DEVELOPMENT OF A THERMOPHYSICAL MODEL FOR THE EXPERIMENTAL ASSEMBLY OF THE VCG-135 TEST BENCH TO STUDY THE INTERACTION OF CORIUM WITH METAL-COOLER IN THE CONDITIONS OF A SEVERE ACCIDENT

This article presents the results of modeling of the temperature field of the experimental assembly of the VCG-135 test bench to study the interaction between model corium and candidate metal-coolers (zinc, antimony and manganese) in the conditions of a severe accident.

The need for modeling is associated with the probability of metal melting in the discharge device due to the heat flow from the heating crucible of the experimental assembly. Thus, the purpose of the modeling was the justification of the integrity of the design of the metal discharge device during the production of liquid corium in the crucible of the experimental assembly.

The thermophysical model was created in the ANSYS software. The temperature field of the experimental assembly was obtained at the moment of obtaining liquid corium as a result of the modeling. An analysis of the results showed that metal in the discharge device wouldn’t reach the melting point. In this regard, the discharge device of the experimental assembly can be used in its current design for experiments conducting at the VCG-135 test bench.

At the same time, after the experiments, the thermophysical model was validated by comparing the calculated temperature values with experimental data. Validation of the model shows that the deviation of calculated and experimental temperature values at control points does not exceed acceptable limits (melting of the studied metal before interaction with corium). Thus, the developed thermophysical model can be used to justify further experiments on the VCG-135 test bench with the current experimental assembly.

1. Journeau, C., Bouyer, V., Cassiaut-Louis, N. et.al. SAFEST roadmap for corium experimental research in Europe. Proceedings of 24th International Conference on Nuclear Engineering (ICONE24). Charlotte, North Carolina, June 26-30, 2016. https://doi.org/10.1115/ICONE24-60916

2. Bechta, S., Ma, W., Miassoedov, A. et.al. On the EUJapan roadmap for experimental research on corium behavior // Annals of Nuclear Energy. – 2019. – Vol. 124. – P. 541–547. https://doi.org/10.1016/j.anucene.2018.10.019

3. Kukhtevich I.V.; Bezlepkin V.V.; Granovskii V.S. et.al. The concept of localization of the corium melt in the exvessel stage of a severe accident at a nuclear power station with a VVER-1000 reactor // Thermal Engineering. – 2001. – Vol. 48, No. 9. – P. 699–706.

4. Sidorov, A.S., Nedorezov, A.B., Rogov, M.F. et.al. The device for core melt localization at the Tyan'van nuclear power station with a VVER-1000 reactor // Teploenergetika. – 2001. – Vol. 9. – P. 8–13.

5. Fischer M. The severe accident mitigation concept and the design measures for core melt retention of the European Pressurized Reactor (EPR) // Nucl. Eng. Des. – 2004. – Vol. 230, No. 1–3. – P. 169–180. https://doi.org/10.1016/j.nucengdes.2003.11.034

6. Ki Won Song, Thanh Hung Nguyen, Kwang Soon Ha et al. Experimental study on two-phase flow natural circulation in a core catcher cooling channel for EU-APR1400 using air-water system // Nucl. Eng. Des. – 2017. – Vol. 316. – P. 75–88. https://doi.org/10.1016/j.nucengdes.2017.03.009

7. Skakov, М. Toleubekov, K., Baklanov, V., et.al The method of corium cooling in a core catcher of a lightwater nuclear reactor // Eurasian phys. tech. j. – 2022. – Vol. 19. – P. 69–77. https://doi.org/10.31489/2022No3/69-77

8. Skakov, М., Baklanov, V., Toleubekov, K. et.al Modeling of the corium and metals – coolers interaction in a core catcher of a light water reactor // NNC RK Bulletin. – 2023. Issue 2(94). – P. 49–57. https://doi.org/10.52676/1729-7885-2023-2-49-57

9. Mukhamedov, N.Y., Kozhakhmetov, Ye. A. and Tskhe, V.K. Microstructure and mechanical properties of the solidified melt obtained by the in-pile test // Annals of Nuclear Energy. – 2022. – Vol. 179. – P. 109404. https://doi.org/10.1016/j.anucene.2022.109404

10. Vurim, A., Mukhamedova N., Baklanova Y. et al. Information and analytical system as a promising database used to justify the safety of nuclear energy // Nucl. Eng. Des. – 2023. – Vol. 415. – P. 112704. https://doi.org/10.1016/j.nucengdes.2023.112704

11. ANSYS Fluent Tutorial Guide, 2016

12. “Elektrotermicheskoe oborudovanie”, spravochnik pod obshchey redaktsiey Al'tgauzena A.P., Moscow.: Energiya. – 1980. (In Russ.)

13. Asmolov V. G., Zagryazkin V. N., Astakhova E. V., Vishnevskiy V. Yu., D'yakov E. K., Kotov A. Yu., Repnikov V. M. Plotnost' UO2–ZrO2- rasplavov // Teplofizika vysokikh temperatur. – 2003. – Vol. 41, Issue 5. – P. 714–719. (In Russ.)

14. Poznyak I.V., Shatunov A.N., Pechenkov A.Yu. Izmerenie elektroprovodnosti rasplava koriuma // Izvestiya SPbGETU “LETI”. – 2008. – Issue.10. – P. 39–45.

15. ICTS project #K-1265 INVECOR (IN-VEssel Corium Retention in accident of water reactor) [Электронный ресурс]. – Режим доступа: https://istc.int/ru/project/DA8802253C138C29C3257052005303CF

16. Baklanov V.V. Pribornoizmeritel'nyy kompleks dlya issledovaniya vzaimodeystviya materialov yadernogo reaktora pri tyazhelykh avariyakh: dissertatsiya na soiskanie uchenoy stepeni kandidata tekhnicheskikh nauk, Yurginskiy tekhnologicheskiy institut TPU, Yugra, 2016 // [Electronic resource]. – Access mode: https://portal.tpu.ru/portal/pls/portal/!app_ds.ds_view_bknd.download_doc?fileid=4000 (In Russ)

17. Chirkin V.S., “Teplofizicheskie svoystva materialov yadernoy tekhniki”, Moscow.: ATOMIZDAT. – 1968. (In Russ.)