Analysis of the danger of structural failure with uncertainties present in the initial data

Initial data on the properties of a material, which are often difficult to determine with prescribed accuracy because of large variances in the experimental results, for example, experimental determination of the characteristics of creep in long-time strength of the materials used in the core of a nuclear reactor, are needed for computational simulation of structural deformation and failure. These conditions engender uncertainty in the results of strength calculations, making them improperly posed from the mathematical standpoint. A mathematical method for taking account of the uncertainties in strength calculations is examined. The method is based on the reconstruction of the distributions of random quantities over a sample of empirical data, using nonparametric method of ordered minimization of risk. The method and the computer codes DENSIT, PTUBL, and PCLAD, which are used for performing the calculation, are described. The application of these codes to the analysis of an accident which occurred in the No. 3 unit of the Leningrad nuclear power plant in 1992 is examined. __________ Translated from Atomnaya énergiya, Vol. 104, No. 3, 152–156, March, 2008.     

Optimization of the characteristics of the heat-insulating layer of quickly removable general-purpose insulation for nuclear power plants with VVÉR reactors

A comparative analysis is made of the computed and experimental data to confirm the adequacy of the heat-transfer model and the suitability of this model for determining the effective thermal condutivity of fibrous heat-insulation materials. Optimization calculations of the heat-insulation layer of quickly removable general-purpose heat insulation are performed for a wide range of the average diameter of fibrous materials as compared with the normative method. __________ Translated from Atomnaya énergiya, Vol. 101, No. 3, pp. 203–208, September, 2006.     

Fast integration of reactor-kinetic equations in problems with large perturbations of the reactivity

Conclusions The method proposed for integrating the point-kinetics equations of a reactor, has a series of positive qualities ensuring its effective application in problems of the numerical analysis of the emergency dynamics of nuclear-power plants with deep perturbations of the reactivity. Its advantages include:simplicity of the programs executing the computational algorithms (the calculation is based on explicit formulas);the presence of an automatic regulator of the computation step;high speed (the number of operations in calculations at the step is comparable with the analogous characteristic of the Euler method).Translated from Atomnaya Énergiya, Vol. 67, No. 4, pp. 246–251, October, 1989.     

Taking account of the effect of the curvature of VVÉR-1000 fuel assemblies on the power of the fuel elements

A comparative analysis is made of the deterministic and statistical methods of taking into account the effect of the curvature of VVéR-1000 fuel assemblies on the power of fuel elements. The fuel-element distribution of the energy release in the core for any random distribution of the gaps between the fuel assemblies is simulated, using the MEX code, on the basis of precise calculations (MCU code) and design calculations (BIPR-7 and PERMAK codes). The Monte Carlo method (Zazor code) was used to model the nominal density distribution of gaps in the core for different degrees of curvature of the fuel assemblies. It is shown that the power gain, obtained for the fuel elements by the probabilistic-statistical method, due to the curvature of the fuel assemblies is smaller and makes it possible to substantiate core safety with large perturbations, in contrast to the deterministic “maximum gaps near the most-energy stressed fuel element” method. 5 figures, 1 table, 3 references. Special Design Office “Gidropress.” Translated from Atomnaya énergiya, Vol. 87, No. 3, pp. 210–213, September, 1999.     

Calculations of the corrections for the true summation of cascade γ rays on the basis of statistical simulation using evaluated nuclear data

A method of calculating the corrections to the true summation of the γ radiation for arbitrary radionuclides for measuring point and volume sources using scintillation and ultrapure germanium detectors is described. The calculations are based on the Monte Carlo method using ENSDF evaluated data on the structure of atomic nuclei. An experimental check made using IAEA test spectra confirms that the calculations of the correction factors are reliable. The method is integrated into the γ spectrometric packages LSRM-2000 and AkWin, where the corrections for the ture summation are used for developing a library. __________ Translated from Atomnaya énergiya, Vol. 100, No. 5, pp. 382–388, May, 2006.     

Calculation of the neutron lifetime by the first-collisions probability method

The calculation of the neutron lifetime by the first-collisions probability method for nuclear power facilities which are described by a universal geometry is described. A derivation of the corresponding matrix approximations is presented. Algorithms for calculating the matrix elements using a universal geometric module, which, for example, is used in programs for performing Monte Carlo calculations of particle transfer, are proposed; MCU serves as an example of such a program. It is shown that the additional time required to calculate the neutron lifetime will not greatly slow down the calculations performed by the first-collisions probability method. Translated from Atomnaya énergiya, Vol. 106, No. 4, pp. 217–220, April, 2009.     

Use of the Monte Carlo method for multi-group reactor calculations

The article describes a multi-group method for two-dimensional reactor calculations in which the solution of the kinetic equation in the transport approximation is found for each group by the Monte Carlo method. The calculated results are in satisfactory agreement with the experimental data.Translated from Atomnaya Énergiya, Vol. 16, No. 2, pp. 119–122, February, 1964     

Computational investigation of the temperature and geometric characteristics of VVÉR fuel in the Kozlodui nuclear power plant (Bulgaria) under normal operating conditions

A method of performing stationary thermomechanical calculations of VVéR-440 and-1000 fuel elements, using the TRANSURANUS computer code to obtain the dependence of the temperature and radius of the fuel elements on the lineal power ensity and burnup, is described. These dependences are intended for use in neutron-physical calculations of the VVéR reactor at the Kozlodui nuclear power plant in stationary and transient regimes. The results obtained with this computer program are compared with calculations performed using the certified TOPRA-s code. The comparison shows reasonable agreement between the results of calculations of the fuel temperature. __________ Translated from Atomnaya énergiya, Vol. 101, No. 5, pp. 336–342, November, 2006.     

Neutron-physical calculations and experiments on a reactor in the first nuclear power plant built in the world

Methods for performing neutron-physical calculations of a reactor and the experiments performed during physical startup and subsequent operation are described. Starting with the simplest one-dimensional calculations during the reactor design process, the operation of the reactor is completed with three-dimensional calculations taking account of the heterogeneous structure of the core.Among the experiments investigations of the effect of water entering the graphite masonry during an accident with the fuel assemblies on the reactivity should be singled out.The method of partial fuel reloading, making it possible to decrease fuel-assembly burnout, has been implemented for the first time for the AM reactor.__________Translated from Atomnaya Énergiya, Vol. 98, No. 1, pp. 13–18, January, 2005.     

Tagged Particles in Monte Carlo Calculations of Radiation Transfer Using the Prizma Computer Program

The additional possibilities due to the introduction into the PRIZMA program, intended for performing Monte Carlo calculations of linear radiation transfer problems, of a particle tagging technique are studied. In calculations, it is often necessary to estimate the contribution of particles which were in some particular situation. The tagged-particle method was developed to obtain such results. A particle is tagged with a marker and the results are recorded as a function of whether or not the particle is tagged. To solve multivariant problems, arising in the calculation of the effects due to small perturbations, a special method was developed for simulating trajectories using tagged particles. This method makes it possible to obtain in a single calculation correlated results for all variants of the problem. The possibilities of the tagged-particle method and the special method of simulation are illustrated for numerical examples.__________Translated from Atomnaya Energiya, Vol. 98, No. 5, pp. 386–393, May 2005.     

Method for assessing the consequences of radiation accidents in the rooms of a power-generating unit of a nuclear power plant with a VVÉR-1000 reactor

Monte-Carlo estimation of the dose rate of external radiation from volume and surface sources of a gas-aerosol radioactive impurity, arising in the rooms of a power-generating unit during a radiation accident at a nuclear power plant, is examined. The volume and surface sources are distributed uniformly. The radial distribution of the dose buildup factor and the dose rate are obtained as functions of the γ-ray energy at different heights. The calculations are performed for distances no more than four γ-ray mean-free path lengths (μd ≤ 4) by the local flux estimation method. Comparisons with similar calculations performed with computer programs employing the discrete-ordinate method show satisfactory agreement. __________ Translated from Atomnaya énergiya, Vol. 102, No. 4, pp. 254–262, April, 2007.     

Probability analysis of the safety of the second energy module of the leningrad nuclear power plant

A special international program was undertaken to estimate the safety of the second power unit at the Leningrad Nuclear Power Plant after reconstruction and the effectiveness of the new safety measures introduced. The analysis is based on the IAEA method, modified for RBMK reactors. The calculations employ a full-scale model of the risk of the power unit and the initial data for a 10-year operating period. After reconstruction, the frequency of serious accidents with core damage is 1.4·10⁻⁵ per reactor-year and is acceptable in terms of the OPB-88/97 requirements. The influence of reconstruction is considerable: according to the estimates, the risk is reduced by a factor of more than 100. The results of the analysis permit suggestions regarding further increase in safety and reduction in cost of the reconstruction without increasing the risk of accident. Scientific-Research and Design Institute of Power Engineering. Translated from Atomnaya énergiya, Vol. 87, No. 2, pp. 113–117, August, 1999.     

Estimation of the Effect of Small Perturbations in Multivariant Calculations Using the Prizma-D Computer Program

In eigenvalue problems, in contrast to problems with a fixed source, the perturbation introduced into the system affects the distribution of the fission points of the source. The method of “small iterations on fissions” was developed to take account of this influence. This method is used together with a special method for simulating trajectories. The crux of the method is that for all variants of the problem the initial source is a source for the unperturbed variant, after which, to establish the source for each variant, iterations on fissions are performed in the second part of the computational cycle using a special method of simulation. The possibilities of using the method of “small iterations on fissions” are demonstrated with calculations performed for several problems and the comparative results are presented. __________ Translated from Atomnaya Energiya, Vol. 99, No. 3, pp. 203–210, September, 2005.     

Models of photonic sources of radiological facilities for use in Monte Carlo irradiation dose calculations

The purpose of the present work is to develop a method for calculating the characteristics of photon bremsstrahlung of a virtual source or developing a model of a source for use in Monte Carlo calculations. Two methods for reconstructing the initial spectrum of the bremsstrahlung of radiotherapeutic facilities are proposed. One is based on analysis of the phase space of particles under the collimator plane, and the other is based on analysis of the experimental dose distribution. The proposed methods for reconstructing the effective spectrum satisfactorily describe the real particle spectrum of radiotherapeutic facilities. The proof for this assertion is a simulated dose distribution, which agrees with the experimental data. __________ Translated from Atomnaya énergiya,Vol. 104, No. 2, pp. 106–111, February, 2008.     

PRIZMA.D computational investigation of physical processes in the IGR reactor

It is shown for one experiment that the physical processes occurring in the-core of a pulsed uranium-graphite reactor IGR can be investigated by a computational method using the PRIZMA.D program and correlated results in multivariant calculations. 7 figures, 1 table, 7 references. Translated from Atomnaya énergiya, Vol. 88, No. 2, pp. 83–88, February, 2000.     

Heterogeneous calculation of hexagonal lattices by the method of passage probabilities

A method is proposed for determining the characteristics of heterogeneous hexagonal cells using first-collision probabilities—probabilities of passage, escape, and collisions, as well as calculations of the flows in cell zones on the basis of balance equations for the current. The spatial-angular dependence of the current at the sides of a cell is described in a discrete approximation, and the nonuniformity of the sources and flows in zones in described using spatial harmonics. The results obtained in one-group calculations of model lattices by the passage probability and Monte Carlo methods are compared. 3 figures, 5 tables. 5 references. IPé, Belarussian Academy of Sciences. Translated from Atomnaya énergiya, Vol. 87, No. 5, pp. 330–335, November, 1999     

Determination of the optimal parameters of a cascade of gas centrifuges

Conclusions The necessary relations which make it possible to implement the numerical-analytical method for determining the optimal parameters of a cascade of centrifuges were obtained. The fundamental distinction from conventional calculations of ideal cascades lies in the fact that high coefficients of utilization of the separation power with strictly fixed external parameters of the cascade are guaranteed. The method can be extended to other formulations of the problem in which the total number of centrifuges appears as a separate component of the target function. Moreover, it can be used for optimizing a real cascade with losses. Since the losses are small, it is sufficient to make several calculations. The first calculation is performed with the formulas presented for estimating N_i and C_i and the corresponding losses q_i of matter and of isotopes q_ui (i=l,...n). The next calculations are performed iteratively according to the prescribed values of q_i and q_ui. In these variants, the expressions for the transit fluxes τ_i′,τ_ui′ are corrected each time. Similar variations of calculations are also possible in other situations, for example, in the case when the number of centrifuges in the steps is small. The possibility of making such changes is due to the flatness of the target function — the target function increases very little when the parameters deviate from strictly best values. This work was supported by a grant from the Moscow Engineering Physics Institute (Goskonvuz) for 1996–1997. Ural State Technical University. Translated from Atomnaya énergiya, Vol. 84, No. 3, pp. 246–253. March, 1998.     

Assessment of the quality and activity of the radionuclides in the fuel cycles of a three-component structure of nuclear power

A three-component structure of nuclear power with a closed nuclear fuel cycle is examined. In addition to the existing thermal and fast reactors the system contains a liquid-salt reactor for closing the fuel cycle with respect to actinides. The quantity and activity of the radionuclides are analyzed for promising variants of the structure of nuclear power and the uranium-plutonium, thorium-uranium, and uranium-plutoniumthorium closed fuel cycle. It is concluded on the basis of calculations and analysis, taking into account the life cycle from production of the fuel to the burial of the wastes, that the uranium-plutonium-thorium fuel cycle is more advantageous than the uranium-plutonium and thorium-uranium fuel cycles. __________ Translated from Atomnaya énergiya, Vol. 101, No. 3, pp. 214–221, September, 2006.     

Computational and experimental investigations of the neutron-physical characteristics of the GT-MHR core

The main results of computational and experimental investigations of the neutron-physical characteristics of the GT-MHR high-temperature gas-cooled reactor are presented: analysis of the possible reasons for the uncertainties in the calculations of the neutron-physical characteristics of a core with plutonium fuel using different computer programs, computational and experimental investigations taking account of the experiments on the Astra critical assembly simulating the ring-shaped GT-MHR core, and the effect of using different nuclear data on the temperature coefficient of reactivity computed with high-accuracy computational codes such as MONTEBURNS-MCNP 5-ORIGEN 2.1. __________ Translated from Atomnaya énergiya, Vol. 102, No. 1, pp. 63–68, January, 2007.     

Analysis of neutron flux noise due to coolant parameter fluctuations in a VVÉR core

The results of investigations of noise appearing in the signals from direct-charge sensors of the in-reactor monitoring system of VVéR reactors as a result of coolant parameter fluctuations are presented. The calculations and experimental data are used to analyze the dependence of the amplitude of the neutron flux oscillations (local intensity) at the locations of sensors as a function of the magnitude and frequency of the fluctuations of a specific coolant parameter. The NOSTRA program was used to perform the calculations. The results of the analysis were used in the in-reactor noise diagnostics system in the No. 3 unit of the Kalinin and the No. 1 unit of the Tianwan (China) nuclear power plants. __________ Translated from Atomnaya énergiya, Vol. 103, No. 5, pp. 283–286, November, 2007.