Introduction

The most important beta emitter-radionuclides applied in the preparation of pain palliatives for metastatic bone include samarium-153 (153Sm)1, strontium-89 (89Sr)2, phosphorus-32 (32P)3, and lutetium-177 (177Lu)4, holmium-166 (166Ho)5, and rhenium-186 (186Re)6. These radionuclides have different physical characteristics and, therefore, different levels of effectiveness in relieving the pain associated with metastatic bone (Table 1)7. Naturally, a longer half-life results in a longer duration of efficacy and pain relief, such as when 153Sm and 177Lu are combined with EDTMP. The relatively low β-energy and also relatively long half-life of 177Lu allow the deposition of an adequate tumor irradiation dose at a constant rate8,9,10. 32P and 89Sr have long half-lives, but due to their high deposited energies, cells absorb high levels of the extra dose, and the bone marrow ablation is probable. Meanwhile, radionuclides with long half-lives present waste disposal issues in hospitals where various therapeutic actions are undertaken.

Table 1 Half-life, \({E}_{\beta max}\), \({E}_{\beta ave}\), and \({E}_{\gamma }\) of the most important beta emitter-radionuclides applied in the preparation of pain palliatives for metastatic bone.
Full size table

The half-life of 143Pr is 2 weeks, which is longer than the half-life of 177Lu, 153Sm, and 186Re, while it is similar to that of 32P. It has pure beta emission with an Eβmax of 0.937 and an MeV lower than 89Sr and 32P too. Pure beta emitter 33P, also shows suitable properties such as Eβmax = 0.2 MeV and a half-life of 25 days; however its production is limited by needing high flux reactor for some countries. Thus, 143Pr is proposed here as a promising radionuclide for preparing bone metastases pain palliatives that allow for an adequate deposition of irradiation resulting in reasonable efficacy, along with a proper half-life and energy. In addition, 143Pr can be produced in a no-carrier-added form in a medium flux reactor, such as the Tehran Research Reactor (TRR). Furthermore, because it is a radiolanthanide; it can be combined with well-known bone-seeking agents such as EDTMP, DOTMP, and zoledronic acid11, which have been used to prepare bone-pain-palliatives, to generate new products. However, one drawback is the time-consuming quality control process involved in making pure beta emitter products using liquid scintillation counting with 143Pr due to a lack of gamma emission.

There are few reports on the production and application of 143Pr. The earliest study was conducted in 1957 by Peppard et al., who reported the efficient isolation of carrier-free 143Pr from neutron-bombarded cerium after applying solvent extraction using a system containing 10 M HNO3 as an aqueous phase into a 0.75-M or 0.30-M solution of di (2-ethyl hexyl) orthophosphoric acid (HDEHP) in n-heptane12. In 1963, following Peppard’s report, J. W. Winchester prepared an extractive chromatographic stationary phase using the same extractant of HDEHP held on dichlorodimethyl-silane-treated diatomaceous silica to separate rare earths. The observed separation factors for adjacent rare earths were similar to those of rare earths in a solvent extraction, uniformly close to 2.513.

In other research, Tomitaro Ishimuri et al. separated 143Pr from neutron-irradiated uranium by solvent extraction using HEDHP/HNO3 system after almost two days of cooling. Then, 143Pr was purified after most of the 143Ce had decayed14. Deptula et al. applied an extraction chromatography (HD2EHPA was used as an extractant on resin) following cation exchange chromatography (using resin 5WX4) to separate 143Pr from cerium dioxide irradiated in the EWA reactor15.

In 1976, Masumitsu Kubota utilized a cation exchange separation method using Diaion SK-1, 100–200 mesh. The chemical yield of the 143Pr was over 99%, and its radiochemical purity exceeded 99.99%16. The latest reports on the preparation of 143Pr were provided by K.V. Vimalnath, who also used an indirect production method by irradiating cerium. The radiochemical separation of 143Pr was performed by precipitation; the separated n.c.a-grade 143Pr had high radionuclidic purity (> 99.9%) and did not contain any detectable impurities17,18,19.

As mentioned earlier there are not many reports on separation of 143Pr. However, solvent extraction, chromatography and precipitation are the most widely used approaches, which are pointed above. Regarding the ease of automation, absence of any organic liquid, and also considering the final chemical/physical-form of product containing 143Pr, authors chose the chromatographic approach.

In this study, natural cerium target material is irradiated in Tehran Research Reactor. The calculated optimized time windows for irradiation and cooling the irradiated target material are applied. The well-established extraction chromatography is then applied to isolate 143Pr from cerium target material. The optimized parameters of extraction chromatography for the separation of praseodymium from the adjacent lanthanide, cerium, are investigated and determined. The well-known Ln-resin as efficient stationary phase and HNO3 as eluent are used20,21. Oxidizing properties of cerium were exploited to differentiate the chemical behavior of cerium and praseodymium22. Consequently, an efficient and convenient separation of no-carrier-added 143Pr from irradiated cerium target material is achieved. Cerium can also be recovered in + 3 oxidation state using ascorbic acid, to be re-used if the costly enriched target material of cerium is used.

Experimental procedure

Materials and instruments

Irradiation was performed at Tehran Research Reactor (TRR). The TRR reactor is a 5 MWt open pool design that uses MTR fuel configurations. It is capable of producing a maximum local thermal neutron flux of approximately 1.0 × 1014 n/cm2 sec within an irradiation box located at the reactor core's center. This irradiation box is used for various activities such as producing radioisotopes, conducting neutron activation analysis, training personnel, and irradiating samples for research purposes. The reactor core itself rests on a grid plate composed of a 9 × 6 array that forms rectangular locations for fuel elements.

Natural cerium (III) nitrate hexahydrate (99.999% trace metals basis) was purchased from Sigma-Aldrich. The extraction resin Ln (100–150 μm) was prepared from Eichrom. All other chemical reagents were in an analytical grade and they were also purchased from Merck. The quartz ampoules, glass chromatography columns, and other disposal fittings were obtained from the NSTRI’s equipment supply workshop. An ultra-low-level liquid scintillation counter (Quantulus 1220) was used for beta counting measurements, the necessary calibrations were performed using 14C standard source as an external standard. All aliquots of the 143Pr and Ce-isotope samples were mixed with 15 mL of HiSafe3, provided by Perkin-Elmer.

Gamma-ray spectroscopy was performed by a p-type coaxial high-purity germanium detector (HPGe, EGPC 80-200R) coupled with a multichannel analyzer and Gamma 2000 software (Silena). Energy and efficiency calibrations were determined by γ-ray sources traceable to the National Institute of Standards and Technology. A high-performance multichannel laboratory peristaltic pump (DYNAMAX, RP-1) was coupled with an Ln-resin-containing chromatography column and was used to apply suction to the column bed and adjust the output solution fractions volume and flow rate. MATLAB software was utilized to calculate radioactivity.

Radioactivity calculations for obtaining praseodymium-143 from the neutron irradiation of cerium target material

There are two routes for the production of 143Pr through (n,γ) reaction, Fig. 1; the direct method uses Pr-target material (red) and the indirect method uses Ce-target material. This study aims to produce non-carrier-added143Pr through the indirect method. The number of nuclei of cerium-143 and praseodymium-143 isotopes created by irradiating 1 mg of natural cerium for 24 h with a neutron flux of 5 × 1013 n.cm−2 s−1 and cooled for 72 h, was calculated using MATLAB software. For calculations, relevant data (cross-section and half-life) from The IAEA Nuclear Data Services were used23.

Figure 1
figure 1

The production of 143Pr applying the direct method using Pr-target material (red) and the indirect method using Ce-target material (green).

Full size image

The time-dependent change in nuclear densities arising from neutron activation and transmutation can be expressed using the following system of differential equations:

$$\frac{d{N}_{i}(r,t)}{dt}= \varphi \left(r,t\right)\sum_{j}{N}_{j}\left(r,t\right){\sigma }_{j\to i}\left(r\right)+\sum_{k}{N}_{k}\left(r,t\right){\lambda }_{k\to i}\left(r\right)- \varphi \left(r,t\right){N}_{i}\left(r,t\right)\sum_{l}{\sigma }_{i\to l}\left(r\right)- {N}_{i}(r,t)\sum_{n}{\lambda }_{i\to n}$$
(1)

Equation (1) constitutes a system of homogeneous first-order ordinary differential equations that describe the time evolution of nuclear densities in the irradiated cerium target.

Preparation and irradiation of the target material

A solution containing 1 mg of natural cerium nitrate target material was heated in a quartz ampoule under a nitrogen environment to dryness. The open end of each quartz tube was welded and fully closed once the solution had dried. The quartz capsules were then deposited in an aluminum cylinder chamber. Samples were irradiated in the reactor for 24 h with a flux of 5 × 1013 cm−2 s−1. After irradiation, the sample was cooled for 72 h. A 10% nitric acid solution was then added to dissolve the dry precipitate, and the fluid in the quartz capsule was drawn with a syringe and emptied into a vacuum vial. The appropriate quantities of this solution were recorded for gamma and beta spectrum analysis, depending on their activity. All sample and dilution stages were carried out in a glove box in compliance with radiation safety standards.

Sample analysis

The intended product, 143Pr, is a pure-beta emitter and the impurities emit gamma rays; therefore, the purity of the samples was determined using gamma spectroscopy and beta analysis (with liquid scintillation counting method). The quantification of beta radioactivity via liquid scintillation counting is a method employed to gauge the level of beta-emitting nuclides within a sample. This process involves utilizing a fluid scintillation medium (cocktail), which is an aqueous solution comprising a scintillating agent capable of emitting light upon intersection with ionizing radiation. When beta particles are discharged from the isotopes inside the sample and collide with molecules of the scintillating agent suspended in the medium, a flash of light is produced. This light yield is then detected and analyzed to determine the radioactive disintegrations occurring per unit time, thereby allowing calculation of the concentration of beta-emitting isotopes in the original sample. In this work, Wallac Quantulus 1220 counter and OptiPhase HiSafe 3 cocktail were used for beta analysis. Each sample was measured three times to minimize the error.

Isolation of praseodymium-143 from the irradiated target material

Extraction chromatography with an Ln-resin and an initial mass of target material containing 1 mg natural cerium was employed to separate 143Pr from the target material of cerium. This approach is a separation strategy that combines the selectivity of an extracting agent with the speed and convenience of a chromatographic column.

A 22-cm-long column with a diameter of 1 cm containing Ln-resin (150–100 μm) was utilized. The resin slurry in nitric acid entered the column uniformly after the glass wool was placed at the bottom of the pipe. Next, the resin was rinsed with 500 mL of distilled water. The sample comprising cerium and praseodymium was loaded on the column at a flow rate of 8 mL/min after a NaBrO3 oxidizing agent had been added. The washing solution, which included 0.01 M nitric acid and 0.05 M oxidizing agent, was then run through the column. This was done to guarantee a higher oxidation state of the cerium of (IV) quantitatively and to perform a column washing step. The radio-lanthanides were eluted from the column by successive changing of the nitric acid through the column as the mobile phase. The eluted fractions were collected from the column in 5-mL quantities.

Results and discussion

The calculated and measured produced activities of Ce-143 and its daughter Pr-143

Result of calculations to determine the number of nuclei of cerium-143 and praseodymium-143 isotopes (activities) created by irradiating 1 mg of natural cerium for 24 h with a neutron flux of 5 \(\times\) 1013 cm−2 s−1 and cooled for 72 h is depicted in Fig. 2. The 72 h cooling time is necessary for Ce-143 to reach equilibrium with Pr-143 and produce the highest possible amount of Pr-143 from the decay of Ce-143.

Figure 2
figure 2

The number of nuclei of 143Ce and 143Pr isotopes created by irradiating 1 mg of natural cerium for 24 h with a neutron flux of 5 × 1013 cm−2 s−1and cooled for 72 h.

Full size image

The estimated and produced activities of cerium-143 and praseodymium-14 for irradiation and cooling under the specified circumstances are provided in Table 2. Given that the discrepancy between the computational and practical findings is less than 15%, the results appear to be in excellent agreement, and the difference can be attributed to measurement uncertainty, analytical sampling error, and the uncertainty of the time and sample location inside the reactor, which is common.

Table 2 The calculated and produced activity for cerium-143 and praseodymium-143.
Full size table

The separation of 143Pr from the irradiated cerium-containing material

After 72 h, the optimized separation method was utilized for a 1-mg cerium-containing target material that had been irradiated for 24 h.

Table 2 shows the separation values, Rs, for the cerium/praseodymium pair related to elution from a column with a diameter of 1 cm and different heights of Ln-resin. The inhibition/retention factor of praseodymium (III) and cerium (III) on Ln-resin is very close to each other in different concentrations of nitric acid. Therefore, by increasing the length of the column and as a result the height of the resin has limited enhancing effect on the separation. In this way, doubling and even tripling the length of the column did not lead to a significant increase in lanthanide separation. In order to achieve an efficient separation, cerium was oxidized to (IV) oxidation state. Cerium is the only lanthanide that is stable in (IV) oxidation state in aqueous solution. For the rest of lanthanides, this oxidation state mainly exists in solid samples24. In this work, cerium was separated from praseodymium through oxidation method with NaBrO3 reagent that will result in better separation of cerium and praseodymium couple.

The distribution coefficient (Kd) values were determined as the ratio of cerium concentrations in the adsorbent and mobile phases. In order to obtain the Kd values, cerium solutions at a fixed concentration of 500 ppb in 0.15 M HNO3 was used. Various concentrations of NaBrO3, ranging from 0.001 to 0.1 M were added to each sample. Cerium was separated on 50 mg of Ln-Resin for 30 min. The concentration of cerium in the filtrate and eluted solutions were determined by ICP-MS. Table 3 shows the Kd values of cerium in the vicinity of the adsorbent (Ln-resin) as a function of NaBrO3 concentration. It can be seen that with increasing concentration of NaBrO3, Kd values related to Ce increase until the concentration reaches 0.05 M of NaBrO3, at higher concentrations, Kd slightly decreases. The increase in retention/inhibition of Ce on the adsorbent, with the increase of NaBrO3 concentration up to 0.05 M, is most likely due to the oxidation of cerium (III) to cerium (IV). Since HDEHP is a cationic extractant, at concentrations higher than 0.05 M NaBrO3 it decreases the retention/inhibition of cerium from Ce on the resin, which is probably the result of the presence of sodium cations that interact with cerium to sit on the extraction sites (Table 4).

Table 3 Resolution values, Rs, for the cerium/praseodymium pair for elution from a column with a diameter of 1 cm and different lengths of Ln-resin height.
Full size table
Table 4 Cerium Kd values as a function of HNO3 concentration.
Full size table

Subsequent column studies showed that the oxidation of cerium (III) to cerium (IV) can be performed directly on the column and the reduction of cerium (IV) to cerium (III) can be performed on the column using 0.05 M ascorbic acid. The significant increase in Kd values as a result of the addition of NaBrO3 shows a significant potential for improving Ce/Pr separation factors using this approach.

Figure 3 shows the separation of praseodymium and cerium under optimal conditions (a column with a diameter of 1 cm and containing resin with a height of 22 cm, resin with dimensions of 100 –200 μm). Separation of cerium base from praseodymium was easily achieved using NaBrO3/ascorbic acid system.

Figure 3
figure 3

Pr/Ce separation in optimal conditions, a column with a diameter of 1 cm and containing resin with a height of 22 cm, resin with dimensions of 100 to 200 μm.

Full size image

The eluted samples from the column were submitted for nuclear analysis. The results of this analysis are shown in the next section.

Quality control: radiochemical and radionuclidic purity

The product, 143Pr, does not emit any gamma-rays, while the impurities, including 141Ce and 143Ce, do. Thus, during the quality control studies, all eluted fractions were analyzed using gamma spectroscopy and beta scintillation methods.

In the first step, gamma spectrometry was used to count all eluted fractions at various stages of separation in equal volumes in containers with similar geometries. Although the presence or absence of 143Pr in a sample could not be confirmed, the quantities of the cerium-isotopes of 141Ce and 143Ce in each sample could be determined, as they have different gamma-ray energies (Fig. 4).

Figure 4
figure 4

Gamma spectra of the solution containing irradiated cerium target material,

Full size image

In the second step, a portion of each sample was extracted and analyzed using the liquid scintillation method. The radioactive sample was delivered in a special 20-ml vial with 10 ml double distilled water. Then, 10 ml of cocktail liquid (the scintillator) was added. The liquid scintillation counting (LSC) is the most accurate method for measuring pure-beta-emitter content. It can be used to count any radioisotope, but it is expensive and requires specific counting solutions to be prepared. Because the beta beam, unlike the gamma-ray, has a continuous spectrum, it is impossible to detect separate beta radionuclides in a sample; in other words, this instrument measures the total beta. Since the samples might have included cerium isotopes (emission of beta plus gamma-ray) and 143Pr (emission of pure beta with no gamma-ray), the sample counting is not correct with LSC alone, so the sample was counted with both gamma spectrometry and LSC.

Cerium isotopes or 143Pr are involved in the measured activity. However, because the presence of cerium in the samples could be determined using gamma spectrometry, the cerium isotopes in the samples were measured first, and then the total beta was determined to measure the 143Pr. The gamma counts of each fraction were measured using gamma spectrometry, and the beta counts of each fraction were measured using the liquid scintillation counter (Table 5).

Table 5 Activity concentrations of the eluted fractions.
Full size table

Figure 5 also shows the resultant chromatogram of non-carrier-added 143Pr separation from the irradiated target material of natural cerium.

Figure 5
figure 5

The chromatogram obtained from non-carrier-added 143Pr separation from the irradiated target material of natural cerium.

Full size image

Results obtained from separation of n.c.a.143Pr from the irradiated target material of natural cerium show the radionuclidic purity of the isolated 143Pr which are greater than 99%.

Thin layer chromatography was applied using Whatman paper and a 0.1 M DTPA solution to assess radiochemical purity. One beta count peak appeared at Rf = 0.9. It was concluded that the product contained only one radionuclidic species, namely 143Pr salt (nitrate), with a purity of above 99%. A sample of 143Pr was passed through a 2-ml cartridge containing Ln-resin, and 143Pr was then eluted with 0.35 M nitric acid to determine its chemical purity.

After cooling, an elemental analysis was conducted to analyze the 143Pr-nitrate salt sample after five to seven half-lives, or about three months (CHN and ICP). ICP/CHN, or Inductively Coupled Plasma/Carbon Hydrogen Nitrogen analysis, is an analytical technique used to determine the concentration of elements in a sample. The sample is first injected into the instrument. The sample solutions are then nebulized with argon gas and injected into a plasma torch, where the atoms and ions are excited to emit electromagnetic radiation at wavelengths characteristic of specific elements. The intensities of these emitted wavelengths are measured and can be correlated to the original concentration of elements in the samples. In this method, no cerium was observed in the sample with the sensitivity of ICP by analyzing the solution. The chemical purity was also over 99%. The primary sources of chemical contamination in the isolation of radionuclides during the chromatographic process were polymer resins, their detached fragments, and the extracting agents attached to them. Fortunately, these contaminants were not observed in the final product of our research.

Conclusion

In this study, we developed and optimized a process for producing high purity non-carrier-added 143Pr via neutron activation of natural cerium oxide and subsequent chromatographic purification. Calculations assisted in determining the irradiation and cooling parameters needed to indirectly produce 143Pr through decay of 143Ce. Extraction chromatography using Ln-resin efficiently isolated 143Pr from the irradiated target material with radionuclidic purity exceeding 99%.

Additional quality control analyses confirmed the radiochemical and chemical purity of the purified 143Pr nitrate product met regulatory standards for potential medical applications. With its half-life of 13.6 days and beta energy of 0.937 MeV, 143Pr shows promising properties for palliative bone pain therapy.

While further preclinical investigations are still needed, the convenient production method presented here establishes a viable path forward for exploring potential clinical uses of 143Pr. Optimizing target material and irradiation parameters could help improve yield. Investigations into complexing 143Pr with bone-seeking agents could yield radiopharmaceuticals with optimized pharmacokinetics.

This work demonstrates the feasibility of utilizing 143Pr as a pain palliative through an optimized production and purification process. Continued research has the potential to translate this radioisotope into clinically valuable targeted radiotherapies for the palliation of bone metastasis pain.

Scaling up the production method described in the study to provide the required amounts of n.c.a. 143Pr for nuclear medical applications would involve optimizing the irradiation parameters, such as target material mass and irradiation time, to increase the yield. The purification process, including extraction chromatography and subsequent purification steps, would need to be adapted for larger volumes, while maintaining high purity. Quality control analysis should be conducted to ensure the chemical and radiochemical purity of the scaled-up production. altogether, with careful adjustments and optimizations, it is feasible to scale up the production of high-purity 143Pr to meet the demands of nuclear medical applications, while considering regulatory and practical considerations.