The complex set of physical phenomena that occur in a gravity environment when a geometrically distinct heat sink and heat source are connected by a fluid flow path can be identified as natural circulation (NC). No external sources of mechanical energy for the fluid motion are involved when NC is established. In a number of publications, including textbooks, the term natural convection is used as a synonym of NC. Within the present context, natural convection is used to identify the phenomena that occur when a heat source is put in contact with a fluid. Therefore, natural convection characterizes a heat transfer regime that constitutes a subset of NC phenomena.

This report provides the presented papers and summarizes the discussions at an IAEA Technical Committee Meeting (TCM) on Natural Circulation Data and Methods for Innovative Nuclear Power Plant Design. While the planned scope of the TCM involved all types of reactor designs (light water reactors, heavy water reactors, gas-cooled reactors and liquid metal-cooled reactors), the meeting participants and papers addressed only light water reactors (LWRs) and heavy water reactors (HWRs). Furthermore, the papers and discussion addressed both evolutionary and innovative water cooled reactors, as defined by the IAEA[1].

NC principles are of fundamental interest to the designers of nuclear power plant systems and components. Making reference to the existing water cooled reactors, the consideration of NC is brought to the design of the layout of the primary circuit. The core is located at a lower elevation with respect to the steam generators and the feed-water inlet location, in the cases of pressurized and boiling water reactors, respectively. In all of the adopted geometrical configurations, NC allows the removal of the decay heat produced by the core, should the forced circulation driven by centrifugal pumps become unavailable. Furthermore, NC is the working mode for the secondary side of most steam generators in existing pressurized heavy and light water reactors. It is essential as well for the core cooling in the unlikely event of loss of primary coolant.

Reactors based on natural circulation during normal operation (e. g. the Dodewaard Reactor in the Netherlands and the VK-50 in Russia) operated for an extended period of time. Most boiling water reactors can operate in the natural circulation mode for power levels below about 40 per cent of full power. Some newly developed designs are based on natural circulation core cooling for normal operation and on the use of the natural convection heat transfer for some safety systems. Reliance on natural circulation can result in simplified systems, reduced costs and — most importantly — a very high safety level.

The accomplishment of the objectives of achieving a high safety level and reducing the cost through the reliance on NC mechanisms, requires a thorough understanding of those mechanisms. Natural circulation systems are usually characterized by smaller driving forces with respect to the systems that use an external source of energy for the fluid motion. For instance, pressure drops caused by vertical bends and siphons in a given piping system, or heat losses to environment are a secondary design consideration when a pump is installed and drives the flow. On the contrary, a significant influence upon the overall system performance may be expected due to the same pressure drops and thermal power release to the environment when natural circulation produces the coolant flow. Therefore, the level of knowledge for the thermal-hydraulic phenomena for the specific geometric conditions and governing heat transfer conditions should be deeper when NC is involved. In addition, the lower driving forces for natural circulation systems might lead to quite large equipment for which the role of 3D phenomena is essentially increased.

Within nuclear technology the renewed interest in NC is a consequence of the above, in combination with the potential for cost savings from increased use of NC mechanisms in plant designs. Relevant experiments directed to the characterization of NC have been carried out in the past because of the importance of the related mechanisms for the safety of existing reactors. Similarly, thermal-hydraulic system codes have been qualified through the comparison of predicted results and experimental data. The quality of recorded experimental data and the precision level of the available system codes, or the expected uncertainty in these predictions, are generally evaluated as satisfactory for the needs of the current reactors. However, the exigencies posed by the more extensive use of the NC in the design of evolutionary and innovative water cooled reactors require a re-evaluation of the experimental data and of the code capabilities considering the new phenomena and conditions involved.

Recent activities completed under the IAEA umbrella, e. g. Refs [1-8], and by other institutions, such as the U. S. Nuclear Regulatory Commission (U. S. NRC) [9], the OECD/NEA/CSNI (Organization for Economic Cooperation and Development/Nuclear Energy Agency/Committee on the Safety of Nuclear Installations), Refs [10-13], and the EC (European Commission), Refs [14-16], show the importance of the subject and constitute the basis for future activities in the area of NC. Potential future international activities could be directed toward:

(a) identification of still unresolved issues from evolutionary and innovative water reactor designs, and

(b) enhancing the quality levels of the available computational tools and experimental databases in relation to design needs.

This report provides an overview of the current state of the art of natural circulation data and methods, and discusses potential benefits of an integrated future effort directed toward the achievement of the two aforementioned objectives. The main attention in this report is paid to the design basis accident phenomena; severe accident issues are considered briefly in their relation to the protection of the containment as the last safety barrier.

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