Determination of the optimal parameters of a gas centrifuge cascade with an arbitrary stage connection scheme

A method for calculating and optimizing a cascade of gas centrifuges with an arbitrary scheme for connecting the stages is proposed. The particulars of its application to separation of a binary mixture of uranium isotopes are examined. Different cascades optimized using the criterion of minimum total number of gas centrifuges are compared. It is shown that the choice of symmetric or asymmetric schemes for designing the cascade depends on the type of gas centrifuge used.     

Method for calculating the optimal parameters of a counter-current cascade with an arbitrary interstage coupling scheme

The characteristic features of the optimization of cascades with an arbitrary interstage coupling scheme are examined. A method for calculating the optimal parameters, which is universal with respect to the type of scheme and optimization functions, is proposed for the case of constant separation coefficients. The method is based on searching through the waste concentrations of the stages and solving systems of algebraic equations for the waste fluxes. The properties of optimal symmetric and asymmetric cascades with various optimization criteria are investigated. It is shown that the method can be extended and that optimal cascades differ from ideal cascades. 3 figures. 2 tables, 6 references. Ural State Technical University. Translated from Atomnaya énergiya, Vol. 88, No. 2, pp. 142–148, February, 2000.     

The effect of certain cycle parameters on the efficiency of a nuclear gas turbine unit

The effect is analyzed of the increase of the effective pressure ratio, the regeneration factors, the initial gas pressure, temperature of the fuel element cladding, hydraulic resistance of the gas circuit on the internal efficiency of a nuclear gas turbine unit, taking into account the characteristics of the active zone of the reactor. The results are given of the effect on the efficiency of a nuclear gas turbine unit (NGTU) of the intermediate heating and cooling of the gas. A possible circuit for a NTGU is discussed, with one intermediate heating and three-stage cooling of the gas;Translated from Atomnaya Énergiya, Vol. 20, No. 5, pp. 412–415, May, 1966.     

Effect of scattering anisotropy on the optimal choice of parameters for tracklength biasing

An approximate theoretical formulation is presented which leads to an analytical expression for the optimal tracklength biasing parameter as a function of the scattering probability and the scattering anisotropy. This shows the sensitive dependence of the optimal parameters on the scattering anisotropy. The implications of these findings on the Monte Carlo simulation of deep-penetration problems are discussed.     

Estimation of research reactor core parameters using cascade feed forward artificial neural networks

The pattern of the core reload program is very important for an optimize use of research reactors. Reactor safety issues and economic efficiency should be considered during pattern studies. In order to find the best core pattern for a research reactor, its reloading program should be solved as a multi-objective and constrained optimization problem. If considered objective functions of the optimization problem can be estimated in very short time, the optimal fuel reloading pattern can be used effectively. In this research a very fast estimation system for suggested core parameters has been developed using cascade feed-forward type of artificial neural networks (ANNs). Four main core parameters are suggested to optimize reactor core adequately. And also to get larger thermal fluxes in the desired irradiation box, a new flexible method was selected. A Software package has been developed to prepare and reform required data for ANNs training. The gradient descent method with momentum weight/bias learning rule has been used to train ANNs. To get the best conditions for considered ANNs training a vast study has been performed. It includes the effects of variation of hidden neurons, hidden layers, activation functions, learning and momentum coefficients, and also the number of training data sets on the training and simulation results. Some experimental convergence criteria are used to study them. A comparison selection rule has been used to adjust desirable conditions. Final training and simulation results show that developed ANNs can be trained and estimate suggested core parameters of research reactors very quickly. It improves effectively pattern optimization process of core reload program.     

Determination of a test section parameters for IRIS nuclear reactor pressurizer

An integral, modular and medium size nuclear reactor, known as IRIS, is being developed by Westinghouse and by research centers. IRIS is characterized by having most of its components inside the pressure vessel, eliminating the probability of accidents. Due to its integral configuration, there is no spray system for boron homogenization, which may cause power transients. Thus, boron mixing must be investigated. The aim of this paper is to establish the conditions under which a test section has to be built for boron dispersion analysis inside IRIS reactor pressurizer. Through Fractional Scaling Analysis, which is a new methodology of similarity, the main parameters for a test section are obtained. By combining Fractional Scaling Analysis with local scaling for the densimetric Froude number and a previously established volumetric scale factor, the values of recirculation orifices, inlet water temperature, time scale factor and recirculation flow for the test section (model) are determined so that boron distribution is well represented in IRIS reactor pressurizer (prototype). Analytical solutions were used to validate the adopted methodology and when the results simulated in the model are compared to those that characterize the prototype, the agreement for both systems is absolute. The thermal power also influences boron distribution inside the test section. This power is determined by condensation laws in the vapor region and by suitable correlations for free convection. The fractions for rising inlet recirculation water enthalpy and vapor formation are also considered.     

Determination of a test section parameters for IRIS nuclear reactor pressurizer

An integral, modular and medium size nuclear reactor, known as IRIS, is being developed by Westinghouse and by research centers. IRIS is characterized by having most of its components inside the pressure vessel, eliminating the probability of accidents. Due to its integral configuration, there is no spray system for boron homogenization, which may cause power transients. Thus, boron mixing must be investigated. The aim of this paper is to establish the conditions under which a test section has to be built for boron dispersion analysis inside IRIS reactor pressurizer. Through Fractional Scaling Analysis, which is a new methodology of similarity, the main parameters for a test section are obtained. By combining Fractional Scaling Analysis with local scaling for the densimetric Froude number and a previously established volumetric scale factor, the values of recirculation orifices, inlet water temperature, time scale factor and recirculation flow for the test section (model) are determined so that boron distribution is well represented in IRIS reactor pressurizer (prototype). Analytical solutions were used to validate the adopted methodology and when the results simulated in the model are compared to those that characterize the prototype, the agreement for both systems is absolute. The thermal power also influences boron distribution inside the test section. This power is determined by condensation laws in the vapor region and by suitable correlations for free convection. The fractions for rising inlet recirculation water enthalpy and vapor formation are also considered.     

Determination of the minimal number and optimal sensor location in a nuclear system with fixed incore detectors

The determination of the minimal number of sensors and the optimal sensor location in a nuclear system with fixed incore detectors, which is represented by a linear stochastic distributed parameter system, was studied in this work. The partial differential equation representing nuclear reactor dynamics was approximated to the finite dimensional ordinary differential equation by the modal expansion. A scalar measure of the covariance matrix error in the optimal filter was minimized with respect to the sensor locations. The necessary conditions for optimal sensor location were derived using the matrix minimum principle, thus making the calculations computationally more attractive. The locations of sensors were guessed initially through sensitivity analysis to reach solutions of the optimal location quickly. A method to determine the minimum number of sensors is suggested based on the observability and admissible error bound. Several numerical simulations are performed to determine the minimal number and optimal sensor location for a one-dimensional slab reactor and a two-dimensional ABB Combustion Engineering type reactor with fixed incore detectors. Through the simulations the possibility of practical implementation and the rapid convergence of the algorithm are verified.     

Determination of plasma parameters in rf glow-discharge plasma

The Langmuir probe measurements were carried out in a planar rf discharge of N₂ and Ar gases. The dc characteristic I-U curves are calculated from the measured rf characteristic I-U curves for frequencies 10-60 MHz. The measured parameters such as electron temperature and electron density are compared with the simulated results. At gas pressures 10-40 Pa, the sheath thickness at the powered electrode was proportional to f^−0.5 for simulation and f^−2/3 by using the electrical parameters of the probe measurements and to p^−0.5 for both cases.     

一种快充式多丝正比计数器测量仪

研制了一种快速充气式多丝正比计数器测量仪器,可大幅减少工作气体消耗,缩短测量时间。能准确测量α粒子和β粒子,探测效率高、计数率线性好、测量结果稳定。样品测试成本大幅度低于纯流气式多丝正比计数器。