Experimental investigation of heat transfer from a circular pipe to lead coolant with controlled oxygen content

The results of experimental investigations of heat transfer from a circular pipe to lead coolant with the oxygen content being controlled and monitored are presented. The heat-transfer investigations are conducted for Peclet numbers 800–3550, Prandtl numbers 0.0123–0.0211, and Reynolds numbers 40,000–190,000 with specific heat flux ~40 kW/m ² and thermodynamically active oxygen content in lead 10^-7 –10⁰ . The experimental dependences of the Nusselt numers on the Prandtl numbers with different oxygen content in the lead coolant are obtained.Translated from Atomnaya Énergiya, Vol. 97, No. 5, pp. 345–349, November, 2004.     

Distribution of coolant flow rate in a steam generator in natural coolant circulation regimes

Conclusions The large hydraulic nonuniformity of steam generator pipes operating in parallel in the natural coolant circulation regime results in a lower efficiency of the heat-transfer surface during emergency cooldown of the reactor plant, and it limits the operational possibilities, specifically, for using this regime at partial power levels. It is obvious that circulation reversal in the pipes of steam generators in the natural circulation regime can have an unfavorable influence on individual structural elements of steam generators as a result of additional temperature stresses appearing in the metal. As one can see from Eq. (6), the conditions of the distribution of the coolant flow rate over pipes in a steam generator can be improved at the design stage. Specifically, they can be realized as an efficient ratio of the “macrogeometric” characteristics of the first loop ΔH and H_sgp as well as by the influence on the ratio of the hydraulic resistance of individual sections of the loop, which determine the numerical value of the parameter m. As m increases, other conditions remaining the same, the character of the distribution of the coolant flow rate in the pipes of a horizontal steam generator improves. Thus, designers of a nuclear power plant have ways to search for optimal solutions. It is obvious that the interrelations of the conditions of operation of a steam generator, examined above, and the natural circulation in the loop require that the distribution of the flow rate in a pipe bundle be taken into account in the physical simulation using special thermohydraulic stands. St. Petersburg State Technical University. Translated from Atomnaya énergiya, Vol. 83, No. 3, pp. 169–174, September, 1997.     

Gas-transfer kinetics and regulation of the gas content of coolant in the first loop of a water-cooled-water-moderated reactor

OKBM. Translated from Atomnaya Énergiya, Vol. 72, No. 6, pp. 554-559, June, 1992.     

Vibration analysis of heat exchanger and steam generator designs

A thorough flow-induced vibration analysis of nuclear components such as heat exchangers and steam generators is essential at the design stage to ensure good performance and reliability. This paper presents our approach and techniques in this respect. In a steam generator, for example, the flow may be liquid or two-phase. In general, parallel and cross-flow exist in the tube bundles of heat exchange components. In cross-flow three basic vibration excitation mechanisms are considered, namely fluidelastic instability, periodic wake shedding resonance, and forced response to random flow turbulence. The latter may need to be considered in parallel flow. These vibration excitation mechanisms and the dynamics of multispan tubes are formulated in a computer model which is used to predict the vibration response of the tubes. The computer model and the parameters required to formulate the vibration excitation mechanism are discussed. Examples of vibration analysis of steam generators and heat exchangers are outlined. It is concluded that most flow-induced vibration problems may be avoided by proper analysis at the design stage.     

Evaluation of a steam generator tube rupture accident in an accelerator driven system with lead cooling

A postulated steam generator tube rupture (SGTR) accident in a lead cooled accelerator driven transmuter (ADT) is investigated. The design of the ADT without intermediate loops bears the risk of water/steam blasting into the primary coolant. As a consequence a nuclear power excursion could be triggered by steam ingress into the ADT core which has a significant positive void worth. A thermal coolant–coolant interaction (CCI) might initiate a local core voiding too and additionally could lead to sloshing of the lead pool with mechanical impact of the heavy liquid on structures. The steam formation will also lead to a pressurization of the cover gas. The problems related to an SGTR are identified and investigated with the SIMMER-III accident code.     

Prediction of the coolant temperature field in a breeder reactor including interassembly heat transfer

A general procedure for investigating the effect of interassembly heat transfer on the temperature field of a breeder reactor is provided. This procedure utilizes a simple multi-assembly code SUPERENERGY and a set of normalized assembly maps derived from this code. This procedure has been applied to a typical liquid metal fast breeder reactor (LMFBR) to study the effect of various concentric rings of coupled assemblies on the temperature prediction of the central assembly. Insignificant perturbation was found from second ring assemblies, while important perturbation was found from first ring assemblies.     

Particulars of radiolysis and mass transfer in two-phase water coolant of a nuclear power facility with steam generation at an interphase surface

A model of an interphase transfer of stable products of the radiolysis of water in boiling coolant is developed taking account of the intensity of their delivery to the interphase boundary in the liquid phase and removal into the vapor phase with vapor generation on the interphase surface. A computational study is made of the radiolysis of the coolant and interphase transfer of the products of the radiolysis of water in the core and on the pulling section of BWR of the Oskarshamn-2 nuclear power plant in Sweden. A comparison of the computational data with the results of the technical measurements of the coolant composition of the BWR at the Oskarshamn-2 nuclear power plant showed that the accumulation of stable products of the radiolysis of water in the vapor-gas phase of the coolant is determined by the kinetics of radiolysis in the liquid phase, the concentration of the oxygen-containing components in the liquid phase is due to the present of hydrogen peroxide in it.     

Effective heat conductivity of fuel element bundles and steam generator tube bundles

Effective heat conductivity of rod and tube bundles is one of thermophysical properties necessary for calculation of thermo hydraulic characteristics of heat producing devices, heat exchange devices and steam generators. This report introduces results of mathematical modeling of effective heat conductivity of transversally anisotropic rod bundles in solid conductive medium. The considered bundles represented cylindrical rods fitted in corners of stretched and compressed in direction of heat transfer rectangular and triangular grids. The calculated results were compared to analytical solutions and previous numerical results.     

Experimental and empirical study of steam condensation heat transfer with a noncondensable gas in a small-diameter vertical tube

An experimental study was performed to investigate local condensation heat transfer coefficients in the presence of a noncondensable gas inside a vertical tube. The data obtained from pure steam and steam/nitrogen mixture condensation experiments were compared to study the effects of noncondensable nitrogen gas on the annular film condensation phenomena. The condenser tube had a relatively small inner diameter of 13 mm (about 1/2-in.). The experimental results demonstrated that the local heat transfer coefficients increased as the inlet steam flow rate increased and the inlet nitrogen gas mass fraction decreased. The results obtained using pure steam and a steam/nitrogen mixture with a low inlet nitrogen gas mass fraction were similar. Therefore, the effects of noncondensable gas on steam condensation were weak in small-diameter condenser tubes.A new correlation was developed to evaluate the condensation heat transfer coefficient inside a vertical tube with noncondensable gas, irrespective of the condenser tube diameter. The new correlation proposed herein is capable of predicting heat transfer rates for tube diameters between 1/2- and 2-in. because of the unique approach of accounting for the heat transfer enhancement via an interfacial shear stress factor.     

Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

A severe accident has inherently significant uncertainties due to the complex phenomena and wide range of conditions. Because of its high temperature and pressure, performing experimental validation and practical application are extremely difficult. With these difficulties, there has been few experimental researches performed and there is no plant-specific experimental data. Instead, computer codes have been developed to simulate the accident and have been used conservative assumptions and margins. This study is an effort to reduce the uncertainty in the probabilistic safety assessment and produce a realistic and physical-based failure probability. The methodology was developed and applied to the OPR1000. The creep rupture failure probabilities of reactor coolant system (RCS) components were evaluated under a station blackout severe accident with all powers lost and no recovery of steam generator auxiliary feed-water. The MELCOR 1.8.6 code was used to obtain the plant-specific pressure and temperature history of each part of the RCS and the creep rupture failure times were calculated by the rate-dependent creep rupture model with the plant-specific data.     

Pressure wave propagation in an LWR downcomer and resultant load on the core shroud due to rupture of a primary coolant loop

Pressure differences and the resultant dynamic load act on the core shroud when pressure waves propagate in the downcomer of a light water reactor (LWR) pressure vessel after rupture of the primary pipe has occurred. An equivalent geometry, i.e. a diverging duct is used to solve by Euler and wave equation for acceleration and velocity of the fluid behind the wave front, that the two-dimensional, time-dependent pressure distribution, induced by the wave propagation, can be calculated. The assumptions lead to an approximate but conservative value of the resultant core shroud load.     

Structure mechanics and thermal-hydraulic behaviour of a PWR coolant loop with a postulated crack like leakage

Thermal hydraulic calculations, using ATHLET, have been used to evaluate pressures and temperatures in primary and secondary circuits, following a postulated leak in the surge line. These were, then, used as input for the structural mechanics calculations with ADINA. The main results of the analyses may be summarised as follows; global deformations are significantly reduced by the drop in pressure and decrease in temperature (e.g. the vertical displacement of the surge line in the region of the crack is reduced by 50%); the leakage area at the end of the transient is about 30% lower compared with the value at the beginning; the leak rate is slightly increased at the end of the transient in comparison with the initial rate; the consideration of the crack surface pressure is important, as it leads to a significant increase in crack load and leakage area by about 50 and 35%, respectively.