The dynamics of nuclear power stations

This paper makes a comparison of the results of eXperimental and theoretical studies that have been carried out on the properties of the engineering model of the Beloyarskii atomic electric station under construction in the USSR, which uses nuclear superheating of the steam. It is shown that a number of the simplifying assumptions are correct which are often used in discussing the dynamics of nuclear power stations.The results of the studies may be used to make a theoretical analysis of the dynamic properties of several types of nuclear power installations, as well as in analyzing and synthesizing the optimum control system.Notation q() specific heat load, referred to length of segment, kcal/hour · m - f(x) distribution function of specific heat load along the length of segment - () heat transfer coefficient, including the thermal resistence of the fuel element, kcal/m² · hour · degree - t_f.e. (x, ) the current value of fuel element temperature, averaged over the corss section, degrees C - t(x, t) current value of coolant temperature, degrees C - p perimeter of fuel element, bathed by coolant, m - m weight of metal per unit length of fuel element kg/m - C_M heat capacity of metal and fuel element, kcal/kg · degree - i(x, ) current value of heat content of coolant, kcal/kg - specific gravity of coolant, kg/m³ - S live cross section of fuel element, m² - D(x, ) current value of flow of steam phase, kg/hour - G(x, ) current value of the flow of water phase, kg/hour - (x, ) current value of the fraction of the cross section occupied by steam - , specific gravity of water and steam at saturation temperature, kg/m² - i, i heat content of water and steam at saturation temperature, kcal/kg - t_S() saturation temperature, degrees C - P_i() pressure in i-th segment, kg/m² - l height, determining the level pressure between segments, m - g acceleration of gravity, m/hour² - w_i() coolant velocity at the i-th segment, m/hour - D_i() steam flow at the i-th segment of the superheating circuit, kg/hour - V_i volume of i-th segment of the superheating circuit, m³ - mean steam temperature at the i-th segment for the superheating circuit, degrees C - k₁,k₂,k₃,k₄ constant coefficients - N/N₀ relative power change in the evaporating channels, % - P_I, P_II pressure change in the first and second loops, atm - t_sps, t_fw change in temperature of superheated steam and feed water, respectively, degrees C Translated from Atomnaya Énergiya, Vol. 15, No. 2, pp. 115–120, August, 1963     

Heat accumulators at nuclear power stations

Conclusions It has been shown to be technically possible and economically desirable to introduce heat accumulators into nuclear power stations and CHPS, particularly FWA, DHWA, and DSWA. The largest power-control range is attained with HA of DHWA, PTA, and DSWA types and is 120–140% for stations with HA of the water class in relation to the nominal reactor power or up to 160% for stations with PTA.The inclusion of HA in a nuclear station provides higher PUF and consequently maintains high rates of production for secondary nuclear fuel with nuclear stations working in variableload mode. The introduction of HA at nuclear stations will enable one to fit turbines of power 1.2–1.4 times larger than the reactor power with HA of the water class or by more than a factor 1.5 for PTA, which provides a reliable means of covering peak loads by means of nuclear stations.Translated from Atomnaya Énergiya, Vol. 56, No. 6, pp. 389–396, June, 1984.     

Certain problems in the operation of atomic power stations

The experience acquired in operating the atomic power station of the Central Committee on the Use of Atomic Energy for the Council of Ministries. USSR, is evaluated. It is assumed that the reader is acquainted with the main characteristics of the reactor and the technological layout of the power station. A method of fuel-channel reloading is described in which only parts of the core are reloaded; using this scheme it is possible to achieve a greater burnup of the nuclear fuel (up to 20%) as compared with that which can be achieved when total reloading is used (11%). The partial reloading scheme increases the useful life of the fuel channels and makes it possible to use them more economically. It is shown that using the partial reloading scheme, in an effective day (24 hours of operation at full power) 1.2 channels are consurned whereas 1.7 channels are consumed when complete reloading is used. Furthermore, the use of the partial reloading scheme results in a considerable improvement in the uniformity of the neutron field at varying distances from the center of the core. The various problems which arise in starting up the atomic power station from zero level to the nominal level are described. The change in reactivity due to the temperature rise is found to be k= 0.027±0.003. The feasibility of dispensing with cooling in the side reflector of the reactor, because of the low heat dissipation is discussed. Heat dissipation in the fuel channels following shutdown is also considered. It has been found that the fuel channels can be removed without cooling two hours after the reactor is shut down. The results of a chemical analysis of the coolant water of the primary circuit after various periods of operation are given and it is found that the insignificant amount of washed down contaminants means that operation can be carried on without the use of bidistillate and the condensate from the turbine condenser can be used. Finally, there is a discussion of questions bearing on biological safety and dosimetry. The operation of the station and its technological plant does not constitute a source of danger to the personnel of the station or the population of the surrounding region.The authors wishes to express a debt of gratitude to D. I. Blokhintsev, A. K. Krasin, and N. A. Nikolaev, whose general scientific and technical guidance made the reliable operation of the station possible.     

Some aspects of operation of atomic power stations

Conclusions The question of conditions of operation of AES in energy systems cannot as yet be considered finally resolved. At the present time, while nuclear energy has not yet attained a scale of use commensurate with conventional energy production, only qualitative comments can be made on this question. It is evident that we must wait some time until data on the operation of a large number of AES will allow us to assess the conditions of their operation in a range of specific circumstances with a large degree of precision.However, it is already clear that objective conditions and the nature of electrical energy requirement in energy systems are evidence as to the unavoidability of using AES with thermal reactors in conditions of moderate coefficients of utilization of installed power. Therefore, even in conditions when powerful and economic energy-producing fast neutron reactors will be available, and it will be possible to use them widely to build up electrical generating power, the use of AES with thermal reactors will also prove to be possible and necessary for the reasons given. All this points to the prospect of improving thermal reactors in the distant future.Translated from Atomnaya Énergiya, Vol. 25, No. 5, pp. 387–394, November, 1968.     

Qualification of equipment for nuclear power generating stations

The seismic qualification of equipment is demonstrated by test or analysis or combined test and analysis. Equipment generally have strong coupling of responses in different directions. In the qualification by analysis, the three components of the specified vibratory motions are considered. In the case of qualification by test, the qualification can be incomplete, as generally only two components of the vibratory motion are applied simultaneously. In the tests, concurrent thermal pressure, mechanical and vibratory loads are not applied simultaneously. As the flexibility of the floor can have a significant influence on the response to vibratory motions, such flexibilities should be considered in the development of the floor response spectra, used in the qualification of equipment.     

Desirability of composites for shielding nuclear power stations

All-Union Experimental Physics Research Institute. Translated from Atomnaya Énergiya, Vol. 75, No. 1, pp. 45–48, July, 1993.