery by extraction chromatography. The solution was divided into 4 equal portions; each was eluted through its own column. Fractions of 5 mL each were gathered during the elution; their gamma-ray spectra were registered using spectrometer with high-purity germanium detector Canberra GC1020 (Canberra Ind, Meridan, CT, USA). Content of hafnium, zirconium, and lutetium in fractions during recovery process was determined using gamma-peaks of the following isotopes: ¹⁷⁷Lu (208.4 keV), ¹⁷⁵Hf (343.4 keV), ⁸⁹Zr (909 keV). Content of yttrium was determined during prolonged registration of spectra using ⁸⁸Y peak (898 keV).

Regeneration of hafnium from the initial solution eluted through the column was conducted by adding ammonia to form precipitate of hafnium hydroxide. This precipitate was separated from the solution by filtration and then was heated for 4 h at 850 °C until the formation of HfO₂. XRD spectra (Miniflex 600, Rigaku Corporation, Tokyo, Japan) of obtained product were compared to the spectra of initial HfO₂ using database PDF-2.

Study of yttrium behavior on LN resin was carried out by determination of distribution coefficients using ⁹⁰Y tracer. Content of ⁹⁰Y in solutions was determined by liquid scintillation spectrometry (LS-spectrometer GreenStar, Moscow, Russia) using liquid scintillation cocktail UltimaGold (PerkinElmer Inc., Shelton, CT, USA), taking into account efficiency calibration for acid concentration.

3.3. Determination of Purification Degree of ¹⁷⁷Lu and ^177mLu Content

1 mL was taken from fractions containing purified lutetium (80 mL) to determine its hafnium and zirconium content using quadrupole mass-spectrometer with inductively coupled plasma X-series II (Thermo Fisher Scientific, Dreieich, Hessen, Germany). Remaining lutetium solution was evaporated to dryness on a round steel plate with a diameter of 2 cm to determine content of ^177mLu and purification degree of ¹⁷⁷Lu by radiometry. Activity of plate then was measured several times for the following 270 days using gamma-ray spectrometer with high-purity germanium detector GC3019 (Canberra Ind). Calibration of count efficiency depending on the energy of registered isotope was conducted using measurements of activity of certified point sources (¹⁵²Eu, ¹³⁷Cs, ⁶⁰Co, ²⁴¹Am) in different location geometries of source and detector and was also modeled in GEANT4. Identification of peak maximum in spectra was carried out using automatic system of spectrum record and analysis, specially created for this purpose. Thus, spectra with duration of 3.5 s each were saved into the database, and analysis system allowed us to summarize them and display total spectrum with assigned duration [15].

Purification degree of ¹⁷⁷Lu from macroquantities of Hf and Zr was calculated by dividing the mass of Hf or Zr in the initial solution by the mass of ones in purified lutetium solution using ICP-MS data. Purification degree of ¹⁷⁷Lu from microquantities of ⁸⁸Y was determined by gamma-ray spectrometry.

4. Conclusions

A method of recovery of carrier-free ¹⁷⁷Lu from macroquantities of hafnium and zirconium and trace amounts of yttrium was developed; the yield of lutetium was not less than 98%. Contents of hafnium, zirconium, and yttrium in the obtained solution of ¹⁷⁷Lu were at least 1.19 × 10¹⁰, 1.7 × 10⁶, and 10⁴ times lower compared to the initial solution. The achieved level of ¹⁷⁷Lu purification allowed us to determine the activity of ^177mLu during the prolonged registration of the gamma-ray spectrum, resulting in determination of the ^177mLu/¹⁷⁷Lu activities ratio that reached a value of (2.87 ± 0.07) × 10⁻⁵ for the photonuclear method at the studied energy; this ratio indicates a high purity of the obtained ¹⁷⁷Lu and the possibility of its use in nuclear medicine. The developed method was successfully applied to obtain ¹⁷⁷Lu after the irradiation of 16 g of HfO₂ in an electron accelerator. It was demonstrated that irradiated HfO₂ could be quantitatively regenerated and later be reused for the production of the medical isotope ¹⁷⁷Lu.

Thus, we demonstrate that it is possible to produce and quantitatively recover ¹⁷⁷Lu for preclinical studies using an electron accelerator. Moreover, the photonuclear production of ¹⁷⁷Lu can also become an alternative method for its obtaining, but to date, an experimental study of the yields of photoproton reactions on enriched targets made of ¹⁷⁸Hf and ¹⁷⁹Hf is required.