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8. YTTRIUM-90 PRODUCTION AND RADIOPHARMACEUTICALS DEVELOPMENT

8.1. INTRODUCTION

Yttrium-90 (T1/2 = 64.1 h, Eβ(max) = 2.28 MeV) is a pure β– particle emitting radionuclide with well established applications in targeted therapy. The high energy β– particles of 90Y are suited for treating solid tumors as well as for ra-diation synovectomy of large joints. The feasibility of production of this radio-nuclide from a 90Sr/90Y generator makes 90Y one of the readily available radionuc-lide in suffi cient quantities. The parent radionuclide 90Sr, which is a long-lived fi ssion product waste, is available in large quantities from spent fuel [1].

There are several well established radiopharmaceuticals based on mono-clonal antibodies, peptides and particulates, which are labeled with 90Y, that are in regular use for the treatment of some forms of primary cancers and arthritis. At present there are no generators for the elution of 90Y that can be set up in a hospital radiopharmacy. The radionuclide is procured from manufacturers and the radiopharmaceuticals are formulated on site. The 64 h life of the 90Y puts constraints in its transportation and hence 90Y radiopharmaceuticals are cur-rently used only in limited number of nuclear medicine departments that are in Europe or in Northern America. The exceptions are the use of 90Y particulates and colloids for the treatment of hepatic cancer and arthritis for which ready to use radiopharmaceuticals are available.

This chapter summarizes the work done by the author along with his col-leagues on the development of 90Sr/90Y generator and the development of 90Y radiopharmaceuticals [2-8].

8.2. INORGANIC CHEMISTRY OF YTTRIUM

Yttrium (Z = 39) is the fi rst d-block element and belong to group 3 and period 5 of the Periodic Table. The ground state electronic confi guration is [Kr]4d15s2. Elemental yttrium was fi rst isolated in 1828 by Friedrich Woehler and named after the village Ytterby in Sweden where the mineral ytterbite was found. Yttrium has no known biological role. Chemically, yttrium resembles with lanthanides particularly with terbium and dysprosium. The chemical similarity of yttrium with lanthanides is attributed to their similar atomic radius, which is due to the lanthanide contraction phenomenon. Yttrium exclusively exists as

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trivalent ion in its compounds. Yttrium forms various inorganic compounds and complexes with organic chelates in +3 oxidation state. Yttrium is mono-nuclidic with the only stable isotope being 89Y. More than thirty radioisotopes of yttrium have been reported with mass number from 76 to 108. Yttrium iso-topes are common products of the nuclear fi ssion. The 91Y, 87Y and 90Y, with half-lives of 58.51 days, 79.8 h and 64 h, respectively are the long-lived iso-topes. The radionuclides, 86Y and 90Y are useful in nuclear medicine, the former as a PET radionuclide and the latter as a therapeutic radionuclide.

8.3. PRODUCTION OF 90Y

8.3.1. DIRECT (n,γ) ACTIVATION OF 89YYttrium-90 can be produced by the irradiation of 89Y in natural yttrium target, 89Y(n,γ)90Y. The major disadvantage associated with this production route is the very low activation cross-section (1 millibarns) of 89Y which requires the use of high fl ux reactors for production of 90Y with reasonable specifi c activity. However, large quantities of 90Y can be produced by increasing the amount of the target. As yttrium is mononuclidic, there is no question of further enriching the target. The radionuclidic purity of (n,γ) activated product is generally very high. However, depending on the epithermal fl ux in that particular reactor used for production 90Y, detectable levels of 89Sr could be present, due to the (n,p) reaction on 89Y, which need to be carefully ascertained before using the 90Y for radiopharmaceutical preparation. Being a pure beta emitter, conventional gamma ray spectrometry cannot be used to detect the presence of 89Sr impurity in 90Y.

The 90Y produced by (n,γ) activation method in low fl ux reactors can be used for the preparation of particulate based radiopharmaceuticals for treating hepatic carcinoma and radiosynovectomy. However, the low specifi c activity 90Y thus produced is not suitable for labeling antibodies or peptides.

8.3.2. THE 90Sr/90Y GENERATORA more favorable route for the production of 90Y for therapeutic applications is through the decay of 90Sr. The decay scheme of 90Sr to 90Y is shown in Fig. 8.1. The 90Sr decays completely by β– decay with no accompanied gamma radiation (T1/2 = 28.78 y, Eβ(max) = 0.546 MeV). Likewise 90Y also decay by β– emission and with no accompanied gamma emission (T1/2 = 64.1 h, Eβ(max) = 2.28 MeV). The decay product of 90Y is 90Zr which is stable.

Strontium-90 is one of the major fi ssion products of 235U with a fi ssion yield of 5.93%. The annual world production of 90Sr in the nuclear reactors amounts to several hundred mega curies. Stable 88Sr as well as long-lived 89Sr (T1/2 = 50.5 d) are also produced during 235U fi ssion. However, 89Sr will decay and will not be present in the 90Sr stock which is generally isolated from fi ssion product waste in the fuel reprocessing facilities. The presence of 88Sr reduces the spe-

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cifi c activity of 90Sr produced. Strontium-90 with specifi c activity 20-50 Ci/g can be commercially obtained. Due to the long physical half-life of 90Sr, a single batch of it can be used for several years to repeatedly isolate 90Y if used in a generator. This is an important factor making 90Y a widely available radio-nuclide. 8.3.2.1. Problems in the development of 90Sr/90Y generatorsIn order to develop a 90Sr/90Y generator there are several issues to be addressed, some related to the separation and some related to the quality control of 90Y. A detailed discussion on the above problem can be seen in the IAEA book ‘Therapeutic radionuclide generators: 90Sr/90Y and 188W/188Re generators’ edited by the author [9].

Despite the convenience of getting 90Y from a 90Sr/90Y generator, a readily usable generator system is not currently available. One of the major problems in making 90Y is the diffi culty to develop a 90Sr/90Y generator that will give 90Y with adequate radionuclidic purity (99.99%).

The separation of 90Y from 90Sr is complex and requires sophisticated chemistry to achieve the level of purity needed for 90Y in order to use it as a raw material for radiopharmaceutical production. The current price of 90Y is 15 euros per mCi and hence there is good business potential for 90Y. Considering the commercial importance of 90Y, a number of internationally valid patents on 90Sr/90Y have been issued as late as 2006 [10-13].

A review on the methods of separation of 90Y from 90Sr described in the above patents reveals that the technologies are not simple enough to be used on a regular basis in a centralized radiopharmacy or a small radiopharmaceu-ticals manufacturing site. In the methods described in the above patents, 90Y is separated from 90Sr by using a combination of several separation techniques such as precipitation, solvent extraction, ion exchange chromatography or by the use of ion selective resins. Most of the methods are cumbersome, requiring handling of large quantities of 90Sr activity in small volumes and involving several steps. The methods also use large amounts of reagents such as acids, ion exchangers etc., thereby introducing the possibility of metallic impurities

Fig. 8.1. Decay scheme for 90Sr/90Y

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and making the 90Y unsuitable for radiopharmaceutical applications. Large quantities of radioactive waste, containing 90Sr, is also generated in the above methods which also require proper handling.

The development of a 90Sr/90Y generator based on an ion exchange column which can be used in hospital radiopharmacy/central radiopharmacy was at-tempted, but with limited success [14-18]. None of these systems can be adapted to make a generator usable in a hospital radiopharmacy of a nuclear medicine department.

The radionuclide generators (e.g. 99Mo/99mTc generator) used in nuclear medicine departments are based on column chromatography as it is easy to elute the daughter radionuclide in ready to use form and leave the parent radio-nuclide in the column, where it decays facilitating subsequent elutions of the daughter activity. The shelf-life of the generator is determined by the half-life of the parent radionuclide. Because of the long half-life of 90Sr, the technology required for the fabrication of 90Sr/90Y generators is considerably different from that used for other generators with short-lived parent radionuclides.

As is the case with the column generators holding short-lived parent radio-nuclides, 90Sr cannot be left in the 90Sr/90Y generator column for prolonged period of time. The high energy β– particles emitted during the decay of 90Sr and 90Y gradually damage the column matrix reducing their capacity to hold 90Sr. Such generators with damaged column matrix will lead to larger than ac-ceptable levels of 90Sr breakthrough in the eluate. Due to this probelm, 90Y production is now confi ned to radiochemical processing laboratories.

Commercial manufacturers separate 90Y from 90Sr using a combination of separation techniques such as precipitation, solvent extraction, ion exchange chromatography or by the use of ion selective resins. These operations are car-ried out at radiopharmaceuticals manufacturing facilities and after quality control the 90Y produced is shipped to the nuclear medicine departments. Due to the short half-life of 90Y, substantial loss of activity is inevitable during transportation.

The development of 90Sr/90Y generator that can be operated at users end or in a central pharmacy could increase the availability of 90Y. An IAEA Coordi-nated Research Project (CRP) ‘Development of generator technologies for therapeutic radionuclides’ was initiated in the year 2004 upon his joining the International Atomic Energy Agency as the Technical Offi cer responsible for radioisotopes and radiopharmaceuticals program. Prior to joining IAEA, the author together with his colleagues at the Bhabha Atomic Research Centre had already made a good start on this program.

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8.4. DEVELOPMENT OF NOVEL 90Sr/90Y GENERATOR SYSTEMS

Two novel generator technologies were developed that are useful for the preparation of 90Y with high radionuclidic purity and these are described be-low.

8.4.1. SUPPORTED LIQUID MEMBRANE (SLM) BASED 90Sr/90Y GENERATOR [2]Ramanujam and his colleagues from the fuel reprocessing program of Bhabha Atomic Research Centre were interested in separating useful fi ssion products from radioactive waste and developed a technology for the separation of 90Sr from PUREX waste [19]. Considering the importance of 90Y as a therapeutic radionuclide, a fruitful collaboration was initiated in the late nineties between the author’s group and the scientists of the above group in order to separate 90Y from 90Sr stock solution which was initially done using extraction chromato-graphy technique [20]. A 90Sr/90Y generator using SLM was one of the fruitful outcomes of this collaborative program [2, 21].

The schematic diagram of the single stage SLM generator employing KSM-17 is shown in Fig. 8.2. The SLM technology involves the use of a semi--permeable membrane with pores where an ion specifi c chelating ligand can be held. The ligand loaded membrane is used to separate the feed and the re-ceiver compartments. The ligand, KSM-17 (bis(ethyhexyl)phosphonic acid, equivalent to PC88A) used as the ion specifi c chelating ligand was synthesized earlier at the Bhabha Atomic Research Centre [22].

The generator consists of two chambers (feed and receiver chambers) of capacity 5 ml each. A PTFE membrane is impregnated with KSM-17 by im-mersing in concentrated solution of KSM-17 for about 12 h prior to use. The membrane is then removed from the KSM-17 solution, washed with deionized water and inserted between the feed and receiver chambers. The feed chamber contains 5 ml of an equilibrium mixture of 90Sr and 90Y adjusted to pH 1-2 and the receiver chamber holds 5 ml of 1 M HCl. The solutions in both the compart-ments are stirred with a small PTFE coated magnetic stirrer bars for about 4 h.

Fig. 8.2. Single-stage SLM generator

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Typically, 40-50% of the 90Y present in the mixture is transported in 4 h to the receiver compartment. If kept for longer, the yields are better but the chances of 90Sr contamination in the product also become higher. The 90Y solution transported into the receiver chamber is withdrawn and used for radiolabeling studies.

The single-stage generator was subsequently modifi ed by the same group to a ‘two-stage SLM generator’. The two-stage SLM generator system is based on the metal extraction properties of KSM-17 and CMPO (octyl(phenyl)-N,N-di-isobutylcarbamoylmethylphosphine oxide) under different acidity conditions. In this system, the generator consists of three chambers viz. feed, intermediate and receiver chambers. A PTFE membrane impregnated with KSM-17 is used to separate the feed and intermediate compartment. Another PTFE membrane impregnated with CMPO is inserted to separate the intermediate and the re-ceiver compartment. The generator is now routinely used for the preparation of 90Y for development of radiopharmaceuticals. The details of the generator can be accessed from the published paper [23].

8.4.2. ELECTROCHEMICAL 90Sr/90Y GENERATOR (KAMADHENU) [7]8.4.2.1. Genesis of ‘Kamadhenu’Kamadhenu (kama-dhenu – ‘wish-cow’ is a miraculous cow in Hindu mythology, who could give her owner whatever he desired) is an automated electrochem-ical 90Sr/90Y generator suitable for daily milking of 90Y in the form of radio-chemical grade solution. Kamadhenu is developed for Isotope Technologies, Dresden, Germany, by Elex Commerce, Serbia. The genesis of ‘Kamadhenu’ is the culmination of a series of interesting events and collaborative efforts.

The electrochemical method followed for the separation of 86Y, a PET radionuclide formed by (p,n) reaction on enriched 86Sr, by Machula’s group [24] was brought to the author’s attention by Dr. Josef Comor of Vinca Institute, Serbia during one of his visits to the IAEA in August 2006. During that period of time Dr. Comor was associated with the cyclotron group at Geneva to auto-mate the production of 86Y prepared by the above method. The author at that time was coordinating the IAEA CRP ‘Development of generator technologies for therapeutic radionuclides’ and one of the objectives of the CRP was to develop 90Sr/90Y generator. The conversation with Dr. Comor was a spark to the author to think about the applicability of the electrochemical technology for the separation of 90Y from 90Sr. The author’s colleagues, Rubel Chakra-varty, Ashuntosh Dash and Meera Venkatesh at the Bhabha Atomic Research Centre, Mumbai, were willing partners to establish the proof of principle and develop the electrochemical methodology. Though the team was enthusiastic to undertake the work, they had little experience in electrochemistry and hence the challenge was not only to demonstrate the technology but also to have hands on experience on radionuclide separation by electrochemistry. Thanks to their untiring efforts, the technology was developed at laboratory scale in a short

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period of time and published in early 2008 [7]. The electrochemical generator developed at the Bhabha Atomic Research Centre, Mumbai, is described below.

Figure 8.3 shows the schematic diagram of the electrochemical generator. An open ended quartz cylinder (50 ml capacity) was used as the electrochem-ical cell. The electrodes were fabricated by spot welding platinum plates of size 75 mm (l) × 10 mm (b) × 0.5 mm (thick) to platinum rods. The electrodes are then fi xed on an acrylic cap which is fi tted on the mouth of the quartz cylinder. Provision for fl ushing gas through the cylinder is made and an outlet (~2.5 mm) is provided in the acrylic cap for venting the gases.

Strontium-90 as 90Sr(NO3)2 in 2 M HNO3 is used as the electrolyte. The electrolysis cell is covered with the acrylic cap together with the two platinum electrodes. Argon gas is gently bubbled through the solution for 10-15 min prior to electrolysis. Electrolysis is performed for 90 min applying a constant potential of -2.5 V using a potentiostat. Upon completion of the electrolysis, the acrylic cap along with the electrodes is removed from the cell while main-taining the potential. The current is then switched off, the cathode plate is re-moved from the acrylic cell and washed with 10 ml of acetone.

Purifi cation of the 90Y deposited in the cathode is carried out by a second electrolysis. A fresh quartz cell of the same dimension as the fi rst electrolysis cell is fi lled with 30 ml of nitric acid at pH 2.5-3. The cylindrical electrode is connected as the cathode and the original platinum plate containing 90Y is con-nected as the anode. Electrolysis is performed for 45 min under a constant potential of -2.5 V and 100 mA current while continuously bubbling argon gas through the solution. At the end of electrolysis, the acrylic cap is removed from the solution while maintaining the potential.

Fig. 8.3. Schematic diagram of the electrolysis cell and assembly used for 90Sr/90Y generator

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The cylindrical platinum electrode with the deposited 90Y is washed with 10 ml of acetone, followed by leaching of 90Y with 500 μl of acetate buffer at pH 4.75. This solution can be used as such for the preparation of radiopharma-ceuticals.

The electrodes are washed with 3 M HNO3 and can be used for subsequent separations.

The feed solution in the fi rst electrolytic cell, containing the 90Sr activity, is preserved for subsequent electrolysis after allowing suffi cient time for the growth of 90Y activity. One half-life decay time (64 h) will result in 50% growth of the 90Y activity. The electrolyte in the second cell as well as the acetone washings are transferred into a waste container for safe disposal.

More than 90% radiochemical yield of 90Y could be obtained after two electrolyses. Generators of 100 mCi activity level were tested repeatedly using this procedure. Experiments were also conducted with 90Sr spiked with 22.5 mg of inactive strontium, equivalent to 1 Ci (37 GBq) of 90Sr, to study the feasibility of scaling up the generator. Studies with carrier added Sr suggested that the generator can be scaled up to prepare curie quantities of 90Y with high radionuclidic purity (> 10–4%).Some of the major advantages of electrochemical method are:

The operational cost of the generator is very low as the basic technology • involved is simple. The method uses minimal amounts of chemicals/resins and hence there is • very little possibility of additional introduction of metal contaminations. The • 90Sr used can be easily stored and repeatedly used. Unlike other radionuclide generators used in the nuclear medicine depart-

ments, a 90Sr/90Y generator will not need frequent replacement, since very little of the radioactivity of 90Sr is lost by decay. A 10% supplementation of the 90Sr activity every 4-5 years will ensure near constant output from the generator.

A radionuclide generator is called ‘cow’ and the elution of the daughter product is milking. The 90Sr/90Y electrochemical generator which is a contribu-tion from Bhabha Atomic Research Centre, Mumbai, was named ‘90Y- Kamadhenu’ after the mythical cow.

An interesting point is that at the time of writing the paper on 90Sr/90Y electrochemical generator for publication, the authors realized that the electro-chemical method for separation of 90Y from 90Sr was not new, but was reported more than a half-century earlier by Lange et al. [25] and by Hamaguchi et al. [26]. However, for unknown reasons this work was neither noticed nor used for separation of 90Y from 90Sr.

8.4.3. AUTOMATION OF 90Sr/90Y GENERATOR (KAMADHENU) The conversion of the electrochemical 90Sr/90Y generator system into a fully automated module was taken up with Dr. Josef Comor as the author saw tre-mendous opportunities in having a fully automated commercial module that

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will make 90Y suitable for radionuclide therapy. A fully automated computer controlled module developed by Comor is shown in Fig. 8.4 and the same is installed in the Isotope Laboratory (Centis) in Havana, Cuba, and operational since March 2010. 8.4.3.1. Principle of operation of Kamadhenu1/

Kamadhenu module is designed for the production of up to a Ci (37 GBq) of 90Y per day. Kamadhenu consists of fi ve main components:fl uid processing electrochemistry module,• • 90Sr stock reservoir,industrial-standard control-unit,• programmable power supply for the electrolysis, • computer based user interface.• The stock reservoir is used for storing the stock solution of 90Sr in 0.2 M

NH4NO3 or similar electrolyte and is fully enclosed in a 15 mm thick poly-ethylene shield, which stops the β– radiation of 90Sr and 90Y completely. The reservoir has a capacity of 500 ml and it is equipped with a number of platinum electrodes, which are used for level sensing. Two electrodes are outside the primary reservoir and they will provide an alarm signal in case of accidental leakage of 90Sr from the stock reservoir.

The module is equipped with two syringe pumps of 25 ml and 5 ml capac-ity; both have identiacal drives and integrated six port valves. The 25 ml syringe pump is intended for exclusive handling of the stock solution (90Sr), while 5 ml syringe pump is used for the transfer of reagents and the fi nished product, 90Y solution. The control system of the pumps allows for independently adjust-ing the fl ow rate during fi lling and emptying of the syringes.

The module is equipped with fi ve 500 ml reservoirs for reagent solutions needed for maintaining the concentration of the electrolyte and its volume in the stock reservoir, as well as for washing and decontaminating the wetted parts

1/ Operation manual of Kamadhenu, Courtesy: Josef Comor

Fig. 8.4. Kamadhenu, 90Sr/90Y electrochemical generator

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of the module and for recovering 90Y from the central electrode after the sepa-ration process. Also there are two 2000 ml waste reservoirs one of which is used for general waste storage while the other is dedicated for the storage of waste containing valuable amounts of 90Sr, which can be recovered. Another reservoir is also provided which can also be a vial with the initial 90Sr solution, a vial for taking samples for process control, a spare reservoir for emptying the stock reservoir, etc.

The electrolysis cell contains three platinum electrodes and the outer two are concentric while the central electrode is a wire electrode. The central platinum electrode (which is a 1 mm diameter Pt wire) is connected to two pneumatic cylinders acting as X-Y linear drives, providing for moving the electrode up, down, left and right. This way, the electrode can be positioned inside the electrolytic cell (for collecting 90Y on its surface) and inside the 250 μl syringe barrel, where 90Y can be recovered into a suitable solution fi lled into the syringe (for instance 0.05 M HCl).

The electrolysis can be performed either under constant voltage or constant current mode of operation. The current is provided by a programmable external power supply, which is controlled by the operating system of the module. The polarity of the three electrodes can be freely selected, excluding the possibil-ity that all three electrodes have the same polarity (in which case there would be no electrolysis).

The module is also built in with a liquid chromatography column which is used for the recovery of 90Sr from the waste reservoir. The module is equipped with three radiation detectors: monitor the activity within the syringe barrel for 90Y recovery, monitor the activity in the electrolytic cell and monitor the break-through from waste recovery.

The module can be operated either in fully automated or manual mode of operation. The manual mode of operation provides full control over the com-ponents of the module, i.e. the user can open or close each valve, set the desired volume for transfer by the syringe pumps, select the inlet and outlet positions of the six port valves for the pumping sequence, move the central electrode by controlling the pneumatic actuators, adjust the current or voltage for electrolysis and change the polarity of the electrodes.

The automated mode of operation provides a comfortable, reproducible and GMP compliant means of operation. The system will constantly log the activity detected by the three radiation detectors, the state of all valves at all times, the current and voltage used for the electrolysis, the position of the pneumatic actuators and the signals detected by the level sensors, and this in-formation is stored in appropriate daily log fi les on the hard disk of the com-puter running the control system.

The operation of the module is controlled by an industrial PLC (power-line communication). Through the SCADA (supervisory control and data acquisi-tion) the user can select between two basic operation modes: 90Y production

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and 90Sr recovery. Users with appropriate credentials can additionally use the module in manual mode of operation. 8.4.3.2. Economic benefi t of KamadhenuThe designed capacity of Kamadhenu is 37 GBq (1 Ci) per day. It needs a feed of 370 GBq (10 Ci) of 90Sr (~30-50 000 euros). The cost of the module to-gether with a GMP compliant hotcell to house the same is expected to be around 200-250 000 euros depending on the design features of the facility. The operat-ing cost for the module is negligibly small. The module needs a supplement of ~20% of 90Sr once every fi ve years to take care of decay and incidental losses. Current market cost of 90Y is 10-15 euros for 37 MBq (mCi). The module can prepare 37 GBq (1 Ci) per day which is equivalent to 10-15 000 euros per day and 200 days run per year is 2-3 million euros worth of 90Y. Even if 25% of the 90Y activity produced can be used or sold, it is a good investment option. The module as designed now is suitable to be placed in radioisotope/radiopharma-ceuticals production laboratory. However, it could also be placed in a central radiopharmacy with suffi cient administrative controls.8.4.3.3. Safety issuesApart from the technological problems associated with 90Sr/90Y generators, there also exists a serious concern on the use/abuse of 90Sr. Strontium-90 is very long-lived radionuclide with maximum permissible body burden of only 74 MBq. Hence great care has to be taken for the physical protection of the 90Sr stock solution in order to avoid accidental or intentional contamination. Con-sidering the current safety/security issues, the 90Sr need to be secured with great care. Hence, any facility which houses Kamadhenu should have the safety features incorporated at the design stage itself. The module is designed to have a continuous monitoring of the activity in the stock solution.

8.4.4. ACCREDITATIONProf. Henry Wagner, in his 31st highlights lecture in the closing of the Annual Conference of the Society of Nuclear Medicine in New Orleans, USA, 2008 had the following remarks about the report on the novel 90Sr/90Y generators [27]: ‘New 90Sr/90Y generators for radiotherapeutic applications, developed with funding by the International Atomic Energy Agency (IAEA) in association with investigators in India, were described by Chakravarty and colleagues from the Bhabha Atomic Research Centre in Mumbai and from the IAEA. These gen-erators can produce up to 100 mCi (3.7 GBq) levels. Radiochemical separation is effective, and radionuclidic purity is 99.998%, suitable for direct labeling of biomolecules. These generators are simple to operate and can be scaled up and automated. This type of collaborative and benefi cial work is an example of what the IAEA continues to do for the fi eld of nuclear medicine, particularly in developing countries’. The development of the generator has been high-lighted in the IAEA Annual Report 2010 and several IAEA member states are interested in setting up the 90Y production facility by installing ‘Kamadhenu’.

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8.5. RADIONUCLIDIC PURITY REQUIREMENT OF 90Y

Strontium-90 is a bone seeker with a maximum body burden of 2 mCi (74 MBq) [28]. The allowed upper limit of 90Sr in 90Y as per pharmacopoeia is 10-3% or 10 ppm [29]. Hence, the quality control of the eluted 90Y is an important issue and the analytical technique should be able to measure 10 Bq of 90Sr or lower in 1 MBq of 90Y. Both 90Sr and 90Y are beta emitters with overlapping β– spectra and hence conventional techniques of radionuclidic purity estimation are not useful for the estimation of 90Sr contamination in 90Y used for therapy.

Multiple separation steps with ion specifi c chromatography resins using large quantities of 90Y activity are done by the manufactures to estimate the radionuclidic impurity of 90Sr in 90Y. However, this radiochemical separation is also cumbersome as the separation procedure should be capable of giving a decontamination factor of 106 or better.

Measurement of 90Sr in a completely decayed sample of 90Y is used as a reliable procedure to ensure the purity of 90Y used for therapy. However, this method cannot be used as a real time QC technique as the 90Y in question, ir-respective of the quality, would have been administered into the patient. De-velopment of a reliable method to estimate the 90Sr contamination in 90Y was the subject of several studies over the last 50 years. A signifi cant contribution by the author was the suggestion of a novel idea to his colleagues who developed the technique ‘extraction paper chromatography (EPC)’ to reliably estimate the 90Sr contamination in 90Y [8]. The EPC technique is described below.

8.5.1. EXTRACTION PAPER CHROMATOGRAPHY FOR ESTIMATION OF THE RADIONUCLIDIC PURITY OF 90Y [8]Extraction paper chromatography is the term coined for a novel paper chroma-tographic technique which combines solvent extraction with paper chromato-graphy. In this paper chromatography technique, one of the analyte is extract-ed using an ion-selective chelate impregnated at the point of spotting of a chromatography paper and rest of the impurities are separated by partition chromatography.

In the case 90Y, the aim was to develop an analytical technique to detect Bq levels of 90Sr in MBq levels of 90Y. In order to attain the separation between 90Sr and 90Y, an ion selective chelating agent for 90Sr (such as a strontium spe-cifi c crown ether, bis[4,4’(5’)-tert-butylcyclohexano]-18-crown-6) or for 90Y (such as KSM-17, 2-ethylhexyl phosphonic acid) could be used for impregna-tion at the point of spotting of the paper. However, in the present method the need was to measure 90Sr in 90Y which is present in a million times higher amount than that of 90Sr. Hence, it was essential to chelate and fi x 90Y at the point of spotting and allow the migration of 90Sr away from the point of spot-ting.

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The extraction paper chromatography method is simple and easy to adapt. In this method, KSM-17 is impregnated at the point of spotting. Owing to the selective chelating of Y+3 by KSM-17, it remains at the point of spotting while Sr+2 moves with the solvent front. The activity at the solvent front is estimated by liquid scintillation counting and compared with the total activity in the paper to estimate the radionuclidic purity.8.5.1.1. MethodA regular chromatography grade paper (such as Whatman 1) of ~10 cm is saturated with KSM-17 at the point of spotting by applying adequate KSM-17 (about 10 μl of KSM-17 is suffi cient to chelate micromolar quantities of Y+3

and keep it bound). The 90Y-acetate test solution, with known radioactive con-centration is diluted appropriately with high accuracy to a concentration of about 1.0 mCi/ml (37 MBq/ml).

A small known aliquot (5 μl) of this test solution is applied on the paper which is then developed in saline. The KSM-17 retains Y+3 tightly at the point of application (Rf = 0) and Sr+2 migrates with the solvent front, resulting in a clear separation. After drying, the paper is cut and counted in a liquid scintil-lation counter (LSC) to measure the associated radioactivity in 90Sr and 90Y regions.

The EPC technique has very high sensitivity and is capable of detecting ≥ 74 kBq (2 μCi) of 90Sr per 37 GBq (1 Ci) of 90Y. The method is now adapted as a routine QC test for 90Y by many research groups engaged in the develop-ment of generator technology as well as development of 90Y radiopharmaceu-ticals.

The name ‘extraction paper chromatography’ for the new technique de-veloped was coined and fi rst reported by us in analytical chemistry [8]. The method of estimation of the 90Sr in 90Y by EPC is named as the ‘BARC tech-nique’. The EPC has wider applicability in conventional as well as radioana-lytical chemistry for the fast and accurate estimation of different metallic species by suitably selecting the ion specifi c chelate and chromatography con-ditions.

8.6. REVIEW OF 90Y RADIOPHARMACEUTICALS

There are several well established radiopharmaceuticals incorporating 90Y used for radionuclide therapy. These radiopharmaceuticals fall into two categories. The fi rst category includes 90Y incorporated carrier molecules, such as peptides and antibodies. The second category contains the 90Y labeled particulates and colloids used for liver cancer therapy and for radiation synovectomy, respec-tively.

Table 8.1 gives a list of 90Y radiopharmaceuticals used for the therapy of a particular disease state. A brief review of the 90Y radiopharmaceuticals, high-

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lighting only select few agents of importance, is also provided in the following section.

8.6.1. YTTRIUM-90 LABELED MONOCLONAL ANTIBODIES FOR TREATMENT OF PRIMARY CANCERDevelopment of a number of 90Y labeled monoclonal antibodies for therapy of a variety of cancers such as lymphomas (non-Hodgkin’s and others), pancreatic, colon, breast, etc. are reported [31-41]. However, as of now there is only a single registered radiopharmaceutical, Zevalin®. This product is based on a monoclonal antibody against the antigen CD20, modifi ed with a bifunctional chelating agent DTPA (diethylene triamine pentaacetic acid) for complexing with 90Y. Zevalin® was approved for human use in 2002 by the US FDA and is effective for the treatment of non-Hodgkin's lymphoma [42-43]. A few cycles of the radiopharmaceutical treatment are needed to achieve good results. Though very effective, a major impediment towards wider usage of this radiopharma-ceutical is its high cost which is about 15 000 euros for the antibody and the radionuclide together. Several other 90Y labeled monoclonal antibody products are in advanced stages of development and are expected to complete clinical trials in near future. Development of 90Y labeled monoclonal antibodies for the treatment of different type of cancers is one fi eld with very high potential in the future.

Table 8.1. Summary of 90Y labeled products of clinical importance

Disease Targeting moleculeProducts already in the market

Non-Hodgkin’s lymphoma 90Y-ibritumomab-tiuxetanRheumatoid arthritis/hemophilia 90Y-eilicate/citrate colloid

Liver cancers 90Y-microspheres (glass or resin)Products with suffi cient clinical data

Lymphoma 90Y-epratuzomabLung cancer 90Y-lanreotideBreast cancer 90Y-anti-CEA MAbLiver cancer 90Y-alpha fetoprotein MabColon cancer 90Y-pretargeted bispecifi c Ab

Pancreatic cancer 90Y-PAM4 AbNeuro endocrine cancer 90Y- DOTA-D Phe1 Tyr3octreotideNeuro endocrine cancer 90Y- DOTA-D Phe1 Tyr3octreotate

Lung, prostate, breast cancers 90Y-bombesin/GRP analog Breast cancer 90Y-biotin/avidin

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8.6.2. 90Y LABELED PEPTIDESPeptides are preferred carrier molecules for 90Y as the synthetic modifi cation incorporating a chelating agent, such as the DOTA (1,4,7,10-tetraazacyclo-dodecane-N,N’,N’’,N’’’-tetraacetic acid), can be easily accomplished without compromising the biological affi nity of the peptide. Though several regulatory peptides are of interest for radiopharmaceutical development [44], so far so-matostatin analogues have demonstrated maximum success due to the effi cient targeting of neuroendocrine tumors [45-50]. Octreotide and octreotate, the two cyclic octapeptides which are metabolically stable analogues of native somato-statin, that binds with somatostatin receptor subtypes (sst 1-5) with varying affi nity has been particularly successful as a carrier molecule for developing therapeutic radiopharmaceuticals [51-59]. The DOTA conjugated peptide, DOTATATE (DOTA-Tyr-octreotate) (Fig. 8.5), labeled with 90Y is a widely used peptide based radiopharmaceutical for treating patients suffering from neuroendocrine tumors. Several other peptides have also been labeled with 90Y for development of targeted agents [60-61].

Fig. 8.5. Pictorial representation of DOTATATE, 90Y binds to the DOTA ligand

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8.6.3. 90Y LABELED PARTICLES OR COLLOIDSYttrium-90 particles, glass [62-67] or resin [68-69] microspheres have estab-lished their useful role in therapy of liver cancers like hepatocellular carcinoma. In order to achieve high target to non-target ratios, these products are administered loco-regionally. The preparations are retained in the target organ due to either physical entrapment or phagocytosis. Both 90Y-glass microspheres as well as 90Y-resin microspheres are now approved commercial products, sold as Thera-spheres™ and Sir-spheres™, respectively.

The 90Y colloids such as citrates and silicates are used for radiation syn-oviorthesis for the past several years with very encouraging results [70-74]. Also there is a growing demand for radiopharmaceuticals for radiosynovec-tomy.

8.6.4. OTHER 90Y BASED RADIOPHARMACEUTICALSThere are several other reports on the development of small molecules (other than peptides) labeled with 90Y for use as therapeutic agents. However, the results are not yet adequately promising to predict emergence of new 90Y based prod-ucts for therapy. The 90Y-EDTMP has been reported as an agent for bone pain palliation [75]. However, the high energy of 90Y is a disadvantage for this ap-plication due to associated bone marrow depletion. In the absence of access to any other, more suitable, radionuclide, 90Y could be an alternative to 32P in bone pain palliation as their β– energy is similar [76].

8.6.5. INTRAOPERATIVE AVIDINATION FOR RADIONUCLIDE THERAPY (IART®)Surgical removal of the bulk of the tumor followed by radiotherapy for destroy-ing the remnant tumor is the usual method for treatment of patients suffering with early breast cancer. As an alternative to the traditional treatment of radio-therapy, the intraoperative radiation therapy (IORT) is used. In intraoperative radiation therapy, a large single dose radiation dose is delivered to the tumor post surgery while reducing the exposure of normal surrounding tissue. A linear accelerator is often used for this mode of treatment.

An interesting and novel application of 90Y reported from the European Institute of Oncology is the use of 90Y labeled biotin to bind with a pre-inject-ed avidin accumulated in the remnant tumor tissues post surgery [77]. The new modality is named IART®. The IART® procedure consists of the fi rst step in which avidin is intraoperatively injected directly into the tumor immediately after surgery. Due to its positive electric charge (pI 10.7) and the infl ammatory reaction after surgery, avidin is retained at site of injection and the remnant cancer tissues and constitutes a sort of new ‘artifi cial receptor’. 90Y-biotin is injected after 24 h intravenously which gets bound to the avidin molecules to deliver the needed radiation dose to the remnant tumor. The IART® approach is simple and easy to perform and it might represent an alternative to intra-

176

operative radiation therapy in early breast cancer patients. The labeling of biotin takes place through the DOTA chelator, which is linked to biotin.

8.7. DEVELOPMENT OF 90Y RADIOPHARMACEUTICALS

The following sections describe the development on 90Y radiopharmaceuticals in which the author was actively involved.

8.7.1. PREPARATION OF 90Y LABELED PARTICLES FOR USE IN RADIOSYNOVECTOMY [3]Yttrium-90 is one of the preferred radionuclides for large joint radiosynovec-tomy. The high energy (Eβmax = 2.28 MeV) β– particles emitted by 90Y is suit-able for treating large joints. The 64 h half-life is adequate for giving the required radiation dose to the joints and long enough to prepare the radiopharmaceutical and ship it to the site of use. The 90Y-ferric hydroxide macroaggregates (FHMA) and 90Y-hydroxyapatite particles were also prepared and tested. The radiolabe-ling of FHMA particles was done in situ during the preparation of the particles itself, whereas 90Y-HA particles were prepared by labeling HA particles previ-ously prepared and characterized. 8.7.1.1. 90Y-FHMA particlesPreparation of 90Y-FHMA particles were carried out by co-precipitation of yttrium and ferric hydroxide under alkaline conditions. 90YCl3 and ferrous sulfate solution were taken together to which sodium hydroxide was added drop wise with vigorous stirring till ferric-yttrium hydroxide precipitates. The particles formed were recovered by centrifugation and resuspended in saline solution. Centrifugation and resuspension step is repeated twice after which the particles were heated to 100oC for 30 min. The particles were then ultrasonicated to loosen the aggregates and the product was ready. The particle size distribution was determined in a laser diffraction particle analyzer using particles prepared under identical conditions but in the absence of 90Y. The particle size was esti-mated to be between 1-30 μm and 80% of the particles were within 5-20 μm size. The particles were found to be stable in human serum for over 6 days. The particles were biologically evaluated by injecting the sterile preparations in the knee joint of adult New Zealand rabbits. Blood samples were drawn at various time intervals up to 4 days to estimate the amount of activity leached out of the joints. Scintigraphic images of the knee joints were also collected at different time points utilizing the Bremstrahlung radiation emitted by 90Y. The imaging studies confi rmed that the particles were retained in the knee joints. 8.7.1.2. 90Y-hydroxyapatite particles Preparation of HA particles of the size 5-40 μm was reported in section 5.7.1 and these particles were used for radiolabeling with 90Y. About 2 mg of HA particles were suspended in 90YCl3 (2-5 mCi) solution adjusted to pH 7.0 and

177

incubated for 30 min. Radiolabeled HA particles were separated by centrifuga-tion and washed twice with saline. The particles were fi nally suspended in 0.5 ml normal saline, heated for 30 min at 100oC, followed by sonication for 2 min to attain even dispersion of the particles. Bioevaluation of the 90Y particles were done as reported for 90Y-FHMA.

Good quality images were obtained with both 90Y-FHMA and 90Y-HA par-ticles with the particles being retained in the knee joint. The blood samples collected at different time points did not show any leaching of the activity at different time points. Both these particles showed good properties for radio-synovectomy.

Preparation and bioevaluation of 90Y-phosphate particles were also subse-quently reported by the group from the Bhabha Atomic Research Centre, Mumbai [78].

8.7.2. 90Y PATCHES SKIN PATCHES FOR THERAPY OF SUPERFICIAL TUMORS [5]Skin cancer is a common malignancy especially affecting the elder population and due to the increasing life span the incidence of skin cancer has been con-tinuously increasing. However, the death rate is not high in skin cancer. Therapeutic modalities for skin cancer are local destruction, radiotherapy and surgery. Treating skin cancer with 125I-seeds on a gold plaque is also reported. The use of 166Ho or 188Re patches for treating skin cancer was reported by Chung et al. and Lee et al. [79, 80].

90Y being a generator produced radionuclide, it could be more widely avail-able than 166Ho and 188Re (availability of 188W/188Re generators are more diffi cult) and hence 90Y patches were prepared for the same application. The 90Y-FHMA particles were immobilized in regular bandages as a simple and effective method to prepare 90Y radioactive patches.

The 90Y-FHMA particles were prepared as mentioned in section 8.7.1.1. The particles were fi ltered through Millipore fi lter (pore size 0.22 mm, diam-eter 1 cm) and the fi lter paper containing 90Y-FHMA particles was covered with another fi lter paper and immobilized between two layers of gauze using acetone. This patch is then applied on the adhesive side of a ready to use stretchable bandage.

The effi cacy of the treatment was evaluated in mice bearing tumor. Super-fi cial tumors were developed in Swiss mice by injecting fi brosrcoma cells sub-cutaneously. Palpable tumors of size 1.5-2 mm diameter were developed in 5 days and the skin patches were applied to the tumors. The dose delivered to the tumor mass was varied by giving exposure to different time periods (1-3 h). The effi cacy of fractionated doses against a single dose was also checked. Tumor growth was followed for two months and compared with tumor growth with untreated animal. Histological examination of the tissues was done after killing the treated and untreated mice. Regression and delay of tumor growth

178

was seen in all the animals as compared to untreated animals. The 90Y skin patches were found to be quite useful for treatment of skin cancer.

The work was further extended with 32P as the use of this long-lived isotope was found to be more convenient rather than short-lived 90Y [81].

8.7.3. 90Y-DTPA AND 90Y-DOTA FOR POTENTIAL APPLICATION IN INTRAVASCULAR RADIONUCLIDE THERAPY [6]

The isotope 90Y is an attractive radionuclide for IVRNT due to its high energy β– emission and potential availability in very high specifi c activity from the 90Sr/90Y generator. 90Y which is generally available as 90YCl3 cannot be used for IVRNT as an accidental rupture of the balloon will result in substantial uptake of the 90Y activity in several organs including liver, intestine, kidneys and bone apart from persistence of the activity in the blood.

Yttrium-90 complexes of several ligands were prepared and evaluated for their stability [4]. The ligands studied included diethylene triamine pentaacetic acid (DTPA), 1,4,7,10-tetraazacyclododecane-N,N’,N’’,N’’’-tetraacetic acid (DOTA) and 1,4,8,11-tetraazacyclotetradecane-N,N’,N’’,N’’’-tetraacetic acid (TETA). The DTPA and DOTA ligands formed stable complexes with 90Y and hence these complexes were further evaluated, in vivo, for their suitability as agents for IVRNT. The in vivo studies indicated that more than 90% of the activity was eliminated through glomerular fi ltration within 1 h post injection. The 90Y-DOTA complex showed marginally better excretion properties than the 90Y-DTPA complex. However, the 90Y complexes were not further pursued for IVRNT due to the preference of drug eluting stents over the IVRNT option.

8.8. CONCLUSIONS

Yttrium-90 is an important radionuclide for therapy and it use is expected to increase drastically over the coming years. The availability 90Y at present is limited and the cost is high. Being short-lived (T1/2 = 64 h) there are logistical problems in transportation of the radionuclide and radiopharmaceuticals. Being separated from 90Sr, a fi ssion product that is available in plenty, there is con-siderable scope for enhancing the availability of 90Y provided suitable separa-tion methods that will give 90Y of high radionuclidic purity are developed. The author’s involvement and good collaborative efforts from group in India and Serbia has resulted in the development of a novel electrochemical separation system, Kamadhenu. The commercial availability of this module will enhance the availability of 90Y. The quality control technique ‘extraction paper chro-matography’ is a signifi cant development as it can be applied not only for the estimation of radionuclidic purity of 90Y, but also to other fi elds.

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