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EXPERIMENTAL STUDIES IN SUBSTANTIATION OF SODIUM COOLED FAST REACTORS SAFETY

https://doi.org/10.52676/1729-7885-2018-3-117-121

Abstract

A large volume of scientific and research activities is currently carried out in substantiation of new projects on nuclear reactor safety, in which high emphasis is placed on “beyond-design basis” accidents, in spite of the fact that applied technical solutions reduce the probability of such accidents occurrence.
To study parameters of severe accidents and develop measures to mitigate consequences an experimental base is necessary, which enables emergency situation modelling maximally close to real conditions.
The National Nuclear Center (NNC) – the leading research organization in the atomic sphere in the Republic of Kazakhstan – possesses such an experimental base, providing scientific and technical support for peaceful uses of nuclear power, involving studies oriented towards the improvement of nuclear reactor safety [1].
The article presents the concept and general results of researches carried out on NNC’s experimental base in substantiation of nuclear reactors safety, realized in collaboration with JAEA, which is a key foreign partner in this area, as well as current plans on new experimental program implementation.

About the Authors

E. G. Batyrbekov
RSE “National Nuclear Center of the Republic of Kazakhstan”
Kazakhstan
Kurchatov


M. K. Skakov
RSE “National Nuclear Center of the Republic of Kazakhstan”
Kazakhstan
Kurchatov


V. A. Vityuk
RSE “National Nuclear Center of the Republic of Kazakhstan”
Kazakhstan
Kurchatov


V. V. Baklanov
Branch “Institute of Atomic Energy” RSE NNC RK
Kazakhstan
Kurchatov


A. D. Vurim
Branch “Institute of Atomic Energy” RSE NNC RK
Kazakhstan
Kurchatov


A. V. Pakhnits
Branch “Institute of Atomic Energy” RSE NNC RK
Kazakhstan
Kurchatov


K. Kamiyama
Fast Reactor Cycle System Research and Development Center, JAEA
Japan
O-arai, Ibaraki


K. Matsuba
Fast Reactor Cycle System Research and Development Center, JAEA
Japan
O-arai, Ibaraki


References

1. N.A. Nazarbayev, V.S. Shkol’nik, E.G. Batyrbekov, etc. Carrying Out Scientific, Technical and Engineering Works to Bring the Former Semipalatinsk Test Site to a Safe Conditions (in 3 volumes) // RSE “National Nuclear Center of RK”, Ministry of Energy RK, Kurchatov – 2016. – Volume 1 – p. 320. Volume 2 – p. 448. Volume 3 – p. 596.

2. GEN IV International Forum, “GIF R&D Outlook for Generation IV Nuclear Energy System”, 21 August 2009.

3. Reactor facilities for fuel pins and FA testing in emergency and transient operation modes / V. P. Burukin, A.V. Klinov, Yu.G. Toropov // Nuclear technology abroad. – 1988. No.6. p. 7–15.

4. Overseas programs of reactor research in emergency and transient modes of NPF fuel pins operation / V. P. Burukin, A. V. Klinov, Yu.G. Toporov // Nuclear technology abroad. – 1988. No.5. p. 3–7.

5. M.N. Devisheva. Overseas research and development programs on NPP safety based on water-cooled reactors: review – М.: CSRIatominform, 1989. – p. 44.

6. Nuclear Fuel Behavior Under Reactivity-Initiated Accident (RIA) Conditions: State-of-the-art Report / Nuclear Energy Agency, OECD. – Paris, 2010. – 210 p. – ISBN 978-92-64-99113-2, NEA/CSNI/R (2010).

7. I. V. Kurchatov, S.M. Feinberg, N.A. Dollezhal. Impulse Graphite Reactor IGR. Nuclear Power. – 1964. – V. 17. – No 6. – p. 463-474.

8. Kotake S, Sakamoto Y, Mihara T, Kubo S, Uto N, Kamishima Y, Aoto K, Toda M. Development of advanced loop-type fast reactor in Japan. Nucl Technol. 2010; 170:133–147.

9. Hishida M, Kubo S, Konomura M, Toda M. Progress on the plant design concept of sodium-cooled fast reactor. J Nucl Sci Technol. 2007; 44:303–308.

10. Kubo, Sh. Et al. Main outcomes and future plan of the EAGLE project / NNC RK Bulletin, issue 1, March 2016, p. 13-18.

11. Konishi, K. et al. The Eagle project to eliminate the recriticality issue of fast reactors; Progress and results of in-pile tests / Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), Jeju, Korea Nov. 2006, pp. 465–471.

12. Kamiyama K, Saito M, Matsuba K, Isozaki M, Sato I, Konishi K, Zuyev VA, Kolodeshnikov AA, Vassiliev YS. Experimental study on fuel-discharge behavior through in-core coolant channels. J Nucl Sci Technol. 2013;50(6):629–644.

13. Kamiyama, K. et al. Experimental studies on the upward fuel discharge for elimination of severe recriticality during coredisruptive accidents in sodium-cooled fast reactors. J Nucl Sci Technol. 2014; 51(9):1114-1124.

14. A.D. Vurim, V.A. Gaidaichuk, Yu.L. Istomin, Yu.V. Aleinikov, Zh.R. Zhotabayev. Experimental studies at the IGR reactor in support of the method determining the spatial position of fuel in experimental devices under conditions simulating disruptive severe accidents and fuel melting. NNC RK Bulletin, No.4, 2010, p. 33–40.

15. A.D. Vurim, Yu.A. Popov, V.A. Vityuk, Zh.R. Zhotabayev. Investigations in support of fuel mass indirect determination technique in the central experimental channel of IGR reactor on the parameters of thermal neutrons field. NNC RK Bulletin, No. 4, 2010, p. 41–49.

16. V.A. Vityuk, A.D. Vurim, V.M. Kotov, and A.V. Pakhnits. Determination of the parameters for fuel assembly tests in a pulsed graphite reactor - Atomic Energy, Vol. 120, No. 5, September, 2016, pp. 323–327, DOI:10.1007/s10512-016-0138-3.

17. Skakov, M., Mukhamedov, N., Deryavko, I. On the issue of substantiation of test modes for ampoule experimental device in research reactor. Bulletin of the Tomsk Polytechnic University-Geo Assets Engineering, vol. 328, issue 7, pp. 51–58.

18. Skakov, M., Mukhamedov, N., Deryavko, I., Wieleba, W., Vurim, A. Research of structural-phase state of natural corium in fast power reactors. Vacuum. 2017;141: 216–221.

19. The ASTRID technological demonstrator. 4th-Generation sodium-cooled fast reactors. –Tome 3, December, 2012., 96 pages.

20. F. Serre, F. Payot, C. Suteau, L. Trotignon, E. Batyrbekov, A. Vurim, A. Pakhnits, V. Vityuk, S. Kubo, A. Katoh, Y. Tobita, K. Kamiyama, K. Matsuba, J. Toyooka. R&D and Experimental Programs to support the ASTRID Core Assessment in Severe Accidents Conditions. – Proceedings of International Congress on Advances in Nuclear Power plants (ICAPP 2016), April 17-20, 2016 – San Francisco (CA, USA), Vol. 3, 2016, pp. 2173–2182.

21. F. Payot, F. Serre, A. Bassi, C. Suteau, E. Batyrbekov, A. Vurim, A. Pakhnits, V. Vityuk. The SAIGA experimental program to support the ASTRID Core Assessment in Severe Accident Conditions. – Proceedings of International Conference on Fast Reactors and Related Fuel Cycles: Next Generation Nuclear Systems for Sustainable Development (FR17), Yekaterinburg, June 26–29, 2017 (Paper ID IAEA-CN245-067).


Review

For citations:


Batyrbekov E.G., Skakov M.K., Vityuk V.A., Baklanov V.V., Vurim A.D., Pakhnits A.V., Kamiyama K., Matsuba K. EXPERIMENTAL STUDIES IN SUBSTANTIATION OF SODIUM COOLED FAST REACTORS SAFETY. NNC RK Bulletin. 2018;(3):117-121. https://doi.org/10.52676/1729-7885-2018-3-117-121

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