RU2276817C2 - Method for producing strontium-89 radioisotope - Google Patents

Abstract

FIELD: nuclear engineering.

SUBSTANCE: proposed method for producing strontium-89 radioisotope includes irradiation of fuel in the form of aqueous solution of uranium salts in nuclear reactor core to obtain fragments of fission of which one is krypton-89 (predecessor of strontium-89 in disintegration chain of fission fragments with atomic mass of 89). Krypton-89 is cleaned of accompanying krypton radioisotopes, including krypton-90 (predecessor of long-living radioisotope strontium-90) due to natural decay. In the process krypton-89 is cleaned by extending residence time of fission fragments in fuel up to disintegration of short-living radioisotopes, including krypton-90, therein due to raising pressure within nuclear reactor vessel and reducing specific energy intensity of fuel. Cleaned krypton-89 is conveyed for weathering till its complete decay to target radioisotope.

EFFECT: facilitated procedure, enhanced reliability of method.

1 cl, 1 dwg, 1 ex

Description

FIELD OF THE INVENTION

The invention relates to a reactor technology for producing radioisotopes.

The present invention can be used for the production of the strontium-89 radioisotope, which has found wide application in nuclear medicine in the treatment of cancer.

State of the art

More than half a century of experience in the use of radioisotopes in nuclear medicine for diagnosis and therapy has made it possible to determine their place and main indications in which the use of radiopharmaceuticals (RFP) for the treatment of cancer and other diseases is effective. The production of medical radioisotopes has become an important industry sector, which accounts for more than 50% of the annual production of radioisotopes worldwide. The scale of the use of radioisotopes in medicine is evidenced, for example, by the fact that today, one in four patients who go to the polyclinic in the United States and one in three who go to the hospital are referred for diagnostic and therapeutic procedures using radioactive isotopes. The total number of such procedures is tens of millions per year.

The rapid development of radioisotope diagnostics and therapy in world practice is due to the ongoing development of new modifications of recording equipment and especially new radiopharmaceuticals.

One of the most effective therapeutic radioisotopes is strontium-89. It is used as an alternative method or addition to external radiation therapy for the treatment of pain in bone metastases.

Strontium-89 has the following radiation characteristics:

half-life T _1/2 = 50.5 days;

the decay method is β ^- (decays into a stable isotope ⁸⁹ Y);

maximum energy of β ^- particles E _β = 1463 keV;

energy of concomitant γ-radiation E _γ = 909.1 keV;

the output of γ-quanta is 9.30 · 10 ^-5 (Bq · s) ^-1 .

Strontium-89 was first used for pain relief in the early 1940s [Pecher C. Biological investigations with radioactive calcium and strontium: preliminary report on the use of radioactive strontium in the treatment of metastatic bone cancer. // Univ. Calif. Publ. Pharmacol 1942; 2, pp. 1117-1149]. Strontium, being a biochemical analogue of calcium, has the same transport mechanism in the human body. With intravenous administration of strontium chloride, its accumulation occurs mainly in bone metastases, where active osteoblastic processes occur. Currently, radiopharmaceutical based on it - strontium chloride-89 is used for medical purposes in the palliative treatment of patients suffering from severe pain due to tumor metastases in the bone system. Malignant neoplasms with a tendency to metastasize to the skeleton include: breast, colon, lung, thyroid, kidney, prostate and skin cancers. The maximum range of β ^- particles of strontium-89 in bone tissue does not exceed 7 mm and thus its radiation exposure can be localized in a sufficiently small region of the skeleton, reducing the dose load on the bone marrow and adjacent areas of soft tissues.

The fraction of the drug remaining in the bones is proportional to the volume of bone metastatic lesion and ranges from 20 to 80% of the administered activity. Being embedded in the mineral structure of the affected area, strontium-89 does not metabolize and remains in it for about 100 days. Normal bone includes an insignificant part of the administered dose and actively loses it during the first 14 days. Clinical trials of the drug showed that 65 ÷ 75% of patients report a significant decrease in pain effect, and in 20% -ax cases we are talking about complete pain relief. Moreover, doctors believe that strontium chloride-89 has a therapeutic effect, that is, it not only blocks metastases, but also suppresses them.

A known reactor method for producing strontium-89, based on a threshold neutron capture reaction with the release of a charged particle ⁸⁹ Y (n, p) ⁸⁹ Sr [Zvonarev A.V., Matveenko I.P., Pavlovich VB and others. "Obtaining strontium-89 in fast reactors" // Atomic energy, vol. 82, issue 5, May 1997, pp. 396-399; Toporov Yu.G., Filiminov VT, Kuznetsov RA et all. "Production of ⁸⁹ Sr Using (n, γ) - and (n, p) -Reactions". // Proc. Third Int. Conffer. "On Isotopes Isotope Production and Applications in the 21 ^st Century", Vancouver, Canada, September 6-10, 1999].

In this method, a target containing the natural yttrium-89 monoisotope is irradiated in a neutron flux of a fast neutron nuclear reactor, and then subjected to radiochemical processing by extraction method. Under optimal irradiation conditions, the yield of strontium-89 can be ~ 10-15 mCi per gram of yttrium. The target is high purity yttrium oxide Y ₂ O ₃ , compressed and calcined at a temperature of 1600 ° C. The method is convenient because during its implementation there is practically no radioactive waste, and the final product does not contain harmful impurities - the amount of the accompanying radioisotope of strontium-90 is less than 2 · 10 ^-4 at.%.

However, this method has an extremely low productivity due to the small cross section of the (n, p) reaction at ⁸⁹ Y, the value of which for the neutrons of the fission spectrum does not exceed 0.3 · 10 ^-27 cm ² , and its implementation is possible only in reactors with a fast spectrum neutrons, the number of which is very small. In addition, for irradiation, it is necessary to use yttrium, which has undergone deep purification from impurity uranium (the uranium content in Y ₂ O ₃ tablets should not exceed 10 ^-5 % by weight).

For the prototype, a method was selected for producing the strontium-89 radioisotope in a solution reactor [RF Patent No. 2155399, IPC G 21 G 1/08, BI No. 24, 2000]. In this method, fuel in the form of an aqueous solution of uranyl sulfate UO ₂ SO ₄ is irradiated in a neutron flux of a nuclear reactor. As a result of uranium fission, fragments are formed, one of which is krypton-89, which is the precursor of strontium-89 in the decay chain of fission fragments with an atomic mass of 89, and is removed from the reactor core by a stream of radiolytic gas - hydrogen and oxygen generated in the fuel solution due to water radiolysis under the influence of fission fragments. The gaseous fragments that have escaped from the fuel are transported to a pre-exposure system, where krypton-89 is held for enough time for the natural decay of the short-lived krypton radioisotopes accompanying it, and then krypton-89 is sent to a capture system, where it is kept until the decay into the target strontium radioisotope -89.

Uranium-233, and / or uranium-235, and / or plutonium-239 can be used as fissile material in the fuel solution.

In this method for the production of the strontium-89 radioisotope, the possibility is used in the process of neutron irradiation of solution nuclear fuel to influence not only the target radioisotope, but also its genetic predecessors, resulting from nuclear transformations of fragments in the decay chain of elements with a mass number of 89 ⁸⁹ Se → ⁸⁹ Br → ⁸⁹ Kr → ⁸⁹ Rb → ⁸⁹ Sr. As can be seen from this scheme, one of the elements of the decay chain is ⁸⁹ Kr, the radioactive isotope of the inert gas of krypton. As a result of the so-called “radiolytic boiling” of the fuel solution under the influence of fission fragments, krypton-89, without entering into chemical interaction with the fuel, leaves the solution, accumulating in free space above the liquid mirror. The transportation of fragmentation krypton to special containers isolated from neutron irradiation, and the shutter organized in two stages, provide not only the complete decay of krypton-89 into the target radioisotope of strontium-89, but also its purification from the accompanying radioisotopes of krypton, including krypton -90, giving the decay of the most regulated impurity radioisotope strontium-90. A significant difference in the half-lives of the krypton-89 and krypton-90 radioisotopes, amounting to 190.7 and 32.2 sec, respectively, allows, by holding the gas phase removed from the fuel solution, the content of strontium-90 impurity in the target radioisotope can be reduced to ~ 10 ^{- 4} at.% And thus ensure high radioisotope purity of strontium-89.

The high productivity of the considered method for producing the strontium-89 radioisotope is due to the high cross section of the fission reaction (n, f) on such nuclei as ²³⁵ U, ²³³ U or ²³⁹ Pu, reaching 600-800 · 10 ^-24 cm ² for thermal neutrons, and the yield Krypton-89 fragment - about 4-5% in the act of division.

The possibility of using homogeneous liquid fuels in nuclear reactors was demonstrated back in the 1950s and 1960s. According to publications of those years, it is known that more than 20 research reactors with fuel based on aqueous solutions of uranium salts operated in the world. These include the reactors HRE-1, HRE-2, LOPO, HYPO, SUPO, etc. [Tishchenko VA, Smirnov Yu.V., Raevsky II US Research Reactors: AINF Review, 588, M., ZniAtominform, 1984]. All these reactors use an aqueous solution of uranyl sulfate UO ₂ SO ₄ with various enrichment in the ²³⁵ U isotope as fuel, graphite is a reflector, water is a moderator and a coolant.

Homogeneous liquid fuel reactors have several advantages over solid fuel reactors. They have negative temperature and power effects of reactivity, which ensures their high nuclear safety. The design of the core is greatly simplified: there are no claddings of fuel elements, spacers and other parts that degrade the neutron characteristics of the reactor. The solution preparation procedure is significantly cheaper than the manufacture of TVE Fishing. Loading (pouring) mortar fuel is also much simpler, which allows you to change the concentration of fissile material in the fuel or the volume of the solution if necessary. In the active zone of the solution reactor, due to the good heat transfer conditions, it is impossible to form local overheating caused by the distortion of the energy release fields. These reactors are simple and reliable in operation, do not require numerous personnel for maintenance.

Today in Russia there are several reactors with liquid fuel - IIN, VIR, IGRIC, Argus. The most powerful among them is the Argus research mortar reactor, which operates in a stationary mode at a power of 20 kW. The reactor was developed at the Russian Research Center Kurchatov Institute and put into operation in 1981 [Afanasyev N.M., Benevolensky A.M., Wenzel O.V. and others. Argus reactor for laboratories of nuclear physical methods of analysis and control. // Atomic energy, t.61, issue 1, July 1986, pp. 7-9].

However, the complex multistage technology for producing strontium-89 according to the method chosen for the prototype, the need to use a highly efficient filtration system to capture volatile and aerosol impurities that need periodic regeneration, are the main constraints in the expanded production of the strontium-89 radioisotope in this way.

Disclosure of invention

The problem to which the invention is directed, is to simplify the method of producing the strontium-89 radioisotope at the stage of fuel irradiation in the active zone of a solution nuclear reactor and its subsequent purification from the impurities of the accompanying radioisotopes, including krypton-90, which gives the most stringent decay regulated strontium-90 impurity radioisotope.

The problem is solved in that in the method for producing a radioisotope of strontium-89, which includes irradiating the core of a nuclear reactor with fuel in the form of an aqueous solution of uranium salts to produce fission fragments, one of which is krypton-89, which is the precursor of strontium-89 in the fragment decay chain fission with an atomic mass of 89, and the purification of krypton-89 due to the natural decay of the accompanying radioisotopes of krypton, including krypton-90, the precursor of the long-lived radioisotope strontium-90, characterized in that the purification of cr pton-89 is carried out by increasing the time spent by fission fragments in the fuel before decaying short-lived radioisotopes in it, including krypton-90, by increasing the pressure in the body of the nuclear reactor and reducing the specific energy density of the fuel, and the purified krypton-89 is sent for exposure to it complete decay in the target radioisotope.

A significant difference in the half-lives of the krypton-89 and krypton-90 radioisotopes, which are 190.7 and 32.2 sec, respectively, allows reducing the content of strontium-90 impurity in the target radioisotope to the level by up to ≈ 400 seconds ~ 10 ^-4 at.% And thus ensure high radioisotope purity of the target product - strontium-89 for medical purposes.

It is known from the experience of solution reactors that radioactive noble gases do not form stable chemical compounds in solution fuel and for the most part leave the solution, remaining in the gas phase above the liquid surface [Loboda S.V., Petrunin N.V., Charnko V.E. ., Khvostionov V.E. Removal of fission products from the fuel of a solution reactor // Atomic Energy, vol. 67, issue 6, December 1989, pp. 432-433]. The mechanism for the removal of noble gases of fragmentation origin from the fuel solution is associated with “radiolytic boiling” in solution reactor systems, which involves the formation of gas bubble nuclei on the tracks of fission fragments. According to this model, which was confirmed experimentally, first vapor bubbles form on the tracks of fission fragments, which, cooling and contracting, pass into gas bubbles containing the products of radiolytic decomposition of water - hydrogen and oxygen. Atoms of radioactive gases - uranium fission products, diffusing into gas bubbles, are carried out from the fuel solution. At normal atmospheric pressure, the release time of radiolytic gas bubbles is only a few seconds [Atomic energy, vol. 67, issue 6, December 1989, pp. 432-433], which allows the removal of relatively short-lived radioisotopes, including those like krypton 89. Once in the gas shell, krypton-89 “pops up” to the “liquid-gas” interface and passes into the free volume above the surface of the liquid fuel.

The drag length of fission fragments in water L is ≈ 0.002 cm [Physical quantities: Reference. Ed. Grigoryeva I.S., Meilikhova E.Z. // Moscow, Energoatomizdat, 1991]. Hereinafter, it is considered that the characteristics of water and fuel solution do not differ. The energy of the fragments (Е _f ≈170 MeV) for a time of ~ 10 ^-10 s passes into the internal energy of the vapor bubble, which at the initial moment is a cylindrical track 2L = 0.004 cm long and a radius of 3 · 10 ^-5 cm [V. Kolesov. Aperiodic pulse reactors, Sarov, RFNC VNIIEF, 1999]. Inside the bubble is N _m molecules of water vapor (H ₂ O) at a temperature of 100 ° C and a pressure of P = P ₀ .

N _M ~ E _f / q ~ 10 ⁸ , (2)

where q = 500 cal / g is the heat of vaporization.

In addition, in the bubble are the radiolytic gas molecules formed as a result of ionization or excitation of water molecules. The composition of the radiolytic gas is mainly determined by hydrogen molecules H ₂ [Pikaev AK Modern radiation chemistry, t.1-3, Moscow, Science, 1985, 1986, 1987]. Along with hydrogen, H ₂ O ₂ and oxygen will be formed.

The rate of hydrogen production, mol / (l · s), can be expressed by the equation [Byakov V.M., Nichiporov F.G. Radiolysis of water in nuclear reactors. M., Energoatomizdat, 1990]:

where G _H2 is the yield of hydrogen; Q - reactor power, kW / l. The indices f, p, and e refer, respectively, to fission fragments, protons arising from neutron scattering, and electrons formed upon absorption of γ radiation and decay of fission products. In a homogeneous reactor, at least about 96% of all the hydrogen generated in the solution is obtained from the energy of fission fragments, 2% from the energy of fast neutrons, and 2% from the energy of γ radiation. Therefore, the last two terms in equation (1) can be neglected.

The number of radiolytic gas molecules in one bubble reaches N _RG ~ 5 · 10 ⁵ . It follows that the specific yield of radiolytic gas is

,

=W/W₀, W - удельное тепловыделение в активной зоне растворного реактора, W₀=1 кВт/л,

Свойства воды практически не меняются с ростом давления, поэтому масса радиолитического газа, образующегося в единицу времени, постоянна, следовательно, объем радиолитического газа обратно пропорционален Р, что и отражено в (3).Where

,

= W / W ₀ , W - specific heat in the active zone of the solution reactor, W ₀ = 1 kW / l,

The properties of water practically do not change with increasing pressure, therefore, the mass of the radiolytic gas generated per unit time is constant, therefore, the volume of the radiolytic gas is inversely proportional to P, which is reflected in (3).

Water molecules hitting the wall of the bubble in the fuel solution are absorbed with a probability of ~ 1 [Y.I. Frenkel Kinetic theory of liquids. Leningrad, Nauka, 1975], and the vapor bubble itself disappears in a time of ~ 10 ^-7 s. From experiments it is known that in its place on average ~ 1.5 initial bubbles remain with a radius of R ₀ = 1.5 · 10 ^-5 cm, filled with radiolytic gas [Kolesov V.F. Aperiodic pulse reactors, Sarov, RFNC VNIIEF, 1999].

The radius of the initial bubble changes according to the law [Zeldovich Ya.B. // JETP 12 (1942), p. 525]:

The first term describes the diffusion of a radiolytic gas from a bubble. Here D ~ 10 ^-5 cm / s is the diffusion coefficient of the radiolytic gas in water, C is the concentration of the radiolytic gas, R _C is the radius of the critical bubble.

=0,035, что соответствует насыщенному раствору радиолитического газа в воде [Киреев В.А. Курс физической химии. Москва. Госхимиздат, 1955]. Радиус критического пузырька определяется выражениемWhere

= 0.035, which corresponds to a saturated solution of radiolytic gas in water [V. Kireev Course of physical chemistry. Moscow. Goskhimizdat, 1955]. The radius of the critical bubble is determined by the expression

- степень пересыщения раствора. Здесь и далее концентрация С определена как отношение объема радиолитического газа, получающегося при полном выделении его из жидкости при данном давлении Р, к объему жидкости. При таком определении согласно закону Генри величина

слабо зависит от давления.where α = 70 days / cm is the coefficient of surface tension of water, X = ΔС /

- the degree of supersaturation of the solution. Hereinafter, the concentration C is defined as the ratio of the volume of the radiolytic gas obtained when it is completely released from the liquid at a given pressure P, to the volume of the liquid. With this definition, according to Henry's law, the quantity

weakly dependent on pressure.

The second term in (4) describes the coagulation (fusion) of bubbles during their ascent under the action of a buoyancy force. Here a ~ 0.5; g = 980 cm / s ² ; ν = η / ρ = 0.01 cm ² / s is the kinematic viscosity of water, η is the usual viscosity of water, ρ is density. For X ~ 1 we have

R _C ~ 2 · 10 ^-4 /

Therefore, for R ~ R ₀ , when the last term in (4) (coagulation) can be neglected, from formula (4) we find that the initial bubble disappears in a time of 2 × 10 ^-5 s. This conclusion remains valid even taking into account fluctuations inherent in diffusion, causing deviations from the average values of R described by equation (4). According to [Frenkel Ya.I. Kinetic theory of liquids. Leningrad, Nauka, 1975] at R ₀ <R _{С the} probability of survival of the initial bubble is vanishingly small, and at R ₀ > R _С it is close to unity, increasing from zero to unity in a narrow range of R ₀ :

Due to (5), (6), the initial bubbles inevitably dissolve, and therefore the spontaneous formation of stable nuclei under normal operating conditions of a solution reactor, for example, an Argus reactor (P ~ 1 ata, X ~ 1, T <100 ° C), is excluded.

Under these conditions, radiolytic gas bubbles are formed by the mechanism of heterogeneous nucleation [Volmer M. Kinetics of the formation of a new phase. Moscow, Nauka, 1986], that is, on the smallest (R ~ R _C ) solid suspended particles that are inevitably present in the fuel solution, on which the barrier-free nucleation of the radiolytic gas occurs due to the adsorption of the radiolytic gas and subsequent film formation on the surfaces of the particles.

In a stationary reactor operating mode, the radiolytic gas balance equation has the form

whence from (1), (3) we obtain the relation for t _e - time of radiolytic gas exit from the liquid

Heterogeneous nucleation is a fast process, therefore, the supersaturation is small (X = 0.1 ÷ 0.5), therefore,

t _e ~ 10 Q, s

This shows, in particular, that at P = 1 atm, W = 1 kW / l, the retention time of the RG in the liquid is ~ 10 s, and at P = 3 atm and W = 0.3 kW / l it already reaches ~ 100 s . In the latter case, the value of t _{e is} optimal for the production of the ⁸⁹ Kr isotope with a low content of ⁹⁰ Kr.

The physical reason for the increase in t _e with increasing Q is as follows. With heterogeneous nucleation, there is no extremely slow stage of subcritical vesicle growth (R <R _C ), which occurs with homogeneous fluctuation nucleation. The subsequent stage of the growth of supercritical vesicles (R> R _C ) occurs in both cases almost identically, since it is completely determined by the volume diffusion of radiolytic gas to growing near-critical (R ~ R _C ) vesicles, which takes place over time

t _D ~ R _C ² / (D · ΔC)

According to formula (4), the bubble growth rate has a minimum at R = R ₁ , where

At R> R _{1, the} bubbles rapidly increase in size and emerge under the action of a buoyancy force, therefore, the duration of the second stage (ascent) is determined by the sizes R ~ R ₁ :

The radius of the pop-up bubbles at the surface of the liquid is

где L - вертикальный размер сосуда, L₀=25 см. Таким образом,

where L is the vertical size of the vessel, L ₀ = 25 cm. Thus,

With increasing parameter Q, the radiolytic gas exit time t _e increases mainly due to the increase in diffusion time t _D. In particular, with W = 0.5 kW / l and P = 1 atm, X = 0.45; t _D = 8 s; t ₀ = 9 s; t _e = 17 s, and already at P = 5 atm, X = 0.09; t _D = 47 s; t ₀ = 18 s; t _e = 65 s. This shows that with increasing Q, the supersaturation of X decreases. According to (4), the bubbles become larger on average, and their concentration decreases. As a result, the ratio of the total surface area to volume decreases, and the diffusion time increases.

As follows from the above estimates, the rate of formation of a radiolytic gas depends only on the power of the solution reactor, and the time for gas bubbles to escape from the fuel solution is a function of pressure, linearly increasing with its growth, which is confirmed by experimental data, for example [Kolosov V.F. Aperiodic pulse reactors, pp. 626, 627]. At very high pressures in the reactor vessel, the "radiolytic boiling" of the solution and, accordingly, the release of radiolytic gases from the fuel are generally suppressed. By varying the pressure in the reactor, it is possible to control the size of the vapor bubbles and their growth rate due to the diffusion of radiolytic gases. By increasing the pressure in the reactor, it is possible to increase the residence time of gaseous fission fragments in the fuel solution. Thus, restraining the “radiolytic boiling” of the solution by varying the pressure in the reactor vessel, the radioisotope composition of the gas in the free volume of the solution reactor can be controlled.

In addition, the rate of accumulation of radiolytic gases in solution reactors depends on the specific energy intensity of the fuel. At the same reactor power, but with different concentrations of uranium in the fuel, the rate of accumulation of radiolytic gases in the reactor can vary several times [Byakov VM, Nichiporov F.G. Radiolysis of water in nuclear reactors, M., Energoatomizdat, 1990, p. 147]. By changing the concentration of uranium in the solution, it is possible to select the optimal specific energy intensity of the fuel, which will correspond to the specified time for the exit of gaseous fragments from the fuel solution, ensuring the purification of strontium-89 from impurity radioisotopes due to their radioactive decay. In particular, by increasing the residence time of gaseous fission fragments in mortar fuel to ≈ 400 seconds, it is possible to minimize the admixture of the krypton-90 radioisotope in the free volume of the reactor, which as a result of decay gives the most regulated radioisotope of strontium-90.

At a pressure in the reactor vessel P = 3 at and a specific energy intensity of 0.3 kW / liter, the residence time of radiolytic gases in the fuel τ will exceed 100 seconds, and at a pressure of ≈10 at τ it will reach 300-400 seconds, which will provide an acceptable level of impurity of the strontium radioisotope 90 in the target radioisotope.

The implementation of the invention

The drawing shows a diagram of an implementation of a method for producing a radioisotope of strontium-89 in a solution reactor under pressure.

The active zone of the reactor 1 is an all-metal welded shell of cylindrical shape with a hemispherical bottom. Case diameter 500 mm, height 1000 mm, wall thickness 25 mm. The casing is connected to a gas circuit containing a circulation pump 2 and a volume 3 for holding gaseous fragments removed from the reactor, as well as valves with remote control 4. Inside the casing there is a cooling coil 5 and a route 6 for supplying fuel solution with valve 7. Elements of the core, in continuous contact with the fuel solution, and the gas circuit is made of stainless steel OX18H10T. The reactor vessel is surrounded by a lateral and lower end graphite reflector.

The reactor fuel is an aqueous solution of uranyl sulfate UO ₂ SO ₄ enriched in the isotope ²³⁵ U to 20%, with a uranium concentration of 73.2 g / l. The volume of the fuel solution is 100 l. The maximum power of the reactor is 20 kW.

An aqueous solution of uranyl sulfate with a pre-selected concentration of uranium is poured along the fuel solution supply path 6 into the core of nuclear reactor 1. The reactor vessel is connected to a loop gas circuit and sealed. The reactor is brought to a predetermined power level, maintaining the conditions of the chain reaction of fission in the solution fuel. As a result of the fission of uranium-235 or the nuclei of other fissile materials in solution fuel, a gaseous radioisotope krypton-89 is produced, which, without entering into a chemical reaction with fuel or fragmentation elements, enters the free space above the surface of the liquid due to the "radiolytic boiling" of the solution.

Removing the gas fraction from the free volume of the reactor is carried out using a circulation pump. With a gas stream, the radioisotopes in a sealed circuit are sent to a holding chamber, in which the target strontium-89 radioisotope is formed and trapped as a result of the decay of krypton-89. Then the gas flow using the circulation pump 2 is returned to the reactor vessel.

The strontium-89 precipitated in the holding chamber is washed off with an acid solution and sent for radiochemical purification to obtain the target product of conditional quality.

For the regeneration of the products of hydrogen and oxygen radiolysis, a catalytic recombiner 8, a heat exchanger 9, and a pipe 10 are used.

To reduce the amount of impurities formed as a result of fission of nuclei and entering the holding volume with a gas stream, a filter system 11 is installed at the entrance to the volume 3.

The selected value of the working pressure in the reactor vessel 1 is supported by the inlet of an inert gas, such as argon, from high pressure cylinders 12 through the valve 13.

Example 1. A solution of a nuclear reactor with a gas circuit.

As a reactor installation, intended for use in the field of production of radioisotopes, a pulse research reactor "Hydra" with a closed gas circuit was selected. The reactor vessel is designed for pressures up to 100 atm.

Let us consider a more detailed scheme for producing the strontium-89 radioisotope in a pressurized solution nuclear reactor.

An aqueous solution of UO ₂ SO ₄ with a pre-selected concentration of uranium along the supply path of the fuel solution 6 through valve 7 is poured into the active zone of nuclear reactor 1.

After transients associated with the output of the reactor at a given power, as a result of fission of ²³⁵ U nuclei in the fuel, gaseous radioisotopes are produced that enter the free space above the solution surface due to the “radiolytic boiling” mechanism. The heat released during the operation of the reactor is removed using a cooling coil 5.

By varying the pressure P in the reactor vessel, the residence time of gaseous fission fragments in the fuel solution τ, reaching ≈ 400 seconds, which is sufficient for the decay of short-lived gaseous fragments, including krypton-90, is selected. Gaseous fission products released from the fuel solution into the free volume of the reactor are removed using a circulation pump 2 and through open valves 4 and the filter element 11 are sent along the gas circuit to the volume 3 for holding and obtaining the target radioisotope of strontium-89.

Precipitated in a volume of 3 strontium-89 is washed off with an acid solution and sent for radiochemical purification to obtain the target product of conditional quality.

The products of the radiolysis of the fuel solution — hydrogen and oxygen — are regenerated using a device including a catalytic recombiner 8, a heat exchanger 9, and a pipe 10, which together with the reactor vessel form a sealed system.

The proposed method for the production of a strontium-89 radioisotope using a research nuclear reactor with solution fuel under pressure is easier to implement and operate than others, maintenance of the reactor and the installation as a whole does not require numerous personnel, and it also ensures the reliability of fulfilling the requirements of radiation safety and the environment.

Claims (1)

A method for producing a strontium-89 radioisotope, comprising irradiating fuel in the form of an aqueous solution of uranium salts in a nuclear reactor core to produce fission fragments, one of which is krypton-89, which is the precursor of strontium-89 in the decay chain of fission fragments with atomic mass 89, and purification of krypton-89 due to the natural decay of the accompanying radioisotopes of krypton, including krypton-90, the precursor of the long-lived radioisotope strontium-90, characterized in that the purification of krypton-89 is carried out by increasing webbings finding fission fuel to collapse it short-lived radioisotopes, including krypton-90, due to the pressure increase in the reactor vessel, and reducing the specific power density of fuel, and a purified krypton-89 is sent to a passage to its complete decay of the target radioisotope.