


Vol 120, No 2 (2016)
- Year: 2016
- Articles: 12
- URL: https://journal-vniispk.ru/1063-4258/issue/view/15429
Article
Surface Pseudo-Source Method of Accounting for Scattering Anisotropy in Many-Group Calculations of VVER Cells
Abstract
A method is developed for solving the many-group neutron transport equation with an anisotropic scattering cross section in cylindrical geometry by a method where the operator is split and the single-group spatialangular parts of the problem are solved by the method of surface pseudosources. The developed procedure is implemented in the RATsIYa option of the WIMS-SH and SVL software systems. It is used to solve a VVER-1000 cell with mixed uranium-plutonium fuel and a VVER-1000 supercell with an absorbing element. The factor K∞calculated in the transport and linear-anisotropic approximation differs by up to 0.4% in the first cell and up to 2.3% in the second cell.



Computational Analysis of a Nuclear Reactor Cell in the Linear-Anisotropic Approximation by the Generalized Method of First Collisions Probabilities
Abstract
This work is devoted to the development of a method for solving the Boltzmann kinetic equation for neutron transport. The method is focused on solving an integral equation for neutron transport in a reactor cell with a complex geometry and different boundary conditions. An algorithm is proposed for solving the problem taking account of anisotropic scattering in the linear-anisotropic approximation. The approach is based on expanding the neutron flux in a system of orthogonal two-dimensional polynomials in each uniform zone of a heterogeneous cell. This expansion reduces the system of linear integral equations to a system of linear algebraic equations. The relations required to calculate the coefficients in the equations (six-fold integrals) as well as the computational algorithm are presented. Calculations of cylindrical and cluster cells are presented. The calculations are compared with the surface pseudosource method. It is shown that the results have advantages over the method of first collisions probabilities.



Analysis of the Correlation Function of the Neutron Field in a Critical Reactor Taking Account of the Neutron Intensity Regulation System
Abstract
The influence of the parameters of the neutron intensity regulation system on the uncertainty of the neutron field in a critical stationary reactor is investigated. The spectral expansion method is used to obtain by means perturbation theory a relation for the neutron flux density distribution. On the basis of the relations obtained an expression for the correlation function of the neutron field is obtained and analyzed. Numerical modeling is used to determine the spatial distribution of the dispersion of the neutron field for different arrangements of the neutron detectors and control organs.



Seismic Margin Assessment Methodology for NPP Buildings, Structures, and Equipment
Abstract
The conditions for using the SMA method based on foreign databases on seismic classification are examined. In contrast to the initial SMA methodology a modified approach does not presume rejection of computationalexperimental evaluation of seismic safety under real conditions. However, the methodology presented examines the possibility of validated reduction of work, including rejection of computational-experimental checking when using the domestic database now under development. Evaluations of the seismic safety margin are used in the validation of service life extension for NPP, reexamination of safety, and probability analysis of safety as well as in establishing the technical specifications for equipment, modernization and reconstruction, and improving safety.



Application of Frequency Regulation of the Main Circulation Pump for Increasing the Nominal Power of VVER
Abstract
The thermal power of NPP with VVER-100 can be increased by increasing the flow of reactor water by increasing the number of revolutions of the main circulation pump. The computational method proposed in the present article makes it possible to determine, for prescribed relations between the reactor water flow and the crisis of heat exchange in the reactor setup with known geometry of flow over the fuel elements and with the prescribed heightened coolant flow, the admissible power increment for constant safety margin of the heat-engineering reliability of the core. The expected effects and problems in implementing the method of operating the reactor are examined and evaluated.



Two-Dimensional Thermohydraulic Module of the Integrated Code Sokrat-BN: Mathematical Model and Computational Results
Abstract
The two-dimensional thermohydraulic model used in the thermohydraulic module of the SOKRAT-BN code is described and the computational results obtained by using it are presented. The thermohydraulic model is based on the solution of two-dimensional equations using empirical correlations for determining the intensity of the turbulent transfer of mass, energy and momentum in the radial direction. The experiments performed in the 19- and 37-pin assemblies on the Siena stand in Japan and in the 37-pin assembly in the KNS stand in Germany were calculated. The average deviations of the computed values do not exceed 10%.



Catalytic Oxidation of Trace Quantities of Hydrogen in Tritium-Containing Gas Flows in the Event of a Fire at Nuclear Facilities
Abstract
The effect of the fume gases, produced when the insulation of chlorine-free cables burns, on the catalytic activity of hydrophobic low-temperature and hydrophilic high-temperature platinum catalysts of hydrogen oxidation, which are to be used in the ventilation-gas detritization system. The temperature regime for the combustion of cables in the presence of an oxygen deficiency is determined and the semiquantitative composition of the products formed is obtained. Upon exposure to the fume gas flow at room temperature the experimental samples of the catalysts were almost completely poisoned. At the initial temperature 473 K of a hydrophilic catalyst thermal activation of the catalyst occurs in the presence of fume gases.



Identification of the Main Dose-Forming Radionuclides in NPP Emissions
Abstract
The results of an analysis of gas-aerosol emissions from the ventilation tubes of European NPP are presented. An assessment is made of the contribution of the controlled radionuclides in the exposure of the public to the radiation from NPP emissions of five types of reactor facilities: AGR, BWR, LWGR, PWR, and CANDU. The radionuclides forming the 95% dose in the critical group of the population are determined for each type of NPP.



Nitrogen Hemioxide: Properties and Neutralization Methods
Abstract
The physical and chemical properties of N2O and methods for its trapping and decomposition are examined. It is found that organic solvents, water, hydrazine hydrate, and a solution of potassium bichromate in sulfuric acid dissolve N2O but its stabilization by this method is unreliable. Catalytic decomposition upon heating is acknowledged to be an effective method of neutralizing N2O. Different composite materials, including those containing d and ƒ elements, were studied as catalysts. It was found that suitable composites for catalytic decomposition of N2O are SiO2–Cu, Al2O3–Ni at 150°C and Al2O3–Cu, Al2O3–Fe, Al2O3–Ni, and SiO2–Cu at 250°C. It was shown that the passage of ethanol vapors above the composites at the indicated temperature increases the degree of decomposition of N2O by a factor of 10. The results obtained were used to develop a gas-purification system in the technology for reprocessing spent nitride fuel.



Trapping of Nitrogen Oxides in Radiochemical Technologies
Abstract
A scheme for trapping the nitrogen oxides released during operations with liquids in radiochemical technologies is develop-ed: the outgoing gases containing nitrogen oxides pass through a dephlegmator (in the process of dissolution of fuel or wastes) or bubbler condenser (in the process of denitration and vitrification), through FSGO and FARTOS filters after water aerosols are introduced into them, and through the RFTBL and then SMOG apparatus after the next water-aerosol portion is introduced. The proposed solution is incorporated in the technological schemes of the spent nuclear fuel reprocessing module in the on-site nuclear fuel cycle of the BREST reactor.



Worker Radiation Exposure
Abstract
This article contains information on professional external and internal radiation exposure in 2013 of workers in organizations, including NPP, on territories serviced by the Federal Medical-Biological Agency of Russia (FMBA). The parameters of the dose distribution are calculated: mode, median, average, standard deviation, quantile, and coeffi cient of variation for institutions on territories serviced by the FMBA. The number of individuals subject to individual dosimetric monitoring, grouped by institution and taking account of the radiation exposure dose is analyzed. Information on the professional external and internal radiation exposure of NPP workers is presented.



Method and Means for Monitoring the Spent-Fuel Pool Subcriticality at the Smolensk NPP
Abstract
Research based on the pulsed neutron experiment method (α-method) of subcriticality monitoring of the spent fuel pool at the Smolensk NPP is presented. The procedure includes stationary and nonstationary neutron-physical calculations and measurements of the main characteristics of a subcritical multiplying system, which the spent-fuel pool is. A description of the STEPAN-KhOYaT software specially developed for the computational support of the pulsed experiment is given. The subcriticality for real loads of spent-fuel pools is calculated and measured using the procedure.


