Thermal hydraulics, fast reactors

Y. S. Tang. Ph.D has more than 35 years of experience in the field of thermal and fluid flow. His research interests have covered aspects of thermal hydraulics that are related to conventional and nonconventional power generation systems, with an emphasis on nuclear reactor design and analysis that focuses on liquld-meta -cooled reactors. Dr. Tang is co-author of Radioactive Waste Management published by Taylor 8 Francis, and Thermal Analysis of Liquid Metal Fast Breeder Reactors, He received a B5. from National Central University In China and an MS. in mechanical engineering from the University of Wisconsin. He earned his Ph.D. 

Determination of the loads through analysis of the reactor operating parameters combined with thermal hydraulics calculations for the specific thermal stratification and fluctuation areas. For this purpose, the recent R D progress made in the EFR (European Fast Reactor) project was used and transposed to a real installation. 

R D items have been (hscussed. The items are classified from two viewpoints. The first is on importance A) items essential for safety design, B) items effective on economy and C) items of desirable demonstration tests. The second is dependency on structure 1) items peculiar to the double pool structure, 2) items peculiar to small reactors and/or modular reactors and 3) items common to all LMRs. Typical items are a thermal-hydraulic test with a scale model (A-1), and the fast pressure release system and demonstration experiment for beyond design basis leak events (B-1). Licensing will be started after completion of detailed design. 

XXII-2] SIENICKI, J.J., MOISSEYTSEV, A.V., SSTAR lead-cooled, small modular fast reactor for deployment at remote sites - system thermal hydraulic development, ICAPP 2005, Paper 5426 (Int. Conf on Advances in Nuclear Power Plants Seoul, May 15-19, 2005). 

Investigation of safety-related physical and thermal-hydraulic characteristics of fast reactor cores. 

Papin, J., Stansfield, R., Thermal-Hydraulic Behaviour of a Fast Breeder Reactor Subassembly During an Undercooling Accident the PHYSURA-GRAPPE Code and its Validations on Scarabee Experiments, NURETH-4, Karlsruhe, Germany, 10-13 October, 1989, FRG. 

Menant, et. al. Detailed Numerical Studies of the Thermal-hydraulics in the Hot Plenum of a Liquid Metal Fast Breeder Reactor. Ibid. 

Ninomiya, S., et. al. Thermal-hydraulic Study Aiming Compact Reactor Assembly, Proceedings of the International Conference on Fast Reactor and Related Fuel Cycles, Kyoto, Japan, 1991. 

Brown, G.A., Three-dimensional Computer Simulation of Flow in a Complex Fast Reactor Geometry, Proceedings of the Fifth International Topical Meeting on Reactor thermal-hydraulics - Salt Lake City, USA, 1992. 

Bums, J., et. al. Numerical Prediction of the Flow in a Sector of a Fast Reactor Hot Pool, Proceedings. Fourth International Topical Meeting on Nuclear Reactor Thermal-Hydraulics - Karlsruhe, Germany, 1989. 

XXIV-2] LEE, I S., SUH, K.Y., HELIOS for thermal-hydraulic behaviour of Pb-Bi cooled fast reactor PEACER, Theoretical and Experimental Studies of Heavy Liquid Metal Thermal Hydraulics (Paper presented at IAEA Technical Meeting, Forschungszentrum Karlsruhe, Karlsruhe, Germany, October 28-31 2003) ORA/PRO 64421. 

Okano, Y., Nakai, R., Kubo, S., 2014. International reviews on safety design criteria and development of safety design guidelines for generation-IV sodium-cooled fast reactors. In The Ninth Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS9), Buyeo, Korea, November 16—19, Keynote Lecture 1. 

Laureau, A., Rubiolo, P., Heuer, D., Merle-Lucotte, E., Brovchenko, M., 2013. Coupled neutronics and thermal-hydraulics numerical simulations of the molten salt fast reactor... 

Kamide, H., et al., 2010. Sodium experiments on decay heat removal systems of Japan sodium cooled fast reactor — start-up transient of decay heat removal system. In Presented at Seventh Korea—Japan Symp. Nuclear Thermal Hydraulics and Safety, Chuncheon, Korea, November 14—17, 2010. 

Nakai, R., et al., December 9—12, 2012. Development of safety design criteria for the generation-IV sodium-cooled fast reactor. In Keynote Lecture at NTHAS8 The Eighth Japan—Korea Symposium on Nuclear Thermal Hydraulics and Safety, Beppu, Japan. 

Rouault, J., CheUapandi, P., Raj, B., et al., 2010. Sodium Fast Reactor Design Fuels, Neu-tronics, Thermal-Hydraulics, Structural Mechanics and Safety. In Handbook of Nuclear... 

Section 2 presents the assumptions and requirements upon which the INEL concept was developed. Section 3 contains an overview of the reactor concept. Section 4 lists the conclusions and recommendations. Most of the technical details and discussions are contained in the appendices. The first task was to examine plutonium destruction rates and isotopics for different neutron spectra, as discussed in Appendix A. This study lead to the adoption of a thermal reactor concept instead of reactors with fast or epithermal neutron spectra. The second task was to study the addition of seed materials for selfprotection from materials diversion. Appendix B illustrates that fission products provide the best. self-protection, and seed materials are not needed for the INEL concept. Various fuel types were investigated and are described in Appendix C. The core neutronics studies presented in Appendix D and thermal-hydraulics studies pre.sented in Appendix E were performed concurrently. An evaluation of potential offsite radiation doses... 

In the event of the LIPOSO break, the reactor thermal hydraulics depends on the core size and on the number of primary coolant pipes. In the case of Superphenix, with 8 primary coolant pipes, a large core, and the pessimistic assumptions of DEGR, the LIPOSO break leads to a fast over-heating of the sodium in the core, which causes both reactivity and a power increase. With all uncertainties taken into account, the accident is first detected by the measurement of the core sodium outlet/inlet temperature increase (AT), which causes fast reactor shutdown 3.6 s after the break. 

J. Yoo, Y. Ishiwatari, Y. Oka and J. Liu, Conceptual Design of Compact Supercritical Water-cooled Fast Reactor with Thermal Hydraulic Coupling, Annals of Nuclear Energy, Vol. 33(11-12), 945-956 (2006)... 

J. Gou, Y. Ishiwatari, Y. Oka, M. Yamakawa and S. Ikejiri, Research and Development of Super Fast Reactor (5) Thermal Hydraulic Analysis of Tight-laiice Subchannels, Proc. 16th PBNC, Aomori, Japan, October 13-18, 2008, P16P1294 (2008)... 

Preliminary reactor core concepts have been proposed at various institutions. Among them, the concept with mixed neutron spectrum proposed by SJTU [95,96] has achieved the special attention of Chinese researchers. The mixed spectrum SCWR core combines the merits of both thermal and fast spectra as far as possible. The basic idea is to divide the reactor core into two zones with different neutron spectra. In the outer zone, the neutron energy spectrum is similar to that of PWRs. To assess the performance of the reactor core, a coupled neutron-physics and thermal-hydraulics analysis was conducted [96]. 

G.J. CALAMAI et al. Steady State Thermal and Hydraulic Characteristics of the FFTF Fuel Assemblies, FRT-1582, 1974. (Cit. by A.E. WALTAR, A.B. REYNOLDS, Fast Breeder Reactors, Perg. Press., N.Y., 1981)... 

Flux traps are useful for reactors of all power levels and a variety of irradiation facilities is desirable (e.g., pneumatic transfer, hydraulic transfer, irradiating baskets in core, or in beam tubes). Similarly, capabilities for thermal interactions and fast neutron irradiations should be available. 

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