Coolant stratification in nuclear power facilities

Computational and experimental studies of the structure of the motion of stratified coolant and its temperature regime are performed on models of different elements of the circulation loops of fast and thermal reactors. The investigations have shown that in the presence of stable stratification thermogravitational forces bring about the formation of stagnant and recirculation formations with large temperature gradients and pulsations at interfaces. The data obtained show that stratification phenomena must be taken into account when validating the reliability of the control, safety, and nominal service lives of nuclear power facilities.     

Transportable nuclear power facilities in the INPRO International Project

Questions concerning safety, nonproliferation, monitoring of nuclear materials, civilian responsibility for nuclear risks, physical protection, transport operations, and others are analyzed within the framework of the INPRO project in application to transportable nuclear energy facilities. Essentially, the operative nuclear law and the experience of world nuclear power make it possible to solve the problems of the legal basis for the life cycle of transportable nuclear power facilities. To attain a system with the optimal accessibility, effectiveness, and safety, the nuclear power facilities will have to be adapted to the new specific requirements and conditions, and the international legal basis will have to be made more precise.     

Engineering of nuclear power facilities for earthquake loads

The state-of-knowledge to engineer nuclear power facilities for earthquake loads is reviewed as it was collectively presented at the fourth SMiRT Conference. All aspects of the design process is critically examined starting with the definition of ground motion. Both past achievements in each of the several areas of endeavor, and the gaps in our knowledge that need further research and study are emphasized. Several alternatives to above ground facilities are reviewed, and issues are raised regarding easy solutions to very complex problems associated with these alternatives. Some questions that must be answered when, and if, earthquakes are predicted, both with regards to existing and to future plants, are raised. It is concluded that major strides have been taken during the last decade, and the future holds many challenges to the profession.     

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     

An investigation of the strength of branch areas in the piping of nuclear power installations

A three-dimensional shell finite-element method and a computer code have been developed in RRC ‘Kurchatov Institute’ for the strength analysis of branch pipe (nozzle) areas, which are among the most critical as regards strength components of pipes in both reactors and other systems. According to the method, the structure under study is broken up into eight-node isoparametric shell-type finite elements, each with eight or 18 integration points. The program was tested with the help of experimental results. The strength analyses of equal and unequal T-pipes were performed. Consideration was given to the variants with loading by pressure and moments (symmetric and asymmetric cases). The effects of fillet wall thickness and radius on the stressed state of the T-pipes were investigated.     

The impact of TMI upon the public acceptance of nuclear power

This paper discusses long-term trends in public attitudes toward nuclear power, focusing on the extent to which the accident at Three Mile Island appears to have affected public acceptance of nuclear energy. Public attitudes towards other energy production options also are considered, particularly in terms of changes that may be related to TMI. Finally, the relationships between attitudes toward nuclear power and perceptions of broader energy, environmental and social issues are examined. The data used in this analysis are from national surveys conducted by major national opinion research organizations from the early 1970s through 1981.

There is considerable evidence that TMI has had a significant impact on public acceptance of nuclear power, in the direction of increasing opposition to and decreasing support for construction of new nuclear power plants. TMI appears to have increased the rates of decline in support and rise in opposition to local construction of nuclear power plants, although a trend of decreasing public acceptance of such local construction had been in evidence since the mid-1970s, prior to TMI. In spite of this decline in public acceptance of new construction, there is substantial support for both completing nuclear power plants currently under construction and for the continued operation of existing plants.     

High-temperature heat pipes for power modules of space nuclear power systems

The results of an experimental analysis of the technology for producing high-temperature heat pipes and choosing their structure so as to ensure a long service life and high performance characteristics required for the power modules of space nuclear power systems with an out-of-core energy conversion system are examined. It is estimated that the operation of molybdenum-lithium heat pipes, fabricated using the technology developed, for the power modules of such nuclear power systems will be stable for >10⁵ h. 2 figures, 1 table, 13 references. State Enterprise Krasnaya Zvezda. Translated from Atomnaya énergiya, Vol. 89, No. 1, pp. 82–86, July, 2000.     

Active seismic response control systems for nuclear power plant equipment facilities

To sustain severe earthquake ground motion, a new type of anti-seismic structure is proposed, called a Dynamic Intelligent Building (DIB) system, which is positioned as an active seismic response controlled the structure. The structural concept starts from a new recognition of earthquake ground motion, and the structural natural frequency is actively adjusted to avoid resonant vibration, and similarly the external counter-force cancels the resonant force which comes from the dynamic structural motion energy. These concepts are verified using an analytical simulator program. The advanced application of the DIB system, is the Active Supporting system and the Active Stabilizer system for nuclear power plant equipment facilities.     

Numerical investigation of acceleration effect on heat transfer deterioration phenomenon in supercritical water

It is important to understand the heat transfer deterioration (HTD) phenomenon for specifying cladding temperature limits in the fuel assembly design of supercritical water-cooled reactor (SCWR). In this study, a numerical investigation of heat transfer in supercritical water flowing through vertical tube with high mass flux and high heat flux is performed by using six low-Reynolds number turbulence models. The capabilities of the addressed models in predicting the observed phenomena of experimental study are shortly analyzed. Mechanisms of the effect of flow structures and fluid properties on heat transfer deterioration phenomenon are also discussed. Numerical results have shown that the turbulence is significantly suppressed when the large-property-variation region spreads to the buffer layer near the wall region, resulting in heat transfer deterioration phenomenon. The property variations of dynamic viscosity and specific heat capacity in supercritical water can impair the deterioration in heat transfer, while the decrease of thermal conductivity contributes to the deterioration.     

Seismic reevaluation and upgrading of nuclear power facilities outside the US using US developed methodologies

Several seismic licensing and safety issues have emerged over the past fifteen years for commercial US Nuclear Power Plants and US Government research reactors, production reactors and process facilities. The methodologies for the resolution of these issues have been developed in numerous government and utility sponsored research programs. The resolution criteria have included conservative deterministic design criteria, deterministic seismic margins assessments criteria (SMA) and seismic probabilistic risk assessment criteria (SPRA). The criteria for SMAs and SPRAs have been based realistically on considering the inelastic energy absorption capability of ductile structures, equipment and piping and have incorporated the use of earthquake and testing experience to evaluate the operability of complex mechanical and electrical equipment. Most of the applications to date have been confined to the US, however there have been several applications to Asian, Western and Eastern Europe reactors. This paper summarizes the major issues addressed, the development of reevaluation criteria and selected applications to non US reactors including VVER reactors of Soviet origin.