Preliminary Investigations of the Feasibility of In-Vessel Melt Retention Strategies for a Small Modular Reactor Concept

[+] Author and Article Information
Lena Andriolo

7 boulevard Gaspard Monge,
Palaiseau 91120, France
e-mail: lena.andriolo@edf.fr

Clément Meriot

7 boulevard Gaspard Monge,
Palaiseau 91120, France
e-mail: clement.meriot@edf.fr

Nikolai Bakouta

7 boulevard Gaspard Monge,
Palaiseau 91120, France
e-mail: nikolai.bakouta@edf.fr

Manuscript received July 31, 2018; final manuscript received December 14, 2018; published online March 15, 2019. Assoc. Editor: Fidelma Di Lema.

ASME J of Nuclear Rad Sci 5(2), 020905 (Mar 15, 2019) (7 pages) Paper No: NERS-18-1060; doi: 10.1115/1.4042360 History: Received July 31, 2018; Revised December 14, 2018

The study presented in this paper is part of the technological surveillance performed at the Electricité De France (EDF) Research and Development (R&D) Center, in the Pericles department, and investigates the feasibility of modeling in-vessel melt retention (IVMR) phenomena for small modular reactors (SMR) with the modular accident analysis program version 5 in its EDF proprietary version (MAAP5_EDF), applying conservative hypotheses, such as constant decay heat after corium relocation to the lower head. The study takes advantage of a corium stratification model in the lower head of the vessel, developed by EDF R&D for large-sized prospective pressurized water reactors (PWRs). The analysis is based on a stepwise approach in order to evaluate the impact of various effects during IVMR conditions. First, an analytical calculation is performed in order to establish a reference case to which the MAAP5_EDF code results are compared. In a second step, the impact of the lower head geometry, vessel steel ablation, and subsequent relocation on the heat flux has been analyzed for cases where heat dissipation through radiation is neglected (in first approximation). Finally, the impact of heat losses through radiation as well as the crust formation around the pool has been assessed. The results demonstrate the applicability of the MAAP5_EDF code to SMRs, with heat fluxes lower than 1.1 MW/m2 for relevant cases, and identify modeling improvements.

Copyright © 2019 by ASME
Topics: Vessels
Your Session has timed out. Please sign back in to continue.


IAEA, 2013, “Nuclear Reactor Technology Assessment for Near Term Deployment,” IAEA Nuclear Energy Series No. NP-T-1.10, International Atomic Energy Agency, Vienna, Austria.
IAEA, 2018, Advances in Small Modular Reactor Technology Development—A Supplement to: IAEA Advanced Reactors Information System (ARIS), International Atomic Energy Agency, Vienna, Austria.
Fauske & Associates, LLC, 2019, “MAAP—Modular Analysis Program,” Fauske & Associates, LLC, Burr Ridge, IL, accessed Jan. 16, 2019, http://www.fauske.com/nuclear/maap-modular-accident-analysis-program
Bakouta, N. , Tellier, R. L. , and Saas, L. , 2015, “ Assessment of Advanced Corium-in-Lower-Head Models in MAAP and PROCOR Codes,” European Review Meeting on Severe Accident Research, Marseille, France, Mar. 24–26, Paper No. 003.
Westinghouse, 2019, “AP 1000 Nuclear Power Plant design,” Westinghouse Electric Company, LLC, Pittsburgh, PA, accessed Jan. 16, 2019, http://www.westinghousenuclear.com/New-Plants/AP1000-PWR/Overview
Granovsky, V. , 2016, “ Experimental Studies for In-Vessel Melt Retention (MASCA, METCOR, CORDEB Projects),” IVR Workshop, Aix-en-Provence, France, June 6–7.
Sehgal, B. R. , ed., 2012, Nuclear Safety in Light Water Reactors: Severe Accident Phenomenology, Academic Press, New York, p. 714.
Caputo, M. , Garcia, J. M. , Gimenez, M. , and Sanchez, S. , 2012, “ MELCOR Severe Accident Simulation for a “CAREM-Like” Integral Reactor,” European Review Meeting on Severe Accident Research, Cologne, Germany, Mar. 21–23, Paper No. 6.6.
Mériot, C. , and Barjot, F. , “ Preliminary Neutronic Design of a Small Modular Reactor Core,” International Youth Nuclear Congress (IYNCWiN18), Bariloche, Argentina, Mar. 11–17. https://www.researchgate.net/publication/327597980_Preliminary_neutronic_design_of_a_Small_Modular_Reactor_core
Jacquemain, D. , ed., 2015, Nuclear Power Reactor Core Melt Accidents, IRSN, EDP Sciences, Waltham, MA, p. 434.
Ma, W. , Yuan, Y. , and Sehgal, B. R. , 2016, “ In-Vessel Melt Retention Pressurized Water Reactors: Historical Review and Future Research Needs,” Engineering, 2(1), pp. 103–111. [CrossRef]


Grahic Jump Location
Fig. 1

Phenomena observed during IVMR [11]

Grahic Jump Location
Fig. 2

Modeling of corium in the vessel lower head with the MAAP5_EDF code [4]. T:temperature(K).

Grahic Jump Location
Fig. 3

Evolution of the decay heat power after SCRAM versus time for a preliminary pressurized water SMR design (540 MWth). Outcome of the APOLLO2/DARWIN2.3.1 calculation.

Grahic Jump Location
Fig. 4

Two phase corium pool power distributions. OX: OXide, LM: Light Metal. Qres: decay power (MW); Qup power transmitted through the upper surface (MW), Qs: power transmitted laterally (MW). Htotal: total height of the pool (m), R: radius (m), Smetal: metal surface (m2).

Grahic Jump Location
Fig. 5

Temperature evolution in pool (case 4). Ox: oxide, LM: light metal. Blue square: zone where oscillations in oxide layer temperature occur.

Grahic Jump Location
Fig. 6

Axial distribution of fluxes in vessel wall, case 5. Elevation 0 corresponds to the end of the cylindrical part of the vessel.

Grahic Jump Location
Fig. 7

Evolution of ingoing and outgoing power versus time, case 5



Some tools below are only available to our subscribers or users with an online account.

Related Content

Customize your page view by dragging and repositioning the boxes below.

Related Journal Articles
Related eBook Content
Topic Collections

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In