Low-Waste and Proliferation-Free Production of Medical Radioisotopes in Solution and Molten-Salt Reactors

There is an old dispute what is more valuable in work of atomic power stations the electric power or by-products of its operation produced stable and radioactive isotopes. Among the experts in reactors physics and techniques, an opinion prevails that nuclear reactors make profit exclusively from electric power production on atomic power stations. However some specialists consider this opinion rather far from a real state of things.


Introduction
There is an old dispute what is more valuable in work of atomic power stations -the electric power or by-products of its operation -produced stable and radioactive isotopes. Among the experts in reactors physics and techniques, an opinion prevails that nuclear reactors make profit exclusively from electric power production on atomic power stations. However some specialists consider this opinion rather far from a real state of things.
In the early nineties on the initiative of the American Council for Energy Awareness, USCEA, detailed research of influence of nuclear and radiating technologies (except for defensive) on national economy of the USA (Rabotnov, 1999) was undertaken. It consisted of two parts. In one part applications of radioactive materials in the industry, medicine and scientific researches were only considered, in the other -actually nuclear power, that is electric power production at atomic power stations. Results have surprised both authors of the report, and customers. First, the scale of figures. It appeared, that the total annual volume of the business connected with radioisotopes application in the United States made 257 billion dollars in 1991. Secondly, this sum appeared three and a half times more than the full cost of the nuclear electric power (73 billion dollars). Radioisotopes are used in eight various branches, 3.7 million persons of almost five hundred specialties are employed in this activity that makes about three percent of full employment in the USA. The total product of 330 billion dollars means that the nuclear-power and radiation complex of the USA looks as the eleventh on size a world industrial state.
The simplest and cheapest way of scale radioisotopes production is based on fission reaction (n,f). The basic medicine radioisotopes formed as a result of uranium-235 fission by neutrons are 137 Cs, 131 I, 133 Xe, 90 Sr and 99 Mo. 99 Mo is one of the most demanded radioisotopes in nuclear medicine. It is used in 99 Mo/ 99m Tc generators widely applied in the world at early diagnostics of oncologic, cardiovascular and a number of other diseases. More than 80 % of radiodiagnostic procedures in the world are carried out by preparations labeled with 99m Тc (IAEA, 1999). From the IAEA data, approximately 15 millions of diagnostic procedures were carried out with 99m Tc annually in the world ten years ago, from them 7 millions in Europe and 8 millions in the USA. The total world consumption of 99 Mo/ 99m Tc for that period was estimated at a level of 6 000 Ci a week (6 day pre-calibration). The prediction of 99 Mo demand till 2006 made in works (Ball, 1999;Ball, Nordyke & Brown, 2002) is illustrated in the diagram in figure 1. About 30 millions of diagnostic procedures are carried out with 99m Tc in the world annually today (Hansell, 2008). The world consumption of 99 Mo/ 99m Tc has exceeded 12 000 Ci a week, and annual volume of the world market of 99m Tc consumption is estimated as US$ 3.7 billion (including cost of medical services).
The basic producers of 99 Mo are Canada, EU and South Africa, which supply about 95 % of this radioisotope in the world. The technology of 99 Mo production is based on irradiation in research reactors and subsequent radiochemical processing of the targets from highlyenriched uranium. Table 1 provides further information on the major 99 Mo producing reactors (Westmacott, 2010).
Recent series of the not planned stops of the reactors producing 99 Mo, mainly due to significant age of these reactors, have shown to the world imperfection of the existing system of supplying this radioisotope and have caused growing concern for the future of nuclear medicine. The faults in 99 Mo production had negative effect on treatment of oncologic patients in Europe and Northern America. For example, during these stops, supply of 99 Mo/ 99m Tc generators in some British hospitals has reduced to 30 % from normal. The doctors had to choose, who from the patients more than others requires the procedures with 99m Tc.
www.intechopen.com 1. Six-day curies end of processing. In some cases, the maximum production can be substantially higher that the values listed here for normal production. 2. Fuel elements and targets are classified as either LEU, containing less than 20% of 235 U, or HEU, which contains greater than 20% 235 U (in some cases greater than 95%). 3. Does not account for increase in capacity since April 2010 with the installation of additional irradiation capacity. This increases BR-2 available capacity to approximately 7 800 six-day curies EOP: however it is not yet clear what "normal" production will be at the facility. 4. These reactors started production in 2010, -so some data is not available yet. 5. The LVR-15 reactor uses fuel elements that are enriched to 36% 235 U. 6. The OPAL reactor started production in 2007 for domestic use but has not yet exported significant amounts. 7. SAFARl-1 is in the process of converting to using LEU targets and expects to have completed conversion in 2010. Besides, the isotope crisis has drawn attention to one more acute problem -production technology. The fission fragment method is a basis of current 99 Mo production. A highlyenriched uranium target is irradiated in a reactor, and then processed by one of the known radiochemical ways. 99 Mo, which specific activity reaches several tens thousand Ci per gram of molybdenum is separated from fission products (Gerasimov, Kiselev & Lantcov, 1989 The weakest feature of this way of 99 Mo production is extremely inefficient use of uranium. Only ≈0.4 % of 235 U is used for production of 99 Mo, and the other part is directed in waste. Moreover, at fission of uranium a wide spectrum of fragments is formed, which total activity at the moment of irradiation end exceeds activity of 99 Mo by two orders of value. That results in necessity of contamination of plenty radioactive waste, including long-lived ones. Fig. 1. Prediction of growth of demand for 99 Mo in the world (Ball, 1999;Ball, Nordyke & Brown, 2002) In Fig. 2 the basic stages of the existing 99 Mo production technology are shown -from manufacturing of a uranium target to delivering the final product to the consumer. The third problem of the traditional 99 Mo production is that its target waste can be easily transformed by known chemical methods to metal HEU -material used in manufacture of nuclear military loads. Orientation of modern 99 Mo production on the use of highlyenriched uranium, against gradual deducing of HEU from a civil turn according to the IAEA "non-proliferation" concept, creates additional risks for 99 Mo consumers. Thereupon expected rapid development of the nuclear medicine using 235 U fission products can lead the next years to serious problems with ensuring the efficiency of the international monitoring system for non-proliferation of nuclear materials.
In the international community there is a consensus about necessity to limit or even to bring to naught HEU turn in the civil sphere. Programs, in particular, the program RERTR (Reduced Enrichment for Research and Test Reactors) on reduction of HEU turn in peace sectors of economy are accepted. The research reactors used for production of 99 Mo will be gradually converted to low-enriched uranium fuel (LEU) and LEU-targets. Risks from use of uranium of high enrichment will start to constrain sooner or later, and then can even stop this production completely.
The world crisis of 99 Mo production gives an exclusive chance for realization of alternative methods of this radioisotope production, free from disadvantages of the modern technologies and ensuring large-scale low-waste production of 99 Mo based on the use of low-enriched uranium.
What could be alternative? The choice, in effect, is limited: building of a new multi-purpose research reactor, development of accelerating methods or building of inexpensive and simple in operation solution reactors. Taking into account inertia of the licensing system and time for designing and building the facilities, the choice needs to be made the next years.
It would seem, the most real alternative to HEU for today is conversion of a research reactor core for work with LEU-fuel and introduction of new types of targets of low-enriched uranium. However the "target" technology characterized by the lowest efficiency of use of uranium, most likely will not give competitive advantages as the use of LEU will inevitably lead to involving for manufacture of a considerable amount of irradiated uranium. It will result in additional loadings on radiochemical processing and in considerable amount of radioactive waste that, obviously, will raise production cost price. The most research reactors now producing the basic amount of 99 Mo are depreciated and for their maintenance in operating and safe condition more and more considerable efforts and expenses are required. The next years these reactors will demand capital reconstruction or will be deduced from operation. Other alternative technologies based on neutron activation of molybdenum isotope 98 Mo or on uranium photo-fission at accelerators, are not effective for large-scale production and can be applied only for satisfying regional needs.
Renunciation of the traditional technology of uranium targets irradiation and transition to technologies based on specific physical-chemical properties of liquid nuclear fuel, ensuring an opportunity to extract 99 Mo directly from the reactor fuel leaving uranium and fission fragments in it, allow to exclude a problem of highly-enriched uranium contamination and to simplify the problem of handling the radiochemical processing waste radically. Water solutions of uranium salts or melts of fluoride salts can be used as liquid nuclear fuel.
The aggregate state of liquid fuel opens the possibility of selective extraction of fragment radioisotopes 99 Mo/ 99m Tc, 89 Sr, 131,133 I, 133 Xe, etc. from fuel not involving both 235 U and the basic group of fragment elements. It will lead to sharp decrease in the flow of nuclear materials in the technological process and in the yield of highly active waste. Estimations show, for example, that a solution reactor of 200 kW power is capable to produce more than 1000 Ci/w of 99 Mo with calibration for 6-days from the moment of delivery.
The technology based on use of homogeneous liquid nuclear fuel possesses a number of obvious advantages in comparison with a "target" method: -All fuel of such a reactor is a "target"; -There is no necessity for creation of special facility for manufacturing and radiochemical processing of targets; -99 Mo is selectively taken from solution fuel by sorption on special filters or spontaneously leaves the molten salt fuel in the form of flying fluorides; -Uranium and the basic highly active fission fragments remain in fuel; -Reactors with homogeneous fuel possess immanent nuclear safety (negative thermal reactivity factor); -Research mini-reactor "Argus" with solution fuel in the form of UO 2 SO 4 has operated in Russia for about 30 years; -Research reactor MSRE of power 8 MW with molten salt fluoride fuel LiF-BeF 2 -UF 4 successfully operated in the USA within four years, when the effect of spontaneous exit of a part of fission fragments in the gas phase was observed (including 99 Mo); - There are no problems with reactor operation on LEU.
Now development and creation of a specialized nuclear-chemical complex for commercial production of fragment radioisotopes 99 Mo, 89 Sr, 133 Xe, 131 I, 133 I, etc. for satisfaction of growing requirements of nuclear medicine look actual.

Production of fragment radioisotopes with help of molten salt fluoride nuclear fuel
Transition from the traditional technology of uranium targets irradiation to the technology based on an unique effect of spontaneous exit of fragment 99 Mo from molten salt fluoride fuel to the gas phase (Grimes, 1970), observed in the late sixties in the experimental reactor MSRE (ORNL) allows to simplify a problem of waste disposal at production of fragment radioisotopes considerably. According to the results of measurements, more than 50 % of all molybdenum formed at fission of uranium leaves melts. The most of other fragment elements, such as Zr, Ba, Sr, Cs, Br, I, rare-earth elements and all uranium having steady well soluble in fuel salt compounds remain in melts.
The idea of use of molten salt nuclear fuel in the reactor technology was proposed at the beginning of the 50-ties at creation of small power autonomous energy sources. A distinctive feature of this kind of fuel is the possibility of continuous correction of the chemical structure and control of nuclear-physical, chemical and thermal processes at its work in the nuclear reactor.
The molten salt reactor concept was realized at creation of experimental reactors ARE (Aircraft Reactor Experiment) by power of 2.5 MW and MSRE (Molten-Salt Reactor Experiment) of 7.3 MW. The first experimental molten salt reactor, ARE, constructed in the USA in 1954, worked with salt composition NaF-ZrF 4 -UF 4 at temperature 860 o C (Bettis e.a., 1957). The basic task of the experiment was confirmation of serviceability and stability of the reactor with circulating fuel. MSRE operated in Oak Ridge National Laboratory (ORNL) within four years in the late 60s (Haubenreich & Engel, 1970) with fuel 7 LiF-BeF 2 -ZrF 4 -UF 4 at temperature 650 o C. The purposes of MSRE construction consisted in working out the molten salt nuclear fuel technology, checking the serviceability of some structure units, study of neutronics characteristics of this reactor type. Corrosion steady nickel alloy Hastelloy-N was developed, some questions of processing and cleaning the salt were solved during reactor operation, and its nuclear safety was shown.
Significant researches of molten salt coolants were performed in the Soviet Union (Novikov et al., 1990). The structural materials on the basis of nickel, corrosion stable in fluorides are created, the pump, heat exchanging devices of various purposes etc are developed and checked in conditions of stand tests. Now these researches proceed, mainly within the framework of the international project "Generation-IV" on creation of new generation nuclear reactors.

Behavior of fission fragments in molten salt fuel
In MSRE, the basic group of fragment elements, such as Zr, Ba, Sr, Cs, Br, I and rare earth elements, having well soluble in salt compounds remained in fuel. Insoluble compounds deposited on the surface of structural materials of the reactor facility (graphite, Hastelloy-N). Krypton and xenon, having low solubility in salt, spontaneously or under bubbling of fuel by inert gas left the salt melt. Moreover it was found, that a small group of elements, socalled "noble" metals Mo, Nb, Te, Ru in the form of volatile fluorides or aerosols entered gas covering free surfaces of salt. About half of these elements deposited on surfaces of reactor structural materials and less than 1% remained in fuel salt.
In a series of ORNL 1966ORNL -1970 reports the results of numerous measurements of fuel structure, researches of structural material samples and gas tests from MSRE contour are published. An analysis of these experimental data has allowed us to reveal certain laws of distribution of some fragment elements in the reactor facility, to put forward a hypothesis of the mechanism of "noble" metals exit to gas volume above the salt melt and on this basis to propose a new way of fragment 99 Mo production.
In particular, in (Rosenthal et al., 1969)  The assumption was made, that these fission fragments arising in extremely electron-scarce states, in the process of their thermalization catch electrons, passing a number of unstable, but volatile valent states (Kirslis & Blankenship, 1967). If the reactor structural materials are rather inert, and the fuel does not contain significant amount of other absorbers of radiolytic fluoride, some part of Mo, Te and Ru will oxide up to maximal fluorides and pass to the gas phase.
In work (Rosenthal et al., 1968), 50 g salt samples were located in hermetic reactionary volumes connected to a gas contour, from which helium or mix of helium with hydrogen could be made to fuel melt. For melting of fuel the reactionary volume was heated up to 600 o C. The gas moved in the tube at the flow rate 10 -15 cm 3 ⋅min -1 , allowing both blowing the salt melt mirror and bubbling the fuel. After passage of reactionary volume the gas was evacuated through a system of filters including few sections of fission fragments catchingmetal-ceramic filter, sorbent on the basis of NaF and chemical sorbent. After reactor operation of 32 000 MW×h the filters were removed, each section was separately dissolved in acid and the solutions were analyzed on the contents of fission fragments and 235 U. Some measurement results of the salt radioisotope structure are given in table 3.
Calculated by us total activities of radioisotopes collected in all elements of filtration system, normalized to initial activity in salt test are shown in the same   140 Ba, though cumulative yields of these fragments differ not more than twice. It is impossible to explain the so large difference by burning up of fragments in the reactor, at the given cross sections. The most probable reason is connected to individual behavior of chemical compounds of these elements in molten salt fuel.
In Molybdenum has a few stable fluorides, which can be formed under -irradiation of the nuclear reactor -MoF 3 , MoF 4 , MoF 5 , MoF 6 . At the fuel melt temperature ≈600 0 C, significant fractions of tetra-and pentafluoride of molybdenum MoF 4 , MoF 5 and also molybdenum hexafluoride MoF 6 (boiling temperature of 36.2 0 (Nikolsky, 1998) are in the gas phase. In case these compounds arise near to borders with the gas phase, they can leave the melt.
In work (Kirslis & Blankenship, 1969), the degree of fission fragments penetration in graphite of MSRE core was investigated. As well as for the previous experiment, we were interested by penetration into graphite of generically connected pair 95 Zr→ 95 Nb. It follows from experimental data that 95 Nb concentration in graphite has appeared approximately by three orders of value higher, than 95 Zr. Their distribution on depth of the graphite sample is shown in figure 3.
It is known, that the radioactive atoms formed as a result of nuclear transformations, partially appear in chemical compound, at which they were formed, as so-called "parent" compound (Nesmeyanov, 1978). Taking into account that zirconium has sole stable fluoride -ZrF 4 , we estimated the possibility of preservation of the initial chemical form of a molecule as tetra fluoride after decay of a parent nucleus 95 Zr and formation on its place of 95 Nb nucleus.
The energy of the exited molecule, which can be spent on break of the chemical binding, is equal: where Е is recoil energy, is mass of the recoil atom, M R -mass of the rest of molecule. 0.44E will be spent on break of the chemical binding for a molecule ZrF 4 , i.e. about a half of 95 Zr recoil energy.
At β --decay of 95 Zr basically two groups of β --particles with average energies of 121 and 110 keV (Burrows, 1993) (Kirslis & Blankenship, 1969) As the binding energy of atoms in molecules is 2-5 eV (Nesmeyanov, 1978), in most cases of 95 Zr decay in a molecule 95 ZrF 4 the formed new atom 95 Nb will be kept in the chemical form of the parent compound, i.e. as a molecule 95 NbF 4 . The further fate of niobium tetra fluoride 95 NbF 4 is determined by thermodynamics and radiation-chemical processes in molten salt fuel. In the reactor γ-field formation of niobium pent fluoride 95 NbF 5 is possible, which at temperature ≈600 o C will be in gas form (Т boiling = 234 o C), and at presence of the phase border "liquid -gas" it will leave fuel melt, penetrating, including, into open gas pores of graphite, which is in contact with molten salt fuel. The boiling temperature of zirconium fluoride is Т boiling = 912 0 , that is much higher than the fuel temperature. Therefore 95 ZrF 4 does not enter gas bubbles and, accordingly, does not penetrate into pores of graphite that explains significant difference of 95 Zr and 95 Nb contents in graphite.
www.intechopen.com In our opinion, the observed experimental results can be explained by formation and exit of easily volatile fluorides of noble metals Mo, Nb, Te from molten salt fuel to the gas phase. In detail the mechanism of 99 Mo exit from molten salt fuel is considered in (Zagryadsky et al., 2008;.
The proposed radiochemical model of "noble" metals exit from fluoride fuel in a gas phase assumes, that chemical processes in fuel LiF-BeF 2 -UF 4 vary sharply in the field of an ionizing radiation and are determined by a number of active particles -a complex ion BeF 4 2 , radicals 4 BeF -,

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Be Fand a number of other ions. The fragment molybdenum formed in molten-salt fuel at uranium fission exists in the melt mainly in the form of complex ions, such, for example, as within characteristic time τ ≈ 3 ⋅10 -4 s, molybdenum ions can form at temperature ≈ 600 o C volatile fluorides MoF 4 , MoF 5 , MoF 6 which, being near to borders with a gas phase will leave the melt. Due to presence in a gas phase of radiolitic fluorine, molecules 99 MoF 4 , 99 MoF 5 are oxidized to a higher fluoride 99 MoF 6 which is delivered by helium flow into catching system along transport communications.

Advantages and possible variants of realization of fragment 99 Mo production on the basis of molten salt nuclear fuel
Basing on the results of experiments performed within the framework of the research program at the reactor MSRE and the hypothesis of the mechanism of 99 Mo exit from molten salt fuel, we proposed a new way of fragment 99 Mo production, using a spontaneous (or stimulated by bubbling of fuel with inert gas) exit of fragment 99 Mo from molten salt fuel to the gas phase. For the first time this idea was mentioned in (Chuvilin & Zagryadsky, 1998).
The aggregate condition of fuel on the basis of Li, Be and U fluorides allows to take 99 Mo, 89 Sr, 133 Xe from salt melt selectively, leaving 235 U, and the basic group of fragment elements. As a result inherent for solid fuel systems radiochemical processing of the irradiated uranium is excluded, what at the end should result in reduction of highly active waste and of the flow of fission materials in the technological process.
Principally the process of 99 Mo production in a research reactor can be organized as follows (Chuvilin & Zagryadskiy, 1997). One of fuel assemblies of the reactor core is replaced with a loop facility filled with molten salt fuel, attached to a gas contour. As a result of 235 U fission, in the salt melt 99 Mo is generated, which moves to the salt -gas border and passes to the gas phase above the salt melt surface in form of fluorides and aerosols. 99 Mo removal can be intensified by bubbling of helium or argon through salt.
The collected in free volume of the loop system gas-aerosol fraction is removed by purge of the cavity with inert gas. The gas -carrier containing aerosols and volatile fluorides of molybdenum enters the catching system, which can be realized as a set of filters, chemical sorbents or freezing traps. Cleaned from fragment elements, inert gas recirculated in the loop system. For 99 Mo removal, the filters periodically are directed to radiochemical processing. Accompanying molybdenum fragment elements deposited in the catching system are utilized after an appropriate procedure of extraction. The basic scheme of fragment 99 Mo production in the molten salt reactor is shown in Fig. 4.
The possible design solutions of the loop system will be determined by the technology of 99 Mo removal from molten salt fuel. At the first stage of its introduction it is expedient to use elementary devices that will allow checking up the basic technical decisions, investigating radionuclide purity of the product, estimating economic parameters of the proposed way of 99 Mo production.

Experimental ampoule loop system "RAMUS"
For demonstration of physical practicability of the proposed new technology of 99 Mo production, in the research reactor IR-8 of the NRC "Kurchatov Institute" experimental ampoule loop system "RAMUS" with molten salt fluoride fuel is created, where the process of 99 Mo production will be realized according to (Chuvilin & Zagryadskiy, 1997). It is planned to determine the efficiency of 99 Mo extraction from fuel, possibility of transporting molybdenum fluoride compounds from the reactor core to the accumulating zone for extraction from the gas -carrier flow, and other parameters. Besides that it is planned to produce as well two other accompanying radioisotopes of medical purpose: 89 Sr, 133 Xe in the loop system "RAMUS".
The loop system "RAMUS" will be placed in the beryllium reflector of IR-8 having neutron flux ≈10 13 n⋅cm -2 ⋅s -1 . The loop system has the following characteristics: The basic scheme of the ampoule part of the loop system "RAMUS" is shown in Fig. 5. The radioisotopes 99 Mo, 89 Sr, 133 Xe will be evacuated from molten salt fuel by bubbling of melt with helium. After passage of gas-carrier along the communications and filters, in which www.intechopen.com Low-Waste and Proliferation-Free Production of Medical Radioisotopes in Solution and Molten-Salt Reactors 151 decay and depositing of short-lived fission products take place, the target radioisotopes are caught in depositing devices, from which they are extracted subsequently.
During design of the loop system "RAMUS" some technical decisions have required to be experimentally substantiated. In particular, the data on formation and carrying over of aerosol particles in the process of molten salt bubbling by inert gases and spontaneous thermocondensation of salt component vapors are practically absent in the literature. At bubbling the large interphase surface on "liquid-gas" border is created that promotes an intensification of mass-exchange processes, and also a more complete chemical interaction of gases with liquids.
The experiments on formation and carrying over of aerosol particles by helium in the process of molten salt fuel bubbling were performed to make more precise the parameters of the unit "RAMUS". The regime of the vertical bubble flow in molten salt 66LiF-34BeF 2 was investigated at temperature 530-630 o C. The experiments are carried out on a model installation reproducing the dimensions of the ampoule "RAMUS". Weight and height of the salt melt made 900 g and 450 mm, respectively. For creation of the gas bubble flow, a tube was immersed in the salt melt, on which bottom end cylindrical sprayers with face or lateral holes were installed. 6 types of sprayers were tested. The stabilized helium flow rate varied in the range of 1.5-4.5 cm 3 s -1 .
The rate of carrying over of aerosol particles was determined with help of the photometer FAN-A. Besides that the aerosols were taken out on filters and then the deposit was weighted and analyzed by ICP-AES and DTA-TDA methods.
Foam was observed on the liquid surface in the experiments with bubbling of the salt melt by helium. It was possible to get the minimal height of the foam layer -15-18 mm with the sprayer of three holes ∅=0.5 mm, L = 0.5 mm located uniformly on the circle. The mass concentration of aerosol particles did not exceed 100 mg⋅m -3 , and the rate of their carrying over -0.1 µg⋅s -1 . The size distribution of aerosol particles has appeared bimodal. The first mode (superfine aerosol, average particle diameter ≈0.05 microns) is caused by spontaneous thermocondensation of salt components vapors at their cooling from 630 o to 60 o . 90÷95 percents of aerosol particles mass consist of beryllium difluoride. That is because BeF 2 vapors pressure at temperature 630 0 more than 50 times exceeds LiF vapors pressure.
The second mode of submicron aerosol is caused by collapsing of bubbles on the boundary surface "foam -gas" and corresponding dispersion of liquid. The diameter of particles taken out from an ampoule does not exceed 4 microns, at average diameter 0.5 microns. Their chemical structure coincides with the structure of the salt melt. It means that the carrying over of salt components in form of aerosols at bubbling of melt is determined by dispersion of liquid at collapsing of bubbles on the surface of two phase boundary, and the contribution of BeF 2 superfine particles generated at vapors spontaneous thermocondensation makes less than 5% of weight.
The received experimental results have allowed making a conclusion that for the planned term of the loop system "RAMUS" operation in the reactor IR-8 channel the convective carrying over of salt components in form of vapors and aerosols will be negligible small and will not influence gas-dynamic characteristics of communications and the gas flow filtration system of the loop system. In other series of the experiments, the possibility of MoF 6 loss at its transport along technological communications of the loop system was studied depending on temperature. Sorption of MoF 6 from the gas mixture flow (He + MoF 6 ) with the contents of molybdenum hexafluoride in the range 10 12 -10 13 cm -3 on granulated NaF was also investigated. The content of MoF 6 in the gas mixture corresponds to the calculated concentration of fragment molybdenum in the loop system "RAMUS". The infra-red Fourier-spectrometer FSM-1202 is used for MoF 6 measurement in the gas mixture. The experiments are carried out for the middle and near infra-red areas of the spectrum, with use of multirunning flask with the optical way 4.8 m long, which has allowed measuring MoF 6 concentration down to 10 12 cm -3 . The measurements have shown that deep extraction of MoF 6 from the gas flow takes place even at high flow rates of the mixture up to 10 cm 3 ⋅s -1 .
The process of MoF 6 desorption from the sorbent surface becomes noticeable at temperature 100 o C. However, as in reactor conditions the temperature at location of the column with sorbent will not exceed 60°C, the effect of MoF 6 desorption in the loop system "RAMUS" could be neglected. The measurements of MoF 6 losses during its transport along metal communications specially prepared in fluoride atmosphere have shown that they are within uncertainty limits of the spectrometer FSM-1202, and the loss of target product could be neglected.
Thus, the carried out pre-reactor experiments have allowed estimating a number of technological parameters of the loop system "RAMUS", choosing the optimal design of pneumatic sprayers and the regime of molten salt bubbling, predicting possible losses of MoF 6 at its transport along gas communications of the loop system.

Production of 99 Мо in solution reactors
The reactor is named "solution" if its fuel is in the form of water solution of uranyl-sulphate UO 2 SO 4 or uranyl-nitrate [UO 2 (NO 3)2 ] (Kolesov, 1999). At production of 99 Mo in the solution reactor, the necessity of target manufacturing and their radiochemical processing disappears. All the reactor fuel is a target. 99 Mo is taken from solution by sorption on special sorbent.
Russell Ball (companies Babcock & Wilcox and Ball Systems, the USA) has proposed a design of the liquid homogeneous reactor MIPR (Medical Isotope Production Reactor) for production of radioisotopes, in particular 99 о (Ball, 1999). This reactor can work on uranylnitrate solution in water with low-enriched 235 U (to 19 %) as fuel composition. At power of 200 kW it is capable to produce by estimations up to 2000 Ci of 99 о a day. The proposed design of a reactor and its characteristics allows receiving the permission to its placing even in the centre of big cities. The power of a solution reactor is more than 10 times less than of a traditional research reactor and demands hundreds times smaller amount of 235 U for production of equal amount of 99 о. According to estimations, building and physical start-up of a reactor will take 36 months with the total cost from 6 to 26 million US dollars.

Production of fragment 99 Mo on basis of solution reactor ARGUS
with uranium concentration 81.3 g/l and enrichment by 235 U 90 % (Pavshouk & Chuvilin, 2005;Afanasev et al., 1986) is used as fuel.
The fuel solution is located in the case -a welded cylinder with a hemispherical bottom and a flat cover in which a cooling coil is located. Vertical "dry" channels are placed in the case: central and two symmetric peripheral channels in which control and safety rods are located. The reactor case is surrounded by a lateral and bottom face graphite reflector. Gaseous radiolysis products of fuel solution are regenerated by means of the system including a catalytic recombiner, which together with the case forms a tight system excluding leak of fission products to the environment. The recombination system is based on the principle of natural circulation of the gas mix along the contour. The solution type of the reactor provides conditions of the maximum safety which is guaranteed by the big negative reactivity effect and the optimum concentration of uranium in solution, leading to selfregulation of the reactor facility.
Since the middle of the 90s years the National Research Centre "Kurchatov Institute" has been working out a technology and creating a demonstration nuclear-technological complex of fragment 99 о production on the solution reactor "Argus". The aggregate state of the reactor "Argus" fuel allows 99 о extraction directly from fuel solution without influence of structure and characteristics of the reactor core. For that, after reactor operation on power solution is pumped through special sorbent on the basis of titanium oxide which provides the first stage of 99 о extraction from all mass of fission products. Thus uranium is not absorbed, and fuel solution completely comes back to the reactor case. All amount of the uranium which has undergone fission reaction, necessary for maintenance of reactor power, is spent for production of the target isotope.
This newest technology of 99 о production allows lowering the necessary reactor power 100 and more times as compared to the traditional technology that not only results in reduction of radioactive waste production, but also supposes reactor locating in the occupied areas and promotes solving the licensing problems. Thus efficiency of 235 U use comes practically to 100 % while in target technologies only 0.4 % of 235 U are spent for getting the product, and its other part after target processing, as a rule, goes to waste (Starkov, 1996). The necessity of uranium targets manufacturing is, besides, excluded. The developed technology allows to solve an actual problem of uranium enrichment reduction in production of 99 о and to eliminate public concern in connection with non-proliferation problem. The reactor which core is water uranylsulphate solution is used for accumulation of 235 U fission fragments. Thus the yield of fragment 99 о per fission of 235 U makes 6 %. Accumulation of 99 о in the core during the reactor operation on power of 20 kW is shown in Fig. 6.
For demonstrating the practicability of 99 о extraction technology directly from fuel solution, the demonstration experimental complex of the reactor "Argus" presented in Fig. 7 and including a reactor loop, transportation means and hot cells equipment for purifying 99 о from fission products has been designed and put into operation.
As a result of reactor experiments the possibility of 99 о extraction from reactor solution fuel has been shown for the first time, and also the following technological process has been developed. The reactor "Argus" operates at power within 5 days. After some endurance, the fuel solution is pumped through sorbent within 3-6 hours. For returning of the rests of fuel www.intechopen.com solution into the reactor case the reactor loop together with the sorption column is washed out by the condensate which has been accumulated in a system of radiolysis products regeneration during power operating time of the reactor. Then the column is remotely placed in a transport container and goes to hot cells, where after sorbent washing by acid solutions radionuclide 99 о is desorbed with alkali. For getting demanded radionuclide purity two stages of purification with help of refinery columns filled with the same sorbent of type "Thermoxid" are taken. Using the created demonstration equipment reactor experiments are performed on production of 99 о samples and working out of solution technologies. Quality certification of the samples was performed in Europe (Institute of radioactive elements, Belgium), and in the USA (Argonne National Laboratory) where the impurity content in the sample more than 10 times less than in the international requirements (see table 4) was confirmed. The maximum activity of samples made ~ 10 Ci.
On the basis of 99 о produced in the solution reactor "Argus", generators of 99m Тс with the standard column filled with sorbent of type "Thermoxid" have been made. Eluted 99m Тс has been investigated on injection suitability in State Scientific Center "Institute of Biophysics" of the Russian Federation and has received a good rating as for radionuclide purity, and absence of macro impurity of stable elements (manganese).  Table 4. Results of 99 о samples quality analysis

Technology of radionuclide 89 Sr production in the solution reactor
One of the most effective ways is production of 89 Sr from uranium fission products, but formation of impurity 90 Sr makes impossible practical application of this way when uranium targets are used. The solution reactor type gives a unique possibility of solving this problem.
The method of fragment radionuclide 89 Sr extraction is based on a possibility during fission reaction in nuclear fuel of solution reactor to influence the genetic predecessors of a target radioisotope resulted in nuclear transformations of fission fragments in a decay chain of elements with the mass number 89: 89 Se→ 89 Br→ 89 Kr→ 89 Rb→ 89 Sr. The earlier experimental researches have shown that practically all long-living isotopes of krypton and xenon leave in the gas phase of solution reactors (Loboda et al., 1989). The basic decay chains of the fission products, leading to formation of strontium radioisotopes with a half-life period of gaseous predecessors more than 1 second, are presented in Fig. 8. www.intechopen.com Apparently from consideration of the basic decay chains leading to formation of radioactive strontium isotopes, the half-life period of the gaseous predecessor ( 89 Kr) of a target radionuclide 89 Sr makes 3.2 minutes, and the half-life period of a gaseous predecessor ( 90 Kr) of the main impurity radionuclide Sr 90 is essentially less and makes 33 s. Using this circumstance, in the isolated volume of a gas after corresponding endurance it is possible to reach demanded value of the activity ratio of target gaseous predecessors and impurity strontium radionuclides, and then to pump over the gas mix through filters in accumulating volume in which after decay of predecessors 89 Sr will accumulate (Loboda et al., 1989;Abalin et al., 2000). An experimental loop system consisting of a closed technological loop with rotation pumps, valves and vacuum-tight connecting units was used for taking gas samples from the reactor free volume. The hydraulic scheme of the experimental installation is shown in Figure 9. The experiments confirming the possibility of 89 Sr extraction from fuel solution of the reactor "Argus" were performed under the following scheme. The reactor operated at power from 5 to 20 kW within 20 minutes. Then after endurance within 5-6 minutes the gasair mix was pumped over through a column filled with Rashig rings. The reactor loop mentioned above was used. Pumping time of the gas mix was 5 minutes, which is enough for filling the column. The columns were removed after 2-3 days of gaseous fission products decay. In the laboratory the aggregate taken from the column was washed out with solution of hydrochloric acid for extraction of 89 Sr and some accompanying fission products.
A series of experiments for confirmation of fragment 89 Sr production using exit of the gaseous predecessor 89 Kr from fuel solution was performed. Samples of 89 Sr chloride of high radionuclide purity are got in which 90 Sr impurity content has made less than 5·10 -4 %. The typical content of the produced 89 Sr chloride solution is given in Table 5.
With the obtained experimental estimations of 89 Sr yield in a solution reactor we can compare the specific productivities of reactors BR-10 (Russia, Obninsk) (Zvonarev et al., 1997), BOR-60, BR2 (Koonen, 1999).) with "Argus" reactor, normalized to the target mass and reactor power. For the sake of comparison, we accept the mass of 235 U in the solution equal to 1600 g, as a target. Table 6 shows specific yields of 89 Sr in various reactor installations. Fig. 9. Scheme of reactor "Argus" with experimental loop. 1 -reactor core; 2 -system of catalytic regeneration; 3 -accumulator of condensate; 4 -heat exchanger; 5 -catalytic recombiner; 6 -valves; 7 -sorption column; P1, P2 -pumps.

Radionuclide
Activity The advantage of the 20 kW solution reactor is obvious and undeniable. The expected "Argus" reactor productivity of (1.5-1.8)·10 3 GBq of 89 Sr a year is comparable with that of the 60-MW BOR-60 Reactor (≈ 4·10 3 GBq/y). However, it is worth noting that 89 Sr in a solution reactor is a byproduct from its "main" production of medical 99 Mo/ 99 Tc. The target radionuclide separation occurs spontaneously; a majority of fission fragments remains in the fuel, and the buildup of radioactive waste with the new technology is very low. The separated impurity radionuclides 133 Xe and 135 Xe can also be used in medicine, while 137 Cs is a valuable radionuclide for technical applications.
The unique properties of the solution reactor fuel open the doors for this type of nuclear installations to the nuclear medicine industry. Separation of radionuclides directly from fuel solution allows using the low power (50 kW) reactor for production of 99 Mo with the minimum yield of 1.85⋅10 4 GBq/day, 700 GBq/week of 133 Xe, as well as a number of other valuable radionuclides (Burrows, 1993;Zagryadsky et al., 2008).
So, the new method for medical 89 Sr production in the reactor with solution fuel is proposed which is characterized by simplicity, high production efficiency and low buildup of radioactive waste. The main advantages of the new technology were validated by numerous experiments. Basic features of the method are as follows: • the mechanism for 89 Sr delivery to the sorption volume of the "Argus" reactor experimental loop is based on transport of gaseous 89 Sr predecessor -radionuclide 89 Kr; • the radionuclide impurity composition of 89 SrCl 2 solution includes elements with nuclear masses 137 , 140 and 141 , which have gaseous elements in their decay lines; • filtration of the gas flow with mechanical aerosol filters reduces the content of radionuclide impurities 10 2 -10 3 times; • cleaning of 89 Sr chloride solution in chromatographic columns with DOWEX-50×8 or Sr-Resin ensures full removal of 137 Cs from the solution and significant reduction of 140 Ba/ 140 La impurity; • 90 Sr impurity in 89 SrCl 2 solution was not detected with the measuring instruments within the sensitivity limits of (≈5⋅10 -4 %).

Conclusions
The key problem of modern medical radionuclide production in research nuclear reactorsconsiderable amount of a highly active waste and use in technology of highly-enriched uranium -can be solved by transition to use of homogeneous liquid nuclear fuel enriched by isotope of uranium 235 U less than 20 %. It is known (Basmanov, E. et al., 1998), that conversion traditional «target» technology of fragment radioisotopes production from HEU to LEU in a target cycle not only does not solve a problem of a radioactive waste (the yield of fission products per unit 99 Mo activity is the same), but also aggravates it: the amount of the irradiated uranium, the radioactive waste accompanying processing of more uranium for the same 99 Mo yield, both accumulation of 239 Pu and other transuranic elements increase in times. That increases the product cost and difficulty of disposal. A target turnover cycle of fission materials is absent in the proposed concept of medical isotopes production. Therefore, on the one hand, the overwhelming part of the waste connected with fission of uranium appears "locked" in a reactor core or a loop installation, and, on the other hand, there is no problem of fission materials proliferation in a target cycle. The question of conversion of the considered reactors with liquid fuel from HEU to LEU does not concern actually technology of medical isotopes production, and should be considered within the general strategy of nuclear reactors safety. Technical possibility of such conversion does not call doubts.
The accumulated operational experience with solution and molten salt fuel, presence of the developed experimental base provide a possibility for performing a broad spectrum of technological researches directed on creation and realization of new methods for production of radionuclides and radiopharmaceuticals for nuclear medicine.
Carried out in the NRC "Kurchatov Institute" model experiments at the solution reactor "Argus" have confirmed basic provisions of the new law-waste technologies of 99 о and 89 Sr production.
This newest technology of the most important medical radionuclide production allows to lower necessary power of a reactor 100 and more times comparing to traditional technology that not only leads to decrease in total radioactive wastes of production, but also supposes placing the reactor in the occupied areas and promotes solving the licensing problems. The efficiency of 235 U use comes practically to 100 % while traditional target technologies use only 0.4 % of 235 U for reception of the final product, and other part of uranium after target processing, as a rule, goes to utilization.