Helical coils are used for different applications across many different industries thanks to their compactness and resistance to thermal expansion respect straight tubes.
Numerous future nuclear reactors, both in fission (Generation III+ and Generation IV) and fusion projects, presently under development, are expected to use helically coiled pipes for the Steam Generator (HCSG). The improvement of such an important plant component is vital as much as the research for new technological solutions for future nuclear power plants. Indeed, they are designed to reach the goal of improved safety, performance and cost established by the world nuclear community. Numerous favorable characteristics justify the renewed interest for the helical tube SG developing in the nuclear field. In particular, helical tubes provide enhanced heat and mass transfer rates, a higher critical heat flux during boiling and evaporation and a better capability to accommodate the thermal expansion, in addition to allowing a more compact design of the SG.
An experimental campaign could improve the experience on the helical geometry. The idea is to implement a new test section into CIRCE facility in order to investigate the thermal-hydraulic behavior of a scale-down helical coil steam generator. CIRCE is a Lead-Bismuth Eutectic (LBE) pool type facility operative at the ENEA Brasimone research center. The HCSG mock-up will be implemented in the actual test section; the mock-up will be composed of a representative module of the full-scale HCSG, which will characterize the heat transfer through a helical coil geometry. For the design of the experiment and for the future design and safety analysis of DEMO reactor a new RELAP5 version will be developed, capable to analyze in a unique transient simulation all relevant phenomena: two-phase water in horizontal, vertical and helical flow maps, molten salt and liquid metals with appropriate heat transfer coefficient (HTC) and pressure drop correlations.
Our research group supports ENEA in the planning and implementation of an experimental campaign to test the operation of a generic HCSG prototype for liquid metal technologies (PbLi for fusion and LBE/Pb for fission). This activity is planned in synergy with the H2020 EUROfusion project and will require the design and construction of a steam generator mock-up will be installed test in the CIRCE experimental plant located in ENEA Brasimone research center, in order to produce experimental data also relevant for the validation of calculation codes. In addition, this is the opportunity to develop a TH transient code suitable for the participation in the next European calls for the developing of safety in GEN IV liquid metal fast reactors.
The RELAP5 code, the main and more widespread two-phase system code, used also by ENEA for the design, the study of the operation and safety analysis for DEMO, does not provide the possibility to correctly simulate the heat transfer in a helical steam generator. The possibility to obtain a transient simulation code capable to analyze both HCSG, molten salts and liquid metals as working fluids is an important innovation. This instrument, actually not available, will allow to reliably support the design of the experimental test section to be installed in the CIRCE facility, mainly focused on the improve the knowledge in these conditions.
RELAP5-MOD3.3 is a most used code for transient analysis that, in addition to calculating the behavior of a reactor coolant system during a transient, can be used for simulation of a wide variety of hydraulic and thermal transients in both nuclear and non-nuclear systems involving mixtures of vapor, liquid, noncondensable gases, and nonvolatile solute. The code can also be used for space reactor simulations, gas-cooled reactor applications and cardiovascular blood flow simulations.
The programming language used in RELAP5 is mainly FORTRAN 90/95. MIL-STD 1753 extensions are used for the manipulation of bits within "packed words". The source code dimension is about 300000 row and, for this the modification is difficult and an accurate knowledge of the structure is needed.
The best correlation for the prediction of the heat exchange outside the helical pipe heat exchanger or steam generator is Zukauskas [1]. This correlation is currently used for water and will be used also for the design of the WCLL molten salt steam generator. This is not so easy for liquid metals (for DCLL concept and for fission GENIV reactors). As a first tentative evaluation, because no correlation in literature exists in this case, also Zukauskas will be used. But the secondary purpose of the experiment will be the evaluation of this HTC and, after the experiments, this new correlation will be implemented in the code to guarantee a better prediction for the design of the full-scale steam generator. This implementation would be expected after the end of the project because the time needed for the experiment could be too long.
Another step needed for the work is the implementation of the state-of-the-art thermophysical properties of liquid metals and molten salts. For this, the methodology described in [2] for RELAP5-3D properties (similar but not equal to the RELAP5mod3.3 properties) will be used. The fluids that will be implemented are:
1) Lithium Lead eutectic (LiPb);
2) Lead-Bismuth eutectic (LBE);
3) Lead;
4) Molten salt.
The HTC correlations preliminary selected for the implementation are presented in [3]. These are the most used for liquid metals.
The study of this type of steam generator could be possible with these implementations and this innovation is an important step for the Italian activities both in fusion and in fission research fields.
[1] Žukauskas, A., "Heat Transfer from Tubes in Crossflow" in J.P. Harnett and T. F. Invine, Jr., Eds., Advances in Heat Transfer, Vol. 8, Academic Press, New York, 1972.
[2] P. Balestra, F. Giannetti, G. Caruso, and A. Alfonsi, New RELAP5-3D Lead and LBE Thermophysical Properties Implementation for Safety Analysis of Gen IV Reactors, Science and Technology of Nuclear Installations, vol. 2016, Article ID 1687946, 15 pages, 2016
[3] Konstantin Mikityuk, Heat transfer to liquid metal: Review of data and correlations for tube bundles, Nuclear Engineering and Design, 239 (2009) 680-687