Neutron-activation method of monitoring burning of spent fuel assemblies of thermal neutron reactors and apparatus therefor
FIELD: physics, atomic power.
SUBSTANCE: present invention relates to means of measuring burn-up of spent fuel assemblies of thermal neutron reactors. A diagnostic container is placed at the bottom of a cooling pond under water. The wall of the housing has an annular cavity with a liquid tracer substance surrounded by polypropylene layers and steel layers, as well as a cadmium layer. The housing has a centre cavity in which spent fuel assemblies are placed. The container is closed with a cover. Water is removed from the centre cavity with the spent fuel assemblies. The tracer substance is activated, drained into a laboratory vessel and stirred. A sample is then collected. The average specific activity of the tracer substance is measured and the neutron radiation intensity of the spent fuel assemblies and the associated burn-up of the spent fuel assemblies is determined. The annular cavity with the liquid tracer substance can consist of multiple rings insulated from each other. Burn-up is then determined for each ring and a burn-up profile for the spent fuel assemblies is then constructed.
EFFECT: enabling all-round encircling of the core region of the spent fuel assemblies with the tracer substance, excluding the effect of water on measurement accuracy, eliminating background effect when measuring specific activity of the tracer substance and high accuracy of determining burn-up of spent fuel assemblies of thermal neutron reactors without retrieving spent fuel assemblies from the cooling pond.
9 cl, 3 dwg
The technical field
The invention relates to non-destructive remote control methods for burnup of spent fuel assemblies (SFAS) reactors on thermal neutrons. In particular, the method and the device are used to measure the burnup (In) SFA of VVER-1000 reactors in the range≈(15÷70) MW·d/kg U in real conditions of their storage in pools filled with water. The burnup of fuel - the amount of released energy in one kilogram of uranium contained in the fuel, during the campaign of the reactor.
Data on the burnup of spent nuclear fuel (SNF) in the active areas of the SFA are of great interest for the analysis and optimization of processes and the implementation of IAEA safeguards on non-proliferation of nuclear materials, which places high demands on the accuracy of this value.
Existing to date methods of controlling burnup SNF can be divided into three main groups: radiochemical methods, gamma (γ) - spectrometric methods and neutron (n) methods.
Radiochemical methods are based on sampling of spent fuel from the spent fuel assemblies, the selection of them actinoid elements (U, Pu, Am, Cm) or indicator nuclides fission products (Cs, CE, Eu and others), mass spectrometry, α with chromaticism and γ-spectrometric determination of isotopes of these nuclides in the samples and the calculation according to this data, the combustion of fuel in each of the samples. The measurement results for samples conducted by these methods are characterized by a relatively high degree of precision.
The disadvantage of radiochemical methods is the long duration of the analysis, the complexity and the need for a bulky and expensive equipment, which is connected with conducting numerous manual operations in contaminated conditions. As all sorts of destructive methods, radiochemical methods are unsuitable for the analysis of combustion of fuel in the entire volume of the SFA.
Gamma (γ) - spectrometric methods based on the measurement line of γ-spectra of one or more radioactive isotopes fission products, the content of which in SFAS correlates with the burnup of nuclear fuel. Gamma-spectrometric methods belong to the category of non-destructive methods to characterize the burnup of spent fuel in the entire volume of the fuel assemblies having, as a rule, very large size parameters and complex constructional structure.
In real conditions of remote spectrometry γ-radiation solid large SFAS (in particular, in the VVER-1000 reactors) achieving high accuracy is problematic primarily because trudnootdelyaemogo effect of self-absorption analyzed γ-quanta. An important factor strongly influencing the accuracy of this method of burnout SFA is also a large γ-background. The dose of γ-radiation near SFA can reach values of ~104R/h. This background caused by other, except analyzed, fission products, as well as multiply scattered γ-rays. It increases the "download" recording equipment, leads to the appearance of false signals in the analyzed energy registration window due to the superposition of signals from γ-quanta with energies other than energy diagnosed γ-lines. As a consequence, the necessity of collimating, scanning, filtering and quotation device, which itself can be a source of additional errors.
Neutron (n) methods based on the strong correlation n-radiation SFAS thermal reactors and burnup, it is obliged to spontaneous fission accumulating in them isotope of curium-244 (244Cm) with a half-life of TCm=18,1 years. When In≥15 MW d/kg U, the exposure time of the spent fuel assemblies removed from the reactor core tin≈(3÷15) years and a fixed initial enrichment of uranium in the fuel. This relationship between the intensity of n-istoriko in SFA (Snand burnups () SFA with good accuracy describes a power-law function:
Sn~Inα·exp(-λCmtin), with α~4÷5 (1) where:
Sn(n/s) is the intensity of the n-sources in SFA,
In (MW d/kg U) - depth of the burn-SFA,
λCm=0.69 (TCm)-1- constant radioactive decay of244Cm,
tin≈(3÷15) years - time SFA after removal from the reactor core, (Vfrolov. Nuclear-physical methods of control of fissile materials. M, Energoatomizdat, 1989).
the α - parameter correlation, which is determined individually for each reactor unit, designed for high-level programs, such as PRISM-RISK (VNIITF) (registered in the branch Fund of algorithms and programs ofap, act No. 687 from 19.11.2009).
It follows that in n-methods achieve high accuracy determination of burnup SNF occurs at the expense of increasing the accuracy of measuring the intensity of the n-sources - Sn.
Dignity n methods are:
- high sensitivity of the method, which provides a high value of the exponent α~4÷5 in the formula (1);
- high representative of control of the average fuel burn-up cross-sectional SFA, which provides a high penetrating power of neutrons fission, i.e. negligible their absorption in the SFA;
- small influence of the history of exposure of the fuel assemblies in the reactor on the measurement results, due to the rather long half-life isotope244Cm(TCm≈18,1 years) - the main source of recorded neutrons is s - and negligible his burnout during the campaign reactor (~3 years);
weak sensitivity of the error to determine the burnout (σinto error neutron measurements (σn): σin~α-1σnwhen α~4÷5, or to the accuracy of the input parameters tinin the formula (1).
The analogue was chosen as the principle of neutron activation of an aqueous solution of manganese salt by analogy with the method of "manganese tank", used for precision calibration of neutron sources (Kbecker, Kwartz. Neutron physics. M, Atomizdat, 1968).
In the measurement by the method of "manganese tank" controlled n-source is placed in a tank with an aqueous solution of manganese sulfate (MnSO4) at time t0during which will be activated nuclei55Mn in the radiative capture slowed down neutrons in water with the formation of a radioactive isotope56Mn, with easy to register a half-life of TMn=2,58 h and the energy created in this γ-quanta: EγMeV=0,847 (99); 1,811 (29); 2,110 (15), where is indicated in parentheses quantum yield in percent for decay. After activation, the solution is thoroughly mixed, take a part (sample) and subjected to γ-spectrometric analysis to determine the specific activity (A) isotope56Mn. Specific activity (A) whether ANO associated with the measured intensity of the (S n) n-radiation controlled source. The accuracy of the measurement of this quantity method "manganese tank" is~(1÷2)%.
The disadvantage of this method are the weight and size characteristics of the diagnosed SFA, its complex structure and a high content of fissile materials. Therefore, direct use of this technique for the measurement of the intensity of the n-dimension SFA in terms of the cooling pool is impossible.
As a prototype for the device was chosen measurement setup that implements the neutron method of controlling burnup of the fuel assemblies. It represents the load-bearing structure in the form of a plate with a Central aperture providing a passage through it SFAS (Chandrasena and other College news. Nuclear energy, No. 2, p.60-70, 2004). On a plate with a Central hole install two or more neutron detectors, usually in the form of fission chambers. Such measuring systems are equipped with the appropriate actuators, which allow axial scanning neutron flux controlled SFA by continuous or discrete pullback through the detecting device directly in the storage pool. Then the results of neutron scanning SFA process, average height SFA and determine the average fuel burnup in a controlled SFA.
in≈(11-14)% (2σ), which corresponds to the error of the n-dimension σn≈30%.
As a prototype of the proposed method was chosen as the method of "Fork", developed at LANL (Fork system) to control the burnout of the fuel from power reactors (R.I.Ewing. Bumnup verification measurements at USA nuclear utilities using the Fork system. ICNC-95. Albuquerque. USA. v.II, p.64-68, 1995). This technique measured n-radiation SFAS cameras division, posted on the levers supporting structure made in the form of a fork. Scan the active zone SFA, holding it between the arms of the fork directly in the cooling pool.
The main disadvantage of the existing metering installations is a strong influence from the external environment, the testimony of the neutron detectors with a typical measurement conditions in the cooling pool, filled, as a rule, Borisovna water. This leads, firstly, to change the multiplication factor of neutrons in the system, consisting of pools of exposure, metering skids, SFA, and, dependent on the burnout of the fuel. Secondly, slowing and absorbing neutrons on their way to the detectors, which complicates the interpretation of the obtained measurement results.
The drawbacks include also significant is th effect of a strong (~10 4R/h) and variable along the axis of SFAS γ-background on the work of the neutron detectors and some of the difficulties associated with achieving the required precision alignment SFA relative to the detector when scanning their neutron flux.
These factors lead to the need for the introduction of numerous and often difficult to estimate corrections in the processing of measurement results, which is reflected in the accuracy of control burnout investigated SFA.
Disclosure of the invention.
The objective of the invention is to improve the accuracy of the neutron activation method and instrument control burnup SFAS energy rectors on thermal neutrons without removing the fuel assemblies from the pool water extracts.
The technical result consists in simultaneous, comprehensive environment of the active zone SFAS indicator substance, the elimination of interference from water between the active area SFA and the indicator substance when it is activated, the elimination of background influences when measuring the specific activity of the indicator substance in the laboratory.
This technical result is achieved by the neutron activation method control the burnout of the fuel of thermal reactors, which consists in moving the fuel assemblies under water to the testing site until the convergence of the active zone SFA with neutron-sensitive element, is the distribution of the intensity (S n) neutron radiation SFA, the calculation of the burn-known relationship between burnout and the intensity B(Sn), according to the invention on the bottom of the pool holding under water establish diagnostic container with the annular cavity in the housing wall, filled with liquid indicator substance, and a Central cavity in which is placed and fixed SFA. Close the container, then remove the water from the Central cavity with SFA and hold the activation of the indicator substance within a fixed time. Merge indicator substance, mix, take out the test, which measure the average specific activity of the indicator substance. Taking into account the calibration of the measurement setup to determine the intensity of neutron radiation (Sn) and related burnups (B) of the SFA.
The location of the indicator substance in the annular cavity in the wall of the container simultaneously and comprehensively to surround the active area SFA on her side. Removing water from DK, not raising SFA on the surface of the water, allows to eliminate the influence of water on the multiplication factor SFA and thus on the measured intensity of neutron sources in SFA in the way of controlling burnup.
The drain of the indicator substance and further his research out of the cooling pool poses which enables you to eliminate strong γ-background diagnosed from SFA. All this allows to increase the accuracy of the method of controlling burnup of the fuel assemblies.
It is possible to measure the specific activity of the indicator substance gamma-spectrometer. This further improves the accuracy of the method of controlling burnup.
You can fill indicator substance annular cavity in the wall of the container, consisting of separate, isolated from each other rings, the inner wall side surface of the cylindrical diagnostic container. After exposure of the indicator substance of each ring is drained separately, determine burnups for each ring, make up the profile of burnup for the active zone SFA. Profile burnout at the same time more accurate than in the prototype, because of the comprehensive coverage of the active zone SFAS indicator layer.
Object of the invention is a device that determines with high accuracy burnups SFAS energy rectors on thermal neutrons without removing the fuel assemblies from the pool water extracts.
The technical result consists in simultaneous, comprehensive grasp of the indicator substance active zone SFA, getting rid of the influence of water on the measurement accuracy, eliminate background effects in the measurement of the specific activity of the indicator substance.
The technical result is achieved that the device is as to control the burnout SFA, containing sensitive to neutron radiation element, surrounded by layers of polyethylene and cadmium, according to the invention the device is designed as a diagnostic of the container, consisting of the cover and the housing with a Central cavity for accommodating the SFA and the annular cavity in the side wall of the housing, which is surrounded by a layer of cadmium and a hydrogen-containing substance. The container is connected to the detecting unit for measuring the specific activity of the indicator substance, while sensitive to neutron radiation element is designed in the form of a liquid indicator substances placed in the annular cavity in the side wall of the housing of the diagnostic container.
The annular cavity is structurally located around the active zone of the fuel assemblies simultaneously and comprehensively covers it. The annular cavity is surrounded by layers of cadmium and hydrogen-containing substances. While sensitive to neutron radiation element is present in the form of a liquid indicator substance filling the annular cavity. The sensing element is a liquid takes the form of an annular cavity into which it is poured, and thus also covers the active area of the SFA.
Indicator liquid substance may be in the form of an aqueous solution of manganese sulfate.
Measuring the intensity of halogen with exceptiona the neutron radiation may be in the form of gamma-spectrometer. This further increases the accuracy of the neutron activation method.
A hydrogen-containing substance can be made from polypropylene.
The side wall of the housing of the diagnostic container may be made of a multilayer consisting of the outer layer of steel, a layer of proprotein, layer steel cavity for the indicator substance, a layer of steel, a layer of polypropylene and an inner layer of steel, a layer of cadmium. Layers give the container strength. Layers of polypropylene slow down fast neutrons. The layer of cadmium skips the fast neutron flux from the SFA to the indicator substance, but reduces thermal neutron flux from the outside at SFA. This reduces the background neutron source in SFA because of the marks on the uranium. In consequence, the indicator solution is activated in strict accordance with the burnup of the fuel assemblies. All of this increases the accuracy of determining the intensity neutron sources in the SFA.
The Central cavity of the body diagnostic container may be provided with a pipe for draining water from the container. The absence of water between the SFA and the indicator substance removes the noise, and thus increases the accuracy of determining the intensity neutron sources in the SFA.
The annular cavity of the container may be provided with a pipe. This solution allows you to put in the annular cavity liquid indicator in the society and remove it for further research. Measurement of the activity of the indicator substance is produced in laboratory conditions, where there is no background effects from the cooling pool and other SFAS. This further improves the accuracy of the device.
Embodiments of the invention.
As shown in figure 1, the diagnostic container consists of a body 1 on which the top cover 2 installed. Inside the housing 1 is a Central cavity 3. The housing 1 is equipped with a pipe 4 for drainage and water supply 5 from the cooling pool. The middle part of the housing 1 is made of a multilayer. Inside the housing 1 a diagnosed SFAS 6, the middle part of which is an active area 7. The housing 1 is at the bottom 8 of the cooling pool.
As shown in figure 2, the multilayer middle part of the housing 1 consists of the following layers. The outer layer of steel 9, polypropylene layer 10, a layer of steel 11, the annular cavity is filled with liquid indicator substance 12, a layer of steel 13, the layer polipropilene 14, a layer of steel 15, a layer of cadmium (Cd) 16.
Layers of stainless steel 11 and 13 form the shell of the annular cavity with indicator substance 12. Layers 9 and 15 form the outer and inner shells diagnostic container.
As the signal substance 12 in a particular embodiment, uses an aqueous solution of manganese sulfate (MnSO4). The indicator substance 12 is placed between the two layers of hydrogenous material, in this case polypropylene 10 and 14. Polypropylene 10 and 14 slows down the neutrons emitted by the active zone 7 SFA 6, and thereby increases the sensitivity of the method. Sensitivity is the number of reactions activated manganese in a particular installation of one neutron emitted from SFA. Polypropylene 10 and 14 slows down the neutrons, thus increasing the probability that the neutron will cause the reaction activation in the indicator substance 12.
The layer of neutron absorber - cadmium 16, is placed on the inner wall of the housing 1 and serves to reduce neutron connection between the active area 7 SFA 6 and its external environment. This weakens the impact of the environment on the measurement results and contributes to the improvement of their accuracy.
For information about the axial distribution of the neutron sources in the active zone 7 SFA 6 and, therefore, the burnup of fuel in it, the annular cavity with indicator substance 12 is divided into several annular sections 17, as shown in figure 3. In this case, averaging the activation solution of the indicator substance 12 is held by the volume of each individual section 17.
The method is as follows. On the bottom 8 of the cooling pool is set housing 1. SFA 6 raise above the bottom 8, without removal from pool water extracts, are loaded into the Central cavity 3 of the housing 1. On top of the building is 1 set the cover 2. From the Central cavity 3 through the pipe 4 remove the water. In the annular cavity of the housing 1 fill indicator substance 12. For time t0=10÷15 min indicator substance 12 is activated.
Fast neutrons emitted by the active zone 7 SFA 6, pass through the layer of cadmium 16, slowed by layers of propylene 10 and 14, fall in the indicator substance 12 and activate it. This increases the sensitivity of the device, and therefore, the accuracy of the measurement of burnup.
The layer of cadmium 16 passes fast neutrons from the active zone 7 in the direction of the indicator substance 12, but does not pass slower polypropylene 10 and 14 neutrons in the opposite direction. This prevents the emergence of fast neutrons from the fission of fissile isotopes that are not associated with a true burnout SFA 6 in the active zone 7 and do not characterize burnout SFA 6. It also increases the accuracy of the burnup of the fuel assemblies 6 and activation of the indicator substance 12.
After the time of activation of the indicator substance 12 is pumped out through a pipe 18 from the annular cavity 12 of the housing 1 in a container located in a radiation safe place. Activated indicator substance 12 is poured into the cylinders-store and mix thoroughly. Take samples of the indicator substance 12. Spend γ-spectrometry identified the E. their specific activity unit for measuring specific activity of 19. Activated indicator substance 12 is maintained in the cylinders storage before reuse within the time required to restore the original properties of the indicator substance 12.
The functional relationship between the measured specific activity of the sample (AMn), and the intensity of the n-radiation controlled SFA (Sn) is determined by the ratio:
AMn, (Bq/l) is the measured specific activity of the sample,
λMn=0,69 (TMn)-1- constant radioactive decay of56Mn,
qnγ(1/n·l) is the number of reactions radiative capture of neutrons by nuclei55Mn per liter of solution of the indicator substance, normalized to one neutron emitted in the spontaneous fission244Cm in SFA (sensitivity of the measurement setup),
Sn(n/s) is the intensity of the n-radiation controlled SFA,
t0(min) - time activation solution of the indicator substance in the diagnostic product is ore.
Sensitive measurement setup includes diagnostic container and a unit for measuring specific activity of 19 (shown in Fig.1).
The value of qnγcan be calculated from the neutron-physical high-level programs or determined by calibration of the measurement setup of the proposed method using the reference neutron sources. Conducted computational studies with the model of the measuring system for monitoring SFA of VVER-1000 showed that the selected layer diagnostic container and accepted the conditions of application of the proposed method, this value is almost invariant (variation in its values did not exceed ~2%) in relation to changes in a wide range of burn-up of spent fuel assemblies and the associated changes of the isotopic composition of spent fuel, the spatial distribution of n-sources on the height of the active zone, including conditions measurement point reference source during the calibration of the measurement setup, the temperature of the water in the cooling pool and concentration in it of boric acid. Its value for the studied model of the measurement setup at a concentration of manganese sulfate in an aqueous solution of the indicator substance=400 g/l was equal to qnγ≈(8,65±0,19)·10-41/n·L.
From which it follows that the control collection is to burn fuel≈(15÷50) MW·d/kg·U, which corresponds to Sn≈(2·106÷4·108) n/s TBC, and t0=10 min for the expected values of the specific activity of the indicator substance will be As≈(102÷2·104) Bq/L. This ensures its existing measurement equipment with an accuracy of ~2%. As a result, according to the formula (2), for a total error of determination of the intensity of the n-radiation SFAS proposed method will get σn~(3÷4)%.
Thus, the proposed neutron activation control method of burnout SFAS significantly reduces the accuracy of neutron measurements in comparison with other methods.
This error (σn) makes the proposed method is promising for solving one of the important tasks of providing IAEA safeguards on nuclear material - tasks separate remote determination of plutonium isotopes (239-242Pu) in the SFA.
In the available sources of information not found technical solutions containing collectively signs, similar to the distinctive features of the proposed neutron activation method. Therefore, the invention conforms to the criterion "novelty".
To have our sources of information there is no information about the effects available in the claimed invention of the distinctive features together on the achievement of the stated technical result. On the basis of this invention meets the criterion of "inventive step".
Implementation of the proposed method is possible, because it is based on the use of technology, well-developed modern industry. The materials used in the proposed method and the device, currently known and used in the nuclear industry. The method and apparatus can be used to measure the burnup of the fuel assemblies of VVER-1000 reactors in the range≈(15÷70) MW·d/kg U in real conditions of their storage in the pools, without removing the fuel assemblies from the water. This confirms the industrial applicability of the proposed method and device.
1. Neutron activation control method of burning the fuel of thermal reactors, which consists in moving the fuel assemblies under water to the testing site until the convergence of the active zone SFA with neutron-sensitive element, the definition of intensity (Sn) neutron radiation SFA, the calculation of the burn-known relationship between burnout and the intensity B(Sn), characterized in that on the bottom of the pool holding under water establish diagnostic container with the annular cavity in the housing wall, filled with liquid indicator substance, and a Central cavity in which is placed and fixed RESP is, close the container, then remove the water from the Central cavity with SFA and hold the activation of the indicator substance for a fixed time, merge indicator substance, mix, take out the test, which measure the average specific activity of the indicator substance, taking into account the calibration of the measurement setup to determine the intensity of neutron radiation SFA (Snand related burnup of the fuel assemblies (In).
2. Neutron activation control method of burnout SFAS according to claim 1, characterized in that the fill indicator substance annular cavity in the housing wall of the container, consisting of separate, isolated from each other rings.
3. Neutron activation control method of burnout SFAS according to claim 2, characterized in that after irradiation the solution of the indicator substance of each ring is drained separately, determine the burnup of the fuel assemblies for each ring, make a profile for burnout active zone of the SFA.
4. Device for controlling the burnout of the fuel assemblies containing sensitive to neutron radiation element, surrounded by layers of polyethylene and cadmium, characterized in that the device is designed as a diagnostic of the container, consisting of the cover and the housing with a Central cavity for accommodating the SFA and the annular cavity in the side wall is ke corps, which is surrounded by a layer of cadmium and a hydrogen-containing substance, the container is connected to the detecting unit for measuring the specific activity of the indicator substance, while sensitive to neutron radiation element is designed in the form of a liquid indicator substances placed in the annular cavity in the side wall of the housing of the diagnostic container.
5. Device for controlling the burnout of the fuel according to claim 4, characterized in that the liquid indicator substance made in the form of an aqueous solution of manganese sulfate.
6. Device for controlling the burnout of the fuel according to claim 4, characterized in that a hydrogen-containing substance is made of polypropylene.
7. Device for controlling the burnout of the fuel according to claim 4, characterized in that the side wall of the housing of the diagnostic container is made of a multilayer consisting of the outer layer of steel, a layer of polypropylene, a layer of steel, the cavity for the indicator substance, a layer of steel, a layer of polypropylene and an inner layer of steel, a layer of cadmium.
8. Device for controlling burnout SFA claim 4, characterized in that the Central cavity of the body diagnostic container equipped with a tube for removing water from the container.
9. Device for controlling burnout SFA claim 4, characterized in that the annular cavity of the body diagnostic container equipped with a pipe for which livki and removal of the indicator substance.
SUBSTANCE: oxides of transuranium elements are mixed with palladium metal powder in the following ratio, wt %: oxides of transuranium elements - 30-70, palladium metal - 70-30, and the obtained mixture is pressed. As a result, a composition for long-term storage of transuranium elements is obtained, which includes oxides of transuranium elements in palladium metal, which provides high chemical stability of the material, safety during indefinite storage, while preserving the capacity to extract transplutonium elements after dissolving the disclosed composition in nitric acid. The invention proposes the use of industrial ("reactor") palladium, which is a nuclear fuel fission product, to produce the disclosed composition.
EFFECT: prolonged and reliable isolation of transuranium elements and preserving the capacity for extraction and use thereof in future, or for further processing using a transmutation process.
2 cl, 2 tbl
SUBSTANCE: invention relates to materials with neutron-absorbing properties for protection against neutron radiation. Claimed is a fire-resistant neutron-protective material, consisting of a magnesium phosphate binding agent (24-33 wt %) and a powder part (76-67 wt %, with the powder part containing titanium hydride TiH2 (90.3-95.5 wt %), magnesium oxide MgO (2.7-4.5 wt %) and boron carbide B4C (1.8-5.2 wt %). Components are mixed to a homogeneous state and poured into a special cavity, and after hardening are subjected to thermal processing.
EFFECT: obtained material possesses long-term mechanical strength, heat resistance to ≈300°C, high heat conductivity, a temperature coefficient of linear expansion, close to the coefficient of construction steels, and high specific density of hydrogen and boron, contained in it, which ensures high coefficients of neutron radiation attenuation.
SUBSTANCE: invention relates to radiochemical technology and can be used in production of "reactor" 99Mo as a generator of 99mTc of a biomedical purpose, as well as in an analysis of technological solutions for preliminary separation of Mo or Mo and Zr in extraction reprocessing of solutions of technology of spent nuclear fuel of nuclear power plants (NPP SNF). Described are versions of methods of selective extractive separation of a considerable part of molybdenum or together molybdenum and zirconium from radioactive solutions with obtaining an extract. A reprocessed radioactive solution is processed with an extractant, which represents poorly soluble in a water phase alcohol, in the presence of an extracted complexing agent. As the complexing agent, hydroxamic acids with a number of carbon atoms 6-12 can be used, which ensures sufficiently complete extraction of molybdenum and zirconium in an organic phase. Molybdenum or molybdenum and zirconium are separated from the extract in the compact form by sublimation or re-extraction.
EFFECT: obtaining the extract, purified from alpha- and gamma-radioactive admixtures more than by 100 times, and further separate extraction of radionuclides from the extract, combined in the final stage of the process with the extractant regeneration.
17 cl, 2 tbl, 12 ex
SUBSTANCE: invention relates to method of determining optimal parameters of dissolution of oxides of transition metals in solutions, which contain complexing agent, and can be applied in atomic energy. As parameters, applied are volume coefficients of distribution of radioactive isotopes of transition metals, which determine composition of oxides, between solutions, which contain complexing agent, and strong-base anionites in form of said complexing agent (complexits) and balanced values of solution pH. Radioactive isotopes of transition metals are introduced into fixed volumes of analysed solutions, after which, fixed volumes of complexits are introduced into solutions. Initial and final activity of solutions is measured. Also measured are balanced values of pH of solutions. Ranges, satisfying optimal parameters, which must be supported in contours of NEP directly in the process of dissolution of oxides in the process of washing and deactivation of NEP contours, are determined by the results of measurements.
EFFECT: increased reliability of determination of optimal parameters of dissolution of oxides of transition metals in complexing agent solutions and absence of necessity of carrying out complex analyses of metals to determine concentration of metal cations in solutions of complexons.
2 cl, 2 dwg, 1 tbl
FIELD: power engineering.
SUBSTANCE: machine for cutting of pipelines, preferably caissons from a storage tank, comprises a body and a rotor covering the cut pipe, with a metal cutting device fixed on it. The machine body is made in the form of a circular ball step-bearing, the fixed part of which is attached to the rotary board of the storage tank together with the drive for rotation of the movable part, coaxially and elastically fixed to the rotor flange. The rotor is made in the form of a cylindrical thin-walled hollow column that covers a caisson, with a rotary stem fixed on it, at the end of which a small-size driving cutting machine is attached with an abrasive disc.
EFFECT: creation of a reliable and comparatively cheap machine for cutting and trimming of pipelines, in particular, caissons in unavailable zones of radiation-hazardous facilities at the depth of at least 3 metres from the outer surface of the facility to provide for nuclear and radiation safety.
4 cl, 7 dwg
FIELD: power engineering.
SUBSTANCE: method includes construction of inspection wells in a reservoir bed, their equipment with facilities of water lift, facilities to measure level and pressure of subsurface water in them, to pump subsurface water from them, performance of physical and chemical tests of pumped subsurface waters. During pumping a volume of subsurface water is extracted from the well, and this volume is less than the one contained in its borehole, density of the extracted subsurface water is measured, afterwards it is supplied back into the well in the interval, from which it was pumped, measurement of subsurface water pressure is carried out along piezometric tubes lowered into a well filter and filled by water with available density, afterwards water density is determined in the reservoir bed.
EFFECT: improved validity of produced data and elimination of hydrodynamic mode disturbances in wells in process of monitoring.
1 cl, 1 dwg
FIELD: process engineering.
SUBSTANCE: invention relates to separation of fluids by evaporation. Proposed method comprises evaporation of solvent from solution film formed on rotary drum inner surface. Solution is fed inside drum to allow displacement of solution toward unloading side. Solvent vapors are collected and condensed. Heat power consumed for solvent evaporation, drum rpm and level of solution inside the drum are selected provided concentrated product is discharged in the form of fluid. Flow rate of solution fed into drum is to be calculated by mathematical tools.
EFFECT: sufficient quality of concentrated product and solvent condensed vapors provided in continuous operation without drop in efficiency in wide range of concentrations of solutions.
11 cl, 1 dwg
SUBSTANCE: invention relates to the technology of recycling nuclear energy materials and specifically to methods of cleaning uranium hexafluoride from ruthenium fluorides, and can be used in returning uranium, extracted from spent nuclear fuel, into the fuel cycle of light water reactors. The method of cleaning uranium hexafluoride from ruthenium fluorides involves reaction of gaseous uranium hexafluoride with a sorbent, the sorbent used being porous granules of sintered metal powder of nickel or nickel which contains up to 10 wt % copper.
EFFECT: invention provides efficient cleaning of uranium hexafluoride from ruthenium fluorides at low temperatures, safety and simplification of the process, and also enables to reuse the sorbent after regeneration.
5 cl, 2 tbl, 5 ex
SUBSTANCE: method of underground disposal of biohazardous waste water in geological formations with no distinct waterproof stratums above production level includes drilling of injection well and pumping of biohazardous waste water in production level. Injection well is drilled inclined directed, in the interval of production level bore of injection well is drilled parallel to bedding at a distance from bottom of production level equal to 0.1-0.2 of production level thickness, and observation well is drilled on water-bearing level located below level with groundwater used for drinking and technical needs.
EFFECT: preventing penetration of disposed biohazardous waste water in above located water-bearing levels.
4 cl, 1 dwg, 1 ex
FIELD: process engineering.
SUBSTANCE: proposed method comprises treatment of radioactive wastes solutions by alkaline metal hydroxides or carbonates to pH 1, extraction of tri-n butyl phosphate in inert diluter to convert rare-earth and transplutonium elements into extracts and their separation from cesium-strontium fraction, flushing of said extract by solution of aluminium nitrate and adding flushing solution to feed water. Hydrogen peroxide is added to radioactive wastes solution at neutralisation stage.
EFFECT: higher reliability of treatment.
FIELD: nuclear engineering.
SUBSTANCE: the proposed method of removal of irradiated material from nuclear reactor plate includes the irradiated material shielding with a shield material and their following removal of both materials. The shielding is fulfilled with a granulated material. Removal of irradiated and granulated materials is carried out by means of an auxiliary tube with a grate, confining the irradiated material and permeable for granulated material. The auxiliary tube is installed into the tube of a reactor fuel channel.
EFFECT: decreasing the idle time of reactor or/and production equipment.
1 cl, 1 dwg
FIELD: soil decontamination from radionuclides by technological methods.
SUBSTANCE: proposed method includes soil plowing and introduction of chemical agents into it. Ground limestone or dolomite in the amount of 5 or 6 tons per ha and potassium fertilizers KCl, KNO3, or KMgCl, *6H2O in the amount of 200 kg per ha are introduced in plowed layer followed by sowing perennial grass which is mowed in autumn and placed in storage excluding migration of radionuclides.
EFFECT: facilitated procedure, enhanced degree of soil decontamination from radionuclides.
3 cl, 1 dwg, 1 ex
FIELD: agriculture, in particular, environment protection, more particular, reduction of 137Cs level in soil.
SUBSTANCE: method involves growing accumulating plants such as lentils and Jerusalem artichoke on contaminated soil during three vegetation periods; alienating the entire plant biomass from soil at the end of vegetation period; determining soil cleaning extent from formula: Cη=(Ca-Cs/Ca)*100(%), where Cη is extent of cleaning soil; Ca is level of 137Cs in soil before planting of said accumulating plants; Cs is level of 137Cs in soil after withdrawal of the entire plant biomass from soil at the end of vegetation period.
EFFECT: reduced specific activity of 137Cs in soil, increased efficiency in removal of radio nuclides and obtaining of ecologically clean plant products, reduced possibility of external and internal radiation of people.
FIELD: chemical industry; building industry; methods and the cleaning gels for the surfaces treatment.
SUBSTANCE: the invention is pertaining to the method of the surface treatment with the cleaning gel and may be used for the surface degreasing and decontamination, and also for removal of the oxide layers from the surface. The method provides for: deposition of the cleaning gel onto the treated surface; aging of the cleaning gel on the treated surface at such a temperature and relative humidity, at which the gel dries and at that there is enough time for the gel to treat the surface before the dry and solid residue forms; removal of the dry and solid residue from the treated surface by suction or by the brush. The gel includes the mixture of the pyrogenetic silica and the sedimentary silica, the cleaning agent and possibly the oxidative agent. The invention ensures regulation of the dimensions of the dry residues of the gel and the time of drying sufficient for the effective treatment of any type surface, and also allows to reduce the quantity of the drainages formed during such treatment of the surfaces.
EFFECT: the invention ensures control over the dimensions of the dry residues of the gel and the time of drying sufficient for the effective treatment of any type surface, and also allows to reduce the quantity of the drainages formed during such treatment of the surfaces.
22 cl, 8 dwg, 5 ex
FIELD: nuclear engineering; preservation of dry, wet, and liquid radioactive wastes.
SUBSTANCE: proposed composition has resin portion of cold-cured compound ATOMIK and filler. Resin portion ingredients are as follows, parts by weight: epoxy oligomer, 100;, curing agent (aromatic amines), 38-50 furfural, 9-11. Used as filler is composition incorporating following ingredients, parts by weight: cement, 50-100; marshalite, 50-100 or bentonite, 50-100, or when they are jointly used: marshalite, 90-100 and bentonite, 90-100. Such composition provides for desired radiation resistance in absence of leaching of alpha-, beta-, and gamma-active radionuclides from preserved specimens of reactor graphite.
EFFECT: enhanced radiation stability of preserved specimens free from pits and voids, and adequate lifetime of preserved wastes; ability of their depreservation.
1 cl, 2 dwg, 2 tbl, 2 ex
FIELD: modeling plutonium dispersion processes in emergency explosion situations at its using entity.
SUBSTANCE: proposed method that can be used to simulate plutonium properties in case of emergency explosion involving escape of plutonium aerosols into atmosphere and to predict degree of radioactive pollution of terrain in emergency situations includes use of metal cerium as plutonium simulator in modeling processes of plutonium dispersion and escape of its aerosols into atmosphere in case of emergency explosion at using entity.
EFFECT: enhanced environmental friendliness of method.
FIELD: immobilization of radioactive wastes.
SUBSTANCE: proposed silicate matrix for conditioning radioactive wastes has SiO2, Na2O, K2O, CaO, Fe2O3, Cr2O3, NiO, Al2O3, ZrO2, oxides of radioactive waste components including nuclear fuel fission products, U, transuranium elements. Proportion of mentioned components used in matrix is as follows, mole percent: SiO2, 60-68; sum of Na2O, K2O, Cs2O, 11-18; sum of CaO, SrO, BaO, 3-6; sum of Fe2O3, Cr2O3, NiO, 2-4; Al2O3, 1-3; ZrO2, 4-7; sum of rare-earth elements, U, and transuranium elements, 1.5; the rest, 3.
EFFECT: enhanced chemical and thermal stability of matrix.
1 cl, 3 dwg, 3 tbl, 8 ex
FIELD: nuclear power engineering; removing radioactive pollutants from surfaces of pieces of equipment or parts by means of circulating solutions.
SUBSTANCE: proposed process for controlling cyclic decontamination involving saturation of decontaminating solution with radionuclides includes chemical treatment of surface pollutants with decontaminating solution, checkup of solution saturation with radionuclides, termination of chemical treatment as soon as saturation of decontaminating solution with radionuclides is brought to limiting value, and removal of saturated decontaminating solution. Ionizing radiation dose rate is remotely measured by means of gamma transducers installed at reference points and chemical treatment is ceased as soon as inequality conditions are satisfied.
EFFECT: enhanced reliability of process due to optimal evaluation of chemical treatment time; enhanced efficiency and quality of decontamination due to reduced secondary sorption of radionuclides.
3 cl, 18 dwg
FIELD: nuclear power engineering.
SUBSTANCE: proposed method designed for controlling cyclic decontamination process by determining optimal time of completing separate decontamination steps involving uninterrupted cleaning of decontaminating solutions in filters and primarily intended for removing radioactive pollutants from surfaces of equipment or separate parts by circulating solution, for instance for decontaminating inner surfaces of nuclear power reactor equipment such as coolant circuits of boiling water reactors (heavy-power pressure-tube reactors RBMK) includes chemical loosening operations and dynamic loosening conducted before and after chemical loosening operation, as well as washing to discharge radioactive pollutants from coolant circuit to filters during each loosening operation. Radioactivity level of pollutants discharged from coolant circuit is periodically checked against reference radionuclides, and each decontaminating operation is completed upon attaining following condition: . In addition, radioactivity level of pollutants discharged from coolant circuit is proposed to be calculated by sum of derived radioactivity of 3-7 reference radionuclides; 58,60Co, 54Mn, 59Fe 95Zr, and 95Nb are used as reference radionuclides.
EFFECT: enhanced reliability of process control due to determining optimal time of completing separate steps, reduced decontamination time, enhanced effectiveness due to reduced secondary sorption of radionuclides.
3 cl, 13 dwg, 3 tbl
FIELD: recovering and degreasing liquid radioactive wastes.
SUBSTANCE: proposed method for decontaminating waste water from radioactive components incorporating in their composition dissolved and/or emulsified mineral oil, dissolved and solid particles of uranium radioactive components, and products of its decay by concentration of radioactive components and mineral oil. Prior to recovery waste water is acidified to pH = 2.5-3.0. Then iron salt based coagulant (III) and modified polyacrylamide based flocculant are introduced. After that waste water is neutralized with alkali to pH > 7 followed by centrifuging purified water and concentrate containing radioactive components and mineral oil. For final procedure concentrate is solidified and buried.
EFFECT: reduced energy requirement, enhanced speed of process.
1 cl, 5 tbl, 5 ex