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.     

Increasing critical heat flows in VVER-1000 model fuel assemblies by inserting additional spacing lattices

One way to increase the margin to critical heat transfer is investigated – closer lattice spacing on the axial section of a fuel assembly with the minimum margin to critical heat emission. Experiments were performed on the KS facility at the Russian Science Center Kurchatov Institute on two 19-rod VVER-1000 fuel-assembly models with lattice spacings 340, 255, and 170 mm. For coolant mass flow rate 4000 kg/(m²·sec) and relative enthalpy 0–0.1, the critical flow rate increases by 15–20% because the lattice spacing decreases from 340 to 170 mm. As the mass flow velocity and relative enthalpy of the coolant decrease, the critical heat flows do not increase as much. A generalizing relation is obtained for the gain in the critical heat flux as a function of the distance to a spacing lattice and the main regime parameters.     

Experimental validation of the cooling loop for a passive system for removing heat from the AES-2006 protective envelope design for the Leningradskaya nuclear power plant site

Equipping new-generation nuclear power plants with passive means for controlling unanticipated accidents is one of the most promising directions for increasing safety, which is being implemented in the AES-2006 design for the site of the Leningradskaya nuclear power plant. An urgent problem is to obtain experimental validation of the passive system for removing heat from the protective envelope during unanticipated accidents with loss of coolant from the first loop in the case where the active systems fail. A particularity of the system is its state of constant readiness. The system functions with natural circulation of the coolant in both loops. Considering the importance of the passive heat removal system for ensuring the localizing properties of the protective envelope, OKBM Afrikantov has developed a large-scale stand and performed experimental investigations on validation of the effectiveness and serviceability of the cooling loop of a passive system for removing heat from the protective envelope. Translated from Atomnaya énergiya, Vol. 106, No. 3, pp. 148–152, March, 2009.     

Experemental investigations of transcritical heat transfer on an electrically heated model of a fuel assembly

The purpose of this work was to obtain experimental information on the temperature state of fuel elements in the transcritical region and to develop methods for calculating the heat-emission coefficient. The following region of regime parameters was investigated: pressure 11.7–16.7 MPa, temperature at the entrance into the fuel-assembly model 200–285°C, mass velocity 700–1900 kg/(m²·sec). The existence of a definite transcritical power reserve was confirmed experimentally. The results of the investigations showed that in the experimental range of regime parameters the coolant temperature at the entrance into the fuel-assembly model has the strongest effect on the transcritical reserve. As temperature increases, the transcritical reserve increases. A method for calculating the heat-emission coefficient in the transcritical region was developed on the basis of the experimental data obtained, 5 figures. 2 references. Translated from Atomnaya énergiya, Vol. 88. No. 4, pp. 257–260, April, 2000. Original article submitted July 8, 1998: resubmitted December 15, 1999.     

Experimental Investigations of the Thermohydraulic Parameters of the Flow Section of a Liquid-Metal Target with a Window for a Proton Beam from an Accelerator

The results of computational and experimental investigations of the thermohydraulic characteristics of a liquid-metal target with a tight-fitting stopper, whose shape ensures a constant energy-release volume in the stopper material, are examined. The investigations on water circulation stands included flow visualization and measurement of the velocity and pressure distributions in the flow part of the target structure. The investigations on the liquid-metal circulation stand with lead-bismuth coolant were performed with coolant working temperature 300°C and maximum flow rate up to 7 m³/h. The temperature and the temperature pulsations in the coolant and in the material of the tight-fitting stopper were determined.     

Some Results of an Investigation and Prospects for the Development of a High-Temperature Power Reactor with a Solid Coolant

A validation of the possibility of developing and the basic advantages of a high-temperature nuclear reactor where the first-loop coolant is a solid are presented. The basic requirements for a solid coolant are formulated, a technology for fabricating spherical graphite particles by gas-phase pyrolytic deposition is developed, and three experimental batches are prepared. The experimental facilities for investigating the motion and heat transfer, including coolant flow stability, heat exchange, and durability, are described. The results of a determination of the heat-emission coefficient during the flow of the solid coolant in a 10-mm in diameter circular channel with warming-wall temperatures in the range 373–1073 K and flow velocities 0.1–0.22 m/sec in vacuum, argon, and helium are presented. The requirements for a 500-kW bench model, on which the basic parameters of the nuclear power system with a solid coolant are to be obtained, are formulated. __________ Translated from Atomnaya Energiya, Vol. 99, No. 5, pp. 358–365, November, 2005.     

Mass Transfer of Lead Coolant Vapors in a BREST-OD-300 Reactor

The goal of the work reflected in this paper is to investigate the mass transfer of lead from the coolant surface in the gaseous cavities of the cooling loop of a BREST-OD-300 reactor. The theoretical and experimental investigations show that the amount of lead evaporating from the free surfaces of the first loop of the reactor is negligible.The results of this work are the recommendations that filters should be placed at locations where the pipelines of the gas system, located outside the reactor block, are connected to the reactor cover, because of the possibility of aerosols – dust-like impurities from other sources with mass flow several times greater than the rate of evaporation of lead in the unit – entering the gas system. 2 figures, 7 references.     

Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for an Advanced Integral Type Reactor

Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed by using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and the auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and the loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and the loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily into the PRHRS loop and that the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable a natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with an operation of the PRHRS.     

Methods for protecting fuel-element cladding from corrosion in lead coolant and the heat-transfer sublayer

The results of investigations of the corrosion of commercial and experimental steels in lead and the possibilities of corrosion protection are presented. The effect of lead coolant and the lead heat-transfer sublayer on fuel-element cladding are examined. Methods based on thermodynamic calculations and experimental data are proposed for protecting fuel element cladding in a lead-cooled reactor from the corrosive effect of the coolant by creating a new corrosion resistant chromium steel and from the corrosive effect of the heat-transfer sublayer by alloying with the components of steel. The results of this work have been implemented in the experimental fuel elements for the BREST-OD-300 reactor which were irradiated in a BOR-60 reactor. __________ Translated from Atomnaya énergiya, Vol. 104, No. 2, pp. 88–94, February, 2008.     

Heat transfer during boiling of a liquid metal during emergency cooldown of a fast-neutron reactor

The physics of the processes, the characteristics, and the stability of different regimes, of boiling (nucleate, projectile, disperse-ring), which are observed in experiments investigating the boiling of liquid-metal coolant in a model of a fuel assembly for a fast-neutron reactor in the emergency cooldown regime with low circulation velocity, are analyzed. The experimental setup, the, methods for performing measurements, and the experimental data on the boiling of a liquid metal are described. A mathematical model of the process of boiling of a liquid-metal, coolant in a natural-circulation loop is described, and the results of test calculations for regimes with an increase in heating and with sharp pressure drop are prresented. 7 figures, 12 references. State Science Center of the Russian Federration–A. I. Leipunskii Physics and Power Engineering Institute. Translated from Atomnaya énergiya, Vol. 87, No. 5, pp. 337–342, November, 1999.     

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.     

Fundamental Experiment of Potassium Heat Exchanger Using Principle of Heat Pipe

Abstract

In order to provide compact and reliable sodium equipments including a steam generator, performance tests are conducted with a potassium heat exchanger, which is featured by the separate construction of primary and secondary coolant systems. A small amount of potassium plays a role as an intermediate media of heat transportation between these two coolant systems. Heat is transfered by evaporation and condensation of potassium on the surfaces of the primary and the secondary coolant pipings, respectively. The tests are performed in the temperature range of 200-300°C and the maximum heat transfer reaches 1.3 kW (heat transfer rate at the primary heating source: 8.6 W/cm² at 300°C). The experimental results are analyzed by using Langmuir's and Schrage's equations and close agreement between experiment and theory is obtained.     

Evaluation of the stability of protective coatings on microfuel in a vapor-gas medium with interaction with structural materials

The serviceability of unirradiated microfuel with protective silicon carbide cladding in contact with corrosion-resistant 08Kh18N10T and EI-847 (05Kh15N16M3B) steels in a vapor-gas medium at 1100–1450°C has been investigated under bench conditions in application to emergency operating regimes of the core of a light-water reactor with loss of coolant in the first loop. It has been established that the cladding of microfuel tested at ~1200°C for 2–4 h in a medium consisting of the products of combustion of propane in oxygen in contact with austenitic steel remained completely sealed and whole. Under these conditions, this temperature is the maximum admissible temperature for microfuel with a protective silicon carbide outer layer. Under the conditions studied at 1300–1450°C, the 08Kh18N10T and EI-847 steels possess the same and comparatively low corrosion resistance. The corrosion depth over 2 h exceeds 0.5 mm or they melt. At 1100–1210°C the corrosion depth in these steels lies in the range 0.15–0.2 mm. Translated from Atomnaya énergiya, Vol. 106, No. 3, pp. 153–158, March, 2009.     

Steam-content and coolant-level emission sensors for a nuclear reactor

Fast sensors for the coolant steam content and level are examined. These sensors are based on the detection of the attenuation of the β radiation passing through a coolant layer. The β-particle emitter is a material containing ⁹⁰Sr and the β-particle detector is a cable with magnesia (MgO) insulation. The current signals as a function of the water level and the average density of the water-air (steam) mixture are obtained using models of the coolant-level and steam-content sensors using β emitters in the form of pellets and cylinders. The density of the air-water mixture was changed by feeding air into the sensitive volume of the steam-content sensor. It is noted that one of the main advantages of the β-emission steam-content and coolant-level sensors is their zero time constant and small size and, for measurements of the steam content of the coolant, the fact that the indications are independent of the temperature and pressure of the medium and the γ-ray dose rate. In addition, the sensors can operate at any point of the reactor vessel or loop. __________ Translated from Atomnaya énergiya, Vol. 101, No. 3, pp. 197–203, September, 2006.     

Natural circulation experiments in a non-uniform diameter lead bismuth loop and validation of LeBENC code

Experimental studies are carried out on natural circulation in a Lead Bismuth Eutectic (LBE) loop. The loop mainly consists of a heated section, air heat exchanger, valves, various tanks and argon gas control system. All the components and piping are made of SS316L. The dissolved oxygen in the LBE is monitored online by an Yttria Stabilised Zirconia (YSZ) oxygen sensor and controlled during the operation of the loop. In this paper the details of the loop and experimental studies carried out with heater power levels varying from 900 W to 5000 W are described. The temperature range of LBE during the experiments was 200 °C–500 °C. The maximum heat loss in the piping is kept less than 20% of the main heater power. Steady state experimental studies are carried out at different power levels and the LBE flow rate was found to be varying from 0.095 kg/s to 0.135 kg/s. The analysis and results of the performance of the heat exchanger with air and water as the secondary coolants are also discussed in the paper. Transient studies were carried out to simulate various events like heat sink loss, step power change and secondary side coolant flow rate change and reported in the paper. In the start up experiments, where the flow is started from stagnant condition of LBE, the time required for starting of natural circulation is found to be 600 s, 400 s and 240 s with power level of 1200 W, 2400 W and 3000 W respectively. The results are compared with available correlation and prediction of computer code LeBENC.     

CFD analysis of a regular sector of the ITER vacuum vessel. Part II: Thermal-hydraulic effects of the nuclear heat load

The 3D Computational Fluid Dynamic (CFD) steady state analysis of the regular sector #5 of the ITER vacuum vessel (VV) is presented in these two companion papers using the commercial software ANSYS-FLUENT®. The pure hydraulic analysis, concentrating on flow field and pressure drop, is presented in Part I. This Part II focuses on the thermal-hydraulic analysis of the effects of the nuclear heat load. Being the VV classified as safety important component, an accurate thermal-hydraulic analysis is mandatory to assess the capability of the water coolant to adequately remove the nuclear heat load on the VV. Based on the recent re-evaluation of the nuclear heat load, the steady state conjugate heat transfer problem is solved in both the solid and fluid domains. Hot spots turn out to be located on the surface of the inter-modular keys and blanket support housings, with the computed peak temperature in the sector reaching ～290 °C. The computed temperature of the wetted surfaces is well below the coolant saturation temperature and the temperature increase of the water coolant at the outlet of the sector is of only a few °C. In the high nuclear heat load regions the computed heat transfer coefficient typically stays above the 500 W/m² K target.     

Characteristics of electronuclear targets containing water

Conclusions The replacement of liquid-metal lead or sodium coolant with water has virtually no effect on the intensity of the neutron flux in the blanket of an electronuclear installation and the formation of²³⁹Pu nuclei. The accumulation of²³⁹Pu nuclei is accompanied by a rapid growth of the heat release in the target and a decrease of the production rate. On account of the block effect, the heterogeneity of the target results in an appreciable increase of the neutron flux and breeding. At the same time, when a hydrogen-containing target is used, the heat release increases considerably. Taking into account the chemical bond in the water molecules and the energy dependence of the neutron cross sections in the thermal region has virtually no effect on the characteristics of an electronuclear installation, and they can be neglected in calculations. Joint Institute of Nuclear Research. Translated from Atomnaya énergiya, Vol. 77, No. 6, pp. 419–424, December, 1994.     

Behavior of Radioactive Iodine Isotopes in the Multiple Forced Circulation Loop in an RBMK Reactor

The flow of iodine, including ¹³¹I, into the coolant water in a nuclear power plant with an RBMK-1000 reactor under normal operating conditions and during transient regimes is analyzed. It is shown that under normal operating conditions the specific activity of ¹³¹I in the coolant is correlated with the iron concentration. During shutdown, its content increases by factors of 30–200. The emission of ¹³¹I into the coolant can be decreased by factors of 10–15 and the degree of unsealing of fuel elements can be decreased if before shutdown the reactor is held for 2–5 days at 50% of the nominal power level. Recommendations are made for decreasing ¹³¹I emissions into the atmosphere. The adoption of these recommendations at the Leningrad nuclear power plant has reduced the ¹³¹I emissions into the atomsphere by a factor of 17. __________ Translated from Atomnaya Energiya, Vol. 99, No. 2, pp. 103–108, August 2005.     

Evaluation of Heat Transfer Augmentation in a Nanofluid-Cooled Microchannel Heat Sink

Present investigation deals with appraising heat transfer enhancement of single phase microchannel heat sink (MCHS) by ultra fine Cu particle incorporation in base coolant fluid. The particle diameter is of nanometer size and base fluid in combination of nanoparticles is called nanofluid. Governing equations for fluid flow and heat transfer are based on well established “porous medium model” and accordingly, modified Darcy equation and two-equation model are employed. Appropriate equations for both fluid flow and heat transfer are derived and cast into dimensionless form. Velocity profile is obtained analytically and in order to solve conjugate heat transfer problem a combined analytical–numerical approach is employed. For heat transfer analysis, thermal dispersion model is adopted and latest proposed model for effective thermal conductivity – which considers the salient effect of interfacial shells between particles and base fluid – is integrated into model. The effects of dispersed particles concentration, thermal dispersion coefficient and Reynolds number are investigated on thermal fields and on thermal performance of MCHS. Additionally, the impact of turbulent heat transfer on heat transfer enhancement is considered.     

Effects of Sputtering of and Radiation by Aluminum on Magnetized Target Fusion Plasmas

We estimate numerically the rate of radiation by aluminum impurities for parameters relevant to magnetized target fusion (MTF) plasmas. We demonstrate that the coronal equilibrium is appropriate for expected MTF plasma parameters. Using the coronal equilibrium, we estimate the power radiated per impurity ion is 0.25–0.5 × 10⁻¹⁶ MW for temperatures and densities relevant to present plasma parameters taken from the FRX-L experiment at Los Alamos National Laboratory and is approximately 75.0 × 10⁻¹⁶ MW for temperatures and densities relevant to anticipated MTF plasmas. We calculate the sputtering rate of aluminum by thermal deuterium and tritium plasma ions is a few percent assuming an impact angle of 45°. Finally, we estimate that with aluminum impurity levels of a few percent, the impurity radiation power density would be approximately 25 kW/cm³ for FRX-L conditions and 2.5 GW/cm³ for anticipated conditions in a MTF plasma. While we have assumed a sputtering model of impurity generation, the results for the power density apply for impurity levels of a few percent, regardless of the generation mechanism.