Thorium-229 for Medical Applications - American Chemical Society

Chapter 13

Thorium-229 for Medical Applications

Downloaded by COLUMBIA UNIV on July 19, 2012 | http://pubs.acs.org Publication Date: November 4, 2003 | doi: 10.1021/bk-2004-0868.ch013

Miting Du, Fred Peretz, Rose A. Boll, and Saed Mirzadeh Nuclear Science and Technology Division, Oak Ridge National Laboratory, O a k Ridge, TN 37831-6385

Medical researchers are assessing the ability of several alpha-emitting radioisotopes to treat cancer and reduce tumor burden. In particular, Bi (t = 45.6 min) attached to tumor specific antibodies has been under clinical investigations for treatment of certain cancers. Bi is a decay product of Ac (t1/2 = 10.0 d), itself an alpha emitter. Currently the domestically-produced supply of Ac, however, is limited by the availability of the parent radionuclide, Th (t = 7340 y). 2 2 9 T h , in turn, is a decay product of highly fissile U, which is currently stored at O R N L and is now scheduled for long-term storage. In this paper we describe the primary separation of Th from U stockpile and further purification o f this radioisotope from trace levels of uranium and plutonium. The chemical separation process is primarily based on the well known chemical behavior of these actinides on anion-exchange resins in both nitric and hydrochloric acid solutions. 213

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© 2004 American Chemical Society In Radioanalytical Methods in Interdisciplinary Research; Laue, C., et al.; ACS Symposium Series; American Chemical Society: Washington, DC, 2003.

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194 2 1 3

2 2 5

The use of the alpha emitters B i and A c is being explored by the medical research community for the treatment o f a variety o f cancers. A n example is the use of B i for the treatment of acute myeloid leukemia proposed by Memorial Sloan-Kettering Cancer Center (/). In this application, the humanized antibody HuM195 is used to target a surface protein on a leukemia white blood ceil, with B i linked to the antibody using a chelating agent (modified D T P A ) . Because of the high linear energy transfer and short range of the alpha particle emitted following the decay of B i , there is a high probability of killing a targeted cancer cell with minimal exposure of other parts of the body to a large dose o f radiation. A supply of T h is maintained at Oak Ridge National Laboratory (ORNL), and its A c daughter isotope (ty = 10.0 d) is extracted on a regular basis and supplied to medical research institutes. Because of its short 45.6 m half-life, B i must be separated from its parent isotope A c at the site o f its use. With its 7,340 y half-life, T h can serve as a source of A c for a long period o f time (Figure 1). A n adequate supply of T h can support basic research and clinical trials for a number of proposed alpha particlemediated radioimmunotherapy. 2 l 3

2 l 3

2 l 3

2 2 9

2 2 5


2

2 1 3

2 2 5

2 2 9

2 2 5

2 2 9

2 3 3

U 1.6E5v α 229

225

Ac 10.0 d

217

Rn 0.5 ms 20!

"Bi Stable

2 , 3

P0 3.7 us 209p

T h

7340v

2 2 1

Fr 4.9 m

k

225

Ra 14.9 d

2 , 7

At 32 ms

213 .

b

B

3.3 h

45.6 m

k

209jj

2.2 m Figure 1: Th-229 Decay Chart (Neptunium Series) 2 2 9

There are two general pathways o f obtaining T h on a significant scale (millicurie levels). One is by neutron transmutation of R a target (successive neutron captures followed by β decays) (2). This approach typically requires a 2 2 6

In Radioanalytical Methods in Interdisciplinary Research; Laue, C., et al.; ACS Symposium Series; American Chemical Society: Washington, DC, 2003.

195 reactor with a high neutron flux and associated facilities for fabrication and postirradiation processing of highly radioactive targets. The second pathway is to separate T h produced by the alpha decay of U . Alternatively, R a and A c can be produced directly by gamma-ray, proton and deuteron induced reactions on Ra targets (2). Separation of T h from U is less complicated than the production by irradiating R a targets, and less expensive especially i f available funding dictates a relatively small program, although, as one may expect, the safeguard of U is rather problematic. About 450 kg of U is stored at O R N L , of which about 250 kg contains relatively low levels of the contaminant U . The U contaminant in U stock was probably produced by U[n,2n] reaction and partly from the decay of ^ ^ u , which in turn was produced by neutron transmutation of T h target. Decay daughters of U include T1 (Figure 2), which emits a 2.6 M e V gamma-ray with high intensity. Thus, containing significant concentrations of U must be handled in a shielded environment (Th separated from U also contains the T h decay daughter of U , thus its 2 2 9

2 3 3

2 2 5

2 2 5

2 2 9

2 3 3

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2 3 3

2 3 3

2 3 2


2 3 3

2 3 2

233

2 3 2

2 3 2

208

2 3 2

2 3 3

2 2 8

2 3 2

M 2

U 68.9 y ^

α

2 2 8

Th 1.91 y 224-

Ra 3.66 d 212r

Rn 55.6 s

Pb Stable 2.8 MeV gamma

i , 2

208JJ

3.1m

2218

Ra 5.75 y

2I6

Bi 60.6 m

Po 0.145 s

/

\

Th

1.4E10 y

Ζ

64% 208

Ac 6.15 h

220

'Po 0.3 us

232

2 M

212

Pb 10.6 h

Figure 2: Th-228 Decay Chart (Thorium Series)


196 2 3 2

radioactivity is also related to the U content of the original uranium). The ORNL U (stored in -1,000 separate packages), has U concentrations ranging from - 2 to -200 ppm, and it has been in storage for up to 35 years. During this priod, significant quantities of T h has grown in (the T h annual in growth rate is -1 mCi per K g of U ) . In April of 2000 the U.S. Department o f Energy (DOE) made $1 million available over three years for the separation of Th from U , to serve as feedstock for the production of additional A c . A set of 19 packages that were recently sent to O R N L from Mound Laboratory, containing 3.3 kg of U with an estimated 72 m C i of T h , were originally identified for separation. This material contains between 2 and 16 ppm of U . A s the project progressed, other packages of U containing about 6 ppm U were identified, and processing of the 16 ppm Mound material was deferred. The amount of T h expected to be present in the revised package sequence remained essentially unchanged, but a variable amount of plutonium, mainly P u , was unexpectly identified, which was not shown on file for the packages (the P u may be a by-product in original V production due to the impurity U inside thorium). A typical process batch contains hundreds of grams of uranium, tens of milligrams of thorium and variable amount of plutonium ranging from milligrams to hundreds of milligrams. Thus, separation and purification of T h from this material became a process of separating a few mgs of thorium and plutonium from a few hundred grams of uranium. 2 3 3

2 3 2

2 2 9

2 2 9

2 3 3

2 3 3

2 2 5

2 3 3

229

2 3 2


2 3 3

2 3 2

2 2 9

239

2 3 9

m

2 3 8

2 2 9

Process for

Th Separation and Purification 2 2 9

2 3 3

The flow sheet for separating T h from U consists of two major processes, Thorium Separation Process and Thorium Purification Process. The Thorium Separation Process identifies the steps for the separation of the thorium and plutonium from bulk of uranium, and for the conversion of uranium back into a stable oxide form for continued storage (Figure 3). The Thorium Purification Process of the flowsheet, lists the steps to further purify the T h product from plutonium and residual uranium contaminants (Figure 4). The Thorium Seperation Process begins by opening packages of U oxide and confirming die uranium content by titrametric method using F e S 0 as reductant (5), and mass analysis (ICP-MS). The oxide (-250 g) is then dissolved in 11 mol/L H N 0 (2.8 m L per gram of U 0 ) , and filtered to remove undissolved impurities. The acid concentration of this solution is then adjusted to - 8 mol/L H N 0 and is loaded onto a 200 m L column containing anion-exchange resin (tertiary amine-based Bio-Rad A G 1 - X 8 , which was later replaced with M P 1) (100-200 mesh, -200 m L bed volume) (4). Nitric acid (7.5 mol/L) is used to elute the bulk of uranium (kj - 14), while T h (k* - 300) and P u (k 1000) 2 2 9

2 3 3

4

3

3

8

3

+ 4

4 4


197 (5) are retained on the column. If significant quantities of Pu are present, a dark green band can be seen near the top of the resin bed. Both thorium (kj - 1) and plutonium (ka « 100) (5) are stripped off the column with 250 m L of 0.1 mol/L H N 0 and then sent for process of Thorium Purification. The bulk of uranium eluted from the anion-exchange column was precipitated as hydroxide by the addition of cone. N H O H , filtered, washed with D I water, and then converted to U 0 by heating in air to 800 °C. Initial batches, with less than 2 ppm U , were processed in a set of glove boxes. As material containing 6 ppm U was handled, most of the process was moved into a shielded manipulator hot cell in the same building. 3

4

2 3 2

3

8


2 3 2

Receive uranium oxides from building 3019 and confirm weight & contents J Dissolution of uranium oxides with 11 M H N O 3 V Adjust solution to 8 M HNO3 and load on an anion exchange column (200 ml)

Bulk U eluted by 8 M H N 0 & then precipitated by fljfEUOH 3

Th & Pu eluted by 0.1 MHNOi

Wash precipitates with 1 M n h 4 o h after filtration

To building 3047 for Thorium Purification

1

Wash precipitates with DI water and then convert to U 0 in air at 800°C 3

*

To building 3019 for storage

8

Figure 3: Thorium Separation Process Flow Chart


198

Stream-2 from Th Extraction (Building 2026) containing Th, Pu and residual uranium t Adjust solution to 7*5 Μ H N 0 and then load onto an MP-1 column (10 ml)


3

Residual U and daughters of 2 2 9 eluted by 7.5 IVI H N 0 Th

yr

3

Th eluted by 10MHC1, dried down and back to 10MHC1

Pu eluted by N H J in 10MHC1

Load onto an MP-1 column (2 ml) to retain trace U & Pu Liquid waste

Liquid waste Th eluted by 10 M HC1 (to Thpool) 229

Figure 4: Thorium Purification Process Flow Chart 2 2 9

In the Thorium Purification Process (Figure 4), T h is separated from plutonium and residual uranium impurities in two consecutive M P - 1 resin columns. These columns are 1 0 and 2 mL bed volumes, and are much smaller than the 2 0 0 m L column used in the Thorium Separation Process. The 0.1 mol/L HNO3 solution containing Th, u and Pu is evaporated to near dryness in a semiclosed distillation apparatus and is then redissolved in a minimum volume (-5 mL) of 7.5 mol/L HNO3. This solution is loaded on a MP-1 anion exchange column ( 1 0 mL bed volume, 1 0 0 - 2 0 0 mesh, pre-conditioned with 7.5 mol/L HNO3) and column is washed with an additional 3 0 mL of 7.5 mol/L HNO3. Similar to the first process, the residual u is eluted with the 7.5 mol/L HNO3 while T h and Pu(IV) are retained on the resin. Thorium is then selectively eluted with - 6 0 mL of 1 0 mol/L H C 1 , while Pu is retained on the resin (Iq - 1 and >


199 1000, respectively, (J,

Thorium-229 for Medical Applications - American Chemical Society

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