Separation of tungsten from LEU fission produced 99mo solution to improve technological performance in both the processes of 99mo and 99mtc generator production

Separation of tungsten from LEU fission-produced 99Mo solutionto improve technological performance in both the processes Van So Le• Cong Duc Nguyen• Minh Khoi Le Received: 4 August 2014

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Journal of Radioanalytical and

Nuclear Chemistry

An International Journal Dealing with

All Aspects and Applications of Nuclear

Chemistry

ISSN 0236-5731

J Radioanal Nucl Chem

DOI 10.1007/s10967-014-3426-1

Separation of tungsten from LEU fission-produced 99 Mo solution to improve

technological performance in both the

processes of 99 Mo and 99m Tc generator production

Van So Le, Cong Duc Nguyen & Minh Khoi Le

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Separation of tungsten from LEU fission-produced 99Mo solution

to improve technological performance in both the processes

Van So Le• Cong Duc Nguyen• Minh Khoi Le

Received: 4 August 2014

Ó Akade´miai Kiado´, Budapest, Hungary 2014

Abstract Method of W separation from fission-99Mo

solution was studied using alumina column and H2SO4/

HNO3eluents The distribution coefficients of WO42-and

MoO42- ions and the column loading/eluting conditions

were investigated to optimize the separation process 4–6 M

H2SO4 solutions were successfully used to elute/separate

99MoO42- ions from the alumina column which strongly

retained WO42- ions without significant W-breakthrough

The developed W/99Mo separation process is fit for being

in-line incorporated/integrated in the alkain-line dissolution-based

process of fission-99Mo recovery currently used in LEU

target-based99Mo production

Keywords LEU/HEU-target
99Mo-production

W-separation
99mTc
Alumina

Introduction

99Mo is a parent nuclide of the99mTc radioisotope

gener-ator which is used in nuclear medicine world-wide Today

the commercial production of99Mo is mainly based on the

235U fission and98Mo (n, c)99Mo reactions [1] The targets

used in235U-fission reaction are highly enriched uranium

(HEU) and/or low enriched uranium (LEU) The mass of

LEU target is around five times larger than that of HEU

Conversion of current technology using HEU targets

([20 %235U) to those using LEU (\20 %235U) requires a

5–6 fold increase in total uranium content to produce irradiation yields of 99Mo equivalent to current HEU tar-gets Consequently, the W contaminant content of 99Mo solution produced from LEU targets is possibly given in approximately five times larger compared to that produced from HEU target, assuming the same purity of the both types

of the targets So, among different technical solutions involved during the conversion of existing facilities using HEU targets, the reduction of W contamination in LEU target and/or the separation of W from 99Mo solution of LEU process should be addressed The large amount of W contaminant is seriously challenging the performance of the

99

Mo and 99mTc generator production technologies Partic-ularly, the W contaminant in LEU may cause a serious problem regarding the decrease of99Mo-retaining capacity

of the ion exchange resin/sorbent columns which are cur-rently used in both the processes of99Mo stock solution and

99mTc generator production This fact is due to the large amount of W element contaminant coming from a large amount of uranium and aluminium metal cladding used in manufacturing LEU targets Moreover, the high similarity in adsorption property of WO42- and MoO42- ions on the currently used sorbents is also the reason causing the reduction in adsorption capacity of MoO42-ions

The above mentioned situation is also challenging the

99Mo production using natural or 98Mo-enriched molyb-denum targets Natural molybmolyb-denum targets used in98Mo (n, c) 99Mo reactions usually contain much more W con-taminant than LEU and98Mo-enriched Mo targets So the natural Mo targets purified to reduce the W content to an extent of less than 10 ppm are needed to produce a medi-cally useful99mTc solution via98Mo (n, c)99Mo reactions The radioactive isotopes 181W (t1/2= 121.2 days),

185

W (t1/2= 74.8 days), 187W (t1/2 = 23.9 h), and 188W (t1/2 = 69.4 days), which are induced from neutron

Medisotec, Gymea, NSW, Australia

e-mail: vansole01@gmail.com

C D Nguyen

ChoRay Hospital, Ho Chi Minh, Vietnam

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DOI 10.1007/s10967-014-3426-1

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activation of natural abundance W element, will present in

the finished99Mo stock solution and then will be retained

by the alumina column along with99Mo in the99Mo/99mTc

generator as the radioactive contaminants, if the separation

of W from the target or target solution could not be

performed effectively Particularly, the natural W contains

a high content of 186W which will be activated by reactor neutron during the target irradiation to produce 188W via

186W (n, n, c)188W reaction Consecutively, the produced

188

W will generate the radioactive 188Re (t1/2 = 16.94 h) via beta particle decay and this188Re-radionuclide will be co-extracted with 99mTc from the 99Mo/99mTc generator and present as a radioactive contaminant in the finished product of99mTc solution

Different methods have been developed using activated charcoal and/or hydrous tin-dioxide sorbents to remove W from a macro quantity molybdate solution Some of them were described in the protocol of natural molybdenum target preparation for (n, c) reaction-based 99Mo produc-tion [2,3] However, the developed methods do not suite the separation of milligram quantities of W from the solution of a comparable amount of Mo (99Mo), which are currently processed in the LEU target-based 99Mo pro-duction plant The above mentioned methods are neither suitable to a process of fission-99Mo production in which the solutions of small volume are usually used Obvi-ously, a suitable and technologically compatible (in-line) W/99Mo separation method should be developed to sub-stantially overcome the disadvantages of above mentioned ones

Experimental

A99Mo/188W-spiked simulator of the fission-99Mo solution was used for W/Mo separation method development The simulator fission-99Mo solution contains 0.0740 mg Mo/

mL, 0.013 mg W/mL, and several given contaminant ions

99MoO42-and188WO42-solutions used in all experiments were prepared by eluting the used 99Mo/99mTc- and

188

W/188Re-generators, respectively, with 1.0 M NaOH solution followed by neutralization with hydrochloric acid All chemicals used were analytical grade Acidic alumina

of Brockmann 1 supplied by Sigma-Aldrich was used for

elements in the simulator

is based on the ICP-EOS

solution eluted from AG1 9

8/AG MP resin column which

was loaded with an alkaline

digestion-solution of 8 CERCA

LEU-target plates used at

facility.)

chromatographic column packing The

element/contami-nant composition of the simulator solution shown in

Table1 is mimicked based on the ICP-EOS analysis

results of a real99Mo solution sample which was obtained

from chemical processing steps in which an alkaline

dis-solution-processed 99Mo solution of LEU (20 % 235U)

target was passed through a strong anion exchange resin

(AG1 9 8 and AG MP) column to remove the majority of

the fission products The related steps of basic

dissolution-based 99Mo- production process is briefly described in

Fig.6 [4]

Kdvalues of the adsorption of molybdate and tungstate

ions on the acidic alumina in the above mentioned simulator

solution of variable acidity and in the pure HNO3and H2SO4

solutions were measured using the procedure described in

our previous work [5] The separation process was studied

using a chromatographic column of 1.0 g acidic alumina

sorbent, on which a given amount of the simulator solution

was loaded The elution of99Mo-molybdate ions was

per-formed by sulphuric and/or nitric acid solutions of given

acidity which is optimised based on the Kd measurement

results Tungstate ions were striped out of the column with

5 mL 1.0 M NH4OH solution for quantity analysis The

elution fractions of 5 mL were collected and their

99Mo/188W radioactivity was measured using an Ortec gamma-ray spectrometer coupled with HpGe detector Results and discussion

The Kdvalues versus acidity of HNO3and H2SO4solutions are shown in Figs 1and2, respectively The Kdvalues of the tungstate and molybdate ions in both HNO3and H2SO4

solutions have a significant difference for the solutions of [1.0 M acidity This difference of three magnitude orders offers a good opportunity to separate tungstates ions from molybdate ions using HNO3 and/or H2SO4 eluent This statement is confirmed by the separation results reported in Figs 4 and 5 For the reason of compatibility with other steps of the existing fission 99Mo separation process, the Mo/W separation using sulphuric acid solution is preferred The suitable acidity of sulphuric acid solutions used for the W/Mo separation is between 4–6 M The elution of alu-mina column with these solutions can be performed with a Mo-elution yield of [98 % and the tungsten is completely removed from the99Mo solution

and tungstate ions uptake from 1.0 mL sulphate solution (Molybdate

and tungstate concentration are 0.074 mg Mo/mL and 0.013 mg

W/mL, respectively): a W-uptake on the alumina of weights from 33.7

to 100.0 mg per mL solution; b Mo-uptake on 100.0 mg alumina per

mL solution; c Mo-uptake on 67.5 mg alumina per mL solution; d

Mo-uptake on 60.0 mg alumina per mL solution; e Mo-uptake on

of molybdate ions for alumina sorbent in the molybdate solution of variable sulphate concentration (Kd measurement was performed in the

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The Mo-uptake conditions of the alumina column in the simulator solutions were investigated to determine the possible influence of different ions existing in the solution used in a real fission Mo separation process The obtained results reported in Fig.3shows that the W and Mo loading

ca99pability of alumina is high and a small size alumina column can be used to retain almost W-content of the99Mo solution The influence of different ions of the simulator solution on the W-adsorption is insignificant Based on the obtained results it is stated that the separation of99Mo from

W contaminant can be effectively performed using an acidic alumina column and H2SO4eluent

The elution profiles of99Mo/W separation are shown in Figs 4and5 This process is conveniently integrated with relevant steps of the alkaline dissolution-based technology

in the process of99Mo production using LEU targets The relevant steps of this99Mo production process are shown in Fig.6 An additional alumina column-based Mo/W sepa-ration step following the step of AG1 9 8 and AGMP anion exchange resin column separation is proposed to eliminate or reduce W contaminant from 99Mo solution before coming into the CHELEX-100 resin column for further purification This typical design of a chromato-graphic column loaded with 5–10 g acidic alumina and the elution of 99Mo with 50–70 mL 4–6 M H2SO4 solutions can be effectively used to remove more than 80 % of W contaminant content (*100 mg W) from a 1.356 9 105 GBq activity (E.O.B) 99Mo solution (*70 mg Mo) pro-duced using 18.4 g235U LEU-target (8 target plates) neu-tron-activated for 120 h in the OPAL reactor at ANSTO (Australia) The99Mo recovery yield will be [96 % The integration of W/99Mo separation step of above mentioned W/99Mo separation process is fit for being in-line incor-porated/integrated in the alkaline dissolution process-based99Mo recovery currently used in Argentina, Australia, and Germany as described in Fig 6[4] This technological step incorporation is feasible with respect to the compli-ance with operational safety requirements and possible update of licensing procedure

The advantage of alumina use for fission99Mo separation

in the field of high radiation dose is the superior radiation resistance of this sorbent compared with the organic ion exchange resins The use of alumina also conforms to an existing final purification step used the process of fission

99Mo production, thus unnecessary to reapply for a 99Mo production licence The use of alumina sorbent for the study

of W/99Mo-separation is also justified based on the com-pliance with an current production facilities licensed by the authority in the majority of countries of the world

(Column: 1 g alumina, Loading solution: 8 mg Mo ? 10 mg W):

The process of W/99Mo separation from a simulator

fis-sion-99Mo solution was developed using acidic alumina

column and H2SO4 eluent 4–6 M H2SO4 solutions were

successfully used to elute/separate99MoO42-ions from the

alumina which strongly retained WO42-ions without

sig-nificant W-breakthrough The use of acidic alumina sorbent

for the W/99Mo-separation conforms to an existing final

purification step of the fission99Mo production process and

being justified based on the compliance with a production

licence for the existing fission 99Mo-production facility

used in the majority of countries of the world The

devel-oped W/99Mo separation process is fit for being in-line

incorporated/integrated in the alkaline dissolution–based

process of fission-99Mo recovery process currently used in

the LEU target-based99Mo production

References

1 Le VS (2014) Science and technology of nuclear installations,

2 Hetherington EL, Boyd RE, Targets for the production of neutron activated molybdenum-99, IAEA-TECDOC-1065

3 Dadachova K, La Riviere K, Anderson P (1999) Improved processes of molybdenum-99 production J Radioanal Nucl Chem 240:935–938

IAEA-TECDOC-1051

5 Le VS, Morcos N (2008) New SPE column packing material: retention assessment method and its application for the radionu-clide chromatographic separation J Radioanal Nucl Chem 277: 651–661

process of LEU target-based

based on the alkaline

dissolution-based process of

target currently used in

Argentina, Australia and

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