Heat transfer of a nuclear reactor containment spray drop

Heat transfer rates to spray droplets under the conditions of a LOCA in a LWR have been evaluated by systematic solution of the governing partial differential equations subject to appropriate initial and boundary conditions. The numerical calculations are based on new correlations. The computations have been facilitated through the use of an efficient hybrid-difference scheme.Results have been provided for the average heat transfer and for the effects of the drop-size, droplet spray angle, initial injection velocity, the containment temperature and pressure on the heat transfer to the drop. The drop fall-heights before attaining thermal equilibrium with the containment atmosphere have been predicted for various conditions. The importance of accurately calculating the drag associated with a moving drop experiencing condensation has been discussed in the context of developing the results.     

Modelling of heat and mass transfer processes during core melt discharge from a reactor pressure vessel

The objective of this paper is to study the heat and mass trasnfer processes related to core melt discharge from a reactor vessel in a light water reactor severe accident. The phenomenology modelled includes the convection in, and heat transfer from, the melt pool in contact with the vessel lower head wall, the fluid dynamics and heat transfer of the melt flow in the growing discharge hole and multi-dimensional heat conduction in the ablating lower head wall. A research programme is underway at the Royal Institute of Technology (Kungliga Tekniska Högskolan, KTH) to (1) identify the dominant heat and mass transfer processes determining the characteristics of the lower head ablation process: (2) develop and validate efficient analytical/computational models for these processes; (3) apply models to assess the character of the melt discharge process in a reactor-scale situation; (4) determine the sensitivity of the melt discharge to structural differences and variations in the in-vessel melt progression scenarios. The paper also presents a comparison with recent results of vessel hole ablation experiments conducted at KTH with a melt simulant.     

Heat transfer in inertial confinement fusion reactor systems

The short time and deposition distance for the energy from inertial fusion products results in local peak power densities on the order of 10¹⁸ W/m³. This paper presents an overview of the various inertial fusion reactor designs which attempt to reduce these peak power intensities and describes the heat transfer considerations for each design.     

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.     

Heat transfer from a totally blocked fuel subassembly of a liquid metal fast breeder reactor: II. Numerical simulation

Fast breeder nuclear reactors used for power generation, have fuel subassemblies in the form of rod bundles enclosed inside tall hexagonal cavities. Each subassembly can be considered as a porous medium with internal heat generation. A three-dimensional analysis is carried out here to estimate the heat transfer due to natural convection, in such an anisotropic, partially heat generating porous medium, which corresponds to the typical case of blocked flow in a fuel subassembly inside the reactor core. Using the finite volume technique, the temperatures at various locations inside hexagonal cavity are obtained. The simulations by the three-dimensional code developed are compared with the results of experiments [Suresh, Ch.S.Y., Sateesh, G., Das, Sarit K., Venkateshan, S.P., Rajan, M., 2004. Heat transfer from a totally blocked fuel subassembly of a liquid metalfast breeder reactor. Part 1: Experimental investigation. Nucl. Eng. Design, present issue] conducted using liquid sodium as the heat transfer fluid. Further, the code is used to predict the maximum temperature in typical liquid metal fast breeder reactors to find the power level where the liquid sodium starts boiling. It helps to decide the power level for initiation of monitoring the temperature for the purpose of reactor control.     

Heat transfer in heat exchangers of sodium cooled fast reactor systems

The present paper describes the heat transfer in heat exchangers of sodium cooled fast reactors. Practical empirical correlations regarding heat transfer coefficients for intermediate heat exchangers (IHXs) and air coolers (ACs) were derived using test data obtained at the fast reactor ‘Monju’ and ‘Joyo’ and also at the 50 MW steam generator facility (50 MW SG). The correlation proposed by Seban and Shimazaki was applicable to estimate the heat transfer coefficients in both flows of IHX, i.e., primary and secondary flows, when the Péclet number was larger than 30. When the Péclet number for shell-side was small, the Nusselt number decreased as a function of the Péclet number. It was clarified that this characteristic was not caused by the heat conduction in flow direction. The heat conduction effect can be neglected even in the natural circulation conditions of the Monju plant. As for the heat transfer coefficient of AC provided in the secondary heat transport system of the fast breeder reactor, data in the above mentioned three facilities were evaluated. As a result, empirical correlations were derived for the average heat transfer coefficients of a large capacity finned air cooler made of stainless steel. These correlations could contribute to analyze the plant dynamics with better accuracy than before.     

Heat transfer from a totally blocked fuel subassembly of a liquid metal fast breeder reactor: Part I. Experimental investigation

The safety issues of liquid metal fast breeder reactors (LMFBR) are crucial due to the fact that a highly reactive and hazardous fluid like liquid sodium is used as coolant. One of the extreme cases, which can occur in a fuel subassembly of an LMFBR, is a total blockage of liquid inside the subassembly, which may lead to boiling of sodium. The present study addresses this problem by conducting experiments on a 19-rod bundle assembly enclosed inside a tall hexagonal enclosure. Liquid sodium is used as the heat transfer fluid. The natural convection mode of heat transfer is the main focus of investigation with a co-flowing air through an annular packed bed to simulate the neighbouring subassemblies. The maximum temperature achieved under different rates of power generations and air flow conditions are observed. Also the radial temperature distributions at different planes under different operating conditions of power and air flow rates have been observed. The results are of significant importance for validating analysis for the purpose of prediction of boiling incipience in an LMFBR subassembly under totally blocked condition.     

An experimental and analytical investigation of uncovered core heat transfer under high pressure, low heat flux conditions

Experimental and analytical results are reported from two series of high pressure core uncovering experiments. It was determined that the uncovered core is cooled primarily by convection and radiation to dry steam and that droplets are confined to the immediate vicinity of the mixture level. Spacer grids substantially increased heat transfer at and downstream of the grid. A simple heat transfer model is presented which accurately predicts uncovered core heat transfer at modified wall Reynolds numbers greater than 2000. Results are expected to be use in modelling small break loss of coolant accidents.     

岭澳核电站计算堆芯功率的热平衡试验分析

介绍了岭澳核电站反应堆功率运行时,为了保证RPN核功率测量系统反应堆堆芯功率测量的正确,利用KME(试验仪表系统)进行热平衡试验测量核反应堆堆芯功率的方法与计算原理,及其与大亚湾核电站试验测量方法的不同点、技术的改进及存在的问题。     

Structural homogenized analysis for a nuclear reactor core

A nuclear reactor core is composed of a great number of tubular beams with periodic structure, which are immersed in an acoustic fluid. In the present paper, a 3-D homogenization model is developed to predict its overall dynamic behavior. An approximate solution to the local problem is given. The application to an 1-D example shows that approximate expressions of the natural frequency, the added fluid mass and the equivalent sound speed can be used in engineering estimation.     

Thermodynamic performance of a gas core fission reactor

In this work we investigated the thermodynamic behaviour of fission products and plutonium as obtained in a gas core fission reactor with graphite walls and operated at 1200 MW thermal power. Equilibrium compositions of the system U-C-F-Pu-fission products were calculated for pressures of 0.1 MPa and 2.5 MPa and temperatures of 1300 K to 10000 K. We found that the reactor can be operated at a pressure of 2.5 MPa and a wall temperature of 2500 K without condensation of any component; no carbides are formed. The main plutonium compound is PuF₄ which, from thermodynamic point of view, can be recycled with UF₄.     

Numerical treatment of pebble contact in the flow and heat transfer analysis of a pebble bed reactor core

This paper studies the numerical treatment of the inter-pebble regions in the modeling of a packed bed geometry for the computational fluid dynamics (CFD) analysis of a pebble bed reactor (PBR) core, where the pebbles are physically in contact with each other. In some studies, the inter-pebble regions have been approximated with gaps, in consideration of the problems on mesh quality or economy of the CFD calculation. To examine such a methodology, a sensitivity analysis for the gap size was conducted with two spherical pebbles, where the inter-pebble region was modeled by means of two kinds of inter-pebble gap and two kinds of direct contact. The cases of direct contact showed numerous differences in the results of the flow regime around the pebbles as well as in the wake, compared to the cases of the inter-pebble gap. No large differences were found between the two cases of direct contact. Based on the result of the sensitivity analysis, the two cases of inter-pebble modeling, i.e., the 1-mm gap and area-contact, were applied to the PBR simulation. It was concluded that the flow regimes and their relevant flow-induced local heat transfer were significantly dependent on the modeling of the inter-pebble region.     

Design of a supercritical water-cooled reactor with a three-pass core arrangement

The Supercritical Water-cooled Reactor (SCWR) is one of the six concepts of the Generation IV International Forum. In Europe, investigations have been integrated into a joint research project, called High Performance Light Water Reactor (HPLWR). Due to the higher heat up within the core and a higher outlet temperature, a significant increase in turbine power and thermal efficiency of the plant can be expected.Besides the higher pressure and higher steam temperature, the design concept of this type of reactor differs significantly from a conventional LWR by a different core concept. In order to achieve the high outlet temperature of over 500 °C, a core with a three-step heat up and intermediate mixing is proposed to keep local cladding temperatures within today's material limits. A design for the reactor pressure vessel (RPV) and the internals has been worked out to incorporate a core arrangement with three passes. All components have been dimensioned following the safety standards of the nuclear safety standards commission in Germany. Additionally, a fuel assembly cluster with head and foot piece has been developed to facilitate the complex flow path for the multi-pass concept. The design of the internals and of the RPV is verified using mechanical or, in the case of large thermal deformations, combined mechanical and thermal stress analyses. Furthermore, the reactor design ensures that the total coolant flow path remains closed against leakage of colder moderator water even in case of large thermal expansions of the components. The design of the RPV and internals is now available for detailed analyses of the core and the reactor.     

IRIS reactor core with thorium fuel

This work is aimed at running the first IRIS reactor core with mixed thorium dioxide fuel（ThO₂-UO₂ and ThO₂-PuO₂）.Calculations are performed by using Dragon 4.0.4 and Citation codes.The results show the multiplication factor(K_eff) for central and peripheral assemblies as a function of burnup.To ensure the proliferation resistance,the value of ²³⁵U enrichment is < 20%.The K_eff is calculated using Dragon 4.0.4 for a single fuel rod and the model developed to fuel assembly,while the whole core was calculated using Citation code.For a fuel burnup,the use of increased enrichment fuel in the IRIS core leads to high reserve of reactivity,which is compensated with an integral fuel burnable absorber.The self-shielding of boron is in an IRIS reactor fuel.The effect of increased enrichment to the burn-up rates,and burnable poison distribution on the reactor performance,are evaluated.The equipment used in traditional light water reactors is evaluated for designing a small unit IRIS reactor.