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Method for evaluation of loads on VVER‑1200 reactor pressure vessel from in-vessel steam explosions
 
PIIS000233100003214-9-1
DOI10.31857/S000233100003214-9
Publication type Article
Status Published
Authors
K. Dolganov 
Affiliation: Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN)
Address: Russian Federation, Moscow
V. Semenov
Affiliation: Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN)
Address: Russian Federation, Moscow
A. Kiselev
Affiliation: Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN)
Address: Russian Federation, Moscow
D. Tomashchik
Affiliation: Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN)
Address: Russian Federation, Moscow
A. Fokin
Affiliation: Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN)
Address: Russian Federation, Moscow
V. Astakhov
Affiliation: OKB “GIDROPRESS”
Address: Russian Federation, Podolsk
A. Nikolaeva
Affiliation: OKB “GIDROPRESS”
Address: Russian Federation, Podolsk
D. Gasparov
Affiliation: OKB “GIDROPRESS”
Address: Russian Federation, Podolsk
S. Pantyushin
Affiliation: OKB “GIDROPRESS”
Address: Russian Federation, Podolsk
N. Bukin
Affiliation: OKB “GIDROPRESS”
Address: Russian Federation, Podolsk
M. Bykov
Affiliation: OKB “GIDROPRESS”
Address: Russian Federation, Podolsk
Journal nameIzvestiia Rossiiskoi akademii nauk. Energetika
EditionIssue 5
Pages42-58
Abstract

This work represents the results of conservative calculations of the loads on flange nozzle of a VVER‑1200 reactor pressure vessel, that occur as a result of fuel–coolant interaction during a hypothetical severe accident. The loads are evaluated using a semi-empirical method. Initial data for the evaluations come from calculations with a severe accident code SOCRAT/V1. The calculated loads may be used for further thermo-mechanical analyses of the integrity of elements that are fastening the upper head of VVER‑1200 reactor.  

Keywordsenergetic interaction, corium, coolant, severe accident, vver, socrat/v1, steam explosion, shock wave, conversion ratio, reactor, core
Publication date10.01.2019
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1. SERENA – Steam Explosion Resolution for Nuclear Applications // Final Report. NEA/
CSNI/R(2007)11. OECD NEA, 2006. P. 126.
2. Main benefits from 30 years of joint projects in nuclear safety // Report № 7073. OECD/
NEA, 2012. P. 127.
3. Tsvetkov P. Nuclear Power – Operation, Safety and Environment. Rijeka: InTech, 2011. 368 p.
4. Grishenko D., Konovalenko A., Karbojian A., et al. Insight into steam explosion in stratified
melt-coolant configuration // 15th Intern. Topical Meeting. on Nucl. Reactor Thermal
Nydraulics, 2013. Paper NURETH15–599.
5. Theofanous T.G. The study of steam explosion in nuclear systems // Nucl. Eng. and Des.
1995. V. 155. Iss. 1–2. P. 1–26.
6. Analysis of In-Vessel Retention and Ex-Vessel Fuel Coolant Interaction for AP1000 // Report
of U.S. Nuclear Regulatory Commission. NUREG/CR‑6849, 2004. P. 138.
7. Magallon D. Status and prospects of resolution of the vapour explosion issue in light water
reactors // Nucl. Eng. and Techn. 2009. V. 41. № 5. P. 603–616.
8. Hicks E.P., Menzies D.C. Theoretical Studies on Fast Reactor Maximum Accident // Argonne
National Laboratory Report. ANL‑7120, 1965. P. 654–670.
9. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power
plants / WASH‑1400 (NUREG‑75/014). App. VIII. U.S. Nuclear Regulatory Commission,
1975. P. VIII‑98 – VIII‑100.
10. STEX. International Steam Explosion Experimental Data Base (Jun. 2008). http://www.nea.
fr/abs/html/csni2007.html. 
11. I. Huhtiniemi, D. Magallon, H. Hohmann. Results of recent KROTOS FCI tests: alumina
versus corium melts // Nucl. Eng. and Des. 1999. V. 189. P. 379–389.
12. I. Huhtiniemi, D. Magallon. Insight into steam explosions with corium melts in KROTOS //
Nucl. Eng. and Des. 2001. V. 204. P. 391–400.
13. Zagorul'ko Yu.I., Zhmurin V.G., Volov A.N., Kovalev Yu.P. Ehksperimental'nye issledo‑
vaniya termicheskogo vzaimodejstviya koriuma s teplonositelyami // Teploehnergetika,
2008. № 3. S. 48–56.
14. Fletcher  D.F.  Steam explosion triggering: a review of theoretical and experimental
investigations // Nucl. Eng. and Des. 1995. V. 155. Iss. 1–2. P. 27–36.
15. Leskovar M., Meignen R., Brayer C., et al. Material influence on steam explosion efficiency:
state of understanding and modeling capabilities / Materials of the 2nd European Review
Meeting on Severe Accident Research (ERMSAR‑2007), 2007. Paper № 5.
16. Ciccarelli G., Frost D.L. Fragmentation mechanisms based on single drop steam explosion
experiments using flash X-ray radiography // Nucl. Eng. and Des. 1994. V. 146. Iss. 1–3.
P. 109–132.
17. Ik Kyu Park et al., Thermal-hydraulic aspects of FCIs in TROI corium/water interaction
tests // Nucl. Eng. and Des. 2013. V. 263. P. 419–430.
18. Song J.H. et al. The effect of corium composition and interaction vessel geometry on the
prototypic steam explosion // Annals of Nuclear Energy. 2006. V. 33. P. 1437–1451.
19. Melikhov V.I., Melikhov O.I., Sokolin A.V. Vzryvnoe vzaimodejstvie rasplava s vo‑
doj. Modelirovanie kodom VAPEX-D // Teplofizika vysokikh temperatur. 2002.
T. 40. № 3. S. 466.
20. Chislennoe reshenie mnogomernykh zadach gazovoj dinamiki / Pod red. C.K. Godunova
M.: Nauka, 1976. 400 s.
21. Ovsyannikov L.V. Lektsii po osnovam gazovoj dinamiki. Moskva–Izhevsk: Institut
komp'yuternykh issledovanij, 2003 g. 336 s.
22. Stanyukovich K.P. Neustanovivshiesya dvizheniya sploshnoj sredy. M.: Nauka, 1971.. 856 s.

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