Reactor production of no-carrier-added ¹⁹⁹Au for biomedical applications

Abstract

This investigation described development of a technology for the reactor production of no-carrier-added (NCA) ¹⁹⁹Au through neutron activation of natural Pt in a research reactor followed by chemical separation of ¹⁹⁹Au employing liquid–liquid extraction (LLX) technique using ethyl acetate. A thorough optimization of experimental parameters was carried out primarily to arrive at the separation conditions resulting in quantitative extraction of ¹⁹⁹Au into the organic phase. The reported procedure has been found to be prolific in providing ¹⁹⁹Au with >95% yields with acceptable radionuclidic and radiochemical purity. An optimized radiochemical processing procedure for obtaining NCA ¹⁹⁹Au of requisite purity in high yields suitable for clinical use has been the positive outcome.

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Introduction

Gold-199 [T_1/2 = 3.14 d, E_βmax = 244 keV (21.5%), 294 keV (72%) and 452 keV (6.5%), E_γ = 158.4 keV (36.9%), 208.2 keV (8.4%)] is a promising therapeutic radionuclide that has attracted considerable attention by virtue of its favorable chemical and nuclear decay characteristics.^1–3 The 294 keV β⁻ particle emitted by ¹⁹⁹Au has a maximum tissue penetration range of 0.8 mm and its average β⁻ particle energy, 84 keV, has a mean free path in tissue of 100 μm, making this radionuclide ideal for delivering energy to small volumes, including micro-metastatic disease, and tumor cells near the surface of cavities and is therefore well suited for radioimmunotherapy.⁴ The emission of accompanying gamma photons of 158.4 keV of 36.9% yield offers the scope for imaging the biodistribution and excretion kinetics with the same radiolabeled preparation used for therapy, and allows dosimetry to be performed before and during treatment as well. The 3.14 day half-life of ¹⁹⁹Au offers extended time periods to radiolabel and purify ¹⁹⁹Au labeled radiopharmaceuticals, provides logistical advantage for shipment to sites distant from the production facility in a clinically relevant time frame, and is compatible with the temporal accumulation of antibodies in tumors.^4,5 Interest in the use of ¹⁹⁹Au was spun primarily due to its ability to form clusters of 11 gold atoms, which could then attach to monoclonal antibodies that selectively recognizes and binds on target cells.^5–9 Gold nanoparticles labeled with ¹⁹⁹Au offers the scope of using as a probe for single photon emission computed tomography (SPECT) imaging to study the biodistribution and clearance of agents as well as to assess the dosimetry and maximum tolerated dose (MTD) of therapeutic gold agents.^10,11

While the radionuclide ¹⁹⁹Au harbor significant therapeutic potential^4–19 and is poised to take radionuclide therapy to a leap forward, cost effective availability of desired quantities and qualities represents the key determinant that underpin its survival, strength and success. In view of this, acquiring local production capability of ¹⁹⁹Au of desired quantities and qualities seemed not only an attractive prospect to meet the local demands but also to empower future development and thus pursued. In this premise, it is imperative to consider all possible ¹⁹⁹Au production options conscientiously. This radionuclide can be produced either using the reactor^{20–27,38,39} or accelerator path.^28–33 In the quest for a viable ¹⁹⁹Au production route, our attention was turned towards reactor path. Both the “direct” and “indirect” reactor production routes can be followed to obtain ¹⁹⁹Au. While reactor production following double neutron capture on ¹⁹⁷Au [¹⁹⁷Au(n,γ)¹⁹⁸Au(n,γ)¹⁹⁹Au] is the least intricate route and generate negligible amount of radioactive waste, the resulting product is a mixture of ¹⁹⁸Au and ¹⁹⁹Au with low specific activity. In light of the perceived need to use ¹⁹⁹Au for radioimmunotherapy (RIT), use of NCA ¹⁹⁹Au constitutes a necessity owing to the limited concentrations of antigen expressed on the tumor cell surface. In this premise, development of an indigenous indirect reactor production route consisting of is not only an interesting prospect owing to its ability to provide NCA ¹⁹⁹Au, but is viewed as a necessity.

With a view to separate ¹⁹⁹Au from the neutron irradiated ¹⁹⁸Pt target, two chemical separation strategies including liquid–liquid extraction (LLX) and ion-exchange separation method are widely followed.^5,27,34–41 In the quest for a viable method to undertake separation of ¹⁹⁹Au from bulk of Pt, our attention was turned towards the use of LLX method owing to operational simplicity, versatility, ready adaptability for remote operation and flexibility to scaling up or down to its level of operation in response to requirements. While the LLX method obviously hold promise as an attractive approach, selection of an extractant to isolate ¹⁹⁹Au of requisite purity represents a key determinant in ensuring successful implementation of this strategy.

A general survey of the literature reveals that the extractants 1-phenyl-3-methyl-4-trifluoroacetyl-pyrazolone-5,³⁵ di-(2-ethylhexyl)phosphoric acid (HDEHP)³⁸ and trioctylamine (TOA)³⁹ have been exploited for the isolation of ¹⁹⁹Au from bulk of Pt. We have chosen ethyl acetate as the extractant owing to its ready availability of requisite purity at low cost, relatively non-toxic, non-hygroscopic and volatile. The prospect of using ethyl acetate seemed appealing as it offers the scope of recovering ¹⁹⁹Au by subsequent removal of the extractant by distillation. In order to tap the potential of ethyl acetate in the isolation of ¹⁹⁹Au from bulk of Pt, a careful scrutiny of the distribution ratios of ¹⁹⁹Au and Pt was hence considered worthwhile investigating to arrive at the optimum separation conditions and thus pursued diligently.

Herein, we report a systematic approach consisting of a number of sequential steps that contributed to the development of an indigenous processing method for the production of NCA ¹⁹⁹Au of requisite purity and yield for clinical applications. Since the inherent success of a ¹⁹⁹Au production scheme requires a thorough knowledge of all the pertinent key factors, the issues underlying neutron irradiation parameters, process equipment design and radiochemical separation procedure optimization to isolate NCA ¹⁹⁹Au were adequately addressed. We described the overall process, an overview of our experience and quality evaluation of ¹⁹⁹Au.

Experimental

Materials

Platinum sponge targets (natural isotopic composition) spec pure grade procured from by M/s. Roviur International, Singapore was used as target for neutron irradiation in the reactor. Ethyl acetate was procured from Sigma-Aldrich India, Mumbai. All chemicals and reagents used in this work were of analytical grade and were procured from Merck India Ltd, Mumbai, India or BDH Industries Limited, Mumbai, India. Paper chromatography (PC) strips were purchased from Whatman, UK. The glass equipments required for the radiochemical separations were designed and fabricated in-house.

Equipments

A high resolution γ-ray spectrometer consisting of HPGe detector coupled to a PC-based 4 K channel analyzer (MCA) obtained from Eurysis Measures, France and Aptec analysing software was used for γ-ray spectroscopic measurements. All the nuclear data used were taken from the publications Table of Isotopes,⁴² Gamma Ray Catalogue⁴³ and Nuclear Wallet Card.⁴⁴ A well type NaI(Tl) scintillation counter of Electronic Enterprises Pvt Ltd, Hyderabad, India was used to measure the activity of chromatography paper strip during radiochemical purity assay. Distribution ratios (D) of Pt and Au at different pH were determined following standard radiometric procedures. Nanalysis' NMReady-60e all-in-one benchtop NMR spectrometer with an operation frequency of 60 MHz (1.4 T) was used to record NMR spectra of ethyl acetate.

Chemical processing of the irradiated target was carried out in a leak tight isotope processing facility of the dimension 1.8 × 1.8 × 0.9 m³ with 100 mm lead shielding wall all-around and equipped with remote handling tongs, lead viewing windows and other associated accessories.

Neutron irradiation of Pt target

The irradiation container (Fig. 1) essentially consists of a 1S aluminum cylinder of 22 mm (ϕ) × 44.5 mm (h) and a lid. About 100 mg of platinum sponge was accurately weighed, encapsulated in a quartz ampoule, wrapped in aluminum foil, placed in an irradiation container, sealed by cold-pressure-weld and irradiated in the Dhruva research reactor of our institute at a flux of 8 × 10¹³ n cm⁻² s⁻¹ for 7 days.

Fig. 1 Irradiation container for neutron irradiation (a) schematic drawing (b) photograph.

Determination of distribution ratios of Au³⁺ and Pt⁴⁺ ions in ethyl acetate

With a view to arrive at the optimum separation conditions, distribution ratios (D) of Au³⁺ and Pt⁴⁺ ions in ethyl acetate were determined at different HCl concentration following radiometric technique using ¹⁹⁹Au and ¹⁹⁷Pt radiotracers. In this procedure, 1 mL of aqueous phase containing varying concentrations of HCl (1–10 M) spiked with either ¹⁹⁹Au or ¹⁹⁷Pt were equilibrated with equal volume of ethyl acetate in stoppered glass tubes in a thermostated water bath (25 ± 0.1 °C) for 25 min.

The two phases were separated, then centrifuged, and 0.1 mL aliquots were taken for radio metric analysis by gamma counting using a well type NaI(Tl) scintillation counter. The distribution ratio (D) was calculated using the expression:

where, A_org and A_aq is the total radioactivity of either ¹⁹⁹Au or ¹⁹⁷Pt. Each experiment was carried out at 25 °C, in triplicate and the average was taken.

Radiochemical processing and separation of ¹⁹⁹Au

The glass ware set up used for the processing of irradiated target is shown in Fig. 2. The irradiated can containing platinum metal target was cut open using a semi-automatic irradiation container opening unit, quartz ampoule containing the irradiated target was cut using a mechanized device, and the radioactive contents were transferred into a 3-necked round bottom flask. 4 mL aqua regia was introduced into the round bottom flask and the target was dissolved by gentle warming to ensure complete dissolution.

Fig. 2 Experimental set up for production of ¹⁹⁹Au.

The solution obtained was evaporated to incipient dryness and then the residue was treated with 5 mL of conc. HCl and heated till evolution of brown fumes ceases. The procedure was repeated thrice to ensure complete removal of HNO₃ and the residue left was reconstituted with 10 mL of 5 M HCl. A measured aliquot from this stock was withdrawn and subjected to high resolution gamma ray spectrometry to assess both total ¹⁹⁹Au activity produced, and co-produced radionuclide impurities.

The solution was then transferred to a 50 mL separating funnel connected in series with the dissolution flask by applying vacuum. 10 mL ethyl acetate was introduced into the main separating funnel through a side neck. Dry filtered compressed gas was passed through a glass tube from the top of the separating funnel to agitate the solution mixture and to provide a shaking action. Extraction was continued (three times) until nearly all ¹⁹⁹Au was extracted into organic phase.

The organic phases (ethyl acetate) were pooled together and heated to remove ethyl acetate by distillation, which was collected and disposed off as radioactive waste. The residue in the flask was treated with 5 mL concentrated HNO₃ to destroy any organic matter and evaporated to incipient dryness, reconstituted in 5 mL of 0.1 M HCl gently heated to leach ¹⁹⁹Au activity and subsequently withdrawn from the processing flask through a sterile PVC tube with leur ends under suction, passed through a 0.22 μ millipore filter and collected in a sterile vial.

Quality control of ¹⁹⁹Au

Activity measurement. The radioactivity content was measured by both a pre-calibrated ionization chamber and by gamma ray spectrometry after appropriate dilution and sampling.

For radioactivity assay in ionization chamber, the sample in a sealed glass vial was remotely placed at the centre of sensitive volume of ionization chamber and ionization current produced in sensitive zone of instrument was recorded and a conversion factor for ¹⁹⁹Au was applied to obtain the value of radioactivity in terms of MBq. The ion current ‘I’ was measured and the radioactivity ‘A’ was calculated by the following relation:

where, ‘ε’ in pA MBq⁻¹ is the conversion factor of ionization chamber for ¹⁹⁹Au.

Energy as well as efficiency calibration of the HPGe detector were carried out using a ¹⁵²Eu reference source prior to the recording of gamma ray spectra of sample. Appropriately diluted aliquots of the processed H[¹⁹⁹AuCl₄] solution were measured for 1000 seconds. The activity of ¹⁹⁹Au in the aliquot was ascertained utilizing the 158.4 keV (36.9%) and 208.2 keV (8.4%) photopeak of ¹⁹⁹Au. The absolute activity of the ¹⁹⁹Au was calculated from the measured count rate and dividing it by the average efficiency.

Radionuclidic purity. Radionuclidic purity was also determined using the gamma spectrometric technique. Trace level of long lived radioactive impurities were determined by recording the gamma ray spectra of the sample, after complete decay of ¹⁹⁹Au activity (9–10 T_1/2 of ¹⁹⁹Au, i.e. for a period of 30 d).

Radiochemical purity. The radiochemical purity of ¹⁹⁹Au was determined by paper chromatography following the reported procedure.⁴⁶

Chemical purity. In order to determine the presence of trace element impurities chemical impurities such as Pt, Ru, Rh, Pd, In, and Os in the H[¹⁹⁹AuCl₄] solution, the samples were allowed to decay for 60 days and subsequently analyzed for each of the trace element impurities by ICP-AES technique. The concentrations are reported as μg of the metal present per GBq of ¹⁹⁹Au.

Results

Target

While platinum metal remain as the obvious choice for target owing to high melting point, good thermal conductivity and stability under neutron irradiation in nuclear reactors, a holistic consideration should be given whether to use natural Pt or isotopically enriched ¹⁹⁸Pt.

A distinct advantage of using isotopically enriched ¹⁹⁸Pt (∼100%) target is the possibility to produce ¹⁹⁹Au with very high radionuclidic purity and offer the prospect of augmenting the yield without significantly altering the radiochemical processing set up. When compared with natural Pt target, use of isotopically enriched ¹⁹⁸Pt requires smaller amount of target for neutron irradiation to produce desired amounts of the activity, reduces radiation dose burden during radiochemical processing, generates minor amount of radioactive waste and offers the scope of using low flux reactor. Although the prospect of using isotopically enriched ¹⁹⁸Pt is relatively more appealing, the need to recover the expensive target in micro- to milli-gram level for recycling is challenging. In view of this consideration, assessing the potential of natural Pt as target is not only an interesting proposition, but deemed worthy of consideration. In this premise, it is imperative to assess the consequence of using natural Pt.

When natural Pt undergoes irradiation in a reactor, it is crucial to consider all the radionuclides produced by thermal neutrons along with their thermal neutron absorption cross section.⁴⁵ Table 1 depicts the nuclear reactions taking place when natural Pt undergoes irradiation with thermal neutrons in a nuclear reactor. It is seen from the Table 1 that among all the possible radionuclides formed by natural Pt neutron irradiation, ¹⁹¹Pt, ¹⁹³Pt, ^193mPt, ^195mPt, ¹⁹⁷Pt and ^197mPt are the most prominent. Concomitant production of these radionuclides not only complicate the irradiated target handling owing to augmentation of radiation dose, but also generate radioactive waste. Production of Pt radionuclides other than ¹⁹⁹Pt during neutron irradiation, are of least concern as they are unlikely to reduce the specific activity of ¹⁹⁹Au because they remain with cold Pt during chemical separation. In view of these considerations, prospect of using natural Pt with some acceptable restrictions is not only an interesting prospect, but deemed worthy of consideration and motivated us to pursue.

Table 1 Isotopic composition of platinum and activation data

Stable isotope	Isotopic abundance (%)	Nuclear reaction	Reaction cross section (b)	Half life
^190Pt	0.012	^190Pt(n,γ)^191Pt	152	2.83 d
^192Pt	0.782	^192Pt(n,γ)^193Pt	10	50 year
		^192Pt(n,γ)^193mPt	2.2	4.33 d
^194Pt	32.86	^194Pt(n,γ)^195Pt	1.44	Stable
		^194Pt(n,γ)^195mPt	0.036	4.01 d
^195Pt	33.78	^195Pt(n,γ)^196Pt	27.5	Stable
^196Pt	25.21	^196Pt(n,γ)^197Pt	0.58	19.89 h
		^196Pt(n,γ)^197mPt	0.044	95.41 min
^198Pt	7.36	^198Pt(n,γ)^199Pt	3.61	30.8 min
		^198Pt(n,γ)^199mPt	0.35	13.6 s

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Natural Pt not only co-produces ¹⁹¹Pt, ¹⁹³Pt, ^193mPt, ^195mPt, ¹⁹⁷Pt and ^197mPt due to competing activation reactions but also other radionuclide impurities such as ¹⁹⁸Au, ¹⁹²Ir and ¹⁹⁴Ir as a result of neutron activation of some of their decay products. Radionuclides that are likely to be produced during the neutron irradiation of natural Pt are shown in Table 2.

Table 2 Radionuclides that are likely to be produced during neutron irradiation of natural Pt and their method of formation

Radionuclide	Half life	Gamma ray energy		Method of formation of radionuclides
		keV	% Iγ	Nuclear reaction	% natural abundance	Reaction cross section σth (barns)
^199Au	3.139 d	158.4	36.9		7.36	3.61
		208.2	8.38
^198Au	2.695 d	411.8	95.5		25.21	0.58
						98.65
^192Ir	73.829 d	295.9	28.7		0.012	152
		308.4	29.7
		316.5	82.9			954
		468.1	48.1
^194Ir	19.28 h	293.5	2.54		0.782	10
		328.4	13.0			111
^191Pt	2.83 d	538.9	13.7	^19078Pt(n,γ)^19178Pt	0.012	152
		409.4	8
^193mPt	4.33 d	66.8	7.38	^19278Pt(n,γ)^193m78Pt	0.782	2.2
		75.7	2.53
		135.5	0.11
^195mPt	4.01 d	98.9	11.4	^19478Pt(n,γ)^195m78Pt	32.86	0.036
		129.7	2.83
^197Pt	19.891 h	77.3	17.2	^19678Pt(n,γ)^19778Pt	25.21	0.58
		191.4	3.68
^197mPt	95.41 min	66.8	23.8	^19678Pt(n,γ)^197m78Pt	25.21	0.044
		364.5	11.2

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Optimization of irradiation parameters

With an aim to produce ¹⁹⁹Au of required quality and quantity, the irradiation time and neutron flux were judiciously optimized.

The ¹⁹⁹Au activity produced at the end of neutron irradiation as a function of irradiation time for different thermal neutron fluxes has been calculated and the results are shown in Fig. 3. There appears to be enticing interest to consider irradiation of Pt at higher thermal neutron flux and shorter irradiation period as it will suffice to obtain requisite ¹⁹⁹Au yield.

Fig. 3 Gold-199 (¹⁹⁹Au) activity produced (MBq) per mg platinum target as a function of the irradiation time at different thermal neutron flux of research reactor.

Owing to the inherent limitation on the availability of limited number of irradiation position at highest neutron flux 1.8 × 10¹⁴ n cm⁻² s⁻¹ in a multi-purpose research reactor and their exclusive utilization for routine production of ⁹⁹Mo and ¹⁷⁷Lu etc., it was not possible to perform irradiation. Consequently, our subsequent efforts were directed towards exploring the possibility to produce ¹⁹⁹Au at a neutron flux of 8 × 10¹³ n cm⁻² s⁻¹. Typical calculated yields of ¹⁹⁹Au from natural and enriched targets for different durations of irradiation in Dhruva reactor at a neutron flux of 8 × 10¹³ n cm⁻² s⁻¹ are shown in Fig. 4.

Fig. 4 Typical calculated yields of ¹⁹⁹Au produced from natural and enriched target of Pt for different irradiation time at a thermal neutron flux of 8 × 10¹³ n cm⁻² s⁻¹.

As expected, the yield of ¹⁹⁹Au in natural target increases marginally with increasing irradiation time. In the case of the 95% enriched ¹⁹⁸Pt target, increase in irradiation time not only increase the yield significantly, but also significantly higher than those obtained from natural Pt targets under similar irradiation conditions.

The irradiation time of 7 d was deemed optimum from the prospective of attaining adequate radioactivity yield of ¹⁹⁹Au and inhibits built up of longer isotopes of Pt (e.g. ¹⁹³Pt) and radioactive impurities. In order to avail GBq amounts of ¹⁹⁹Au per batch, irradiation of 100 mg nat. Pt target is necessary. It is pertinent to point out that 2.7 mg 95% enriched target of Pt (I.E ¹⁹⁸Pt) under similar conditions of irradiation would be able to provide the same quantity of ¹⁹⁹Au.

Target processing

In view of the need to dissolve the neutron irradiated platinum quantitatively for subsequent radiochemical separation, the scope of using aqua-regia seemed attractive.

The initial chemical reactions of Pt with aqua-regia, led to the formation of mixture of chloroplatinous acid (H₂PtCl₄) and nitrosoplatinic chloride [(NO)₂PtCl₄] in solution.

2Pt_(s) + 2HNO_3(aq) + 8HCl_(aq) → (NO)₂PtCl_4(s) + H₂PtCl_4(aq) + 4H₂O_(l)

When this irradiated Pt was treated with aqua-regia, it steadily dissolves, but left a solid residue due to formation of nitrosoplatinic chloride, which subsequently dissolved with time due to the formation of chloroplatinous acid.

(NO)₂PtCl_4(s) + 2HCl_(aq) → H₂PtCl_4(aq) + 2NOCl_(g)

The chloroplatinous acid (H₂PtCl₄) was oxidized to chloroplatinic acid with chlorine while heating.

H₂PtCl_4(aq) + Cl_2(g) → H₂PtCl_6(aq)

The overall equation of Pt dissolution can therefore be written as

2Pt_(s) + 4HNO_3(aq) + 12HCl_(aq) + Cl_2(g) → 2H₂PtCl_6(aq) + 6H₂O_(l) + O_2(g) + 4NOCl_(g)

The corresponding equation for Au is

Au_(s) + 3HNO_3(aq) + 4HCl_(aq) → H[AuCl₄]_(aq) + NO_2(g) + 3H₂O_(l)

Excess of residual nitric acid was removed by repeated treatment with conc. HCl and heating.

The gamma spectrum of the solution as shown in Fig. 5 shows gamma peaks pertaining to only Pt and Au radionuclides, which indicated that the target used for ¹⁹⁹Au production was free from other metallic impurities. The gamma spectrum further revealed that radionuclides formed apart from ¹⁹⁹Au were ¹⁹¹Pt, ^193mPt, ^195mPt and ¹⁹⁷Pt. It was further observed that ¹⁹²Pt, ^197mPt, ¹⁹⁸Au, ¹⁹²Ir and ¹⁹⁴Ir which are expected to be formed during neutron irradiation were below detectable levels. Their absence is primarily attributed due to combination of multiple factors such as low abundance of corresponding target nuclide, low neutron absorption cross section, short irradiation time and long half-life of the product radionuclide etc.

Fig. 5 Gamma ray spectrum of the feed solution before radiochemical separation.

Following the aforementioned procedure, several batches of ¹⁹⁹Au were produced by irradiating 20 mg nat. Pt target for 7 d at various neutron flux in Dhruva research reactor, activity of ¹⁹⁹Au produced were measured following gamma spectrometric technique and the result obtained are depicted in Table 3. As expected, the activity produced increases with increasing the neutron flux. Activity measurement from gamma spectrometric technique revealed that in a typical batch, about 290–936 MBq (7.8–25.3 mCi) of ¹⁹⁹Au could be produced by irradiating 20 mg platinum target.

Table 3 Production yields of ¹⁹⁹Au at various neutron flux

Batch^a	Thermal neutron flux n cm^−2 s^−1	Activity produced^b MBq (mCi)
a Average of five batches.b Activity corrected to EOI-end of irradiation.
1	2.3 ± 0.2 × 10^13	288 ± 7 (7.8 ± 0.2)
2	3.4 ± 0.2 × 10^13	407 ± 74 (11 ± 2)
3	4.2 ± 0.3 × 10^13	540 ± 7 (14.6 ± 0.2)
4	5.6 ± 0.2 × 10^13	740 ± 111 (20 ± 3)
5	7.5 ± 0.1 × 10^13	936 ± 11 (25.3 ± 0.3)

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Radiochemical separation of ¹⁹⁹Au from neutron irradiated Pt

Following the aforementioned procedure, D values of Au and Pt ions were determined at different HCl concentrations and the results are summarized in Table 4.

Table 4 Distribution ratio (D) of ¹⁹⁹Au and ¹⁹⁷Pt

HCl concentration (M)	D
	^199Au	^197Pt
1	5.3 ± 0.1	0.06 ± 0.01
2	7.0 ± 0.1	0.06 ± 0.01
3	10.3 ± 0.2	0.07 ± 0.02
4	11.6 ± 0.4	0.05 ± 0.01
5	20.0 ± 0.5	0.07 ± 0.02
6	7.7 ± 0.6	0.08 ± 0.01
7	6.1 ± 0.8	0.01 ± 0.01
8	6.1 ± 0.1	0.08 ± 0.02
9	4.2 ± 0.3	0.07 ± 0.01
10	1.2 ± 0.1	0.11 ± 0.03

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A careful scrutiny of the D values of Au indicates that it increases with increasing acidity and attains a maximum at 5 M HCl and thereafter decreases. Pt has negligible D values in ethyl acetate at all the acidity of HCl studied. The most striking feature of Au extraction in ethyl acetate is its extremely high D values compared with that of Pt in all acidity studied. From the perspective of isolation ¹⁹⁹Au, the excellent distribution ratio (D) values for ¹⁹⁹Au over Pt at 5 M HCl is particularly heartening and conducive for the selective recovery of ¹⁹⁹Au. This assumption has been simplified and assiduously exploited in our work.

While the distribution ratio values are beneficial to achieve clean separation, it is obligatory to evaluate the behavior of ethyl acetate in the presence of intense radiation environment with the radiolytic products generated as a result of radioactive isotopes of Pt and ¹⁹⁹Au. In view of these considerations, validation of the separation procedure was carried out using 20 mg of irradiated Pt target following the procedure in the Experimental section.

While the use of single extraction of ¹⁹⁹Au had tangible benefits, but following multiple extractions (3 times) was considered a trustworthy proposition. Experimentally, it was seen that it was possible to extract >75% of ¹⁹⁹Au in a single extraction, where as >95% yield was achievable in 3 extractions. A reproducible separation efficiency of 94–97% was found to be achieved in 3 extractions. With an aim to verify the efficacy of the 3 step extraction procedure, gamma spectrum of an aliquot withdrawn from the aqueous fraction was taken as shown in Fig. 6. The gamma spectrum obtained shows photo-peaks pertaining to Pt radionuclides only. Absence of photo-peak corresponding to ¹⁹⁹Au in the aqueous phase confirms quantitative transfer of ¹⁹⁹Au into the organic phase.

Fig. 6 Gamma ray spectrum of the feed solution after radiochemical separation of ¹⁹⁹Au.

Having successfully completed the feasibility demonstration studies, scaling-up of the optimized radiochemical separation procedure was then undertaken by gradual increase of the weight of the target and it was observed that with increasing the batch size, no major problems in term of reproducibility of purity or yield of ¹⁹⁹Au were encountered during the production process.

The aforementioned developed radiochemical separation process was repeated in 5 different batches using 100 mg of neutron irradiated Pt target irradiated for 7 d at 8 × 10¹³ n cm⁻² s⁻¹ and results obtained is shown in Table 5.

Table 5 Recovery yield of ¹⁹⁹Au in different batches

Batch no.	Activity before chemical separation (GBq)	Activity recovered (GBq)	Recovery yield (%)
1	5.55 ± 0.25	5.16 ± 0.21	92.97
2	5.72 ± 0.22	5.49 ± 0.26	95.97
3	5.63 ± 0.28	5.41 ± 0.21	96.09
4	4.81 ± 0.23	4.63 ± 0.25	96.25
5	5.75 ± 0.19	5.56 ± 0.27	96.69

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Experimental results demonstrated that it was possible to produce about 5.6 GBq ¹⁹⁹Au per batch from 100 mg of irradiated Pt using the optimized procedure.

Quality control of ¹⁹⁹Au

Activity measurement. Experimentally, it was found that the maximum observed percentage differences between the activity measurement values obtained by gamma spectrometric and ionization current measurement techniques lies within ±5%. The activity of the ¹⁹⁹Au solution as determined using well type ionization chamber was therefore adopted for routine measurements of ¹⁹⁹Au activity owing to operational simplicity.

Radionuclidic purity. The γ-ray spectra of the ¹⁹⁹Au obtained after chemical separation, is shown in Fig. 7 which showed no traces of extraneous photo peaks except those due to that of ¹⁹⁹Au, thereby confirming the absence of gamma emitting radioactive impurities. The corresponding radionuclidic purity was found to be ≥99.99% ¹⁹⁹Au.

Fig. 7 Gamma spectra of separated ¹⁹⁹Au–HAuCl₄ solution.

The gamma spectrum of the decayed sample did not show any photo peaks thereby confirmed the absence of any long lived extraneous radionuclide in the product.

Radiochemical purity. The radiochemical purity of ¹⁹⁹Au in chloride form as determined by paper chromatography is shown in Fig. 8, which indicate >95% purity.

Fig. 8 Paper chromatographic pattern of ¹⁹⁹Au–HAuCl₄ using saline in 0.02 M HCl as the eluting solvent.

Chemical purity. The level of metallic ions present as chemical impurity in the decayed ¹⁹⁹Au samples as determined by ICP-AES was below detectable limits.

Gold-199 (¹⁹⁹Au) product specification

Chemical form the product obtained was HAuCl₄ in 0.1 M HCl, appearance was a clear solution, and the radioactive concentration was 370–740 MBq (10–20 mCi) per mL. Radionuclidic purity attained was >99.9. Radiochemical purity of the ¹⁹⁹Au obtained was greater than 99% in all the lots analyzed.

Radioactive waste management

The yields of various coproduced Pt radionuclides that constitute the major radionuclide impurities are given in Table 6. Although, concomitant generation of radioactive waste containing medium energy gamma emitting radionuclides is a matter of concern from the prospective of waste management point of view, but surmountable owing to their relatively short half-lives. Aqueous radioactive waste were collected in a labeled container kept in the shielded enclosure of radiochemical processing facility and allowed to decay for 30 days and disposed.

Table 6 Observed yields of ¹⁹⁹Au and *Pt radionuclides by neutron activation of 20 mg nat. Pt target irradiated for 7 d at 2.3 × 10¹³ n cm⁻² s⁻¹

Quantity of Pt irradiated (mg)	Radio nuclide	Activity at EOI MBq (mCi)	Activity per mg Pt at EOI MBq (mCi)
20	^199Au	288 ± 7 (7.8 ± 0.2)	14.4 (0.39)
	^191Pt	11.8 ± 0.4 (0.32 ± 0.01)	0.59 (0.016)
	^195mPt	14.4 ± 0.7 (0.39 ± 0.02)	0.72 (0.019)
	^197Pt	331.5 ± 19.9 (8.96 ± 0.54)	16.6 (0.448)
100	^199Au	1554 ± 185 (42 ± 5)	15.5 (0.42)
	^191Pt	66.6 ± 7.4 (1.8 ± 0.2)	0.66 (0.018)
	^195mPt	77.7 ± 7.4 (2.1 ± 0.2)	0.77 (0.021)
	^197Pt	1676.1 ± 88.8 (45.3 ± 2.4)	16.7 (0.453)

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Discussion

While both the reactor and accelerator production ¹⁹⁹Au routes are pursued, either of them is not pitted against each other, but instead, capable of providing clinical grade NCA ¹⁹⁹Au. Each institution needs to assess options, weigh pros and cons and select the one based on the available technical and economic resources to suit its needs. Our pursuit to develop a method for the routine production of ¹⁹⁹Au is driven due to the availability of operating research reactor of neutron flux of the order 10¹⁴ n cm⁻¹ s⁻¹ as well as expertise and shielded radiochemical facilities to undertake chemical processing of neutron irradiated target in our institution.

The prospect of using a volatile extractant for the selective extraction of ¹⁹⁹Au seemed sagacious as it provides the possibility of recovering ¹⁹⁹Au by evaporation and subsequent leaching from glass vessel.

Ethyl acetate is susceptible to radiolytic degradation, especially when GBq activity of ¹⁹⁹Au are processed. NMR spectra of ethyl acetate before and after radiochemical separation were recorded and depicted in Fig. 9. As seen from the NMR spectra, there is significant change in the chemical composition of ethyl acetate on exposure to GBq levels of radioactivity. In order to circumvent the radiolytic degradation of ethyl acetate, the extraction is repeated three times with fresh ethyl acetate each time.

Fig. 9 NMR spectrum of ethyl acetate (a) before radiochemical separation (b) after radiochemical separation.

The flow sheet of the radiochemical separation process is depicted in Fig. 10.

Fig. 10 Flow sheet of the radiochemical separation process to isolate ¹⁹⁹Au from neutron irradiated Pt.

Given the simplicity of producing ¹⁹⁹Au by reactor activation of natural Pt targets and ease of adaptation of radiochemical separation strategy, the reported procedure would be of considerable value for accessing NCA ¹⁹⁹Au.

Conclusions

The objective of developing a facile and robust procedure to undertake routine production of NCA ¹⁹⁹Au using natural Pt target has been successfully achieved. The developed technology is sufficed to provide 5.6 GBq (∼150 mCi) of NCA ¹⁹⁹Au per batch. Using this procedure it is possible to avail ¹⁹⁹Au in >95% yields with acceptable radionuclidic and radiochemical purity. The method can be upscale to the desired level if situations arise. It is envisaged that the reported production procedure would serve in good stead for ensuring local availability of ¹⁹⁹Au. The ubiquitous cost effective availability of ¹⁹⁹Au would facilitate more research on new radiopharmaceuticals with ¹⁹⁹Au for therapy.

Conflict of interest

The authors declared that they have no financial or other conflict of interest relevant to the subject of this article.

Acknowledgements

Research at the Bhabha Atomic Research Centre is part of the ongoing activities of the Department of Atomic Energy, India and is fully supported by government funding.

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