Problems related to axisymmetric deformation of a truncated conical shell of constant thickness under the effect of a constant pressure are encountered when calculating the linear stress-deformed state of shell structures used in the nuclear fuel cycle within the framework of a model of the Kirchhoff–Love parameters. In practice, an analytic solution of such boundary-value problems leads to certain mathematical difficulties related to integration of one of the two differential decision equations, which is of fourth order, reduced to second order by means of a complex transformation. The differential decision equations for a truncated conical shell obtained following Meisner were transformed without any reduction in order to a new form more convenient for practical integration by means of infinite power series. The solution found in terms of power series was tested in compiling a test example of a calculation of the parameters of a deformed annular apparatus one of whose platforms is in the form of a truncated conical shell. The test example contains the results of an analytic solution of the above boundary-value problem and finite-element modeling using the CAN program. The error of the calculated stresses is estimated for an analytic solution, making it possible to consider the solution as a standard for comparison.

It is proposed that the familiar physical phenomenon of the equilibrium of a system subjected to internal static loads be used to evaluate the error of solution of boundary problems by analytical methods of the theory of thin plates and shells under conditions when, as a rule, there is no rigorous solution, and exact numerical values of parameters that can be balanced or combined using formulated boundary conditions are unknown. Measures are presented for evaluation of the imbalance of mechanical forces and bending moments in an individual structural element, and the structure as a whole, which is caused by the error of the functions that closely approximate particular solutions of differential equations that resolve the problem, and by the error of arbitrary constants; a method is also presented for evaluating of the lower and upper boundaries of the error generated in determining internal force factors and stresses, which is based on analysis of the residual of differential and integral equations.

Physicochemical properties of magnesium potassium phosphate (MPP) matrices for conservation of highly saline high-level liquid wastes (HLW) after their solidification at room temperature were studied comprehensively. The matrices showed high chemical stability to leaching of radionuclides and other components at various temperatures. The leaching indices of 239 Pu, 237 Np, 241 Am, Sr, and Cs are in the range 12–14, and those of Tc, I, and Se, in the range 10–11. The mechanical strength (>20 MPa) and radiation resistance of the matrices and the chemical yield of radiolytically generated hydrogen (0.004 molecules/100 eV) were determined. The phase composition of the matrices and the character of radionuclide distribution in their bulk were determined. An enlarged installation was developed and fabricated at the Mayak Production Association, and solidification of simulated liquid radioactive wastes from this plant was performed in volumes of up to 2001.

The new phosphate Cs 2 Mn 0.5 Zr 1.5 (PO 4 ) 3 was synthesized for the first time and characterized by X-ray diffraction. Its crystal structure was refined in space group P 2 1 3, Z = 4 at 25°C ( a = 10.3163(1) Å, V = 1097.93(1) Å 3 ), by the Rietveld method using the powder X-ray diffraction data. The structure is built of an octahedral-tetrahedral framework {[Mn 0.5 Zr 1.5 (PO 4 ) 3 ] 2− } 3∞ with cesium atoms being located in large cavities. The hydrolytic stability of the powdered phosphate containing 137 Cs radionuclide was studied. The minimum achieved 137 Cs leaching rate was 4 × 10 −8 g/cm 2 day.

Perovskites of the compositions xLa2O3·Al2O3 and xLa2O3·Fe2O3 were prepared by cold pressing and sintering at temperatures of up to 1500°C, and their properties were studied. Perovskites based on lanthanum aluminate are more porous than those based on lanthanum ferrite, whereas the rate of radionuclide leaching from the former compounds is lower than from the latter compounds. The parameter x varied in a relatively wide range does not affect the leaching rate.

Matrices of the compositions 0.2La 2 O 3 ·Al 2 O 3 and 0.4La 2 O 3 ·Fe 2 O 3 , containing weighable amounts of Pu (13 mg g −1 ) and Am (0.336 mg g −1 ), were prepared, and their leaching rates were studied. The leaching rates calculated from the surface area determined by low-temperature nitrogen adsorption (BET method) virtually coincide with those found previously for the matrices with tracer amounts of the radionuclides, which indicates that these parameters are intrinsic characteristics of the matrices. The leaching rates calculated from the geometric surface areas are considerably higher and depend on the material porosity and hence on the procedure of its synthesis. The rates of the matrix corrosion were calculated from the 147 Pm leaching rates. They show that the matrix will not disintegrate into small fragments for 100 thousand years even under the emergency conditions of repository flooding with water.

An analytical solution of the boundary problem, which is one of the different forms of the approximate method of elastic solutions, is obtained within the framework of the deformation theory of small elastoplastic strains for the Kirchoff-Love model of a shell structure loaded by constant pressure. Completely stable approximations of partial solutions of the differential equations resolving the problem of the deformation of a cylindrical shell with compulsorily assigned initial strains are plotted using specially selected basis and weighting functions, and also boundary conditions formulated for the elasto-plastic section. In an example of an annular vessel, the permissible shaping of which is restricted by safety concerns, the variation in the width of the gap between cylindrical shells is analyzed elasto-plastically.

The results of investigations of the nuclear safety of the masonry in the AI uranium–graphite (Industrial Association Mayak) reactor are presented. It is concluded on the basis of these results that the masonry is nuclear safe and a radiation certificate is composed. The radiation examination made it possible to determine the level, composition, and distribution of the radioactive contamination of the masonry as well as the level and distribution of the neutron and γ radiation, and to construct a forecast of the change in the activity of radionuclides in graphite as a function of the holding time. These data are necessary for safety analysis and for making decisions about the subsequent stages of decommissioning of the reactor.

Features of the formation of a nonuniform temperature field and the appearance of thermal stresses in linked elements of structures that are washed with media at substantially different temperatures were studied on a radiochemical apparatus. The level and dynamics of changes in temperature stresses in the body of the studied apparatus were found from solution of a static boundary-value problem of thermal elasticity in an axially symmetric formulation by the finite-element method. To specify one of the boundary conditions necessary for solving this problem, the temperature field on the apparatus’s exterior surface was determined experimentally with a NEC Thermo Tracer TH-9100 thermal imager. Other boundary conditions were determined from the results of physical modeling and from approximate analytical solutions. It has been established that the maximum thermal stresses arise during the initial short period when the apparatus is heated by overheated steam supplied into it.

It is proposed that a database on the characteristics of research reactor fuel be constructed. This database will permit determining the unknown fuel and the unknown parts of the core. Arduous and precision methods are not needed to identify the fuel parts of pulsed reactors. Inspection and measurements of the geometric dimensions are sufficient in most cases. If a fuel part of a reactor is destroyed and has been converted to scrap, then the enrichment of the fuel, the isotopic composition of the uranium, and the impurity composition must be measured, although even this does not always solve the problem. Consequently, the selection and validation of the identifying indications must be the subject of special investigations.

A graphic comparison is presented for results of analysis of stresses determined from approximateanalytical solution and the finite-element method using the CAN program.The validity of results derived from analytical calculations is characterized by the magnitude of the residual of the equilibrium equations of an infinitely small region of the median surface of structural elements, which solve the problem of the differential and boundary equations.The trial example makes it possible to confirm the performance of programs employed for calculation of the SSS of geometrically complex shell structures.

The results of an examination of the radiation state of the graphite masonry in three commercial uranium-graphite reactors at the Mayak Industrial Association are presented. On the basis of these results, conclusions are drawn about the nuclear safety of the masonry and radiation licenses are composed. The comprehensive radiation examination made it possible to determine the level, composition, and distribution of radioactive contamination of the masonry and the level and distribution of neutron and γ radiation and to predict the variation of the radionuclide activity in the graphite as a function of the holding time. These data are required to make design decisions about further reactor decommissioning stages.

The precipitation of poorly soluble Pu(IV) and Np(IV) dibutyl phosphates on acidifying Pu(IV)-and Np(IV)-containing carbonate solutions contaminated with dibutyl hydrogen phosphate was studied. The degree of precipitation of Pu and Np is dependent on the final acidity and the storage time of the resulting suspension. The presence of iron in carbonate solutions (<0.5 g/l) does not noticeably affect the distribution of the actinides in the neutralized solutions.

In regeneration of Np(IV)-and Pu(IV)-containing recycled solvent by treatment with aqueous sodium carbonate contaminated with iron, some portion of Np(IV) and Pu(IV) coprecipitates with hydrolyzed iron. The degree of coprecipitation of Np and Pu depends on both the iron and sodium carbonate concentrations. The presence of n-dibutyl hydrogen phosphate in the recycled solvent before its regeneration does not noticeably affect the coprecipitation of Np and Pu. The possible mechanisms of coprecipitation of actinides with hydrolyzed iron in carbonate solutions are discussed.

The potential harm of long-lived radionuclides estimated as the product of the activity of a radionuclide and the dose coefficient is examined. The potential harm of actinides in high-level wastes is calculated taking account of the harm due to their decay products. At the same time, an analogous calculation is performed for the uranium isotopes 238 U, 235 U, and 234 U consumed in the reactor. The value obtained, which depends on the time, can be regarded as the averted harm. The time for establishing radiation equivalence between the high-level wastes and the consumed uranium is determined as the time in which the potential harm from actinides becomes equal to the averted harm. It depends on the holding time of the spent fuel before radiochemical reprocessing. For a 5 yr holding period, it is ∼49000 yr.

Sorption of Tc(VII) on fibrous sorbents filled with finely dispersed anion exchangers (AV-17-n and AN-31-n) and on a chelating sorbent bearing 1,3(5)-dimethylpyrazole groups (POLYORGS-17-n) is studied. The Tc(VII) distribution coefficients ( K d , cm 3 g −1 ) are determined in sorption from simulated nitric acid, alkaline, and saline solutions. The highest distribution coefficients ( K d ∼10 3 − 10 5 cm 3 g −1 ) were obtained in sorption of Tc from weakly acidic (≤0.1 M HNO 3 ) and alkaline (0.1–1.0 M NaOH) solutions. In sorption of Tc from alkaline and nitric acid solutions containing NaNO 3 , POLYORGS-17-n demonstrated the highest efficiency. Tc is totally recovered with the sorbents studied in 10 min. Then it can be eluted with 1 M HNO 3 . The sorbents AV-17 and POLYORGS-17-n can be used for Tc(VII) recovery from saline solutions like groundwater (pH ∼8).

It is shown that technetium can be extracted from acidic media by the fibrous complexing sorbent POLIORGS 35, containing hydrazidine groups, and from neutral and nitric-acid solutions with a complex salt composition (with nitrate-ion concentration not exceeding 10 g/liter) by strongly basic anionite POLIORGS AV-17 with quaternary ammonium bases. The effect of the pH of the solutions on the sorption of technetium is studied. It is shown that the sorption depends on the nature and concentration of the salt background of the solutions. The kinetics and mechanism of sorption of technetium by fibrous sorbents and the possibility of their desorption and reuse in sorption–desorption cycles are investigated. It is shown that the fibrous sorbents POLIORGS 35 and POLIORGDS AV-17 are promising for extracting technetium from waters with a high mineral content (up to 20 g/liter).