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RIAR reactor produced radionuclides
Appl. Radiat. lsot. Vol. 49, No. 4, pp. 299-304, 1998 © 1998 Publishedby ElsevierScienceLtd. All rights reserved Printed in Great Britain PII: S0969-8043(97)00041-9 0969-8043/98 $19.00+ 0.00
Pergamon
RIAR Reactor Produced Radionuclides YE. A. K A R E L I N
a n d Y U . G. T O P O R O V
State Scientific Center of Russia, Research Institute of Atomic Reactors, Dimitrovgrad-10, 433510, Russia
Introduction One of the significant R I A R activities is the development of manufacturing processes for reactor radionuclides and production of ionizing sources. Research nuclear reactors play first fiddle in high specific activity radionuclide production. They have a high density of neutron flux and allow a wide range of varying spectrum hardnesses. They permit us to increase the specific activity of radionuclides, reduce impurity formation, and use start-up materials effectively. Research reactors allow the use of flexible irradiation facilities, easily re-adjustable for the production of various radionuclides. The R I A R research reactor complex (Table 1) consists of a unique high flux SM reactor having thermal neutron density up to 5 x 1015 cm -2 s -~, a loop channel material research MIR reactor, and three basin-type RBT reactors. Besides, R I A R has an experimental fast neutron BOR-60 reactor which may be used for radionuclide production too. The Institute has available a complex of radiochemical and engineering facilities, including chains of hot cells for reprocessing radioactive materials up to 100,000 Ci, manipulator-type and glove boxes, and a system for collection, reprocessing and disposal of radioactive wastes. Technologies developed and applied by R I A R are generally intended for the production of radionuclides with low cross-sections or produced by a sequence of neutron captures. First of all these are the transplutonium elements 243Am, 2~24SCm, 249Bk,249Cf, and 252Cf. Practically valuable quantities of these radionuclides can be produced only using high flux reactors presently operating in Russia and the U.S.A. only. Within the last 15 years R I A R has dealt with producing 'light' radionuclides that have resulted in developing processes for phosphorus-33, gadolinium153, iridium-192, cobalt-60, tungsten-188, nickel-63, iron-55,59, tin-ll3,119m, strontium-89 and others (Table 2).
Transplutonium Elements and Ionizing Sources The technology of transplutonium element (TPE) production has been developed, until the start of
the 70th element. It has already been reported more than once (see for example KnHnoB et al., 1989; 5IenHHa, 1986), so I would like to dwell only on the list of TPEs currently produced. These are 243Am, 2"-24SCm, 249Bk, 249Cf, and 252Cf. Their specifications are shown in Table 3. TPEs are used for the production of neutron sources, alpha- and gamma-radiation sources, for investigation of fundamental physical and chemical properties, and for heavy nuclei (Z > 104) synthesis. Californium-252 neutron sources (Kape.rmnH and Ky3neraoB, 1991; Kapea~n and Ky3neIXOB, 1992)
The californium-252 neutron source production process consists of preparing an active core, i.e. fixing a radionuclide on an inert carrier and sealing the core in a capsule, eliminating direct contact of the radionuclide with the environment. Four methods of fixing californium have been developed by RIAR. They are as follows: • impregnation of porous materials (e.g. foam alundum); • californium electroplating; • preparation of Pt~Cf203 cermet; • glass bead production. By means of these methods, cores as wire, plate, cylinder or micro-sphere may be prepared. Several types of neutron source, either of industrial or medical use, were developed on the basis of these processes (Table 4). Industrial sources are of double encapsulation, providing safe maintenance under elevated temperature, pressure, and vibration. Small source dimensions (6 mm diameter, 15 and 25 mm length) and high neutron flux (from 2 x 106 to 2 x 10 ~° s ~) allow their use while deciding practical problems over a wide area. Medical neutron sources are of three main types: dowel type sources for interstitial therapy, cylindrical ones for intracavitary treatment, and flexible assemblies for treatment of complicated forms of tumours. Source capsule design makes both simple and remote afterloading possible. Dowel type sources and flexible assemblies are 1 mm in diameter, cylindrical ones 2 mm. The neutron flux varies,
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Ye. A. Karelin and Yu. G. Toporov
300
Table I. Parametersof RIAR research reactors Parameter SM-3 MIR-M1 Power (thermal) (MW) 100 100 Maximal thermal neutron flux density (cm -2 s-~) 5 x 10'5 5 x 10~d Moderator Water Beryllium and water Reflector Beryllium Beryllium Cooler Water Water
correspondingly, within the regions 106-107 s - t and 108-109 s - i . Since neutron source production was created, more than 1200 medical sources have been prepared and used at various cancer centres in our country. Neutron radiation therapy has been used as a treatment for more than 2000 patients, and as a result this method has been shown to be effective for treatment of some cancers.
Curium-244 alpha sources (Paaqenro et al., 1990) Two methods of curium-244 alpha source preparation have been developed. The coupled reduction procedure consists of the high temperature annealing of curium oxide on metal substrates in hydrogen atmosphere. As an initial substrate, metal iridium, rhodium, and also composite substrates such as rhodium and platinum, iridium and platinum, iridium and palladium, rhodium and palladium, stainless steel and platinum are used. Another method consists of metal curium vapour high temperature condensation onto platinum metal or silicon substrates. 5-8 mCi sources having alpha-line half-width less than 3% have been prepared. These sources are used in particular for making up the set of X-ray, alpha and proton spectrometers of the Alpha-X complex, designed for element assay of space body surfaces (Phobos, Mars).
High Specific Activity Radionuclides and Ionizing Sources Phosphorus-33 This radionuclide (see Table 2) is routinely produced by irradiation of enriched sulphur-33 in the reactor, separation of phosphorus-33 and sulphur by distillation of the latter, and ion exchange purification of phosphorus-33 from impurities. The
Radionuclide 153Gd ~3p "~Sn "3Sn ~SsW 63Ni 5SFe SgFe S'Cr ~Mn *gSr
RBT-10 10 1.5 x 1014 Water Beryllium Water
RBT-6 6 1.4 x 10~4 Water Beryllium Water
distinctive feature of R I A R ' s technology is the use of a high flux SM reactor for irradiation. This increases the yield of phosphorus-33 up to 4-6 Ci/g (compared to 0.6-0.7 Ci/g for reactors with a neutron flux density about 10 '4 cm -2 s - ' ) , and significantly improves the preparation quality due to decreasing a m o u n t of impurities. Carrier-free ortho-phosphoric (33p) acid has a specific activity from 4200 to 5000 Ci/mmol (theoretical value 5200 Ci/mmol), polyphosphates have a fractional activity less than 0.1-0.2%, the total non-radioactive impurity concentration is less than 0.02 g/1.
Gadolinium-153 (TapacoB et al., 1996) Natural europium is used as a target material to produce this radionuclide. Computer modelling based on experimental data has allowed optimisation of irradiation conditions resulting in a ~53Gd specific activity at the end irradiation of not less than 120-150 Ci/g, the yield of gadolinium-153 is 1012 Ci/g europium. Reprocessing of irradiated europium targets is based on cementation of europium with sodium amalgam followed by ~53Gd affinage using extraction chromatography. The preparation may be supplied to a customer either as gadolinium oxide or its chloride. A significant a m o u n t of the preparation is used for manufacturing low radiation photon sources applied for bone densitometers. The nominal activity of these point-size sources is l Ci, the minimum yield of Eu K-line (44keV) is equal to 1.2 x 109photons/ s sterad. Within the last two years, R I A R has dealt with developing Gd-153 linear sources of 50 mCi to 1 Ci activity. An active core is made of AI-Gd203 cermet. Overall dimensions are as follows: 350-500 mm length, 2 - 6 r a m diameter. The gadolinium-153 distribution uniformity along the source is not less than 10-15%.
Table 2. High specificactivity radionuclidesproduced by RIAR Specific activity(Ci/g) Chemical form 100 mGd20~ pellets;t~3GdCl3 4500 Ci/mmol H333PO4,solution in HCI 0.8-1 "~SnCI4, solution in HCI;"~=Snmetal "3SnCl4,solution in HCI;mSnmetal 50 5 tssWO3;Na2t~sWO4,solution in NaOH 63NICI2 10-12 55-60 55FeC13,solution in HCI 100 59FeC13,solution in HCI ferric citrate solution, pH 5-6 3500 5~CrCI3,solution in HCI;Na25~CrO4,solution, pH 6-8 Carrier-free ~*MnCl2,solution in HCI Carrier-free ~9SrC12,solution in HCI
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Tungsten-188 (Tonopoa et al., 1996) This radionuclide results from enriched tungsten186 through double neutron capture. So, production of ~88Wof high specific activity (greater than 5 Ci/g) is possible only by irradiation of target metal tungsten or its oxide in a high flux reactor, in our case the SM-3 reactor. As a target material we usually use tungsten oxide WO3 enriched with ~sw up to 98.5%. Post-reactor reprocessing includes tungsten oxide dissolution in sodium hydroxide solution, purification from impurities by cation exchange (if necessary), and, depending on customer requirements, preparation of either tungsten oxide or sodium tungstate solution in sodium hydroxide.
Tin-119m (TonopoB et al., 1996) The original procedure of reprocessing of irradiated tin targets was developed during the last year. It is based on tin reduction to metal state from citrate solutions with metal aluminium. Satisfactory purification from large numbers of impurities, including antimony radionuclides, has been obtained in this procedure. For additional purification from antimony radionuclides, anion exchange in a Dowex-lHCI system may be used. The tlgmSn specific activity is equal to 0.5M).8 Ci/g, the total radionuclide impurity content is less than 0.1%. The developed procedure is notable for its simplicity of remote design. It may be used for the preparation of other tin radionuclides, such as 113Sn and 117mSn.
Strontium-89 The routine procedure of preparing this radionuclide, applied for palliative treatment of bone cancers, is based on the irradiation of strontium carbonate enriched with 88Sr. The main disadvantage of this technology is that the 895r low specific activity does not exceed 0.3 Ci/g, even when irradiating in a high flux reactor. This results in a necessity to irradiate significant amounts of expensive target material. In total it causes a high price for the final product, limiting its practical application. A relatively new method of 89Sr production is irradiation of yttrium in a fast reactor, producing carrier-free strontium-89 by the sgY(n,p)agSr reaction. In this case, inexpensive natural yttrium oxide is used as a target material. The procedure of 895r carrier-free preparation has been developed at RIAR. The irradiation is performed in a fast neutron reactor BOR-60, providing fast neutron flux of more than ! × l0 ts c m - : S-'. The yield of 895r is 10-15 Ci/kg yttrium. The procedure of yttrium irradiation and reprocessing yttrium targets provides preparation of strontium-89 with the following specifications: strontium-90 content, less than 1 x 10-4%; gammaemitting radionuclide content, less than 0.01%.
The quality of carrier-free strontium-89 is higher than that "traditionally" produced from strontium-88.
Other high specific activity radionuclides Besides the above, RIAR produces radioactive preparations that appear to be produced in high flux reactors only (Table 2). First are nickel-63 (SA --- 1015 Ci/g), iron-55 (55-60 Ci/g), iron-59 (80-100 Ci/g), carrier-free cobalt-58 and manganese-54. These radionuclides may be supplied in various chemical forms, for example, as chlorides and nitrates. Iron-59 is used for biological research and is available as citrate. Chromium-51 of specific activity up to 1000 Ci/g production technology has been developed, this radionuclide may be supplied either as chromium(Ill) chloride or sodium chromate. RIAR also has experience of preparing barium133, zinc-65, calcium-45, cadmium-ll5, antimony124 and some others (Table 5). We are ready to discuss any proposal concerning preparation of high specific activity reactor radionuclides.
Large Scale Production of High Specific Activity Iridium-192 Iridium-192 is the radionuclide widely applied for production of gamma sources for flaw detectors. The world annual demand in this radionuclide reaches 2 x 106Ci. The preparation must have specific activity about 400 Ci/g. This is possible only while irradiating target iridium in reactors with high neutron flux density. Special research in optimising iridium-192 production conditions in MIR and CM reactors, pre-irradiation treatment of target material, post-reactor treatment of irradiated material (cutting of irradiation assemblies, iridium disc activity measurement, preparation for transportation) have allowed the development of large-scale iridium-192 production technology at RIAR (TonopoB and (TapacoB, 1996). It is supplied as irradiated iridium discs of various dimensions according to customer requirements.
Industrial Gamma Sources Using high specific activity radionuclides for production of gamma sources allows a significantincrease in their quality. A series of gamma sources, Table 5. HSA radionuclides available from RIAR Radionuclide Specific activity(Ci/g) mBa 5 6~Zn 40 ~24Sb 40 45Ca 60 b°Co 400 ~921r 600 ~69Yb 1000 JT°Tm 1000 75Se 1200
RIAR reactor produced radionuclides
303
Table 6. Parametersof gamma sources, produced from high specificactivityradionuclides Radionuclide Source size (mm) Core size (ram) Activity(Ci) Diameter Length Diameter Length 7~S¢ 6 12-20 1.0-3.5 1.0-3.5 0.8-120 ~°Co 6-10 12-20 1.5-6.0 1.5~.0 6-350 ~69Yb 2.5 7 0.~0.6 0.3-0.6 0.2-1.8 ~r°Tm 6 12 1.0-3.5 1.0-3.5 1.6-80 mlr 4-6 5-12 0.5-4.0 0.5-4.0 0.5-240
Radionuclide ~°Co ~°Co ~9-'lr
Table 7. Parametersof medicalgammasources Source size (mm) Core size (ram) Activity(Ci) Application Diameter Length Diameter Length 1 25-35 0.5 20-30 0.001-2.0 Interstitial brachytherapy,manualafterloading 2 16 0.5-0.8 10 0.0094-5.4 lntracavitarybrachytherapy,manualafterloading 1.6 10.6 0.8 4 1.5-10.5 Interstitial brachytherapy,manualafterloading
mainly intended for flaw-meters, were designed at RIAR. They are Co-60, Ir-192, Se-75, Yb-169, and Tm-170 sources; their parameters are presented in Table 6.
Gamma Sources for Medical Applications The trend of developing radiation therapy consists in increasing the differentiation of radiation influence on tumours and sound tissues. It can be achieved by source miniaturization and variation of emitted energy. To conform to these requirements, RIAR has developed some new 6°Co and t92Ir source modifications for contact radiation therapy using both manual and remote afterloading with domestic devices of AGAM, AGAT-T, and AGAT-VU type (Table 7). The Ministry of Public Health has permitted the use of these sources in clinical practice. Annually, from 400 to 600 sources accompanied with appropriate instruments providing self-operation conditions for medical personnel are shipped to clinics in Russia and the CIS.
Control and Measurement System The control and measurement system covers all stages of source and preparation production from incoming inspection of irradiated raw material to final certification of products. The system includes 17 measuring devices, with 28 measuring procedures under operation. The following stages of the production process are under control: • incoming inspection of irradiated material according to activity and isotope composition; • on-line inspection when reprocessing irradiated material by radiochemical separation according to activity and impurity content through the process; • inspection of the preparation loading into each source according to activity; • certification of final sources according to activity or particle flux, radioactive impurity content, tightness and surface contamination.
Methods applied are as follows: alpha, gamma, and X-ray spectrometry; alpha and beta counting; exposure dose rate measuring; measurement of photon and neutron fluxes. The procedures provide measures of both solution and source geometry. The region of activity measured is from 10 Bq to 100,000 Ci. The measurement error when inspecting technological operations is 10-15%. Measuring devices and procedures are certified by the State Standard Board of Russia. The staff of the laboratory of measurements consists of 26 persons. All leading specialists have been trained in central offices of the State Standard Board.
Quality of Radionuclide Products: Source Safety The quality and reliability of RIAR radionuclide products are ensured by a Quality Management System based on the ISO-9001 International Standard. Production of all types of source and preparation is under current inspection by the Quality Service. The Quality Service inspects raw material quality, technological processes on-line, methods and equipment for checking. While developing a procedure of source production, their safety is given significant consideration. Source safety means the ability of sealed sources to prevent contact of radioactive material with the environment and ensure the integrity of a sealed capsule both under standard working conditions and in the event of an accident. Source safety is checked and confirmed by special tests for conformity with Strength Grades stipulated by the ISO-1677 and ISO/TO 4826 International Standards. When producing sources, continuous control for tightness is performed. All sources or Packing Sets used for transportation of sources are checked for conformity with "Special Form Radioactive Material", that according to IAEA provides for the safety of their transportation to users.
304
Ye. A. Karelin and Yu. G. Toporov
Pa~qearo, B. M., A ~ p e i ~ r o s , B. M., EepxyTos x ~p, B. JI. (1990)//A~s~ba-HcToqar~ ~ a npoe~Ta "Oo6oc" KaHnOB, A. B. and MaMeaaa, 10. F. (1989) Tonopos Pa~HOXHMH~I, NO. 3, C.142-145. HccJ1e~osaTe~bcrHe peaKTopbl HHHAPa ~ Tapacos, B. A., TonopoB, IO. F. and O r ~ M o a o s n ~p, napa6oTrn paarlonyKal'~oa, // ATOMHaa 9Heprn~l, T.69, B. T. (1996) Texao~iorHa noay~eH~a npenapaTa Fa~oJIHHH~I-153, BhICOKOfiy~e~IbHO~aKTHBHOCTH.//Proc. No. 3, C.186--190. of IRRMA'96. :Ienaua, B. H. (1986) CocToarlne n TeI~eHRHB pa6oT no no~lyqeaaio, nccae~osanam CBfiCTB 14 rIpHMeHtflO Totlopoa, IO. F., Tapacoa, B. A., Ky3neitoa, P. A. and Fonqapoaa, F. B. (1996) IIo~yqenHe npenapaTa rpaacnayToHneabIX 3JIeMeHTOB B HFIAP HM. // soasqbpaua-188. Proc. of IRRMA'96. PagaoxaMna, No. 4, C.533-539. Tonopoa, IO. F., BaxeTos, ~. 3., Ky3neuoa, P. A. and Kapennn, E. A. and Ky3uettos, P. A. (1991) Ait~eeB, O. I4. (1996) Pa~noHyraam,i H3Sn, 119mSn: KaanqbopnaeBue HCTOqHItKI4 nefiTpOnOa noJlyqenae a peagTope n xaM~qecraa nepepa6oTra o6menpoMbmgqcaHoro Haanaseana.: HpenpnaT o6~iyenaux Masmene~. // Proc. of IRRMA'96. HHHAP-20(823). ~UMaTpoarpa~. Tonopoa, IO. F. and Tapacos, B. A. (1996) Kapeaaa, E. A. and Kyaneuoa, P. A. (1992) l-Ipoqbaanposaaae IIJIOTHOCTH norora nefiTpono6 npa KaaaqbopnaeB~e ilCTOttHHKItH¢fiTpOHOB MeaHIIHHCKOrO aa3naaeaas.: HpenpanT HHHAP-20(828). O6Jlyqennn npnaaa-192. Proc. of IRRMA'96. ~nMnrpoarpaa.
References
Pergamon
RIAR Reactor Produced Radionuclides YE. A. K A R E L I N
a n d Y U . G. T O P O R O V
State Scientific Center of Russia, Research Institute of Atomic Reactors, Dimitrovgrad-10, 433510, Russia
Introduction One of the significant R I A R activities is the development of manufacturing processes for reactor radionuclides and production of ionizing sources. Research nuclear reactors play first fiddle in high specific activity radionuclide production. They have a high density of neutron flux and allow a wide range of varying spectrum hardnesses. They permit us to increase the specific activity of radionuclides, reduce impurity formation, and use start-up materials effectively. Research reactors allow the use of flexible irradiation facilities, easily re-adjustable for the production of various radionuclides. The R I A R research reactor complex (Table 1) consists of a unique high flux SM reactor having thermal neutron density up to 5 x 1015 cm -2 s -~, a loop channel material research MIR reactor, and three basin-type RBT reactors. Besides, R I A R has an experimental fast neutron BOR-60 reactor which may be used for radionuclide production too. The Institute has available a complex of radiochemical and engineering facilities, including chains of hot cells for reprocessing radioactive materials up to 100,000 Ci, manipulator-type and glove boxes, and a system for collection, reprocessing and disposal of radioactive wastes. Technologies developed and applied by R I A R are generally intended for the production of radionuclides with low cross-sections or produced by a sequence of neutron captures. First of all these are the transplutonium elements 243Am, 2~24SCm, 249Bk,249Cf, and 252Cf. Practically valuable quantities of these radionuclides can be produced only using high flux reactors presently operating in Russia and the U.S.A. only. Within the last 15 years R I A R has dealt with producing 'light' radionuclides that have resulted in developing processes for phosphorus-33, gadolinium153, iridium-192, cobalt-60, tungsten-188, nickel-63, iron-55,59, tin-ll3,119m, strontium-89 and others (Table 2).
Transplutonium Elements and Ionizing Sources The technology of transplutonium element (TPE) production has been developed, until the start of
the 70th element. It has already been reported more than once (see for example KnHnoB et al., 1989; 5IenHHa, 1986), so I would like to dwell only on the list of TPEs currently produced. These are 243Am, 2"-24SCm, 249Bk, 249Cf, and 252Cf. Their specifications are shown in Table 3. TPEs are used for the production of neutron sources, alpha- and gamma-radiation sources, for investigation of fundamental physical and chemical properties, and for heavy nuclei (Z > 104) synthesis. Californium-252 neutron sources (Kape.rmnH and Ky3neraoB, 1991; Kapea~n and Ky3neIXOB, 1992)
The californium-252 neutron source production process consists of preparing an active core, i.e. fixing a radionuclide on an inert carrier and sealing the core in a capsule, eliminating direct contact of the radionuclide with the environment. Four methods of fixing californium have been developed by RIAR. They are as follows: • impregnation of porous materials (e.g. foam alundum); • californium electroplating; • preparation of Pt~Cf203 cermet; • glass bead production. By means of these methods, cores as wire, plate, cylinder or micro-sphere may be prepared. Several types of neutron source, either of industrial or medical use, were developed on the basis of these processes (Table 4). Industrial sources are of double encapsulation, providing safe maintenance under elevated temperature, pressure, and vibration. Small source dimensions (6 mm diameter, 15 and 25 mm length) and high neutron flux (from 2 x 106 to 2 x 10 ~° s ~) allow their use while deciding practical problems over a wide area. Medical neutron sources are of three main types: dowel type sources for interstitial therapy, cylindrical ones for intracavitary treatment, and flexible assemblies for treatment of complicated forms of tumours. Source capsule design makes both simple and remote afterloading possible. Dowel type sources and flexible assemblies are 1 mm in diameter, cylindrical ones 2 mm. The neutron flux varies,
299
Ye. A. Karelin and Yu. G. Toporov
300
Table I. Parametersof RIAR research reactors Parameter SM-3 MIR-M1 Power (thermal) (MW) 100 100 Maximal thermal neutron flux density (cm -2 s-~) 5 x 10'5 5 x 10~d Moderator Water Beryllium and water Reflector Beryllium Beryllium Cooler Water Water
correspondingly, within the regions 106-107 s - t and 108-109 s - i . Since neutron source production was created, more than 1200 medical sources have been prepared and used at various cancer centres in our country. Neutron radiation therapy has been used as a treatment for more than 2000 patients, and as a result this method has been shown to be effective for treatment of some cancers.
Curium-244 alpha sources (Paaqenro et al., 1990) Two methods of curium-244 alpha source preparation have been developed. The coupled reduction procedure consists of the high temperature annealing of curium oxide on metal substrates in hydrogen atmosphere. As an initial substrate, metal iridium, rhodium, and also composite substrates such as rhodium and platinum, iridium and platinum, iridium and palladium, rhodium and palladium, stainless steel and platinum are used. Another method consists of metal curium vapour high temperature condensation onto platinum metal or silicon substrates. 5-8 mCi sources having alpha-line half-width less than 3% have been prepared. These sources are used in particular for making up the set of X-ray, alpha and proton spectrometers of the Alpha-X complex, designed for element assay of space body surfaces (Phobos, Mars).
High Specific Activity Radionuclides and Ionizing Sources Phosphorus-33 This radionuclide (see Table 2) is routinely produced by irradiation of enriched sulphur-33 in the reactor, separation of phosphorus-33 and sulphur by distillation of the latter, and ion exchange purification of phosphorus-33 from impurities. The
Radionuclide 153Gd ~3p "~Sn "3Sn ~SsW 63Ni 5SFe SgFe S'Cr ~Mn *gSr
RBT-10 10 1.5 x 1014 Water Beryllium Water
RBT-6 6 1.4 x 10~4 Water Beryllium Water
distinctive feature of R I A R ' s technology is the use of a high flux SM reactor for irradiation. This increases the yield of phosphorus-33 up to 4-6 Ci/g (compared to 0.6-0.7 Ci/g for reactors with a neutron flux density about 10 '4 cm -2 s - ' ) , and significantly improves the preparation quality due to decreasing a m o u n t of impurities. Carrier-free ortho-phosphoric (33p) acid has a specific activity from 4200 to 5000 Ci/mmol (theoretical value 5200 Ci/mmol), polyphosphates have a fractional activity less than 0.1-0.2%, the total non-radioactive impurity concentration is less than 0.02 g/1.
Gadolinium-153 (TapacoB et al., 1996) Natural europium is used as a target material to produce this radionuclide. Computer modelling based on experimental data has allowed optimisation of irradiation conditions resulting in a ~53Gd specific activity at the end irradiation of not less than 120-150 Ci/g, the yield of gadolinium-153 is 1012 Ci/g europium. Reprocessing of irradiated europium targets is based on cementation of europium with sodium amalgam followed by ~53Gd affinage using extraction chromatography. The preparation may be supplied to a customer either as gadolinium oxide or its chloride. A significant a m o u n t of the preparation is used for manufacturing low radiation photon sources applied for bone densitometers. The nominal activity of these point-size sources is l Ci, the minimum yield of Eu K-line (44keV) is equal to 1.2 x 109photons/ s sterad. Within the last two years, R I A R has dealt with developing Gd-153 linear sources of 50 mCi to 1 Ci activity. An active core is made of AI-Gd203 cermet. Overall dimensions are as follows: 350-500 mm length, 2 - 6 r a m diameter. The gadolinium-153 distribution uniformity along the source is not less than 10-15%.
Table 2. High specificactivity radionuclidesproduced by RIAR Specific activity(Ci/g) Chemical form 100 mGd20~ pellets;t~3GdCl3 4500 Ci/mmol H333PO4,solution in HCI 0.8-1 "~SnCI4, solution in HCI;"~=Snmetal "3SnCl4,solution in HCI;mSnmetal 50 5 tssWO3;Na2t~sWO4,solution in NaOH 63NICI2 10-12 55-60 55FeC13,solution in HCI 100 59FeC13,solution in HCI ferric citrate solution, pH 5-6 3500 5~CrCI3,solution in HCI;Na25~CrO4,solution, pH 6-8 Carrier-free ~*MnCl2,solution in HCI Carrier-free ~9SrC12,solution in HCI
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Tungsten-188 (Tonopoa et al., 1996) This radionuclide results from enriched tungsten186 through double neutron capture. So, production of ~88Wof high specific activity (greater than 5 Ci/g) is possible only by irradiation of target metal tungsten or its oxide in a high flux reactor, in our case the SM-3 reactor. As a target material we usually use tungsten oxide WO3 enriched with ~sw up to 98.5%. Post-reactor reprocessing includes tungsten oxide dissolution in sodium hydroxide solution, purification from impurities by cation exchange (if necessary), and, depending on customer requirements, preparation of either tungsten oxide or sodium tungstate solution in sodium hydroxide.
Tin-119m (TonopoB et al., 1996) The original procedure of reprocessing of irradiated tin targets was developed during the last year. It is based on tin reduction to metal state from citrate solutions with metal aluminium. Satisfactory purification from large numbers of impurities, including antimony radionuclides, has been obtained in this procedure. For additional purification from antimony radionuclides, anion exchange in a Dowex-lHCI system may be used. The tlgmSn specific activity is equal to 0.5M).8 Ci/g, the total radionuclide impurity content is less than 0.1%. The developed procedure is notable for its simplicity of remote design. It may be used for the preparation of other tin radionuclides, such as 113Sn and 117mSn.
Strontium-89 The routine procedure of preparing this radionuclide, applied for palliative treatment of bone cancers, is based on the irradiation of strontium carbonate enriched with 88Sr. The main disadvantage of this technology is that the 895r low specific activity does not exceed 0.3 Ci/g, even when irradiating in a high flux reactor. This results in a necessity to irradiate significant amounts of expensive target material. In total it causes a high price for the final product, limiting its practical application. A relatively new method of 89Sr production is irradiation of yttrium in a fast reactor, producing carrier-free strontium-89 by the sgY(n,p)agSr reaction. In this case, inexpensive natural yttrium oxide is used as a target material. The procedure of 895r carrier-free preparation has been developed at RIAR. The irradiation is performed in a fast neutron reactor BOR-60, providing fast neutron flux of more than ! × l0 ts c m - : S-'. The yield of 895r is 10-15 Ci/kg yttrium. The procedure of yttrium irradiation and reprocessing yttrium targets provides preparation of strontium-89 with the following specifications: strontium-90 content, less than 1 x 10-4%; gammaemitting radionuclide content, less than 0.01%.
The quality of carrier-free strontium-89 is higher than that "traditionally" produced from strontium-88.
Other high specific activity radionuclides Besides the above, RIAR produces radioactive preparations that appear to be produced in high flux reactors only (Table 2). First are nickel-63 (SA --- 1015 Ci/g), iron-55 (55-60 Ci/g), iron-59 (80-100 Ci/g), carrier-free cobalt-58 and manganese-54. These radionuclides may be supplied in various chemical forms, for example, as chlorides and nitrates. Iron-59 is used for biological research and is available as citrate. Chromium-51 of specific activity up to 1000 Ci/g production technology has been developed, this radionuclide may be supplied either as chromium(Ill) chloride or sodium chromate. RIAR also has experience of preparing barium133, zinc-65, calcium-45, cadmium-ll5, antimony124 and some others (Table 5). We are ready to discuss any proposal concerning preparation of high specific activity reactor radionuclides.
Large Scale Production of High Specific Activity Iridium-192 Iridium-192 is the radionuclide widely applied for production of gamma sources for flaw detectors. The world annual demand in this radionuclide reaches 2 x 106Ci. The preparation must have specific activity about 400 Ci/g. This is possible only while irradiating target iridium in reactors with high neutron flux density. Special research in optimising iridium-192 production conditions in MIR and CM reactors, pre-irradiation treatment of target material, post-reactor treatment of irradiated material (cutting of irradiation assemblies, iridium disc activity measurement, preparation for transportation) have allowed the development of large-scale iridium-192 production technology at RIAR (TonopoB and (TapacoB, 1996). It is supplied as irradiated iridium discs of various dimensions according to customer requirements.
Industrial Gamma Sources Using high specific activity radionuclides for production of gamma sources allows a significantincrease in their quality. A series of gamma sources, Table 5. HSA radionuclides available from RIAR Radionuclide Specific activity(Ci/g) mBa 5 6~Zn 40 ~24Sb 40 45Ca 60 b°Co 400 ~921r 600 ~69Yb 1000 JT°Tm 1000 75Se 1200
RIAR reactor produced radionuclides
303
Table 6. Parametersof gamma sources, produced from high specificactivityradionuclides Radionuclide Source size (mm) Core size (ram) Activity(Ci) Diameter Length Diameter Length 7~S¢ 6 12-20 1.0-3.5 1.0-3.5 0.8-120 ~°Co 6-10 12-20 1.5-6.0 1.5~.0 6-350 ~69Yb 2.5 7 0.~0.6 0.3-0.6 0.2-1.8 ~r°Tm 6 12 1.0-3.5 1.0-3.5 1.6-80 mlr 4-6 5-12 0.5-4.0 0.5-4.0 0.5-240
Radionuclide ~°Co ~°Co ~9-'lr
Table 7. Parametersof medicalgammasources Source size (mm) Core size (ram) Activity(Ci) Application Diameter Length Diameter Length 1 25-35 0.5 20-30 0.001-2.0 Interstitial brachytherapy,manualafterloading 2 16 0.5-0.8 10 0.0094-5.4 lntracavitarybrachytherapy,manualafterloading 1.6 10.6 0.8 4 1.5-10.5 Interstitial brachytherapy,manualafterloading
mainly intended for flaw-meters, were designed at RIAR. They are Co-60, Ir-192, Se-75, Yb-169, and Tm-170 sources; their parameters are presented in Table 6.
Gamma Sources for Medical Applications The trend of developing radiation therapy consists in increasing the differentiation of radiation influence on tumours and sound tissues. It can be achieved by source miniaturization and variation of emitted energy. To conform to these requirements, RIAR has developed some new 6°Co and t92Ir source modifications for contact radiation therapy using both manual and remote afterloading with domestic devices of AGAM, AGAT-T, and AGAT-VU type (Table 7). The Ministry of Public Health has permitted the use of these sources in clinical practice. Annually, from 400 to 600 sources accompanied with appropriate instruments providing self-operation conditions for medical personnel are shipped to clinics in Russia and the CIS.
Control and Measurement System The control and measurement system covers all stages of source and preparation production from incoming inspection of irradiated raw material to final certification of products. The system includes 17 measuring devices, with 28 measuring procedures under operation. The following stages of the production process are under control: • incoming inspection of irradiated material according to activity and isotope composition; • on-line inspection when reprocessing irradiated material by radiochemical separation according to activity and impurity content through the process; • inspection of the preparation loading into each source according to activity; • certification of final sources according to activity or particle flux, radioactive impurity content, tightness and surface contamination.
Methods applied are as follows: alpha, gamma, and X-ray spectrometry; alpha and beta counting; exposure dose rate measuring; measurement of photon and neutron fluxes. The procedures provide measures of both solution and source geometry. The region of activity measured is from 10 Bq to 100,000 Ci. The measurement error when inspecting technological operations is 10-15%. Measuring devices and procedures are certified by the State Standard Board of Russia. The staff of the laboratory of measurements consists of 26 persons. All leading specialists have been trained in central offices of the State Standard Board.
Quality of Radionuclide Products: Source Safety The quality and reliability of RIAR radionuclide products are ensured by a Quality Management System based on the ISO-9001 International Standard. Production of all types of source and preparation is under current inspection by the Quality Service. The Quality Service inspects raw material quality, technological processes on-line, methods and equipment for checking. While developing a procedure of source production, their safety is given significant consideration. Source safety means the ability of sealed sources to prevent contact of radioactive material with the environment and ensure the integrity of a sealed capsule both under standard working conditions and in the event of an accident. Source safety is checked and confirmed by special tests for conformity with Strength Grades stipulated by the ISO-1677 and ISO/TO 4826 International Standards. When producing sources, continuous control for tightness is performed. All sources or Packing Sets used for transportation of sources are checked for conformity with "Special Form Radioactive Material", that according to IAEA provides for the safety of their transportation to users.
304
Ye. A. Karelin and Yu. G. Toporov
Pa~qearo, B. M., A ~ p e i ~ r o s , B. M., EepxyTos x ~p, B. JI. (1990)//A~s~ba-HcToqar~ ~ a npoe~Ta "Oo6oc" KaHnOB, A. B. and MaMeaaa, 10. F. (1989) Tonopos Pa~HOXHMH~I, NO. 3, C.142-145. HccJ1e~osaTe~bcrHe peaKTopbl HHHAPa ~ Tapacos, B. A., TonopoB, IO. F. and O r ~ M o a o s n ~p, napa6oTrn paarlonyKal'~oa, // ATOMHaa 9Heprn~l, T.69, B. T. (1996) Texao~iorHa noay~eH~a npenapaTa Fa~oJIHHH~I-153, BhICOKOfiy~e~IbHO~aKTHBHOCTH.//Proc. No. 3, C.186--190. of IRRMA'96. :Ienaua, B. H. (1986) CocToarlne n TeI~eHRHB pa6oT no no~lyqeaaio, nccae~osanam CBfiCTB 14 rIpHMeHtflO Totlopoa, IO. F., Tapacoa, B. A., Ky3neitoa, P. A. and Fonqapoaa, F. B. (1996) IIo~yqenHe npenapaTa rpaacnayToHneabIX 3JIeMeHTOB B HFIAP HM. // soasqbpaua-188. Proc. of IRRMA'96. PagaoxaMna, No. 4, C.533-539. Tonopoa, IO. F., BaxeTos, ~. 3., Ky3neuoa, P. A. and Kapennn, E. A. and Ky3uettos, P. A. (1991) Ait~eeB, O. I4. (1996) Pa~noHyraam,i H3Sn, 119mSn: KaanqbopnaeBue HCTOqHItKI4 nefiTpOnOa noJlyqenae a peagTope n xaM~qecraa nepepa6oTra o6menpoMbmgqcaHoro Haanaseana.: HpenpnaT o6~iyenaux Masmene~. // Proc. of IRRMA'96. HHHAP-20(823). ~UMaTpoarpa~. Tonopoa, IO. F. and Tapacos, B. A. (1996) Kapeaaa, E. A. and Kyaneuoa, P. A. (1992) l-Ipoqbaanposaaae IIJIOTHOCTH norora nefiTpono6 npa KaaaqbopnaeB~e ilCTOttHHKItH¢fiTpOHOB MeaHIIHHCKOrO aa3naaeaas.: HpenpanT HHHAP-20(828). O6Jlyqennn npnaaa-192. Proc. of IRRMA'96. ~nMnrpoarpaa.
References