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The application of poorly crystalline silicotitanate in production of 225Ac

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Actinium-225 (225Ac) can be produced from a Thorium-229/Radium-225 (229Th/225Ra) generator, from high/low energy proton irradiated natural Thorium or Radium-226 target. Titanium based ion exchanger were evaluated for purification of 225Ac. Poorly crystalline silicotitanate (PCST) ion exchanger had high selectivity for Ba, Ag and Th. 225Ac was received with trace amounts of 227Ac, 227Th and 223Ra, and the solution was used to evaluate the retention of the isotopes on PCST ion exchanger. Over 90% of the 225Ac was recovered from PCST, and the radiopurity was >99% (calculated based on 225Ac, 227Th, and 223Ra). The capacity of the PCST inorganic ion exchange for Barium and 232Th was determined to be 24.19 mg/mL for Barium and 5.05 mg/mL for Thorium. PCST ion exchanger could separate 225Ac from isotopes of Ra and Th, and the process represents an interesting one step separation that could be used in an 225Ac generator from 225Ra and/or 229Th. Capacity studies indicated PCST could be used to separate 225Ac produced on small 226Ra targets (0.3–1 g), but PCST did not have a high enough capacity for production scale Th targets (50–100 g).
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The application of poorly crystalline
silicotitanate in production of 225Ac
Jonathan Fitzsimmons1, Alyson Abraham2, Demetra Catalano3, Ali Younes1, Cathy S. Cutler1 &
Dmitri Medvedev1
Actinium-225 (225Ac) can be produced from a Thorium-229/Radium-225 (229Th/225Ra) generator, from
high/low energy proton irradiated natural Thorium or Radium-226 target. Titanium based ion exchanger
were evaluated for purication of 225Ac. Poorly crystalline silicotitanate (PCST) ion exchanger had high
selectivity for Ba, Ag and Th. 225Ac was received with trace amounts of 227Ac, 227Th and 223Ra, and the
solution was used to evaluate the retention of the isotopes on PCST ion exchanger. Over 90% of the
225Ac was recovered from PCST, and the radiopurity was >99% (calculated based on 225Ac, 227Th, and
223Ra). The capacity of the PCST inorganic ion exchange for Barium and 232Th was determined to be
24.19 mg/mL for Barium and 5.05 mg/mL for Thorium. PCST ion exchanger could separate 225Ac from
isotopes of Ra and Th, and the process represents an interesting one step separation that could be used
in an 225Ac generator from 225Ra and/or 229Th. Capacity studies indicated PCST could be used to separate
225Ac produced on small 226Ra targets (0.3–1 g), but PCST did not have a high enough capacity for
production scale Th targets (50–100 g).
Ion-exchange chromatography has been successfully used to separate radioisotopes for medical applications, nuclear
fuel reprocessing and other applications1,2. However in many instances the ion-exchange material lacks desired
selectivity. Current methods of separation rely on commercially available ion-exchange resins that preferentially
bind the element based on charge3. Extraction chromatography methods have been developed for some separa-
tions4, but the extraction chromatography resins sometime have slow ow rates, and the extractant can be eluted.
Oen the process to purify a radioisotope requires the use of multiple columns and result in consuming more time
and labor. Chemical separations used in accelerator isotope production process at Brookhaven National Lab (BNL)
present interesting challenges. e target masses for production scale targets irradiated at BNL are over 50 grams,
and the mass of the radioisotopes produced is less than 0.0001 grams5,6. e separation presents challenges if no
ion exchange resins are available that have more selective for the isotope of interest rather than the target material.
Production of 225Ac has been of particular interest recently since ecacy of this material has been demon-
strated in a treatment of certain types of cancer. 225Ac can be made available by several routes: separation from
229/225Ra generator, by high energy (100–200 MeV) proton irradiation of natural orium target, or by irradi-
ation 226Ra target with low energy proton (10–24 MeV)7,8. Chemical separation of 225Ac from thorium irradiated
with high energy protons is especially challenging. e irradiation results in ssion of the thorium in the target
and, in addition to 225Ac, produces a variety of potentially useful radionuclides such as 111Ag, 105Rh, 140La, Ra
isotopes, and 140Ba. As previously mentioned, 225Ac (t1/2 = 9.9 days) and its daughter; 213Bi (t1/2 = 45 min) are
emerging as important isotope for targeted alpha therapy911. Other isotopes in the list can be used for beta ther-
apy (111Ag, 105Rh, and 140La), or as a parent in medical isotope generators, for example Radium isotopes and 140Ba.
The use of inorganic ion-exchange materials which selectivity stems from the crystal structure of the
ion-exchanger could provide a more attractive mode of separation. Crystalline silicotitanate (CST) and the poorly
crystalline CST (PCST) have been synthesized hydrothermally in alkaline media12. e reduced crystallinity was
obtained by either shortening the reaction time at the same synthesis temperature for CST (200 °C) or reducing
the temperature to 170 °C and changing the hydrothermal reaction time. e CST inorganic ion exchangers
were initially developed in the 1960s and have been evaluated for nuclear waste treatment due to the materials
remarkable selectivity toward Cs and Sr1315. Studies showed that Cs and Sr2+ cations demonstrate higher rate of
uptake by poorly crystalline CST compared to the crystalline form. is was attributed to the higher surface area
and smaller particle size, which was highlighted to account for the increased rate even though there would be
1Medical Isotope Research & Production Laboratory, Collider-Accelerator Division, Brookhaven National
Laboratory, Upton, NY, 11973, USA. 2Chemistry Department, Stony Brook University, Stony Brook, NY, 11794, USA.
3Biochemistry Department, Stony Brook University, Stony Brook, NY, 11794, USA. Correspondence and requests for
materials should be addressed to J.F. (email: jtzsimmons@bnl.gov)
Received: 8 February 2019
Accepted: 25 July 2019
Published: xx xx xxxx
OPEN
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otherwise slow diusion through the channels. e high selectivity of CST ion exchanger for Cs and Sr was used
to decontaminate the Fukushima site16.
Titanium-based ion exchange materials have been demonstrated to be selective for strontium and actinides in
highly alkaline environments17. e studies reported herein seek to synthesize, characterize, and evaluate PCST
inorganic ion exchanger in the separation process of 225Ac from irradiated  target. During the accelerator
production of 225Ac from a 232 target 227Ac and the daughters of 227Ac are coproduced18. e daughters of 227Ac
(227 and 223Ra) can be separated from 225Ac during the purication process19. However aer purication the
227Ac would grow in and reduce the purity of 225Ac18. Various literature studies to purify 225Ac from orium and/
or radium have focused on organic cation or anion ion exchange resins1,2. e PCST ion exchanger has been syn-
thesized and evaluated for the removal of Sr, Cs and other isotopes from waste streams20. In this manuscript the
purication of 225Ac from 227 and 223Ra with PCST ion exchanger was developed. e utility of the purication
process was evaluated in the following applications: 225Ac production from proton irradiated  or Ra, 225Ac pro-
duced from the 229/225Ra generator, and to purify 225Ac from the daughters of 227Ac.
Results
Inorganic ion-exchangers were synthesized by hydrothermal synthesis, puried and the phases were conrmed
by comparing XRD patterns to published results20. Figure1 outlines the study design to evaluate the inorganic ion
exchangers. Initially, ion-exchange properties of the synthesized materials were evaluated by the batch method,
and distribution coecients (Kd) were determined for 225Ac and . Subsequent studies determined the Kd of
several elements (, La, Ce, Rh, Ag, and Ba) on PCST ion exchanger. Optimal conditions for the separation on
PCST inorganic ion-exchanger were evaluated with a representative sample containing 225Ac, 227, and 223Ra.
Capacity studies were performed with barium and thorium on PCST to determine if the ion exchanger could be
used for radium (0.3 g) and/or thorium (50–100 g) production targets. e stability of the PCST ion exchanger
was determined in: ammonium acetate buer from pH 5 to 1, hydrochloric or nitric acid at 0.1,1, 2 and 3 M.
Evaluation of inorganic Ion-Exchanger Selectivity. e evaluation of the selectivity of poorly crystal-
line silicotitanate (PCST) for Rh, Ba, La, Ce, , and Ag were performed with batch studies, and the data for ,
Ag, and Ba are presented in Fig.2. e results of the PCST ion exchanger show an insignicant selectively at low
pH for Ba, La, Ce,  and Rh, with a higher selectivity for Ag (3465 ml/g) at a pH of 1. As the pH goes from 1 to 5
the PCST ion exchanger increases selectivity for Ba, and Ag (46305 ml/g, 6796 ml/g respectively). e PCST ion
exchanger had low selectivity for Ce, La and Rh (100 ml/g, 71 ml/g and 85 ml/g respectively). e distribution
coecients for the trivalent cations Ce and La increased as the pH went 1 to 5.
Column studies. To increase the ow rates buer was added to the 100–200 mesh PCST material and the solution
was decanted. is was able to remove ne particles of PCST material, and the ow rates increased to 0.25–1 ml/min.
PCST breakdown study. e quantication limit for Ti on the ICP-OES was dened by analysis of diluted stand-
ards and the acceptance criteria for the true concentration value and the % RSD was within 10%. e quanti-
cation limit for Ti by ICP-OES was determined to be 0.005 ppm. ICP-OES analysis of all samples indicated Ti
breakthrough (Fig.3). e Ti breakthrough was lowest in 0.5 M ammonium acetate pH 5 with 0.05–6 µg of Ti
present in the load, 0.5 M ammonium acetate pH 5 and 3 rinses. Rinsing the PCST material with ammonium
acetate at pH 1 resulted in 352–405 µg of Ti. In subsequent rinses with 0.1 M HCl or nitric acid the PCST material
showed a slight higher amount of Ti breakthrough in HCl (91–110 µg) versus nitric acid (63–86 µg). Higher con-
centrations of acid resulted in more breakthrough of the Ti from the PCST with 1 M (480–700 µg Ti per fraction)
2 M (750–1022 µg Ti per fraction) and 3 M (703–1135 µg Ti per fraction). In all elutions with HCl or nitric acid
breakthrough of the Ti from the PCST material was higher for HCl.
Separations of , 225Ac and other metals. In summary the Kd data indicated: the PCST ion exchanged has
favorable properties to capture orium at pH values of 1 to 2 while 225Ac would be less favorable; the resin has
1) Synthesis of ion
exchangers
:HCST,
PCST
2) Characterization:
XRD
6) PCST Capacity Studies
on PCST
a) Thorium, b) Ba
3) Initial Screening studies
a) Kd of La, Lu, Ce on HCST,
b) Kd of 225Ac, Th on HCST,
c) Kd of Rh, Ag,Ba, La, Ce, Th
on PCST
8) Optimize conditions
for225Ac, separation
from 227Th, 223Ra
a) examine pH of rinse
steps: 5, 4.5, 4, 3.5, 3, 2.5,
2, 1.5, 1
b) examine rinse steps at
pH 5, 3 and 1
5) Stability of PCST in
buffer and acids
a) Evalua
te PCST in
b
uffer pH 1-5
b
) Evaluate PCST in 0.1,
1, 2 and 3 M HC
l
c) Evaluate PCST in
0.1, 1, 2 and 3 M HN
O3
4) Column studies
7) Evaluate separation
with surrogates
a) use Ba andLa in
separation
Figure 1. Flow chart illustrating the sequence of studies reported in the manuscript.
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high anity for Ba at pH values of 3 to 5, and is more selective for Ag at pH values of 1, 4 or 5 than , La, Ce, and
Rh. A column of PCST ion exchanger was used to capture orium and barium from a solution containing ,
Ba, 225Ac, Rh, La, Ag, and Ce in 0.5 M NaOAc at pH 2 (see Supplementary Fig.S4). e eluted solution contained
95% of Ac, 90% of Ce, 73% of La and 77% of Rh. e data indicates 15% of La, 5.8% of Rh, and , Ag and Ba
were totally absorbed and retained on the PCST column. e absorbed metals except Ag were recovered from
the PCST column using 3 M nitric acid solution. e capacity of the PCST inorganic ion exchange materials for
Barium and 232 was determined to be 24.19 mg/mL for Barium and 5.05 mg/mL for orium.
Ba and La column studies. e PCST ion exchanger was able to retain Ba in 0.5 M ammonium acetate at pH 5
while eluting La (see Supplementary Fig.S2). Combining the load and rinse 1–3 provided 85% recovery of La
while only 0.15% of the Ba was present. Rinse solutions 4–5 and elution solutions 1–2 recovered 99.8% of the
barium with 92.8% of the barium eluting in 0.5 M ammonium acetate at pH 1.
223Ra, 225Ac, 227 studies. pH study & PCST breakdown: A study was performed with PCST material with 225Ac,
223Ra and 227 in 0.5 M ammonium acetate at pH 5 and the column was rinsed with the buer at 4.5, 4, 3.5, 3, 2.5,
2.0, 2.5, 1.5, 1 (Fig.4). e 225Ac eluted with two peaks at pH 5 and pH 3 with 96% eluting in the load and fractions
with a pH from 5–3.5. e 223Ra and 227 were both retained on the column and began to elute from the column
at pH 1.5 with 98–100 percent of the isotopes eluting in the buer at pH 3 to 1. e activity retained on the col-
umn was not measured. e breakdown of the PCST was evaluated and Ti was below quantication limits in the
eluted load and pH 5 solutions, but Ti was quantied in all other fractions. e amount of Ti breakthrough in
buer at pH values from 4.5–2.5 was less than 1 μg per fraction. Eluting the PCST column with 0.5 M ammonium
acetate buer at pH 2, 1.5 and 1 resulted in higher breakthrough of Ti (11.7, 80, and 89 μg).
Initial optimization of the separation with PCST: Initial optimization of the separation examined using a rinse
sequences with 6 bed volumes (BV) or column volume of 0.5 M ammonium acetate at both pH 5 and 3, and 12
0.E+00
5.E+03
1.E+04
2.E+04
2.E+04
3.E+04
3.E+04
4.E+04
4.E+04
5.E+04
5.E+04
12345
Kd values of metals on PCST
ion exchanger
Th Ag Ba
Kd (mg/L)
pH
0.E+00
1.E+03
2.E+03
3.E+03
4.E+03
5.E+03
6.E+03
7.E+03
8.E+03
12345
Kd values of metals on PCST
ion exchanger
Th Ag
Kd (mg/L)
pH
Figure 2. Distribution coecients (Kd) of Ba, , and Ag on PCST ion exchanger at pH values from 1 to 5 in
0.5 M ammonium acetate. La, Ce, and Rh had Kd values less than 100 mg/L.
0
200
400
600
800
1000
1200
load 10 20 30 33 36 39 42 45 48 51 54 57 60 63 66 69 72 75 78
Mass of Ti (µg)
volume eluted (mL)
PCST breakdown in buffer and acids
HCl
Nitric acid
0.5 M NH4OAC
pH 5 pH 3 pH 1 0.1 M 1 M 2 M 3M
acid concentraon
Figure 3. Acid study examining the breakdown of PCST in the presence of various concentrations of 0.5 M
ammonium acetate and hydrochloric or nitric acid. 10 mL (20 bed volumes) was used for the 0.5 M ammonium
acetate load and rinse steps then each acid concentration was rinsed with 4 × 3 mL rinses.
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BV at pH 1 (see Supplementary Fig.S3). Combining the load, pH 5 and 3 rinses resulted in 91% of the 225Ac, and
approximately 39–41% of the 225Ac was eluted in pH 5 and 34–42% of the 225Ac was eluted in the pH 3 rinse. e
0.5 M ammonium acetate pH 1 rinse step eluted 44–81% of the Ra and 95% of the orium. e column retained
0.1–1.4% of the 225Ac and 19–56% of the 223Ra.
Optimized purication and PCST breakthrough: e rinse sequence used: 3X3BV of 0.5 M ammonium ace-
tate at pH 5, 2X3BV of the buer at pH 3, 2X3 BV of the buer at pH 1 (Fig.5). Combining the load, pH 5 and the
rst pH 3 rinse (heavy black line) resulted in the elution of 91% of the 225Ac. Based on the activity of 225Ac, 223Ra
and 227 the initially radionuclidic purity of the 225Ac was 78.4%, and the radionuclidic purity of the puried
225Ac in the combined fractions was 99.3%. In the pH 5 rinse step 62.9% of the 225Ac was eluted and in the pH 3
rinses 21.3% of the 225Ac was eluted with 18.9% eluting in the rst pH 3 rinse and only 2.3% eluted in the second
pH 3 rinse step. e second pH 3 rinse and the pH 1 rinses eluted 95% of 227. e pH 1 rinse contained 42%
of 223Ra and 53.7% was retained on the column. Ti breakthrough was checked in the second pH 5, rst pH 3 and
both pH 1 fractions, and the Ti breakthrough was 0.025, 0.29, 120.7 and 142.5 µg per fraction consistent with
previous studies.
Discussion
Dierent separation methods and materials are being evaluated to purify 225Ac from orium, so that the US
Department of Energy can supply 225Ac to researchers and clinicians on clinical scales19,2123. In this manuscript
PCST ion-exchanger was evaluated to determine if the material can be utilized for the purication of 225Ac from
orium or radium targets or in a 229/225Ra generator.
0
10
20
30
40
50
60
70
80
90
0.0
10.0
20.0
30.0
40.0
50.0
60.0
70.0
80.0
90.0
9.512.515.518.521.524.527.530.533.536.539.5
Mass Ti (μg)
Percenteluted(%)
Volume eluted (mL)
Elu on profile of Ac-225,Ra-223, Th-227 and Ti from PCST
Ac-225 Ra-223 Th-227 Ti breakthrough
0.5 Mammoniumacetate
Figure 4. pH studies: A Separation of 225Ac, 223Ra and 227 using a PCST ion exchanger in 0.5 M ammonium
acetate. 223Ra and 227 and 225Ac is retained at pH values greater than 3. 223Ra and 227 are eluted at more acidic
pH values. Titanium present in fractions associated with study A, illustrating the breakdown of PCST material
at lower pH (in green).
0.000
0.500
1.000
1.500
2.000
2.500
3.000
3.500
9.512.515.518.521.524.527.530.5column
Ac vity (uCi)
Volume (mL)
Op mizedsepera on of Ac-225 from Ra-223 andTh-227onPCST
Ac-225 Ra-223 Th-227
0.5Mammonium acetate
Figure 5. Optimized separation of 225Ac, 223Ra and 227 using a PCST ion exchanger in 0.5 M ammonium
acetate. 225Ac solution containing 223Ra and 227 using a 0.5 mL column of PCST ion exchange material. e
load, pH 5 and the rst pH 3 rinse (heavy black line) contained 91% of 225Ac with a radiopurity of 99.3%. e
second pH 3 rinse and pH 1 rinses eluted 95% of 227. e pH 1 rinse eluted 42% of 223Ra and 53.7% was
retained on the column.
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Column studies. Column studies with PCST ion exchanger were plagued with poor (1 ml/15 min) or no ow
rates. Sieving the PCST material through various mesh lters did little to increase the ow rates. In all column
studies 100–200 mesh PCST was used and reasonable ow rates were achieved by soaking the PCST in buer and
decanting the buer.
Acid study of PCST. Ammonium acetate buer at pH 1 and both hydrochloric acid and nitric acid at a concen-
tration of 0.1 M lead to breakdown of PCST material. e breakdown of the PCST at pH 5–3 is far less then at
more acidic pH values. Using the PCST material at pH 1 or in 1, 2, and/or 3 M acid may require a cleanup column
to remove Ti breakthrough.
Separation of Ba from La. e Ba-La separation was conducted to assess the potential of a separation of Ra
and Ac radioisotopes with a PCST column, with the stable isotopes serving as surrogates for the radioisotopes.
e elution of both metals at separate pH values was clear, and La eluted at pH 5 while Ba eluted at pH 1 (see
Supplementary Fig.S2). Application: is purication approach could be used to separate 140La from 140Ba in a
generator system. e study indicates all trivalent lanthanides will be eluted at pH 5 with 140La and 225Ac.
Evaluating PCST ion exchangers for purication of 225Ac. e elution of Ba and La on PCST was repeated with
223Ra, 227 and 225Ac, and the separation was optimized to purify 225Ac.
e separation of 225Ac from 223Ra and 227 had good reproducibility. In ve studies (the pH study, three
optimization studies with 225Ac, 223Ra and 227, and the column studies 225Ac, , Ag, Ba, Rh, Ce, La) greater than
90% of the 225Ac eluted in the load, pH 5 and/or pH 3 solutions. e optimized elution method to separate 225Ac
was selected based on the elution of the highest percentage of 225Ac with the lowest percentage of impurities, in
this case, 223Ra and 227. e optimized method rinsed the PCST column with more 0.5 M ammonium Acetate
at pH 5 and 3 resulted in a shi in the 225Ac elution peak. e result was more 225Ac was eluted earlier with 92%
eluted in the pH 5 and rst pH 3 rinse step. e process produced very pure 225Ac with 223Ra and 227 eluting
at low pH, and this data indicates the optimized method with PCST material could be used in several dierent
production approaches to purify 225Ac from orium and/or Radium radioisotopes.
Applications of separation. 225Ac from a 225Ra/229 generator: 225Ac has been produced at Oak Ridge
National Lab (ORNL) from a 225Ra/229 generator and they produce 5.5 × 1010 Bq (~1.5 Ci) per year24. e one
column separation of 225Ac from 223Ra and 227 with PCST could simplify the multicolumn approach used
at ORNL to purify 225Ac in the 225Ra/229 generator. e ORNL process is a four step chemical process with
two MP1/NO3 columns to separate 225Ac and 225Ra from 229. en the 225Ac is puried from 225Ra with two
AG50X4/1.2 HNO3 purication steps. e high selectivity of PCST for both thorium and radium would simplify
the purication of 225Ac, and the process would be one column, making the purication shorter than the ORNL
process. Accelerator produced 225Ac: Accelerator produced 225Ac can be produced at high energy (>100 MeV)
with a natural thorium target or at low energy (10–25 MeV) from a 226Ra target. PCST for the purication of 225Ac
from 232 targets: Although the PCST column worked to separate 225Ac from orium, Ra and some ssion
products the approach is similar to published separations that capture orium on an MP1 column and let 225Ac
pass through the column24. is separation strategy would work for smaller thorium target. For larger scale clin-
ical production of 225Ac with 50 to 100 g orium targets and potential orium stack targets could be required
resulting in 100–600 grams of orium in the separation. e small capacity of PCST for orium indicates a
large mass of PCST ion exchanger would be needed to capture all the orium. To process one 50 gr orium
target it was estimated 10 L PCST column would be needed, so the material does not have a reasonable capacity
to purify 225Ac from orium targets. PCST for the purication of 225Ac from 226Ra targets: Low energy protons
irradiating a 0.3 g 226Ra target can produce clinical scales of 225Ac (~ 1 Ci/target)7,25. e amount of PCST ion
exchange material to process a 226Ra target was estimated from capacity studies with barium, and the data indi-
cates a minimum of a 12.5 mL PCST column would be needed to retain 226Ra in the target. e PCST separation
method to purify 225Ac could be used to purify 225Ac from a 223Ra targets. 226Ra could be eluted from the PCST ion
exchanger in pH 1 buer, and the process could be used to recycle the 226Ra.
Removal of radio impurities (223Ra and 227) in accelerator produced 225Ac from  targets: Accelerator
produced 225Ac from a proton irradiated thorium targets has 0.1% abundance of 227Ac (t1/2 = 21.8 years) at end
of bombardment. 227Ac decays to 223Ra (t1/2 = 11.4 days) and 227 (t1/2 = 18.5 days), and 227Ac and the daughters
represent the major radio- impurities for 225Ac. e expiration of a batch of accelerator 225Ac is dened by the
puried 225Ac failing one of its specications, which are currently being determined by the Trilab 225Ac team. e
radiopurity of 225Ac will likely be the rst specication that fails. e Trilab team has estimated the radiopurity
post purication in the presence of 227Ac and 227Ac and daughters for BLIP produced 225Ac from  targets18.
Removal of 223Ra and 227 would increase the radiopurity of 225Ac produced from a  target at BLIP. During
the development of the PCST separation the original radiopurity of 225Ac was 78.4% (calculated from 225Ac, 227
and 223Ra), but aer performing the purication with PCST column the radiopurity was 99.3% and the recovery
of 225Ac was 92.4%. is improvement in radiopurity indicates that this separation can be used to extend 225Ac
shelf life by removing 227Ac daughters, 223Ra and 227. A PCST column run on the 225Ac sample would remove
223Ra and 227; increasing the radiopurity to greater than 98% and this could be done out to 30.1 days. For 225Ac
produced from a  target at BLIP the radiopurity falls below 98% aer 14.9 days aer the last separation step.
A radiopurity greater than 98% can be achieved with a PCST column out to 27.8 days. In a clinical setting, the
collection of the 18 BV of the pH 5 rinse and rst 6 BV of the pH 3 rinse would recover 92.4% of eluted 225Ac and
a negligible percentage of impurities. is 12 mL sample of 225Ac would be easy to evaporate.
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Conclusion
e impact of developing a material with specic isotope and/or metal selectivity would potentially be invaluable
in assisting with eorts in medical isotope production. e studies herein evaluated titanium ion exchanger and
examined if the material could be used for the purication of 225Ac. Examination of the eects of acid rinses on
PCST indicated that even 0.1 M acid, either nitric or hydrochloric, breaks down the material and resulted in the
elution of the material (Titanium). is led to the conclusion that acid more dilute than 0.1 M is needed when
working with PCST. For the separation of 225Ac from radioactive Ra and , the optimized method used 18 BV of
buer at pH 5 and 6 BV of buer at pH 3. is lead to a high recovery of eluted Ac (92.4%) and high radiopurity
(99.3%). An 225Ac 229 generator can be established based on this separation. Capacity studies of Barium and
orium on PCST indicated that the material did not have a high enough capacity for a production scale thorium
target (50 g) of PCST. However, PCST could be used to purify 225Ac from smaller production scale 226Ra targets
(0.3 g).
Materials and Method
Reagents were used from manufacturer without additional purication: phosphoric acid, fumed silica, titanium
isoproproxide and KOH Pellets were purchased from Sigma Aldrich. Sodium hydroxide (98%), nitric acids (70%
optima) and trace metal grade hydrochloric acid were purchased from Fisher. La, Ce, Lu ICP standards were
purchased form Fluka in 1000 mg/L concentrations. ICP single elemental certied standards of: , Ag, Ba, Rh,
Ce, and La were purchased from SPEX Certiprep. All solutions were prepared using Milli-Q water and all experi-
ments were conducted at room temperature. All the chemicals used were of analytical reagent grade. Buers were
prepared from previously prepared 0.5 N Sodium Acetate buer and adjusted with 8 N HCl or 10 M NaOH. Initial
and equilibrium pH readings were obtained using a Denver Instruments UB-10 pH/mV meter calibrated at pH
2.0, 4.0 and 7.0. Since Ba chemically behaves similarly to Ra and La is chemically similar to 225Ac for some studies
Ba and La can be used as surrogates. 225Ac radiotracer was supplied by Oak Ridge National Laboratory as a dried
sample, and the sample was dissolved in 0.1 M HCl solution prior to use. 223Ra and 227 were present in some
225Ac samples as a result of the decay of 227Ac.
Synthesis of PCST inorganic ion-exchanger. PCST were synthesized according to published meth-
ods26,27. Phase and purity was conrmed by powder X-Ray diraction (patterns collected using a Rigaku MiniFlex
II Desktop X-Ray diractometer sampling at 0.040 degrees at a speed of 1degree/min, starting at 5 degrees and
ending at 60 degrees.
Determination of Kd values for dierent inorganic ion-exchangers. 225Ac and . Experimental
solutions, consisting of 20 mg L1 of  and 3.7 × 104 Bq 225Ac in 0.5 M sodium acetate (NaOAc), were adjusted
to a pH of 1, 2, 3, 4 or 5 with 10 M NaOH or 69% HNO3. Inorganic ion-exchangers HCST (100 ± 0.5 mg) were
added to 10 mL of the metal containing solution. e tubes were shaken (using ermo Scientic compact dig-
ital microplate shaker) for 12 hours at room temperature. e tubes were centrifuged (4000 rpm, 2598 RCF) for
4 minutes and the aqueous phases were separated using 0.2 µm micro syringe lter. An aliquot from the aqueous
phases was diluted in 2% nitric acid and analysis performed.
Rh, Ag, Ba, La, Ce and  on PCST inorganic ion-exchangers. A solution consisting of: 30 mg L1 of each of Ag,
Ba, Rh, Ce, and La; and 150 mg L1 of  in 0.5 M NaOAc. Ba was used as a surrogate of Ra, and La was used as a
surrogate for 225Ac as the chemistries are similar. e PCST inorganic ion-exchangers (30 ± 0.3 mg) were added
to the 10 mls of the metal containing solution. e samples were processed and analyzed as described in the
previous section.
Analysis and calculations of Batch studies. Actinium activities in the initial, intermediate and nal
solutions were determined by using a gamma spectrometry (ORTEC) with a calibrated high purity germanium
detector28. e separation fractions containing 225Ac, 227 and 223Ra were quantied aer 24 hours by gamma
spectroscopy at 236 (227), and 269.6 keV (223Ra). Aer 24 hours the original activity associated with 213Bi and
221Fr has decayed away (24 hours >10 half-lives of 213Bi and 221Fr). e presence of 213Bi and 221Fr in the samples is
the result of in growth from 225Ac, and activity associated with 213Bi and 221Fr would be at equilibrium with 225Ac.
At the time of analysis, 225Ac and its daughters (specically 221Fr and 213Bi) were at equilibrium with 225Ac, and the
gamma peak at 440 KeV for Bi-213 was used to quantify 225Ac. e 218 KeV gamma peak for 221Fr was used to
quantify 225Ac and similar results were obtained.
e concentration of thorium and other metals were measured using ICP-OES (Perkin Elmer Optima 7300
DV spectrometer), and the instrument was calibrated according to published methods28. e wavelength (λ) used
for thorium, barium, lanthanum, cerium, rhodium, silver and titanium analysis were 283.73, 233.527, 408.672,
413.764, 343.489, 338.289 and 334.94 nm respectively.
Kd values were calculated based on the following equation:
()
KCCCVg[/]/
diff
=−
Where Ci = Stock concentration metal, Cf = Final concentration metal, V = Volume of stock solution, and
g = Measured weight of ion-exchanger. Kd values were plotted versus pH, and the standard deviation of triplicate
samples was calculated.
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Column studies. Column Preparation Method. e inorganic ion exchange material was sied through
>50, 50–100, 100–200, 200–400 and <400 µm mesh sier. For all column studies the particles between 100 and
200 µm in size were used and approximately 1 gram of material was added to 20 mL 0.5 M ammonium acetate at
pH 5. e sample was mixed vigorously then the mixture was allowed to settle before decanting the supernatant.
is was repeated three times to remove any small residual particles that impeded column ow and caused poor
ow rates. en the inorganic ion exchange material was used to prepare columns with 0.4–0.5 mL bed volumes.
e columns were rinsed with 20 BV of 0.5 M ammonium acetate at pH 5 prior to use, and the pH of the solution
eluting the column was at pH 5. Columns were timed to monitor the ow rate.
Acid Study of PCST. Two 0.5 mL BV columns of poorly crystalline silicotitanate (PCST) were prepared in 0.5 M
ammonium acetate at pH 5. One column was rinsed with 10 mL of 0.5 M ammonium acetate at pHs 5, 3, and
1. en the column was rinsed with: 4 × 3 mL of each of the following concentrations of nitric acid: 0.1 M, 1 M,
2 M and 3 M. A 500 microL aliquoted of each rinse solution was diluted with 4.5 mLs of 2% Nitric Acid to make
an ICP-OES sample, and the amount of titanium was quantied. e experiment was repeated with the second
column with the exception hydrochloric acid was used instead of nitric acid.
Capacity studies to Evaluate resin for purication of 225Ac from production targets or isotope generators. Barium:
A 0.4 mL column of PCST was prepared in 0.5 M ammonium acetate at pH 5. A barium solution was prepared by
evaporated 25 mg of Ba metal ICP standard and resuspended in 5 mL of 0.5 M ammonium acetate buer at pH 5.
e Ba solution was loaded on the PCST column and the column was rinsed with 3 × 10 mLs of 0.5 M ammonium
acetate at pH 5. e amount of barium in the load and rinse solution was quantied by ICP-OES. orium: e
study was repeated for thorium with the exception that a 0.5 mL PCST column was used and the load solution
contained 100 mg of thorium.
225Ac, 227, 223Ra Separation. pH Elution with either PCST ion exchanger: A 0.5 mL PCST column was prepared
in 0.5 M ammonium acetate at pH 5, and a 10 mL load solution of the buer was prepared with tracer quantities
of 225Ac (4.00 × 105 Bq), 227 (8.136 × 104 Bq) and 223Ra (2.886 × 104 Bq). A HPGE counting sample was prepared
by diluting 500 µL of the prepared load solution to 3 mL. e remaining load solution was added to the PCST
column. en the column was rinsed with 3 mL of 0.5 M ammonium acetate at each of the following pH values: 5,
4.5, 4, 3.5, 3, 2.5, 2, 1.5, and 1. e radioactivity in the load and rinse samples were quantied by HPGE analysis
and the titanium was quantied in the samples by ICP-OES analysis. e ICP-OES samples were prepared by
diluting the 3 mL fractions to 5 mLs with 2% nitric acid.
Optimized Elution method: A 0.5 mL PCST column and a solution containing 225Ac, 227, and 223Ra in 0.5 M
ammonium acetate at pH was prepared as described above. e previous experiment was repeated with the fol-
lowing rinse sequence of 0.5 M ammonium acetate: 3 × 3 mL at pH 5, 2 × 3 mL at pH 3, and 2 × 3 mL at pH 1.
Radioactivity in the fractions was determined by HPGE analysis and ICP-OES was conducted on one of the pH
5 and 3 rinses and both pH 1 rinses.
Data Availability
All data generated or analyzed during this study are included in this published article (and its Supplementary
Information les).
References
1. Hyde, E. . e adiochemistry of orium. National Academy of Sciences (1960).
2. Massart, D. L. Cation-Exchange techniques in radiochemistry. National Academy of Sciences (1971).
3. Pacage inserts for resins from BIO-AD: Chelex-100 and chelex 20 chelating ion exchange resin (LIT200 evB), AG 50W and AG
MP-50 cation exchanger resins (LIT203 ev B).
4. https://www.eichrom.com/eichrom/products/accessed 1-28-2019.
5. Fitzsimmons, J. & Mausner, L. Development of a production scale purication of Ge-68 from irradiated gallium metal. adiochimica
Acta 103, 117–123 (2015).
6. Medvedev, D. G. et al. Development of a large scale production of 67Cu from 68Zn at the high energy proton accelerator: Closing the
68Zn cycle. Applied adiation and Isotopes 70(3), 423–429 (2012).
7. Apostolidis, C. et al. Cyclotron production of Ac-225 for targeted alpha therapy. Applied adiation and Isotopes 62(3), 383–387
(2005).
8. Meeting Isotope Needs and Capturing Opportunities for the Future: e 2015 Long ange Plan for the DOE-NP Isotope Program.
NSAC Isotopes Sub committee. 07–20, https://science.energy.gov/~/.../np/.../2015/2015_NSACI_eport_to_NSAC_Final.pdf
(Accessed 2/13/2018) (2015).
9. ratochwil, C. et al. Morgenstern. 225Ac-PSMA-617 for PSMA-Targeted a-adiation erapy of Metastatic Castration-esistant
Prostate Cancer. J. Nucl. Med. 57, 1–4 (2016).
10. Morgenstern, A., Bruchertseifer, F. & Apostolidis, C. Bismuth-213 and actinium-225 - generator performance and evolving
therapeutic applications of two generator-derived alpha-emitting radioisotopes. Curr adiopharm. Jul 5(3), 221–7 (2012).
11. ratochwil, C. e t al. 213Bi-DOTATOC receptor alpha-radionuclide therapy induces remission in neuroendocrine tumors refractory
to beta radiation: a rst in human experience. A. Eur J Nucl Med Mol Imaging, Nov 41(10), 2106–19 (2014).
12. Cle areld, A. et al. ates of exchange of Cs+ and Sr2+ for poorly crystalline sodium titanium silicate (CST) in nuclear waste systems.
Solvent Extraction and Ion Exchange. 30(3), 229–243 (2012).
13. Miller, J. & Brown, N. Development and properties of crystalline silicotitanate (CST) ion exchangers for radioactive waste
applications. Sandia eport: SAND97-0771 UC-721, UC-510. April 1997.
14. National esearch Council 2000. Alternatives for High-Level Waste Salt Processing at the Savannah iver Site. Washington, DC: e
National Academies Press, https://doi.org/10.17226/9959.
15. Behrens, A., Sylvester, P. & Cleareld, A. Assessment of a sodium nonatitanate and pharmacosiderite-type ion exchanger for
strontium and cesium removal from DOE waste Simulants. Environ. Sci. Technol. 32, 101–107 (1998).
16. Tusa, E. Cesium and strontium removal with highly selective ion exchange media in Fuushima and cesium removal with less
selective media – 14018. WM2014 Conference, March 2–6, Phoenix, Arizona, USA (2014).
Content courtesy of Springer Nature, terms of use apply. Rights reserved
8
SCIENTIFIC REPORTS | (2019) 9:11808 | https://doi.org/10.1038/s41598-019-48021-7
www.nature.com/scientificreports
www.nature.com/scientificreports/
17. Hobbs, D. T. et al. Strontium and Actinide Separations from High Level Nuclear Waste Solutions Using Monosodium Titanate 1.
Simulant Testing. Sep. Sci. Technol. 40(15), 3093–3111 (2005).
18. Fitzsimmons, J., Griswold, J., Medvedev, D., Cutler, C. S. & Mausner, L. Dening processing times for accelerator produced 225Acand
other isotopes from proton irradiated thorium. Molecules 24, 1095, https://doi.org/10.3390/molecules24061095 (2019).
19. Fitzsimmons, J. et al. Evaluation of SynPhase Lanterns for capturing Ac-225 from bul orium and comparisons to equivalent
functionalized resins. J of adioanal. and Nucl. Chem. 317, 477–483 (2018).
20. Medvedev, D. G., Tripathi, A., Cleareld, A., Celestian, A. J. & Parise, J. Hanson. Crystallization of Sodium Titanium Silicate with
Sitinaite Topology: Evolution from Sodium Nonatitanate Phase. Chemistry of Materials 16, 3659–3666 (2004).
21. adcheno, V. et al. Application of ion exchange and extraction chromatography to the separation of actinium from proton-
irradiated thorium metal for analytical purposes. J. Chrom. A. 1380, 55–63 (2015).
22. Harvey, J. et al. Production of Actinium-225 via High Energy Proton Induced Spallation of orium-232. United States: N. p., Web,
https://doi.org/10.2172/1032445 (Accessed 2/13/2018) (2011).
23. McAlister, D. & Horwitz, E. Selective separation of radium and actinium from bul thorium target. Appl. adiat. Isot. 140, 18–23
(2018).
24. Boll, ., Malemus, D. & Mirzadeh, S. Production of actinium-225 for alpha particle mediated radiotherapy. Appl. adiat. Isot. 62,
667–679 (2005).
25. Niedzielsi, D. & Yip, . Using Monte Carlo methods to estimate occupational dosage during Ac-225 production. Summer
Internship esearch Symposium. Broohaven National Laboratory (August 9, 2018).
26. Behrens, E. A. & Cleareld, A. Titanium silicates, M3HTi4O4(SiO4)3.4H2O (M = = Na+, +), with three-dimensional tunnel
structures for the selective removal of strontium and cesium from wastewater solutions. Micropourous Mat. 11, 65–75 (1997).
27. Behrens, E. A., Poojary, D. M. & Cleareld, A. Syntheses, Crystal Structures, and Ion-Exchange Properties of Porous Titanosilicates,
HM3Ti4O4(SiO4)3·4H2O (M = = H+, +, Cs+), Structural Analogues of the Mineral Pharmacosiderite. Chem. Mater. 8, 1236–1244
(1996).
28. Fitzsimmons, J. M., Medvedev, D. B. & Mausner, L. F. Specic activity and isotope abundances of strontium in puried strontium-82.
Journal of Analytical Atomic Spectrometry 31, 456–463 (2016).
Acknowledgements
e research described in this paper was funded by the United States Department of Energy, Oce of Science via
funding from the Isotope Development and Production for Research and Applications subprogram in the Oce
of Nuclear Physics. is project was supported in part by the US Department of Energy, Oce of Science, Oce
of Workforce Development for Teachers and Scientists (WDTS) under the Science Undergraduate Laboratory
Internships Program (SULI). e isotope(s) used in this research was supplied by the Isotope Program within
the Oce of Nuclear Physics in the Department of Energy’s Oce of Science. Accelerator-produced 225Ac is
available through the Department of Energy, Oce of Nuclear Physics, subprogram Isotope Development and
Production for Research and Applications, and can be ordered from the National Isotope Development center,
who can be contacted at: phone: (865) 574-6984; fax: (865) 574-6986; email: contact@isotopes.gov or online at:
https://www.isotopes.gov/catalog/product.php?element=Actinium. https://www.isotopes.gov/catalog/product.
php?element=Actinium&type=rad&rad_product_index=87.
Author Contributions
J.F. and D.M. designed the experiments and prepared safety documents. Experiments were carried out by A.A.,
D.C., A.Y. and J.F. Data analysis was carried out by J.F., A.A., A.Y. and D.M. Figures were prepared by J.F., A.A. and
A.Y. e main manuscript text was written by J.F. and A.A. All authors reviewed the manuscript.
Additional Information
Supplementary information accompanies this paper at https://doi.org/10.1038/s41598-019-48021-7.
Competing Interests: e authors declare no competing interests.
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Proton irradiation of a thorium (Th) target can produce Actinium-225 (Ac-225), but the irradiated thorium fissions results in the production of 100’s of other isotopes. SynPhase Lanterns containing sulfonic acid groups were evaluated for the purification of Ac-225 from Th and other fission metals. The SynPhase Lanterns were able to quantitatively and selectively capture Ac-225 and lanthanum (La) over Th. A second purification step would be needed to purify Ac-225 from trace amounts of Th and other fission products. The results from these studies were compared to similar studies performed with cation exchange resins, and the resins were superior at removing impurities.
Article
Methods: (68)Ga-PSMA-11 PET/CT validated a PSMA-positive tumor phenotype. Activities of 100 kBq/kg body-weight of (225)Ac-PSMA-617 were administered bi-monthly. PSA response and hematological toxicity were measured at least every four weeks. Restaging was again performed with (68)Ga-PSMA-11 PET/CT. Results: Both patients experienced a PSA decline below measurable and presented with complete imaging response. No relevant hematological toxicity was observed. Xerostomia was the only mentionable clinical side-effect. Conclusion: Targeted alpha therapy with (225)Ac-PSMA-617, yet experimental, obviously offers a high potential to provide significant benefit to patients in critical condition, e. g. with diffuse red marrow infiltration and resistance against other therapies.
Article
A linear accelerator was used to irradiate a rubidium chloride target with protons to produce strontium-82 (Sr-82), and the Sr-82 was purified by ion exchange chromatography. The amount of strontium associated with the purified Sr-82 was determined by either: ICP-OES or method B which consisted of a summation of strontium quantified by gamma spectroscopy and ICP-MS. The summation method agreed within 10% to the ICP-OES for the total mass of strontium and the subsequent specific activities were determined to be 0.25–0.52 TBq mg−1. Method B was used to determine the isotope abundances by weight% of the purified Sr-82, and the abundances were: Sr-82 (10–20.7%), Sr-83 (0–0.05%), Sr-84 (35–48.5%), Sr-85 (16–25%), Sr-86 (12.5–23%), Sr-87 (0%), and Sr-88 (0–10%). The purified strontium contained mass amounts of Sr-82, Sr-84, Sr-85, Sr-86, and Sr-88 in abundances not associated with natural abundance, and 90% of the strontium was produced by the proton irradiation. A comparison of ICP-OES and method B for the analysis of Sr-82 indicated analysis by ICP-OES would be easier to determine total mass of strontium and comply with regulatory requirements. An ICP-OES analytical method for Sr-82 analysis was established and validated according to regulatory guidelines.
Article
Molecular imaging plays important role in detection and staging of hematological malignancies. Multiple myeloma (MM) is an age-related hematological malignancy of clonal bone marrow plasma cells characterized by destructive bone lesions and is fatal in the vast majority of patients. Traditional skeletal survey and bone scans have sensitivity limitations for osteolytic lesions manifested in MM. Progressive biomedical imaging technologies such as low dose CT, molecularly targeted PET, MRI, and the functional-anatomical hybrid versions (PET-CT, PET-MRI) provide incremental advancements in imaging MM. Imaging with PET and MRI using molecularly targeted probes are promising precision medicine platforms that might successfully address the clinical ambiguities of myeloma spectrum diseases. The intent of this focus article is to provide a concise review of the present status and promising developments on horizon such as the new molecular imaging biomarkers under investigation that can either complement or potentially supersede existing standards.
Article
Germanium-68 (Ge-68) is produced by proton irradiation of a gallium metal target, purified by organic extraction and used in a medical isotope generator to produce Gallium-68 PET imaging agents. The purpose of this work was to implement a production scale separation of Ge-68 and Zn-65 that does not use organic solvents and uses a limited number of columns. The current separation approach was modified to use AG1 resin and/or Sephadex(©) G25 with zinc spikes to purify Ge-68 with near quantitative recovery. The purified Ge-68 meets DOE specifications. Methods utilizing zinc spikes resulted in the purist Ge-68 produced at Brookhaven National Lab with no other impurities by ICP-OES. During process optimization approximately 2.5Ci of Ge-68 was purified utilizing the different processing methods, and the material was sold to the Nuclear Medicine community between 2012-2013. Copyright © 2015. Published by Elsevier Ltd.
Article
Germanium-68 (Ge-68) is produced by proton irradiation of a gallium metal target and purified by organic extraction. The Ge-68 can be used in a medical isotope generator to produce Gallium-68 (Ga-68) which can be used to radiolabel PET imaging agents. The emerging use of Ge-68 in the Ga-68 medical isotope generator has caused us to develop a new purification method for Ge-68 that does not use toxic solvents. The purpose of this work was to develop a production scale separation of Ge-68 that utilizes a leaching step to remove a bulk of the gallium metal, followed by purification with Sephadex© G25. Production scale (300 mCi) purification was performed with the new method. The purified Ge-68 contained the highest radioactivity concentration of Ge-68 produced at BNL; the sample meet Department of Energy specifications and the method had an excellent recovery of Ge-68.