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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. 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