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244
Current Radiopharmaceuticals, 2012, 5, 244-252
Production of High-purity Radium-223 from Legacy Actinium-Beryllium
Neutron Sources†
Chuck Z. Soderquist, Bruce K. McNamara and Darrell R. Fisher*
Isotope Sciences Program, Pacific Northwest National Laboratory, Richland, Washington, USA
Abstract: Radium-223 is a short-lived alpha-particle-emitting radionuclide with potential applications in cancer treatment. Research to develop new radiopharmaceuticals employing 223Ra has been hindered by poor availability due to the
small quantities of parent actinium-227 available world-wide. The purpose of this study was to develop innovative and
cost-effective methods to obtain high-purity 223Ra from 227Ac. We obtained 227Ac from two surplus actinium-beryllium
neutron generators. We retrieved the actinium/beryllium buttons from the sources and dissolved them in a sulfuric-nitric
acid solution. A crude actinium solid was recovered from the solution by coprecipitation with thorium fluoride, leaving
beryllium in solution. The crude actinium was purified to provide about 40 milligrams of actinium nitrate using anion exchange in methanol-water-nitric acid solution. The purified actinium was then used to generate high-purity 223Ra. We extracted 223Ra using anion exchange in a methanol-water-nitric acid solution. After the radium was separated, actinium and
thorium were then eluted from the column and dried for interim storage. This single-pass separation produces high purity,
carrier-free 223Ra product, and does not disturb the 227Ac/227Th equilibrium. A high purity, carrier-free 227Th was also obtained from the actinium using a similar anion exchange in nitric acid. These methods enable efficient production of 223Ra
for research and new alpha-emitter radiopharmaceutical development.
Keywords: Alpha emitters, actinium-227, thorium-227, radium-223, anion exchange.
INTRODUCTION
Actinium is one of the rarest, naturally occurring elements, and sufficient quantities of actinium radioisotopes are
difficult to obtain for research. Actinium-227 is the longestlived isotope of actinium (half-life = 21.8 years) and is the
only isotope of actinium that can be isolated in milligram or
greater amounts. Actinium-227 can be obtained in trace
amounts from uranium minerals because it belongs to the
natural 235U decay chain. Pure 227Ac metal has a specific
activity of 2.66x103 GBq per gram. Actinium-227 can be
produced synthetically by neutron irradiation of 226Ra in an
isotope production reactor via the nuclear reaction
226
Ra(n,)227Ra (half-life = 42.2 minutes) 227Ac + -, and
by subsequent separation and purification of 227Ac. The thermal neutron cross section for the 226Ra(n,)227Ra reaction is
about 13 barns (National Nuclear Data Center, Brookhaven
National Laboratory; http://www.nndc.bnl.gov/).
Actinium-227 decays by beta-emission to 227Th (half-life
= 18.7 days), which in turn decays by alpha-emission to
223
Ra (half-life = 11.4 days). The remaining members of the
decay chain have shorter half-lives and grow-in completely
within a few hours. The first daughter of 223Ra is 219Rn, a
noble gas that must be carefully handled since it can cause
airborne alpha contamination in the work space. Most of the
members of the 227Ac decay series have gamma emissions
that may cause high radiation fields. Actinium-227 also decays by alpha decay through a side chain (1.4% abundance)
to 223Fr (half-life = 21.8 minutes). Francium-223 de-
cays by energetic beta (Emax = 1.15 MeV) to 223Ra. This
side chain grows in quickly and gives the actinium chain a
high beta dose-rate. Because of potentially high dose and
high airborne contamination, curie quantities of 227Ac (milligrams) must be handled and processed in a hot cell rather
than in a glove box.
The daughter product 223Ra has important medical applications as a therapeutic radionuclide for cancer treatment.
Commercial availability of 227Ac is poor [1], and only a few
small sources of 227Ac are known to exist in the world. Both
223
Ra and 227Th can be generated from highly purified 227Ac.
Möller et al. [2] explored the use of crystalline hybrid organic/inorganic ion exchangers based on zirconium phosphate and phosphonate compounds for constructing a convenient 227Ac/223Ra generator system. In contrast, we investigated radiochemical separations to produce 223Ra from
227
Ac using anion exchange chromatography in methanolwater-nitric acid solutions.
Study Objectives
The purpose of this study was to develop improved radiochemical separation methods for obtaining high-purity
223
Ra from 227Ac. Medical research on new applications for
223
Ra has been hampered by poor availability of the parent
227
Ac. To improve the availability of 223Ra, we recovered and
purified 227Ac from legacy neutron sources. The long-lived
parent 227Ac decays through a chain of short-lived members
(Fig. 1). The short-lived decay products, after 223Ra decay,
contribute alpha and beta particle radiation and increase the
therapeutic effectiveness of 223Ra in medical applications.
*
Address correspondence to this author at the Isotope Sciences Program,
Pacific Northwest National Laboratory, 902 Battelle Blvd., P7-27, Richland,
Washington 99354 USA; Tel: 509-375-5098; Fax: 509-375-5099;
E-mail: [email protected].
†
Pacific Northwest National Laboratory in Richland, Washington, is operated by Battelle for the U.S. Department of Energy.
1874-47/12 $58.00+.00
Alpha-emitter Attributes for Cancer Treatment
In many instances, alpha-particle radiation excels for
cancer treatment [3] compared to beta emitters and gammaemitters, since short-range alpha particles impart straight,
© 2012 Bentham Science Publishers
Production of High-purity Radium-223 from Legacy Actinium-Beryllium
Current Radiopharmaceuticals, 2012, Vol. 5, No. 3
245
tional chelates and cell-targeted protein carriers [6]. A bifunctional chelate for 223Ra has been developed but has not
been tested [7] in laboratory animals. Alpha emitters should
have reasonably short physical half-lives, and should be
compatible with targeting moiety biokinetics of the protein
carriers to maximize radiation dose to cancer cells and
minimize dose to the remainder of the body. Optimal physical half-lives for cancer uptake and clearance biokinetics
may be on the order of one to six days, and very short halflives (such as the 46-minute bismuth-213) may be too short
for many cancer-treatment applications. For commercial
preparation and shipment to hospitals, physical half-lives
should be longer than two or three days. Decay-chain daughters should have half-lives of a few minutes or less to minimize daughter migration away from target sites. The alpha
emitters should have imageable photon emissions to aid in
gamma-camera imaging and dosimetry for treatmentplanning. Only a few alpha-emitters meet these criteria and
can be provided with excellent availability at reasonable
cost. Toward these objectives, we have studied 223Ra and
227
Th as candidate radionuclides for cancer treatment. Radium-223 and 227Th from 227Ac may have advantages over
other possible alpha-emitter choices in terms of half-life,
relative cost, availability, and chemistry (Table 1).
Fig. (1). Actinium-227 decay chain.
short-range (40 to 70 μm) high-ionization-density (90-150
keV/μm) tracks through targeted cells. In addition, one or
two direct alpha-particles traversing a cell nucleus may be
sufficient to cause cell death, and repair of sub-lethal damage
is minimized. The practical advantages of these properties
distinguish alpha-emitters from beta-particle emitters used in
radiopharmaceuticals for targeted radionuclide therapy [4].
Examples of Clinical Application
Alpha emitters provide effective treatment of cancer metastases, which often lead to cancer progression. For example, the more common beta-emitters used in radionuclide
therapy cannot as efficiently irradiate and sterilize single
cells and small metastatic lesions compared to alphaemitters. Alpha-emitters also minimize the toxicity associated with radiation therapy by focusing a more localized
energy distribution pattern within targeted tissues. By comparison, radiation from beta/gamma emitters extends greater
distances from targeted lesions, and to a greater extent, can
damage normal organs and tissues--thereby limiting the
amount of activity that may be administered. Alpha emitters
are effective in cell-killing at low dose rates and in low tissue-oxygen environments characteristic of cancerous
growths [5]. These properties establish a scientific rationale
for identifying and employing alpha-emitting radionuclides,
having suitable chemical and physical properties, for applications in cancer treatment.
Radium-223 can be administered in the simple, unchelated chloride form (Alpharadin™, Algeta ASA; Oslo,
Norway) for treating skeletal metastases associated with advanced prostate and breast cancer [8]. Thorium-227 can be
complexed by DOTA and other ligands and linked to monoclonal antibodies for cell-targeted cancer therapy. Our Laboratory is also investigating the use of new nanoparticle constructs for 227Th and 223Ra.
Acting as a calcium mimic, 223Ra-chloride targets new
bone growth in and around skeletal metastases [9]. The
short-range alpha particles destroy nearby cancer cells but do
not uniformly traverse pockets of active red marrow in trabecular bone. A treatment involves intravenous infusion of 50
KBq per kg body weight, once per month, for six months;
this protracted infusion scheme provides maximum treatment
benefit without exceeding normal tissue (bone marrow) toxicity. More patients have been treated with 223Ra than with
any other alpha-emitter during clinical trials for Alpharadin™ [10]. About 3000 patients have been treated in 20
The chemical properties of alpha emitters must be such
that they can be administered to cancer patients using bifunc-
Table 1. Physical and Chemical Properties of thorium-227 and radium-223
227
Properties
223
Th
Ra
Advantages
Physical half-life
18.7 days
11.4 days
Shipping over great distances
Total decay energy
34 MeV
28 MeV
High dose to target tissues
(5 alphas)
(4 alphas)
Imageable photons
Lines at 84, 154, and 269 keV from 223Ra
Dosimetry and treatment planning
Availability and production of
parent 227Ac
Single neutron capture on 226Ra in a nuclear reactor, followed by radiochemical separation and purification of 227Ac, from which both 227Th and
223
Ra may be obtained
Lower cost and better availability
than other alpha-emitters proposed
for radioimmunotherapy
Conjugation to radiopharmaceuticals
Binding to chelating agents (such as
DOTA)
Multiple applications
treatment
Administered as the salt or complexed with nanoparticles
in cancer
246 Current Radiopharmaceuticals, 2012, Vol. 5, No. 3
countries, including nine medical centers in the U.S. and
several others in Europe. Patient imaging shows the accumulation of activity in the intestines after clearance of 223Ra
from the blood, indicating that simple cationic 223Ra not
bound to skeleton clears the patient via the small intestine
into bowel contents [11] and is excreted, which represents a
distinct advantage over other radionuclides that clear through
the renal pathway and subject kidney tissues to localized,
high-dose radiation.
In contrast to the use of 223Ra chloride, the delivery of
Ra to cancer cells in radioimmunotherapy for treating softtissue malignancies requires a suitable bi-functional chelating agent. Ionizable calixarene-crown ether ionophores were
found to exhibit high selectivity for radium over lighter alkaline earth metal ions [7]. In addition, calix[4]arene-crown-6
exhibited high kinetic stability for radium in the presence of
competing, serum-abundant metal ions, including sodium,
potassium, magnesium, calcium, and zinc [7]. The combination of high selectivity and high binding constant predicted a
useful chelating agent for radium. However, further work is
needed to develop water-soluble bifunctional chelating
agents having ionizable calixarene-crown moieties at the
lower rim as a radium ionophore, together with an isothiocyanate group at the upper rim for linkage to monoclonal antibodies. Current research at our Laboratory focuses on alternative radium-binding inorganic nanoparticles employing
bifunctional properties. Such constructs should enable further development of 223Ra-labeled antibodies.
223
Thorium-227 has been considered an alternative to 223Ra
in various cancer-treatment applications because 227Th can
be linked to monoclonal antibodies using bifunctional
DOTA chelates. Thorium-227-labeled antibodies decay in
situ to 223Ra and decay-chain daughters. Thorium-227labeled antibodies have been considered for targeting and
treating a variety of cancer types, such as non-Hodgkin’s
lymphoma [12] breast cancer and ovarian cancer [13]. Thorium-227-polyphosphonate compounds have been proposed
[14] as bone-seeking agents for treating skeletal metastases
from breast and prostate cancer.
RADIOCHEMISTRY
Actinium-227 Recovery for Beneficial Re-use
As an alternative to waste disposal, and on behalf of a
small isotope company (AlphaMed, Inc., Acton, Massachusetts), Pacific Northwest National Laboratory (Richland,
Washington) acquired two excess 227Ac/Be neutron sources
from a U.S. oil company to recover 227Ac for beneficial reuse. The neutron sources were encased in multiple layers of
steel. The origin and pedigree of the sources were unavailable and unknown. The source activities were expected a
priori to contain about 44 GBq and 5.5 GBq, respectively,
but lacking documentation, no specific information was
available concerning method of encapsulation, encasement
alloy materials or thicknesses, weld materials, or chemical
forms of the 227Ac and beryllium contents.
In various calculations that follow in this paper, we have
assumed that the total inventory of actinium in the two neutron sources was about 50 GBq. However, we did not accurately assay the actinium in this phase of the work.
Soderquist et al.
Recovery of Actinium from the Neutron Sources
This section describes the method used to open, recover,
and purify 227Ac present in two neutron sources. Source descriptions, chemical forms, and capsule drawing were not
available for either orphan source. The two 227Ac/beryllium
neutron sources were first examined, weighed, photographed, and radiographed to ascertain their construction.
One capsule was clean, one was rusted, and both sources
were welded. The x-ray images showed no details inside
because the images were distorted by the high gamma radiation dose from 227Ac and its decay-chain daughters. The density of the radioactive sources was about 7 grams per cubic
centimeter, implying that the sources were solid steel tightly
enclosing the 227Ac/Be source material.
The intact neutron sources were transferred into a hot cell
and were mounted in a metallurgical saw. The sources were
cut open by successively taking slices off the end, so that the
saw blade would not cut into the beryllium inside. The
sources were constructed of welded cylinders inside welded
cylinders. One source was triply encapsulated; the other was
quadruply encapsulated.
The first 227Ac/Be source contained two pellets in the
core that were too small to mount in the metallurgical saw
using hot-cell manipulators. The two pellets weighed 6.7
grams. Instead of cutting these pellets, we dissolved them in
a sulfuric-nitric acid mixture. The mass of the iron (together
with chromium and nickel) was small enough to handle in
ordinary analytical operations and laboratory equipment. The
two pellets dissolved slowly at first, providing a dark, greencolored solution. After the outer layer was breached, the material inside dissolved within minutes, effervescing furiously.
The beryllium inside these two pellets was metal, rather than
the oxide form. The two pellets were completely dissolved
within one hour, leaving no shell. Sulfuric acid with nitric
acid will readily dissolve titanium, chromium, manganese,
iron, nickel, cobalt, zinc, copper, the lanthanides, and actinium without forming precipitates. Beryllium sulfate is highly
soluble. The green solution looked like and behaved like
dissolved stainless steel.
When we opened the second source, the final cut went
through the end of the innermost capsule and grazed the contents, exposing gray beryllium metal inside. The innermost
capsule, with one end cut off, was added to the green-colored
sulfuric acid solution obtained from the first neutron source.
The beryllium metal bubbled and dissolved in a few minutes.
We weighed the capsule before and after the beryllium metal
dissolved, and the difference was 0.782 g.
Separation of Crude Actinium
When water was added water to dilute the sulfuric acid to
a larger volume, a heavy white precipitate formed. The green
solution was decanted, leaving the white precipitate. The
white precipitate exhibited a high gamma dose-rate, indicating presence of an undetermined radioactive species.
Stainless steel and any other alloys of iron, chromium,
manganese, cobalt, and nickel will not form precipitates in
dilute sulfuric acid. However, lead and silver have sparingly
soluble white sulfates. If the pellets had been soldered with
silver solder, then a white precipitate would form when we
dissolved the pellets in sulfuric acid. The presence of lead
Production of High-purity Radium-223 from Legacy Actinium-Beryllium
and silver would also explain the high dose-rate, since lead
sulfate coprecipitates radium [15]. Kohltoff and Elving [16]
reported that lead sulfate will coprecipitate actinium. Silver
and lead sulfates can be dissolved by conversion to acidsoluble oxides and hydroxides by metathesis with strong
base, then dissolving the product in an acid. Silver converts
to dark brown silver oxide, and lead converts to a white hydroxide. With excess sodium hydroxide, lead dissolves as
sodium plumbite. The sulfate anion goes into solution. Assuming that the precipitate was lead and silver sulfates, we
warmed the precipitate with dilute sodium hydroxide solution. The solids turned a dark brown color. We assumed that
the following chemical reactions occurred:
Ag2SO4 + 2NaOH Ag2O + Na2SO4
(Ag2SO4 is white and Ag2O is dark brown)
+
H2 O
PbSO4 + 2NaOH Pb(OH)2 + Na2SO4 + H2O
(with dilute NaOH only)
The dilute solution was filtered, and the solids and the filtrate were acidified with dilute nitric acid. The dark-brown
solids dissolved on contact in dilute nitric acid. The first
treatment with sodium hydroxide produced a small amount
of white sulfates, and the treatment was repeated to remove
the white sulfates.
To remove the lead and silver from these two solutions
without removing actinium, we bubbled hydrogen sulfide
gas through the solution. A black sulfide precipitate formed
(Ag2S and PbS), which we filtered out of solution. The black
sulfide precipitate had low activity after several days, and
was discarded. We recovered 227Ac from the dark-green sulfuric acid solution and the two filtrates after removing silver
and lead by coprecipitation on thorium fluoride. Each solution was transferred to a polyethylene jar and about 20 milligrams of thorium was added to each. Several mL of concentrated hydrofluoric acid were added to each one, and the solution was allowed to stand for the precipitate to form. The
ThF4 precipitates were filtered out of solution on cellulose
nitrate-acetate membrane filters. This precipitation separates
actinium from beryllium, since beryllium is soluble in hydrofluoric acid solution. The green color (presumably iron,
nickel, and chromium) stayed in solution and was separated
from the actinium.
Current Radiopharmaceuticals, 2012, Vol. 5, No. 3
247
Actinium-227 plus decay
products, sodium, iron,
and other contaminants
Load onto anion exchange resin
in 2M HNO3 + 70% CH3OH
passes through
loads onto column
Ac, Th, Ra, Pb, Bi
Na+
Fe2+
Al3+
Po+4
other
Wash column with
80%CH3OH + 1M HNO3
(column volumes 3-10) Elute Ra2+
223Ra
Wash column
with 8M HNO3
Elute Ac3+
227Ac
Wash column
with 0.5 M HCl
Elute Th4+
227Th
Fig. (2). Process flow-chart for actinium, radium, and thorium separation and purification by anion-exchange in methanol-water-nitricacid solution.
perature to dissolve the aluminum, convert the sulfates to
acid-soluble compounds, and convert the insoluble fluorides
to acid-soluble hydroxides. Thorium hydroxide is highly
insoluble under these conditions, and carries actinium. This
solution was filtered through a membrane filter. The filter
cake then contained all of the thorium and actinium (accompanied by a trace of silver). The sulfate, fluoride, aluminum,
and lead are soluble under these conditions and pass through
the filter with the filtrate.
Thorium carrier was removed by anion exchange. The
filter cake of thorium and actinium hydroxides was dissolved
in 8 M nitric acid and was passed through an anion exchange
column. Thorium loads onto the resin, but actinium does not.
The column effluent with actinium was collected and evaporated dry to yield a crude actinium nitrate.
The crude actinium, coprecipitated with thorium fluoride
on several membrane filters, was taken to a smaller, cleaner
hot cell for final purification and production of 223Ra. The
filters holding thorium fluoride cakes were placed in a
beaker with aluminum nitrate and nitric acid to dissolve the
thorium fluoride. Aluminum nitrate-nitric acid solutions dissolve thorium fluoride, but failed to dissolve the activity on
the filters. The remaining insoluble matter had a high doserate, possibly because some of the white sulfate precipitate
had been decanted with the green-colored solution.
The crude actinium nitrate appeared as a white crust of
crystalline matter evaporated onto the bottom of a beaker.
We expected it to be contaminated with thorium from the
thorium carrier, incompletely removed by anion exchange,
and with aluminum, iron, nickel, lead, silver, sulfate, and
fluoride. These contaminants are separated from the actinium
by anion exchange in methanol-water-nitric acid. Sulfate was
reported to not interfere [36], but the effect of fluoride in the
anion exchange is unknown. The aluminum would presumably take up the free fluoride and prevent it from interfering
with actinium loading, if enough aluminum happened to be
present. Thorium would likewise scavenge fluoride.
The solution with the filters and insoluble debris were recombined in a beaker, evaporated to near-dryness, and boiled
with concentrated nitric acid at length to destroy the filters
and expel fluoride. The wet-ashed residue was dissolved in
0.01M nitric acid, and was then changed to 1M in sodium
hydroxide. The solution was warmed to near-boiling tem-
To produce a chemically pure actinium source, we separated actinium by anion exchange in a methanol-water-nitric
acid solution (Fig. 2). The crude actinium nitrate was dissolved in nitric acid, evaporated to a damp residue, and taken
up in 6 mL of 2M nitric acid in 80% methanol. This solution
was loaded onto a column of AG1-X8, 100-200 mesh, nitrate
Actinium Purification
248 Current Radiopharmaceuticals, 2012, Vol. 5, No. 3
Soderquist et al.
activity in the exhaust stack monitor. The activity appeared
about 30 minutes after evaporations started, and decayed
with a 30-minute half life. The 219Rn has a half-life of only 4
seconds and would be expected to completely decay in the
minute or so transit time for air to move from the hot cell to
the HEPA filter banks. The 30-minute delay and 30-minute
half life indicated that the activity was 211Pb and 211Bi.
Measurement of Actinium
Fig. (3). Bottom of beaker (top-down view) containing 40 mg ultrapure actinium nitrate (56 GBq). This quantity evaporated to dryness
represents a substantial fraction of the world supply of purified
actinium-227, perhaps one of the rarest and most valuable of all
chemical radioelements.
form anion resin, 10 mL column volume, in 2M nitric acid,
80% methanol. After the load solution had passed through
the column, a wash solution of 1M nitric acid, 70% methanol
was added in 1 column volume (10 mL) increments. Each
column volume of effluent was collected separately and
evaporated dry for storage. The column developed gas bubbles during the separation, possibly caused by radiolysis
from the high activity. The bubbles slowed, but did not stop
column flow. After 14 column volumes of wash had passed
through, a clean beaker was placed under the column, and
actinium was stripped with 80 mL of 8M nitric acid. Thorium was then stripped with 60 mL of 0.5M hydrochloric
acid.
The separated actinium evaporated down to a thin white
crust of Ac(NO3)3 (presumably 6-hydrate) on the bottom of
the beaker. Fig. (3) shows actinium nitrate photographed
through a hot-cell window. The eluted thorium evaporated
down to a viscous liquid which solidified into a white mass.
The first ion exchange separation produced 15 fractions
that could, in theory, be used to draw an elution curve for
223
Ra. However, the high dose and high airborne alpha contamination from 219Rn prevented us from handling any of the
vials outside the hot cell. The elution curve would have been
distorted by gas bubbles that formed in the column and by
the chemical contaminants which were present. We evaporated all 15 fractions and the load solution dry for storage.
The residue in the load vial was blue-green. The first several
column volumes had decreasing amounts of black residue.
The last two fractions (column volumes 13 and 14) had
white solids, which may have been the start of the 227Ac elution peak.
The dried vials were capped and placed into a steel can
with a tight fitting lid to prevent escape of radon, to allow
223
Ra to decay off for several weeks. After the 223Ra had
mostly decayed off, we measured 227Ac in the fractions by
chemical separation.
When we evaporated the methanol-water-nitric acid solutions from the ion exchange run, the 219Rn gas that evolved
during the evaporation caused measurable alpha and beta
Actinium mass and activity are difficult to measure-particularly in disequilibrium with its decay chain. The beta
emission from 227Ac has a low (45 keV) end-point energy
and cannot be measured in the presence of alphas and more
energetic betas of the decay chain. Actinium-227 has low
abundance gammas that are impossible to measure in the
presence of the hundreds of gamma peaks from the other
members of the decay chain. It is possible, though not easy,
to measure 227Ac by measuring the gamma emission twice
and calculating 227Ac by ingrowth of the 227Th. Many of the
gamma abundances for 227Ac, 227Th, and 223Fr are not accurately known. Our stock of actinium nitrate amounted to tens
of milligrams and could be weighed, but the amount of
bound water was unknown. A small amount of actinium in
the presence of a large amount of 227Th and 223Ra can be
measured by chemically separating the actinium, such as by
solvent extraction or ion exchange. After separation, the
227
Ac can be measured by gamma by ingrowth of 227Th.
Radiochemical Separation of 223Ra from 227Ac
Direct production of 223Ra from 227Ac requires that 227Th
be left in equilibrium with 227Ac, because the ingrowth time
is about 50% longer if one waits for 223Ra ingrowth after
227
Th ingrowth. Various methods for radiochemical separations of 223Ra and 227Th have been published [17-20]. Preparation of high purity 223Ra and 227Th from 227Ac is chemically straightforward, but is mechanically complex due to
intense alpha and gamma emission. Radiolysis produces reactive species (solvated electrons, hydroperoxide ions,
atomic hydrogen, and free hydroxyl) that can quickly decompose ion exchangers and solvents and interfere with radiochemical separations. Ion exchange resins degrade under
high alpha dose. Radiochemistry must usually be performed
in shielded facilities. Radon-219 must be contained to prevent general-area work place contamination.
Preparation of High Purity 223Ra
To generate an actual 223Ra elution curve, separated actinium was allowed to stand for three days to allow the decay-chain products to grow in, and then we performed a second ion exchange separation identical to the first separation.
After only three days, little 227Th and 223Ra had ingrown, and
the radiation dose-rate was low enough to handle the radionuclides outside the hot cell. No gas bubbles formed in the
ion exchange column. Figure 4 shows the elution curve
measured by gamma spectrometry 58 days after the ion exchange separation. Actinium-227 was not directly measurable because its gammas are too weak in the presence of
other gammas associated with the actinium decay chain. Instead, we measured 227Ac by ingrowth of 227Th. Since thorium does not come off the column under the conditions
used, 227Th in the final product grew in from 227Ac. The 223Ra
Production of High-purity Radium-223 from Legacy Actinium-Beryllium
Current Radiopharmaceuticals, 2012, Vol. 5, No. 3
249
Fig. (4). Elution curve for nitrate-methanol anion exchange 58 days after separation (activity in MBq versus column volume eluted). This
figure shows the number of column volumes (3 to 10) required to remove 223Ra from the column before 227Ac starts to elute (at column volume 16). The elution curve shows a constant, low-level of ingrowth 223Ra from 223Fr over all column volumes.
peak is much smaller than the 227Ac peak (most of which lies
off the right side of the elution curve), partly because the
ingrowth time was short before the ion exchange separation,
and because the 223Ra decayed through five half lives before
counting.
Radium appeared between column volumes 3 and 10,
centered about column volume 6. Actinium began to appear
in column volume 13, but the actinium peak was centered to
the right of column volume 20, off the chart and far from the
223
Ra peak. If the trailing tail of the 223Ra were rejected, then
pure 223Ra could be produced in good yield. The 223Ra was
free of 227Th and 227Ac.
The elution curve shows the 223Ra after five half-lives of
decay. The 223Ra peak area decay-corrected to the separation
date was 662 MBq. The leading tail of the 227Ac peak shown
in the plot (column volumes 13 to 19) represented 38.8 MBq
227
Ac, or about 0.09% of the total actinium.
The elution curve has a constant, low level of 223Ra
(which is too small to show on this chart) from decay of
223
Fr. About 1.4% of the 227Ac decays by alpha to 223Fr,
which has a 22-minute half-life, and which will grow in during the ion exchange separation. The 223Fr decays by betaemission to 223Ra.
Radon gas quickly diffuses through plastics. Dried fractions from the ion exchange run were stored in glass liquid
scintillation vials with polyethylene liners in polyvinyl chloride caps. Radon will diffuse through the caps and cause
alpha contamination on the outside of the vials. Our standard
practice for gamma counting is to tape each vial around the
cap with vinyl tape, place the vial in a small polyethylene
bag, and tape the bag shut. When the 223Ra fractions were
counted this way, alpha activity appeared on the outside of
the glass, under the bag, within about a day. Counting vials
are stored in a steel slip-lid can. After storing these vials for
a few days, alpha activity appeared on the inside of the steel
can. Since 219Rn has only a 4-second half-life, but appeared
outside the vial, its diffusion rate through polyethylene was
rapid.
Analysis of Actinium in Elutions
Elutions from each ion exchange separation were analyzed for actinium by extraction into thenoyltrifluoroacetone
(TTA) in xylene [21]. The dried effluents from the ion exchange fractions were dissolved in dilute nitric acid, combined to make fewer samples, and adjusted to pH 2. Each
sample was shaken with 0.25M TTA in xylene to extract
thorium, and the xylene layer was then removed. The pH
was readjusted to 6.0 and fresh 0.25M TTA-xylene was
added to extract actinium. The xylene layer was transferred
to a clean vessel and back-extracted with 0.1M nitric acid to
recover the actinium. The actinium extract was removed
from the hot cell for gamma counting. The 227Ac activity was
measured by ingrowth of 227Th.
In the first ion exchange (to purify the crude Ac(NO3)3),
we measured 2690 MBq 227Ac in the load solution (about 5%
of the assumed actinium inventory), and 5610 MBq 227Ac in
column volumes 1 through 14, combined (about 11% of the
assumed actinium inventory). Column volumes 13 and 14
had white solids, which may have been the beginning of the
actinium elution peak. We stopped the separation at column
volume 14. Chemical contaminants in the column may have
caused the actinium to elute early.
In the TTA extraction of the second ion exchange product, we obtained 0.0121 MBq 227Ac in column volumes 1
through 10 (223Ra peak), and 27.1 MBq 227Ac in column volumes 11 through 19 (227Ac peak). The second ion exchange
produced highly pure 223Ra. The 0.0121 MBq of 227Ac found
in the 223Ra peak (column volumes 1 through 10) may have
been hot cell contamination. The TTA extraction yielded an
incomplete actinium recovery. The integrated area of the
second peak (the 227Ac peak, column volumes 11 to 19) was
38.8 MBq 227Ac, but the TTA extraction recovered only 27
MBq 227Ac.
Separation and Recovery of Thorium
The anion exchange separation that we used to produce
Ra can also be used to produce 227Th, with small adjustments. If one omits methanol and adjusts the nitric acid con223
250 Current Radiopharmaceuticals, 2012, Vol. 5, No. 3
centration to 7.5M, then thorium will load onto the anion
exchanger and will allow actinium and radium to pass
through. We used this chemistry to remove thorium carrier
from the actinium stock. This separation can also be used to
separate pure 227Th. The pure actinium can be dissolved in
7.5M nitric acid and passed through an anion exchanger as
described above. Thorium-227 alone loads onto the resin.
The 227Ac accompanied by the remainder of the decay chain
appears in the column effluent. Thorium is eluted with 0.5M
HCl. The thorium separation can be repeated to generate
high-purity 227Th, free of parent 227Ac.
Anion Exchange in Methanol-Water-Nitric Acid
Anion exchange in nitrate solution represents a versatile
separation for the light lanthanides and many of the actinides. This separation retains actinium and thorium on the
column, but allows radium to be eluted. In this method, the
actinium/thorium sample is dissolved in methanol with 20%
water and 2M in nitric acid. The sample is them loaded onto
a nitrate-form anion exchanger. The tetravalent actinides
(including thorium) load firmly onto the resin as anionic
nitrate complexes and do not move down the column. Actinium and the light lanthanides also load, but not as tightly,
and move slowly down the column. Most other elements,
including radium, pass through the column faster than actinium and the light lanthanides. The anion exchange method is
a convenient way to chromatographically separate the light
lanthanides. One advantage of this method is that the load
solution and eluting agent are completely volatile and leave
no residue.
Soon after commercial anion exchange resins became
available, thorium was observed to load from aqueous nitric
acid [22-26]. Other metals, particularly the lanthanides, were
found to load onto an anion exchanger from a weakly acidic
solution of a nitrate salt such as lithium nitrate or aluminum
nitrate [27-34]. A number of chemists developed chromatographic separations that used a nitrate salt solution. Danon
[28] used anion exchange in LiNO3 solution to separate actinium from lanthanum. The disadvantage of the lithium nitrate method is the high concentration of nitrate salts in the
eluted product. Other investigators found that metals load
more tightly to the anion exchanger if a water-miscible organic solvent is added. Thorium was found to load more
tightly from a light alcohol-nitric acid solution than from
aqueous nitric acid alone [35, 36]. Uranium was also found
to load more tightly to the resin if an alcohol is added [37].
Metals frequently load more strongly from a mineral acid
mixed with any of many organic solvents [38-42]. The
choice of organic solvent makes only a minor difference, and
results are similar using any light alcohol or acetone.
The lanthanides, especially the lighter elements, load
onto an anion exchanger from an ethanol- or methanol-nitric
acid solution with useful differences in distribution coefficient from one lanthanide to the next [36, 43, 44]. The alcohol-nitric acid solution permits convenient chromatographic
separation of the early lanthanides [45-47]. These separations
perform as well as other chromatographic separations for the
lanthanides, but use only volatile reagents. Other common
chromatographic methods for separating the lanthanides use
cation exchange with non-volatile complexing agents such as
EDTA, citrate, or -hydroxyisobutrate, which must be re-
Soderquist et al.
moved from the separated product (such as by wet-ashing in
concentrated nitric acid). Anion exchange in methanol with a
low concentration of water and nitric acid gives a product
that evaporates down to carrier-free, essentially massless,
trace-level radionuclides. The concentration of nitric acid
used in these separations is too low to oxidize the alcohol.
The eluted product evaporates quietly dry without oxidation,
even boiling on a hot plate.
Investigators have tabulated distribution coefficients for
metals in alcohol-nitric acid solutions [45, 48-51]. Thorium
and the other tetravalent actinides have the highest distribution coefficients, followed by the trivalent actinides and the
early lanthanides. Lead and bismuth have distribution coefficients comparable to the early lanthanides. Other elements,
including aluminum, iron, cobalt, nickel, copper, zinc, and
silver, have much lower distribution coefficients, and can be
removed from the column by washing. Most authors do not
mention actinium, with one exception [52]. We found that
actinium behaves like the early lanthanides. Guseva et al.
[52] used an ion exchange separation to construct a 223Ra
generator, similar to the process that we describe in the following section.
The heavier alkaline earth elements also load onto the
column, but not as well as the early lanthanides. The order of
distribution coefficients is [Be, Mg] << Ca < Sr < Ba [51].
This property has been used to separate magnesium from
calcium [53], and calcium from strontium [54]. We found
that radium elutes well before actinium.
Quality Assurance and Quality Control
Quality assurance and quality control are essential to ensuring delivery of safe radioisotope products for human use
and for compounding or radiolabeling as the products are
developing into human-use radiopharmaceuticals. The process chemistry described in this paper has not been reduced to
practice in the form of a shielded, compact generator system
for use by radiopharmacists. The relatively long half-lives of
227
Th (18.7 days) and 223Ra (11.4 days) argue for direct distribution of these two radionuclides from centralized laboratories rather than from distributed compact generator systems. This paper shows that the process chemistry is straightforward, but that the mechanical details and steps involved
are more complex than those involved in simple generator
elution. A standard generator based on our process chemistry
would require 219Rn control, a means for removing solvent
from the product, and a means for storing 227Ac parent between generator milkings.
No industry standard for 227Ac breakthrough has been described in the literature. Actinium contamination in the 223Ra
product cannot be measured directly because the 223Ra gammas interfere with the measurement. However, one may control product purity and maintain a very low level of 227Ac in
the 223Ra product. To measure actinium, radium first must be
removed. In the event that one were to find 227Ac in excess
of one part per million (radioactivity), an additional an ionexchange separation step would be performed. Conditions
for radium-actinium separation were selected to place the
radium and actinium elution peaks apart, so that the tails of
the peaks had negligible overlap to prevent contamination of
223
Ra with parent actinium. The final ion exchange separa-
Production of High-purity Radium-223 from Legacy Actinium-Beryllium
tion is relatively fast and can be repeated to remove 227Ac to
non-detectable levels. Chemical contaminants in the final
radium product are best controlled using high-purity reagents
and non-contaminating vessels, such as quartz glassware.
Thorium-227 and 223Ra produced by the process described in
this paper were free of other radionuclides (other than natural
decay products) and non-radioactive contaminants.
SUMMARY AND CONCLUSIONS
223
Current Radiopharmaceuticals, 2012, Vol. 5, No. 3
[4]
[5]
[6]
[7]
227
The improved availability of Ra and Th will enable
research to develop new radiopharmaceuticals using these
radionuclides. On this project, we recovered, isolated, and
purified approximately 50 GBq from 227Ac from two legacy
actinium-beryllium neutron sources.
An ion exchange in methanol-water-nitric-acid method
was demonstrated for fast, carrier-free separation of actinium, thorium, and radium. This anion-exchange method uses
completely volatile eluting agents so that carrier-free 223Ra
and 227Th can be readily separated from parent 227Ac. A highpurity, carrier-free 223Ra may be separated from the parent
227
Ac by anion exchange with little contamination from 227Th
or 227Ac. This separation constitutes an economical, singlepass 223Ra production method that does not disturb the
227
Th/227Ac equilibrium. After 223Ra separation, the 227Ac can
then be eluted off the anion resin to avoid problems from
high radiation dose to the resin.
Methanol-water-nitric-acid anion-exchange uses gravityfeed columns but no mechanical pumps or other highly specialized equipment. This approach is relatively simple and
less costly than other typical ion-exchange methods. However, radon-219 control is essential to reduce the spread of
alpha contamination in the laboratory. High-purity 227Th may
also be prepared by this method. Radium-223 and 227Th are
now available from our Laboratory when purchased through
the Department of Energy’s Isotope Program.
ACKNOWLEDGEMENTS
This research was supported by the U.S. Department of
Energy, Office of Science, Office of Nuclear Physics, Isotope Development and Production for Research Applications
Program under Contract DE-AC05-76RL01830.
[8]
[9]
[10]
[11]
[12]
[13]
[14]
[15]
[16]
[17]
[18]
[19]
[20]
CONFLICT OF INTEREST
Declared none.
[21]
[22]
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