Zhdanov V.S.1, Mosunova N.A.2, Pribaturin N.A.1, Strizhov V.F.2, Usov E.V.1
1 Nuclear Safety Institute of the Russian Academy of Sciences, Novosibirsk Branch, Novosibirsk, Russia
2 Nuclear Safety Institute of the Russian Academy of Sciences, Moscow, Russia
The analysis of the published experimental data on modelling of accidental situations with destruction and melting of core components of fast neutrons reactors with liquid metal coolant for the analysis of fuel pins and fuel assemblies (FA) behaviour in the conditions of severe accidents with such initial events, as FA full instant blockage, uncontrolled loss of the coolant flow (ULOF) or uncontrolled overtop of reactor power (UTOP) is presented. In-pile and out-of-pile experiments, results of measurements of parameters of experiments and post-experimental researches are considered.
Integral codes used at all design stages use results of experimental modelling of the phenomena occurring in reactor facilities both during normal operation, and in the conditions of incidents and severe accidents. Extensive experimental researches in 1970-1995 now have restarted in the several countries in connection with announcement of programs of the closed fuel cycle.
The analysis of a experimental database condition of shows that modern development of new fuel types, a material of fuel pins cladding and as a whole fuel pin designs demand refinement of experimental data, and also obtaining new data, for correct verification of integral codes.
1. Proceedings of the International Meeting on Fast Reactor Safety and Related Physics. Chicago, Illinois, 1976.
2. Proceedings of Fourth CSNI Specialist Meeting on Fuel-Coolant Interaction in Nuclear Reactor Safety. Bournemouth, United Kingdom, 1979.
3. Proceedings of the International topical Meeting on Fast Reactor Safety. Knoxville, Tennessee, 1985, vol. 1.
4. Proceedings of Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors. O-Arai, Ibaraki, JAPAN, 1994.
5. Experimental Facilities for Sodium Fast reactor Safety Studies. Task Group on Advanced Reactor Experimental Facilities (TAREF). NEA/CSNI/R (2010)12. IRSN 978-92-64-99155-2.
6. Hofmann P., Hagen S.J.L., Schanz G., Skokan A. Reactor core materials interactions at high temperature. Nuclear technology, 1989, vol. 87, pp. 146—186.
7. Walter A., Reynolds А. Reacktory-rasmnogiteli na bystryh neytronah [Fast neutron breeders]. Moscow, Energoatomizdat Publ., 1986. 624 p.
8. Vassiliev Yu.S., Vurim A.D., Zhdanov V.S., Zuev V.A., Kenjin E.A., Kolodeshnikov A.A., Pahnitc A.V. Experimental’nye issledovanija po modelirovaniu procceccov, harackternyh dlja tjagelyh avarii jadernyh reaktorov, provedennye v IAE [Experimental studies on modeling processes characteristic of severe accidents of nuclear reactors conducted in IAE]. Vestnik NJATC RK - Bulletin NNC RK, 2009, no. 4 (40), pp. 26–54.
9. Konishi K., Toyooka J., Kamiyama K., Tobita Y., Sato I., Kubo S., Kotake S., Koyama K., Vassiliev Yu., Kolodeshnikov A., Vurim A., Zuev V., Pakhnits A., Gaidaichuk V. Progress in Establishment of the Innovative Safety Logic for SFR Eliminating the Recriticality Issue with the EAGLE Experimental Program. Proc. Int. Sci. Pract. Conf. “Nuclear Power Engineering in Kazakhstan“. Kurchatov, Kazakhstan, 2008.
10. Wright S.A., Schumacher G., Henkel P.R. In-pile observations of fuel and clad relocation during LMBFR Core Disruptive Accidents in the STAR Experiments. Nuclear Technology, 1985, vol. 71, pp. 187–216.
11. Dickerman C.E., Rothman A.B., Klickman A.E., Spencer B.W., DeVolpi A. Summary of TREAT experiments on oxide core-disruptive accidents, (1979). Available at: https://inldigitallibrary.inl.gov/ Reports/ANL-79-13.pdf (accessed 27.10.2018).
12. Barts E.W., Deitrich L.W., Eberhart J.G., Fischer A.K., Meek C.C. Summary and evaluation - fuel dynamics loss-of-flow experiments (tests L2, L3, and L4), (1975). Available at: inis.iaea.org/ collection/NCLCollectionStore/_Public/07/237/7237075.pdf (accessed 27.10.2018).
13. Sato I., Lemoine F., Struwe D. Transient fuel behavior and failure condition in the CABRI-2 experiments. Nuclear Technology, vol. 145, 2004, pp. 115–137.
14. Wright S.A., Worledge H., Cano G.L., Mast P.K., Briscoe F. Fuel-Disruption Experiments Under High-Ramp-Rate Heating Conditions. NUREG/CR-3862, SAN081-0413, 1983.
15. Butov A.A., Zhdanov V.S., Klimonov I.A., Mosunova N.A., Pribaturin N.A., Strizhov V.F., Usov E.V., Chuhno V.I. Modelirovanie processov plavlenija tvela I zatverdevanija rasplava, obrazujushegosja pri termicheskom razrushenii tvela bystrogo reactor, s pomoshju modulja SAFR/V1 integralnogo coda EVKLID/V2 [Modeling the processes of melting the fuel element and solidification of the melt formed during the thermal destruction of the fuel element of the fast reactor using the SAFR/V1 module of the integrated code EVKLID/V2]. Atomnaya energiya - Atomic Energy, 2018, vol. 124, no. 3, pp. 123–126.
16. Butov A.A., Zhdanov V.S., Klimonov I.A., Mosunova N.A., Pribaturin N.A., Strizhov V.F., Usov E.V., Chuhno V.I. Modelirovanie peremeshenija rasplave po poverhnosti tvela bystrogo reactor pri tjazoloj avarii s pomoshju modulja SAFR/V1 integralnogo coda EVKLID/V2 [imulation of melt movement on the surface of a fast reactor fuel element in case of a severe accident using the SAFR/V1 module of the EVKLID/V2 integral code]. Atomnaya energiya - Atomic Energy, 2018, vol. 124, no. 4, pp. 197–200.
17. Alipchenkov V.M., Anfimov A.M., Afremov D.A., Gorbunov V.S., Zeigarnik Yu.A., Kudryavtsev A.V., Osipov S.L., Mosunova N.A., Strizhov V.F., Usov E.V. Fundamentals, Current State of the Development of, and Prospects for Further Improvement of the New-Generation Thermal-Hydraulic Computational HYDRA-IBRAE/LM Code for Simulation of Fast Reactor Systems. Thermal Engineering, 2016, vol. 63, no. 2, pp. 130–139.
18. Veprev D.P., Boldyrev A.V., Chernov S.Y., Mosunova N.A. Development and Validation of the Berkut Fuel Rod Module of the EUCLID/V1 Integrated Computer Code. Annals of Nuclear Energy, 2017, vol. 113, pp. 237–245.
19. Kayser G., Charpenel J., Jamond C. Summary of the SCARABEE-N Subassembly Melting and Propagation Tests with an Application to a Hypothetical Total Instantaneous Blockage in a Reactor. Nuclear Science and Engineering, 1998, vol. 128, pp. 144—185.
20. Aberle J., Borms L., Homann Ch., Maschek W., Schmuck I., Schneisiek K., Rahn A., Romer O., Schmidt L., Verwimp A. The Mol-7C in-pile local blockage experiments: main results, conclusions and extrapolation to reactor conditions. Nuclear Science and Engineering, 1998, vol. 128, pp. 93–143.
21. Zhdanov V.S., Vurim A.D., Zverev V.V., Pivovarov O.S., Kulinich Yu.A. Resultaty ispytanij modelnyh tvelov reactora tipa BREST-300 v reactore IGR [The test results of model fuel rods of the BREST-300 reactor in the IGR reactor]. Vestnik NJATC RK - Bulletin NNC RK, 2000, no. 1, pp. 25–30.
22. Baklanov V.V., Zhdanov V.S. Facility for LWR Core Materials Studies at High Temperature. Proc. ICAPP’05. Seoul, Korea, 2005. Paper 5242.
23. Plevacova K., Journeau C., Piluso P., Baklanov V., Poirier J., Zhdanov V. Zirconium Carbide Coating for Corium Experiments related to Water-cooled and Sodium-cooled Reactors. Journal of Nuclear Materials, 2011, vol. 414, no.1, pp. 23–31.
24. Bottomley P.D.W., Journeau Ch., Zhdanov V.S., Miassoedov A., Tromm T.W., Altstadt E., Clement B., Oriolo F. Study of the processes of corium-melt retention in the reactor pressure vessel (INVECOR). Proc. ICAPP 2011. Nice, France, 2011. Paper 11375.
25. Zhdanov V.S., Kojanbajev E.T., Okapbajev R.A., Skakov M.K., Utkelbajev B.D., Shapovalov G.V. Izuchenie zashitnyh svojstv pokrytij iz carbidov niobija i tantala, nanesennyh na poverhnost grafitovogo tiglja, v experimentah po plavleniju toplivnyh smesej. Vestnik NJATC RK - Bulletin NNC RK, 2002, no. 4, pp. 69–72.
26. Plevacova K., Journeau C., Piluso P., Baklanov V., Poirier J., Zhdanov V.S. Zirconium Carbide Coating for Corium Experiments related to Water-cooled and Sodium-cooled Reactors. Journal of Nuclear Materials, 2011, vol. 414, no. 1, pp. 23–31.
27. Mosunova N.A. The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models. Thermal Engineering, 2018, vol. 65, no. 5, pp. 304–316.
