Preparation and characterization of silver coated ...

JEMT 2 (2014) 1-9

ISSN

2053-3535

Preparation and characterization of silver coated alumina for isolation of iodine-131 from fission products A. Mushtaq1*, S. Pervez1, S. Hussain1, M. Asif1, M. Usman Siddique1, J. A. Mirza1, M. M. Khan1, U. Khalid1, M. A. Haq2 and K. Shahzad2 1

Isotope Production Division, Pakistan Institute of Nuclear Science and Technology, P. O. Nilore, Islamabad, Pakistan. 2 Physics Division, Pakistan Institute of Nuclear Science and Technology, P. O. Nilore, Islamabad, Pakistan.

Article History Received 06 February, 2014 Received in revised form 22 February, 2014 Accepted 06 March, 2014

Key words: Iodine-131, Molybdenum-99, Fission products, Silver-coated alumina.

Article Type: Full Length Research Article

ABSTRACT Silver coated alumina was prepared and characterized by scanning electron microscopy (SEM) and X-ray diffraction (XRD). Compositional changes were investigated with energy dispersive system (EDS). A method of Iodine-131 separation from uranium, actinides and fission products was developed. Uranium aluminum alloy target plates were irradiated in the core of Pakistan Research Reactor-1. Iodine-131 was separated as a byproduct of the molybdenum-99 production in Molybdenum-99 Production Facility PINSTECH, Islamabad. After basic dissolution of the targeted material, uranium and actinides along with some fission products were precipitated, while the desired isotopes of molybdenum and iodine remained in filtrate. The filtrate was passed through silver coated alumina column and only iodine was adsorbed on the column. This was desorbed from the column using sodium thiosulfate solution. Nearly 90% radioiodine was eluted. However, for medical grade 131I, distillation is required. ©2014 BluePen Journals Ltd. All rights reserved

INTRODUCTION Silver and its compounds find wide range of applications in our daily life. One of the most effective water purification systems contains silver ions used to kill microorganisms therein. However, silver ions level should be maintained between 0.01-0.1 ppm. A water purification composition including silver and metallic zinc and aluminum has been reported (Denkewicz et al., 2001). Compared to macroparticles, silver nanoparticles have a wide range of applications due to its highly improved properties. These applications include extensive utilization as an antibacterial agent in hygiene and in medical fields, catalyst in chemical reactions and

*Corresponding author. E-mail: [email protected].

as biosensors. Several methods have been used to synthesize and stabilize nano-silver particles such as irradiation, chemical reduction, and thermal treatment. One of the common methods about the synthesis of silver nanoparticles is using a chemically reductive agent (NaBH4) in the pores of a porous material or on the surface of functionalized material (Park et al., 2007; Lee et al., 2007). Porous alumina filters were coated by vacuum deposition with silver metal and used as substrates for surface-enhanced Raman scattering (SERS) (Walsh and Chumanov, 2001). A range of silvercoated or impregnated dressings are now commercially available for use (Margaret et al., 2006). Numerous studies for removal of radioiodine from off-gas stream of nuclear facilities have been performed with various silver impregnated inorganic adsorbents (Pence and Stapples,

J. Eng. Manuf. Technol.

1974; Funabashi et al., 1995; Scheele and Burger, 1987). Different methods have been described for the 131 separation of I from fission products and uranium (Abrashkin and Radicella, 1964; Case and Acree, 1966; Soenarjo et al., 1999; Mondino et al., 1999; Wilkinson et al., 2003; Braker et al., 2002). 131 In a method reported, the removal of I from heated uranium oxide was facilitated by flowing air through the system at 800°C and 131I was collected in a solution containing NaOH (0.2 M) and sodium bisulfite (0.2 M) (Abrashkin and Radicella, 1964). In another method, the irradiated uranium aluminum alloy is dissolved in 4.5 M NaOH followed by wet distillation. 4.5 M sulfuric acid is drop wise added to the filtrate, the solution becomes strongly acidic, at which time Na131I is converted to H131I (Case and Acree, 1966). Yet In another procedure, the irradiated 235U target capsule was connected to a copper wool column to trap the iodine. A mixture of 0.1 M sulfuric acid and concentrated nitric acid (80:5 v/v) was injected into target capsule for dissolving the irradiated target. The radioiodine was distilled into the copper wool column (iodine trap) by heating the target capsule (95°C, 30 min) (Soenarjo et al., 1999). In the reported procedure of Mondino et al. (1999), Al/U targets was transferred to a stainless steel dissolver containing 1.5 L of 3.0 M NaOH with the addition of 3 mg Mo carrier. The precipitate of U was filtered. The filtrate was loaded on a column filled with a bed composed of alternate layers of porous metallic silver and glass microsheres or glass wool. Then it was washed with 100 ml of 1 M NaOH. The elution of 131 I from silver column was performed with 50 ml of 0.2 M Na2S. Wilkinson et al. (2003) described the iodine retention in silver coated alumina and molybdenum non-retention using 131I and 99Mo tracers, respectively. For elution of iodine, 0.2 M and 0.8 M sodium sulfide was used. Radioiodine can be adsorbed on a small column filled with platinum powder from an acidified aqueous solution. With an alternated flow of hydrogen and solvent, the iodine can be desorbed from the platinum into an aqueous or organic solvent (Braker et al., 2002). Pakistan has been producing 131I for 3 decades by wet and dry distillation of natural tellurium or tellurium dioxide irradiated at PARR-1 of PINSTECH. Fission 99Mo is also produced by irradiating U/Al alloy and alkaline digestion and purification by column chromatography. High activity 131 235 of I is also produced during fission of U, which 99 interferes in the Mo separation. In this paper, bulk separation of 131I from uranium, actinides and fission products using a characterized silver coated alumina is reported. The technique of radioiodine separation from Mo-99 is part of fission of Mo-99 in production plant and column chromatography technique was used due to its simplicity, efficiency and easy handling with master slave manipulators.

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MATERIALS AND METHODS The experiments were performed by using guaranteed reagents and deionized water. Acidic alumina, ascorbic acid, silver nitrate and other chemicals were purchased from Merck, Germany. Natural and HEU/Al alloy plate targets were irradiated inside the core of the 10 MW swimming pool type Pakistan Research Reactor-I (PARR-I) for up to 12 h at a neutron flux of >1x1014 cm-2 s-1. After 2 days of cooling, the irradiated target was 99 transferred for separation of fission Mo to Molybdenum99 Production Facility. Preparation of silver coated alumina Acidic alumina (60-200) mesh was washed with deionized water and the fine particles were discarded by decantation, and then dried in an oven. Dried alumina (100 g) was impregnated with 55 ml of 4 M silver nitrate solution. The material was dried again. Afterwards, it was mixed with 100 ml of 2 M ascorbic acid solution heated to 50°C. Both solid and liquid components of this mixture were heated for 15 min. After discarding the supernatant, the material was first washed with hot deionized water and then with cool water up to complete removal of nitrate ions. Finally, the material was heated in an oven at 200°C for 4 h and sieved. Characterization of silver-coated alumina Scanning electron microscopy (SEM), LEO 440i was employed to investigate the morphology of silver coated alumina particles and alumina powder 90 active acidic (MERCK) as received. Operative conditions of the microscopy were kept under applied voltage10 Kevh under high vacuum. Image of the particles of the both samples were taken. Compositional changes were investigated with energy dispersive system (EDS). X-ray diffraction (XRD) using CuKα radiation with wavelength, ʎ= 1.542 Å, was used to characterize the coated and uncoated alumina. The scanning parameters of step time 5 s and step size 0.1 were used over the range of 30-90 degrees. Separation of iodine-131 The neutron irradiated U/Al target plates were dissolved in a mixture of sodium hydroxide and sodium nitrate. Filtration was carried out for separation of uranium residue and solvent containing 99Mo and 131I and some fission products. The filtrate was boiled to remove ammonia and passed through silver coated alumina column. The total amount of silver coated alumina filled

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Irradiation in reactor 3 Al/U target plates

Off-gas, N2, radioactive noble gases

Dissolution and filtration

Solvent: 3 M NaOH, 4 M NaNO3 Water washing

Filter cake: Na2U2O7, ~ 60% fission products

Molybdenum-99, traces of some fission products

Adsorption of radioiodine on silver coated alumina column

Desorption of radioiodine

1 M Na2 S2O3

Wet distillation of radioiodine

Figure 1. Schematic diagram of fission radioiodine separation.

was 80 g, making a bed size of 10 cm height. The column height and internal diameter was 20 cm and 2.8 cm, respectively. Nearly 3300 mL filtrate was passed through silver-coated alumina and washed with 500 mL deionized water. Flow rate was 75 mL/min. Elution of radioiodine was carried out with 300 mL of 1 M Na2 S2O3. Figure 1 shows the schematic diagram for separation of fission 131 I. Determination of radionuclidic purity Radionuclide determination was performed using Highpurity Germanium (HPGe) Detector coupled with Canberra 4096 multichannel analyzer. The characteristic

Gamma peaks of different radionuclides were taken from NuDat 2.5 (IAEA Nuclear Data Services; www.nds.iaea.org). The system was calibrated using point sources of 57Co (122 keV), 137Cs (661 keV) and 60 Co (1332 keV) from Gamma Source Set Model S-13 (Oxford Instruments Inc.). Determination of α-emitter content in the purified radioiodine was performed using an Alpha Counter and a α-standard source of 239Pu.

Determination of radiochemical purity The radiochemical purity was determined by ascending paper chromatography on a piece of 1.5 cm × 25 cm


Table 1. Particle size of alumina and silver-coated alumina.

Particle size (micron) of silver-coated alumina 119 153 118 179 144 139 148 136 139 122 158 135 125 140 153 116 222 129 151 193 165 194 204 189 221 170 Average = 156

Particle size (micron) of alumina 104 111 129 152 94 132 123 68 138 129 111 74 135 137 165 116 147 111 94 195 156 138

Average = 125

Whatman 1 paper using methanol-water mixture (75:25, v/v) as the mobile phase. Iodide and iodate carrier were also used. The chromatogram paper was then dried at room temperature and scanned by the 2 Scanner (Berthold, Germany). The Rf values of different species are given below. Rf value of I- = 0.75 3Rf value of IO = 0.50 Rf value of IO4- = 0.00 RESULTS AND DISCUSSION EDS, SEM and XRD techniques have been widely used to analyze films and coated materials (Pint et al., 2006; Cabanillas et al., 2004; Malinovschi et al., 2006; Passoni et al., 2010). Compositional changes were investigated with EDS. Table 1 gives the particle size of both the coated and uncoated alumina to be 125 and 156 micron

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respectively. SEM image of uncoated as well as Ag-coated Al2O3 particles is shown in Figures 2 and 3. The morphology of both types of powder is round. The round particles have good flow properties. The thickness of coating is observed to be around 15.5 microns from SEM analysis. Inter planner spacing of the XRD pattern were compared with d-value of JCPD cards. Figure 4 shows XRD pattern of Ag coating on Al2O3 particles through chemical deposition at 50°C; whereas Figure 5 shows XRD pattern of Al2O3 powder particles as received. The differences between the Al2O3 d-values and JCPD card values are due to the peak broadening. The peak broadening is also indicative of fine particles size of Al 2O3 powder. Table 2 shows the XRD data and JCPD data of coated and received alumina. Various batches of silver coated alumina were prepared by chemical reduction method and characterized as mentioned previously. Ascorbic acid was used as a reducing agent. The practical yield of silver coated alumina was ~95%. Iodine-131 has played a major role in nuclear medicine as a therapeutic agent. The common sources of 131I are the neutron capture in a 130Te and the fission of 235U. Most of the producers use neutrons capture method, while fission 99Mo producers generally extracted 131I as a byproduct. During fission of U-235, 15 radioactive iodine nuclides are generated; in which eleven radionuclides have half-life of seconds to nearly one hour. Table 3 presents the fission product yields per 100 fissions for 235 U thermal neutron induced fission decay (England and Rider, 1994). The half saturation yield of 131I from fission is ~ 2.2 GBq mg-1 (~ 60 mCi mg-1) of 235U at a neutron flux of 1x1014 n cm-2 s-1, with a specific activity of ~ 0.67 GBq μg-1 (~18 mCi μg-1) at EOB. The theoretical specific activity of 131I is 4.58 GBq μg-1 (124 mCi μg-1); however fission produced 131 I is contaminated with stable 127I and 129I (T½=1.57x107y) as the fission yields leading to masses 127 and 129 are 0.157 and 0.54% respectively, versus 2.89% leading to mass 131. The actual ratio of 131I to 127I and 129I however will be a function of neutron flux, irradiation time and post irradiation decay time (the age of 131 I preparation). Also, fission yields for shorter lived iodine isotopes are substantial (most importantly mass 133 with a yield of 6.70%), requiring post irradiation 131 133 decay to increase the radionuclidic purity of I ( I has a half-life of 20.8 h). Production of 131I by neutron activation of a Te target 131m 131g proceeds through β decay of 30 h Te and 25 m Te intermediate radionuclides. The reported reactor cross section to 131mTe is only 20 mb, more than a factor of 10 smaller than the cross section through the 131gTe path. 130 131 Since the natural abundance of Te is 33.8%, the I 127 produced from a natural Te target always contains I and 129 I created by neutron captures by 126Te and 128 Te

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Figure 2. Scanning electron microscopy image of silver coated alumina.

Figure 3. Scanning electron microscopy image of alumina powder as received (Merck).


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XRDplot PlotofofAg Agcoated Coated Alumina XRD Alumina 2500

Ag

Intensity )AU)

Intensity (AU)

2000

1500

Ag 1000 Ag

Al2O3

500

Ag

Al2O3

Ag

0 30

35

40

45

50

55

60

65

70

75

80

85

90

Angle (2 Theta)

Angle (2 Theta) Figure 4. X-ray diffraction of silver coated alumina.

XRD Plot of Alumina 1000 900

Al2O3

Al2O3

800 Al2O3

(AU) Intensity Intensity (AU)

700 600 500 400 300 200 100 0 30

35

40

45

50

55

60 Angle (2 Theta)

Angle (2 Theta) Figure 5. X-ray diffraction of alumina as received (Merck).

65

70

75

80

85

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Table 2. X-ray diffraction and JCPD data of silver-coated and non-coated alumina.

Al2O3 as-received d-value(Å)

Hk1

JCPD card asreceived d-value(Å)

2.402

110

2.3606

1.983

113

2.068

1.393

214

Ag silver coated Al2O3 d-value (Å) 2.354 2.030 1.987 1.443 1.393 1.23 1.178

111 200

JCPD card Ag d-value (Å) 2.3587 2.0427

220

1.4444

311 222

1.2318 1.1793

Hk1

1.393

Table 3. Independent fission product yields and fission product yields per 100 fissions for

Radio-iodine 129 I 131 I 132 I 133 I 134 I 134m I 135 I 136 I 136m I 137 I 138 I 139 I 140 I 141 I 142 I

Half-life 7 1.57x10 y 8.02 d 2.295 h 20.8 h 52.5 m 3.60 m 6.57 h 83.4 s 46.9 s 24.5 s 6.49 s 2.28 s 0.86 s 0.43 s ~0.2 s

Thermal 3.92x10-3 1.83x10-2 1.65x10-1 5.00x10-1 -1 3.64x10 2.93x100 1.32x100 1.25x100 2.62x100 1.42x100 -1 7.71x10 1.37x10-1 4.07x10-2 5.86x10-3

having natural abundances of 19% and 32%, respectively. Similar to fission produced 131I, the specific activity of Te produced131I not only is dependent on neutron flux, irradiation time and 131I age, but also highly depends on the degree of Te target enrichment. Reported 131I yields from large natural Te metal and oxide -1 targets (20-100 g) range from 22-33 GBq g (600-890 -1 130 mCi g ) of Te irradiated for 5-21 d in a flux of 1x1014 n cm-2 s-1 with a specific activity of ~0.74 GBq μg-1(~20 mCi μg-1) (Roesch, 2003; Handbook of Nuclear Chemistry). 99 131 Figure 6 shows the yield of Mo and I at different 235 irradiation time by irradiation of 4.65 g U at a neutron 14 -2 -1 flux of 1x10 n cm s in the core of PARR-1. After 120 h irradiation which is optimum for generation of fission 99 131 Mo (615 Ci), the activity of I would be 134 Ci. Hence, 131 extraction of I as a by-product during separation of

Fast 1.08x10-3 1.02x10-2 3.84x10-1 7.45x10-1 -1 3.38x10 3.60x100 1.85x100 1.45x100 2.40x100 1.25x100 -1 4.60x10 1.11x10-1 1.69x10-2 2.26x10-3

235

Phases identified Ag Ag Al2O3 Ag Al2O3 Ag Ag Al2O3

U thermal neutron induced fission decay.

Independent 0.003.92 0.01.83 0.165 0.500 0.364 2.93 1.32 1.25 2.62 1.42 0.771 0.137 0.0407 0.00586

Cumulative 2.89 4.31 6.70 7.83 0.364 6.28 2.64 1.26 3.07 1.49 0.778 0.154 0.0407 0.00586

99

main radionuclide of Mo makes the process of fission more economical. Figure 7 shows the elution behavior of 99Mo and 131I on silver coated alumina column. Molybdenum-99 quantitatively passes through the column while 131I is retained almost completely (> 99%). Washing with 99 deionized water removes the traces of Mo from column. 131 Desorption of I with water is negligible. More than 90% 131 I was desorbed with 300 mL of 1 M Na2S2O3 solution. The sorption and desorption mechanism of iodine on silver coated alumina may be explained by the following equations. Sorption = Ag+ + I- → AgI (ppt) Desorption = AgI + 2Na2S2O3 → NaI + Na3[Ag(S2O3)2]

Activity (Ci)


1000 900 800 700 600 500 400 300 200 100 0

Mo-99

I-131

0

24

48

72

96

120

Time (hrs) Time (h) Figure 6. Yield of Mo-99 and I-131 by irradiation of 3 target plates (U-235; 4.65 g) at 1x1014 n cm-2 s-1 with different irradiation time.

25 131

I in Na2S2O3--------------------------------

% Activity

20

15 99

Mo in NaOH and NaNO3----------------

99

Mo in H2O------

10

5

Eluant (L) Figure 7. Elution behavior of Mo-99 and I-131 on silver coated alumina column.

0. 3

0. 26

0. 22

0. 18

0. 14

0. 1

0. 06

0. 02

0. 5

0. 3

0. 1

3. 3

3

2. 6

2. 2

1. 8

1. 4

1

0. 6

0. 2

0

8

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In basic medium, iodine reacts with silver and precipitated as NaI. With the addition of sodium thiosulfate, the sodium iodide is dissolved, whereas silver forms a soluble complex. The half life of 133I is 20.8 h; hence to eliminate its activity in final product of 131I, the elution of column was performed after 10 days cooling. No alpha counts were detected. Gamma spectrometry analysis revealed no other isotope of iodine except 131I, while Ru-103 (T1/2 = 39.35 d) was the main impurity in 131I. It was < 0.5% of 131 I activity while radiochemical purity as iodide was 103 >90%. The removal of Ru and other chemical impurities, such as silver was achieved by wet distillation technique. Conclusion Silver coated alumina was prepared by chemical reduction method. It was characterized by SEM, XRD and EDS techniques. Several batches of silver coated alumina were prepared and used in different batches of 99 Mo/131I separation. Silver coated alumina prepared in isotope production laboratory is selective for quantitative 99 retention of radioiodine. The sorption of Mo on silver coated alumina is negligible. The sodium thiosulfate can be used for elution of radioiodine. However for medical applications, final purification of 131I would be achieved by wet distillation technique. Silver coated adsorbents not only work trapping of gaseous radioiodine but also effective in removal of radioiodine from solutions. REFERENCES Abrashkin S. & Radicella R. (1964). Preparation of Iodine-131 from irradiated uranium oxide. Int. J. Appl. Radiat. Isot. 15:695. Braker A. H., Moert F. P., van der Zwart R. E., Eersels J. L. H. & Herscheid J. D. M. (2002). Adsorption of radioiodine on platinum; a fast and simple column method to obtain concentrated and pure radioiodide in either water or anhydrous solvent. Appl. Radiat. Isot. 57:475-482. Cabanillas E. D., L_opez M., Pasqualini E. E. & Cirilo L. D. J. (2004). Production of uranium–molybdenum particles by spark-erosion. J. Nucl. Mater. 324:1-5. Case F. N. & Acree E. H. (1966). Production of Iodine-131. US patent office, 3, 282, 655. Denkewicz R. P., Rfter J. D., Bollinger M. A., Grenier J. W. & Souza T. R. (2001). Silver self-regulating water purification compositions and methods. United States Patent No. US 6,254,894 B1. England T. R. & Rider B. F. (1994). Evaluation and compilation of fission product yields. LA-UR-94-3106, ENDF-349. Los Alamos National Laboratory. Funabashi K., Kukasawa T. & Kikuchi M. (1995). Investigation of silver impregnated alumina for removal of radioactive methyl iodide. Nucl. Technol. 109:366-372.

Lee J. M., Kim D. W., Kim T. H. & Oh S. G. (2007). Facile route for preparation of silica–silver heterogeneous nanocomposite particles using alcohol reduction method. Mater. Lett. 61: 1558-1562. Malinovschi V., Ducu C., Aldea N. & Fulger M. (2006). Study of carbon steel corrosion layer by X-ray diffraction and absorption methods. J. Nucl. Mater. 352:107-115. Manual for Reactor Produced Radioisotopes (2003). IAEA TECDOC1340, IAEA, Vienna. Margaret I. P., Lui S. L., Poon V. K. M., Lung I. & Burd A. (2006). Antimicrobial activities of silver dressings: An in vitro comparison. J. Med. Microbiol. 55:59-63. Mondino A. V., Kols H. J., Cristini P. R., Furnari J. C. & Radioanal J. (1999). Separation of iodine produced from fission with a porous 99 metal silver column in Mo production. Nucl. Chem. 240:371-374. Park J. H., Park J. K. & Shin H. Y. (2007). The preparation of Ag/mesoporous silica by direct silver reduction and Ag/functionalized mesoporous silica by in situ formation of adsorbed silver. Mater. Lett. 61: 156-159. Passoni M., Dellasega D., Grosso G., Conti C., Ubaldi M. C. & Bottani C. E. (2010). Nanostructured rhodium films produced by pulsed laser deposition for nuclear fusion applications. J. Nucl. Mater. 404:1-5. Pence D. T. & Stapples B. A., (1974). Solid adsorbents for collection th and storage of iodine-129 from reprocessing plant. Proc. 13 AEC Air Cleaning Conference. CONF-740807. Atomic Energy Commission, Washington, DC. Pp. 758-764. Pint B. A., Moser J. L., Jankowski A. & Hayes J. (2006). Compatibility of multi-layer, electrically insulating coatings for vanadium–lithium blankets. J. Nucl. Mater. 352:107-115. Roesch F. (2003). Handbook of nuclear chemistry: Manual for reactor produced radioisotopes. Vol 4, Editor F. Roesch. Kluwer Academic Publisher, London. Scheele R. D. & Burger L. L. (1987). Evaluation of silver modernite for radioiodine retention at the Purex process facility modification. PNL6261. Pacific National Lab. Richland, Washington. Soenarjo S., Gunawan A. H., Purwadi B., Wisnukaten K. & Sriyono A. S. (1999). Analysis of radioiodine fraction separated from production process of fission molybdenum-99. Atom Indonesia. 25(2):1-6. Walsh R. J. & Chumanov G. (2001). Silver coated porous alumina as a new substrate for surface-enhanced Raman scattering. Appl. Spectr. 55:1695-1700. Wilkinson M. V., Mondino A. V., Manzini A. C. & Radioanal J. (2003) Separation of iodine produced from fission using silver coated alumina. Nucl. Chem. 256: 413-415. Www.nds.iaea.org. IAEA Nuclear Data Services, NuDat 2.5.