Nuclear Safety
NEA/CSNI/R(2017)15
March 2018www.oecd-nea.org

Status Report on Ex-Vessel
Steam Explosion

NEA/CSNI/R(2017)15

THE COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONSThe Committee on the Safety of Nuclear Installations (CSNI) shall be responsible for Nuclear Energy
Agency (NEA) programmes and activities that support maintaining and advancing the scientific andtechnical knowledge base of the safety of nuclear installations, with the aim of implementing the NEA
Strategic Plan for 2017-2022 in its field of competence.
The Committee shall constitute a forum for the exchange of technical information and forcollaboration between organisations, which can contribute, from their respective backgrounds in research,development and engineering, to its activities. It shall have regard to the exchange of information between
member countries and safety research and development (R&D) programmes of various sizes in order tokeep all member countries involved in and abreast of developments in technical safety matters.
The Committee shall review the state of knowledge on important topics of nuclear safety science andtechniques and of safety assessments, and ensure that operating experience is appropriately accounted forin its activities. It shall initiate and conduct programmes identified by these reviews and assessments inorder to confirm safety, overcome discrepancies develop improvements and reach consensus on technicalissues of common interest. It shall promote the co-ordination of work in different member countries thatserve to maintain and enhance competence in nuclear safety matters, including the establishment of joint
undertakings (e.g. joint research and data projects), and shall assist in the feedback of the results toparticipating organisations. The Committee shall ensure that valuable end-products of the technical reviews
and analyses are provided to members in a timely manner, and made publically available when appropriate,
to support broader nuclear safety.
The Committee shall focus primarily on the safety aspects of existing power reactors, other nuclear
installations and new power reactors; it shall also consider the safety implications of scientific andtechnical developments of future reactor technologies and designs. Further, the scope for the Committee
shall include human and organisational research activities and technical developments that affect nuclear
safety.
The Committee shall organise its own activities. Furthermore, it shall examine any other mattersreferred to it by the Steering Committee. It may sponsor specialist meetings and technical working groups
to further its objectives. In implementing its programme the Committee shall establish co-operative
mechanisms with the Committee on Nuclear Regulatory Activities in order to work with that Committee
on matters of common interest, avoiding unnecessary duplications.
The Committee shall also co-operate with the Committee on Radiation Protection and Public Health,
the Radioactive Waste Management Committee, the Committee for Technical and Economic Studies on
Nuclear Energy Development and the Fuel Cycle, the Nuclear Science Committee, and other NEA
committees and activities on matters of common interest.

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LIST OF ABBREVIATIONS AND ACRONYMS

ALISA

Access to Large Infrastructures for Severe Accidents

BWR

Boiling water reactor

CEA

Commissariat à l’Energie Atomique et aux Energies Alternatives

CSNI

Committee on the Safety of Nuclear Installations

DEFOR

Debris bed formation

DNS

Direct Numerical Simulation

DROPS

FCI- Droplet Experimental facility

EC

European Commission

EDF

Électricité de France

EPR

A PWR design by Areva NP and EDF

EVSE

Ex-Vessel Steam Explosion

FARO

FCI-Experimental facility-JRC-Ispra-Italy

FCI

Fuel-coolant interaction

FP-6, FP-7

Framework Program (European Commission research programmes)

ISP

International Standard Problem

IRSN

Institut de Radioprotection et de Sûreté Nucléaire (France)

KAERI

Korean Atomic Energy Research Institute

KROTOS

FCI-Experimental facility CEA-Cadarache-France

LWR

Light water reactor

NEA

Nuclear Energy Agency

O/M

Oxide-to-metal ratio

OECD

Organisation for Economic Co-Operation and Development

PLINIUS-2

CEA future large-mass experimental platform

PWR

Pressurised water reactor

RCS

Reactor coolant system

RPV

Reactor pressure vessel

R&D

research and development

SAFEST

Severe Accident Facilities for Europe Safety Targets

SARNET

Severe Accident Research Network of Excellence

SE

Steam explosion

SIGMA

FCI- Droplet Experimental facility

TOP

Technical opinion paper

TROI

Experimental facility KAERI, Dejon, Korea

WGAMA

Working Group on the Analysis and Management of Accidents

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TABLE OF CONTENTS

THE COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS ...............................................1
LIST OF ABBREVIATIONS AND ACRONYMS ........................................................................................2
TABLE OF CONTENTS ................................................................................................................................3
EXECUTIVE SUMMARY .............................................................................................................................4
ACKNOWLEDGEMENTS ............................................................................................................................6
I. INTRODUCTION........................................................................................................................................7
II. STATUS OF THE UNDERSTANDING OF THE EX-VESSEL STEAM EXPLOSION (EVSE) ...........8
II.1

Short historical perspectives .............................................................................................................8

II.2

Status of the EVSE knowledge gained in the frame of SERENA-Phase 2 .......................................9II.2.1 Experimental task ...........................................................................................................................9II.2.2 Analytical task ..............................................................................................................................11
II.2.3 Reactor application task................................................................................................................12
II.3

Ranking and knowledge of EVSE key phenomena ........................................................................12

II.4

EVSE containment building consequences .....................................................................................16

II.5

EVSE national approach after Fukushima ......................................................................................19

II.6

EVSE improvements and needs for reactor application..................................................................20

REFERENCES ..............................................................................................................................................23

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EXECUTIVE SUMMARYThe Status Report on Ex-Vessel Steam Explosion (EVSE) summarises the significant progress made in thefield of Fuel Coolant Interaction/Steam Explosion (FCI/SE) knowledge during these last ten years,
especially in the framework of the OECD/SERENA-Phase 2 program (2007-2012).
Key phenomena for steam explosions have been ranked for their knowledge base and possible
consequences for the containment building. From this ranking, it can be stated that jet orientation andfragmentation, triggering, solidification and oxidation phenomena are considered to have limited
knowledge, but also high importance on SE energetics.
The consequences on the integrity of the containment building due to an ex-vessel SE have been also
reviewed. The consequences of ex-vessel steam explosion can be divided in two categories: directconsequences (mechanical constraints on the concrete structures, typically the reactor pit and on the reactor
vessel) and indirect consequences (motions of heavy reactor circuit components such as reactor vessel orsteam generators that can affect leak tightness of the containment). It is also important to consider impact
on medium and long term severe accident management e.g. with regard to possible damage of equipment
responsible for long term cooling of the debris. The methodologies used to estimate the consequences of
ex-vessel steam explosions in pressurised water reactors (PWRs) and boiling water reactors (BWRs) andthe maximum acceptable impulse load on the containment are at least slightly different among countries,
which emphasises the difficulty in reaching a common international approach on this point.
After Fukushima-Daiichi severe accidents, the regulatory bodies in the participating countries weresurveyed on the existing regulatory requirements for steam explosion. Main observations from this survey
responses were: (1) the regulatory requirements varied between countries; (2) in most countries, if not all,the consideration of ex-vessel steam explosion remains an open issue, mainly due to unresolved
uncertainties; (3) the methodologies available for steam explosion risk assessment have the same basic
elements but the complexity and sophistication varied among countries: the existing methodologies used adeterministic assessment of steam explosion loads (impulse and/or pressure) with a structural mechanics
assessment of containment fragility and a probabilistic framework to evaluate the risk of containment
failure associated with a steam explosion; and finally, (4) the need for and scope of continued research on
the subject varied between the countries.
In spite of the important progress made in the frame of the OECD/SERENA-Phase 2 program, theknowledge is still limited on some important aspects of EVSE and therefore the impact of this issue on
containment building integrity.
There are ongoing new FCI/SE programs, even after the OECD/SERENA-Phase 2 program: in 2017 atnational level in France ICE-Post-Fukushima program, and in Sweden; and at international level in Europe,
the SARNET network in the frame of NUGENIA program, providing exchanges among the experts andpromoting access to some of the facilities in Europe in the frame of the European SAFEST (2014-2018)
and ALISA (2015-2018) programs. Korea finished their national FCI program in early in 2017.
Small scale experiments are suitable to improve models currently used in FCI codes, and these
improvements can then be validated using large scale experiments. Among the experiments that are needed
to improve analytical models, experiments dedicated to study the oxidation of single drops, the4NEA/CSNI/R(2017)15
solidification of single drops, the fragmentation of prototypical drops in reactor conditions, and film
boiling at very high temperature are more useful.
In the short term (2017-2021), it seems useful to build an international program in order to addresssome of the required improvements and the needs identified for reactor applications: additional tests could
be performed in the TROI and KROTOS facilities with a focus on the impact of more realistic corium
composition on SE energetics and debris bed cooling (debris coming from either premixing or explosion).
Among the potential prototypical corium tests, the useful tests are with corium compositions with largesolidification intervals and sub-stoichiometric composition. These tests could be performed taking into
account conditions more representative of ex-vessel scenarios (for example higher initial velocity of themelt, “real” water with boric acid and painting aerosols, and submerged vessel). Experimental tests in“stratified” configuration could also be considered.
In the middle term (2022-2026), main FCI/SE phenomena at large scale (80-120 kg) with different
corium compositions -including metallic corium- would allow reliable qualification and validation of the
EVSE modelling for the extrapolation to reactor case. Configurations which have scarcely been evaluatedpreviously (i.e. pressurised and inclined jets, partial melt stratification, lateral jets impacting on walls)
could be studied. These experiments will also be representative of intermediate scale between
KROTOS/TROI facilities and reactor scales, and thus would allow more reliable qualification andvalidation of the EVSE modelling for the extrapolation to reactor case. The future CEA/PLINIUS-2
experimental platform should be able to answer this final and indispensable step of validation forpremixing and steam explosion for reactor case code application. Additionally, international collaboration
on such large scale programs would be desirable in order to rationalise research efforts.

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ACKNOWLEDGEMENTS

Compiled by P. Piluso, CEA
Leading authors: Martina Adorni (BELV), Garo Azarian (AREVA), Kwang-Hyun Bang (KOREA
Maritime and Ocean University), Sudhamay Basu (US-NRC), Michael Buck (IKE), Philippe Dejardin(TRACTEBEL- ENGIE), Erik De Malmazet (EDF), Wiktor Frid (SSM), Akitoshi Hotta (NRA), SeongWan Hong (KAERI), Pavel Kudinov (KTH), Matjaz Leskovar (JSI), Renaud Meignen (IRSN),
Thambiayah Nitheanandan (AECL), Pascal Piluso (CEA).
This work has been done in the frame of OECD/NEA/CSNI/WGAMA organisation. This Status
Report includes contributions from: Renaud Meignen; Sudhamay Basu; Matjaz Leskovar; Claude Brayer;
Nathalie Cassiaut-Louis; Pascal Piluso; Pavel Kudinov; Michael Buck; Wiktor Frid; Patrick Isaksson;
Thambiayah Nitheanandan; Akinobu Morita; Harutaka Hoshi; Yu Maruyama; Akitoshi Hotta; Bojan
Niceno; Erik De-Malmazet; Seong-Wan Hong; Garo Azarian; Tristan Conti; Manfred Fischer; KwangHyun Bang; Martina Adorni; Stephane Picchi; Philippe Dejardin; Thibaut Helman; Mike Corradini.

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I. INTRODUCTION

The interaction of the reactor core melt with water for light water reactors (LWRs) (Generation 2&3),
known as molten fuel-coolant interaction (FCI), is one of the most complex technical phenomena,
involving a coupling between thermal-hydraulic and physico-chemical phenomena. FCI may occur for ExVessel configuration, when the reactor vessel lower head fails and a molten corium jet is ejected into aflooded cavity in the reactor pit. Each of the severe accident scenarios with vessel failure and water in thecavity may lead to an energetic FCI, commonly known as “steam explosion” (SE), which represents
potentially serious challenge to the reactor containment integrity and severe accident management. Bothenergetic and non-explosive FCIs result in melt fragmentation, quenching and formation of a debris bed in
the reactor pit.
This Status Report is devoted to ex-vessel steam explosion (EVSE) status of knowledge for LWRs
reactor applications. It has been prepared by the main experts involved in SERENA-2 program (Steam
explosion resolution for nuclear applications), belonging to Committee on the Safety of Nuclear
Installations/Working Group on the Analysis and Management of Accidents of OECD
/NEA/CSNI/WGAMA. The expressed experts’ opinions are summarised here in support of regulatory
decision making process for the nuclear power regulators in addressing the FCI issues consistent with theirown regulatory philosophy.
The EVSE Status Report begins with a brief historical perspective of the progress made on FCI/SE, inparticular, the knowledge gained on the subject through decades of research. Then, it summarises thefindings obtained through the framework of the OECD/SERENA-Phase 2 program. Expert’s ranking of the
EVSE key phenomena and evaluation of their status of knowledge are then summarised. In another section,consequences of EVSE on containment and reactor building integrity are presented, in particular toevaluate plant specific potential margin in case of ex-vessel steam explosion and the degree of accuracy
necessary for the evaluation of Steam Explosion energetics. Approaches to addressing the EVSE issues by
various national regulatory bodies after Fukushima Daiichi severe accident are then summarised. Finally,
main improvement needs for reactor applications are identified.

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II. STATUS OF THE UNDERSTANDING OF THE EX-VESSELSTEAM EXPLOSION (EVSE)

II.1

Short historical perspectives

Historically, operating light water reactors (LWRs) in many OECD countries were commissioned at a time
when steam explosion (SE) did not surface as an issue. Consequently, there were no regulatory
requirements set forth for these reactors at the time.
However, since the identification of the issue in WASH-1400 (USNRC, 1975), the safety analysis
practice in most of all OECD countries included an assessment of SE risk. Specifically, containment failurefrom an impact generated by the failed reactor upper head as a result of an in-vessel steam explosion eventand subsequent missile generation was designated as the alpha-mode failure.
In 1985, the first Steam Explosion Review Group (SERG-1) (USNRC, 1985) was convened todeliberate on the probability of alpha-mode failure, based on the then available information.
The second Steam Explosion Review Group (SERG-2) (NUREG, 1996) was convened in 1995 torevisit the deliberations from SERG-1 in light of the technical progress made and new findings therein. An
important conclusion from the SERG-2 (NUREG, 1996) meeting was that the alpha-mode failure issue wasconsidered to be resolved from a risk point of view, meaning that this mode of failure is of very low
probability. The alpha-mode failure is considered to be of little or no significance to the overall risk from anuclear power plant, and that any further reduction in residual uncertainties is not likely to change theprobability in an appreciable manner.
The CSNI ISP-39 International Standard Problem (NEA, 1998) exercise has been completed on thetest FARO L-14 to benchmark the predictive capabilities of computer codes used in the evaluation of FCI
and quenching phenomenologies of relevance in water cooled reactors accident safety analyses. The
exercise demonstrated a large spread between FCI code predictions for the first time.
A previous Technical opinion paper (TOP) devoted to FCI status was published by OECD/NEA (NEA,
1999) in March 2000. Different topics relevant to nuclear safety were analysed in this paper including
quenching and coolability of corium, and energetic FCI. While it was recognised that the experimental database with prototypic reactor core material was insufficient, concerning energetic FCI the general opinion
was that the consequences of steam explosion under reactor conditions may not be as severe as previously
envisioned in the 1990’s. The March 2000 TOP recommended performing integral small-scale and largescale experiments with prototypical materials to verify and to assess the models and codes performances,
and also to apply the FCI codes at the reactor scale. It was believed that the material characteristics play afundamental role in SE explosivity and energetics. Furthermore, it was advised to focus on improving bothprobabilistic and deterministic methodologies, and on better understanding of the physics of corium jet
break-up and fine fragmentation. Finally, there was a general consensus that fully validated codes, while
desirable, were not an absolute requirement for reactor applications and that a fit for purpose validation
approach would be sufficient.
In the framework of the OECD/SERENA analytical program (NEA, 2007) (SERENA-Phase 1, 20022005), FCI integral codes have been applied to experimental tests and reactor cases. Initially, selected premixing and explosion tests were calculated and the results compared, for verification of the tools and for8

NEA/CSNI/R(2017)15
identification of the common areas where uncertainties impacted the explosion results. Then, two reactor
cases for “PWR-like” configurations were chosen: in-vessel and ex-vessel steam explosions (EVSE). For
the in-vessel configuration, whatever the modelling and numerical approaches may be, all codes were able
to calculate the reactor events of concern. Despite the different choices for setting the code parameters forthe reactor applications and respective spread of the predicted pressures and impulses, all the calculatedloads were below the mechanical failure of a nuclear reactor vessel.
These results showed that the safety margin for in-vessel FCI might be sufficient for the configurationand scenario selected in the frame of SERENA-Phase 1. For the ex-vessel configuration, the situation was
different: the calculated loads were in some cases above the load bearing capacity of typical cavity walls
for the reactor containment. The wide scatter (1-2 order of magnitudes) of the predicted loads raised the
problem of the quantification of the safety margin for EVSE. The major sources of uncertainties concern
the characteristics of the flow regime in the pre-mixing phase, especially void behaviour, and of thefragmentation of corium melts in the explosion phase (due to material effect). In particular, it was planned
that confirmation for an extended range of corium melts that the void fractions in the mixing area andfragmentation characteristics would confirm the “mild” characteristic of explosion with reactor melts. At
the end, the SERENA-Phase 1 report further confirmed the conclusion of SERG-2 but recommended
prototypic material experiments to provide data base for improvement of analytical tools.
II.2

Status of the EVSE knowledge gained in the frame of SERENA-Phase 2

The OECD/SERENA-Phase 2 (NEA, 2015) was carried out in 3 tasks (experimental, analytical, andreactor application) and focused on EVSE studies, especially on the two following phenomena, partially
understood and identified as having impact on SE energetics:


formation of the void (steam content and spatial/time distribution) on pre-mixture/fragmentation
stage and its impact on steam explosion;

physical relationships between corium melt materials compositions and-pre-mixture/
fragmentation and steam explosion.

II.2.1 Experimental task
Complementary experiments were performed in the two FCI facilities and using prototypical corium1:
KROTOS, at the Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA) in Cadarache(France), and TROI at Korea Atomic Energy Research Institute (KAERI) in Daejeon (Korea). The
KROTOS facility features one-dimensional behavior of mixing and explosion propagation, and allows thecharacterisation of premixing (melt and void distribution), escalation, and propagation behavior. The TROI
facility is more suited for testing the FCI behaviour in multi-dimensional geometry.
The experimental test matrix (see Table 1) was composed of two series of 6 tests: one in the KROTOS
configuration with about 3-5 kg of prototypical corium melt, and the other in TROI configuration with
about 15-20 kg of prototypical corium melt. To guarantee the occurrence of a steam explosion, all tests inboth facilities were artificially triggered just before melt-bottom contact of the test section.
Four corium material compositions representative of reactor case conditions were chosen (seeTable 1). The first composition was the 70% UO2-30% ZrO2 so-called “eutectic” composition, a referencecase identified as having spontaneous and more energetic explosions in TROI conditions (Song et al.,2003) than the second composition 80% UO2-20% ZrO2 so-called “non-eutectic” composition. Two newprototypical corium compositions, more representative of reactor conditions, were chosen: one suboxidized composition (so-called “corium C-50”: 50% of the zirconium assumed to be oxidized in the1.

Prototypical: corium with nuclear fuel UO2.

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vessel, final oxide-to-metal ratio O/M =O/(U+Zr)<2.00), and the last one including low volatile oxide
fission products (source term representative of PWR spent fuel) and oxidized iron (internal components,
vessel).
The void fractions were measured in both facilities using different measurement techniques: in the
KROTOS facility, an X-ray source2 able to distinguish water, void (mainly steam) and corium, and in the
TROI facility a differential pressure transducer system. For the first time, it was possible to have access to
the local void and corium volume fractions (spatially and in time) in the KROTOS facility during thepremixing/fragmentation stage.
Table 1: OECD/SERENA-Phase 2. KROTOS and TROI experimental test matrix
Series1

Challenging conditions

KROTOS (KS)TROI (TS)Standard geometrical conditions

High system pressure (0.4 MPa)

High system pressure (0.4 MPa)

Reduced free fall (melt jet velocity) and thickmelt jetMaterial 1: 70% w/o UO2-30% w/o ZrO2
2

Geometry effect
Effect of geometry bycomparison between KROTOS
and TROI

Standard conditions: jet of diameter 3 cm

Large jet at penetration (5 cm)

3

Reproducibility tests

Idem Test 2

Idem Test 2

4

Material effectOxidic composition

Standard conditions

Large jet at penetration (5 cm)

Material 1: 70 %w/o UO2-30% w/o ZrO2Material 2: 80% w/o UO2-20% w/o ZrO2
5

Material effectOxidation/composition

Standard conditions

Large jet at penetration (5 cm)

Material 3: 80% w/o UO2-11.5% w/o ZrO2+8.5%Zr
6Material effectLarge solidus/liquidus T

Standard conditions. Effect of fission product:
higher melt superheat

Large jet at penetration (5 cm).
Effect of fission product: higher melt
superheat

Material 4: 73% w/o UO2-20% w/o ZrO2 +4% FeO+ 3%low volatile fission products

The SE efficiency (expressed in terms of conversion ratio η, representative of a given geometry)3
spanned a range between 0.1% and 0.6%. Compared to previous tests performed in KROTOS and TROI
facilities using prototypical corium, the maximum value of the conversion ratio was higher. This was
attributed to the melt pouring conditions in both facilities in the newer experiments, allowing larger

Ekinetic
Ethermal

TROI:.
Linatron
X-Ray device-high energy: E = 6-11MeV - 0.33Gy/s.


2
3

.

Steam explosion efficiency ratio (%), a criterion for SE energetics qualification for a given geometry:

The corium temperature considered for the calculation of thermal energy is : Liquidus temperature +50 K
1

KROTOS :  

Ekinetic
Ethermal

The difference between corium temperature and water considered for the calculation of thermal energy is : 2500K

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overheated melt masses to be delivered in the test section and participate in a steam explosion, all the otherparameters being equal. It can be noted that these values are lower than for those obtained in previous
KROTOS experiments with simulant material (e.g. alumina, maximum efficiency: η ~2%), and considered
as conservative reference case in the absence of oxidation effects and neglecting the so-called “material
effect”. Experts have concluded: prototypical corium compositions used in previous tests did notencompass sufficient variations to support conclusively the “mild” explosive behavior of prototypic corium
observed up to now for the same geometry.
TROI and KROTOS tests series 1, 2, 3, 4 have shown same behavior for “eutectic” (Material 1) and“non-eutectic” (Material 2) corium compositions for the steam explosion. TROI and KROTOS tests
series 5 (Material 3) have shown very low energetics release, stressing the importance of the oxidation
mechanisms (heat release and hydrogen production) on SE and its potential to limit the mechanical loads.
TROI and KROTOS tests series 6 (Material 4) have shown different behaviour, stressing the importance ofsolidification phenomena on SE energetics: for TROI experiment, the highest energetic SE of the
SERENA-Phase 2 series program has been observed (η~ 0.6%), whereas it was not the case for KROTOS
test.
These two last series of tests, argued to be more prototypic of reactor situations, have shown therelevance of the oxidation and solidification processes as well as the importance of the corium composition
on SE energetics.
The OECD/SERENA-Phase 2 program generated important new data and understanding of ESVE,
showing the impact of more realistic corium compositions on SE energetics. At the end of the
OECD/SERENA-Phase 2 program, it was not possible to propose a definitive modelling of the oxidationand solidification phenomena in SE due to the limited number of experimental tests (only one integral test
in each facility). Complementary experimental data using prototypical corium (current TROI and
KROTOS facilities) would be desirable as they would likely improve our understanding of oxidation andsolidification phenomena, and help develop appropriate models for implementation in FCI codes, and in
the future, at larger scale (CEA-PLINIUS-2 project) development and validation of respective models in
FCI integral codes.
II.2.2 Analytical task
The main results obtained can be summarised in the following points:

Jet fragmentation and void behavior in subcooled pool conditions, is confirmed once again as a
key phenomenon that determines the amount of melt mass participating in explosion. Experts
noted that improvements in modeling jet fragmentation are highly desirable if further reductionof uncertainties in energetics is warranted for specific safety applications. At the same time, someexperts also opined that a simplified approach as is currently adopted, based on global
fragmentation correlations, might be adequate for some reactor applications.

Melt solidification is a major contributor to the limitation of energetics for oxidic and substoichiometric corium melts. Attempts to incorporate the solidification effect in FCI codes led toimportant model improvements, e.g., modeling of crust formation and its impact on fine
fragmentation, and modeling of drop size distributions to capture effect of different solidification
times. It is also necessary to confirm the limiting effect of solidification for a wider range ofmaterials, especially prototypical corium material with large temperature interval of solidificationsuch as the material 4 of the SERENA-Phase 2 experimental test matrix, more representative ofreactor case.

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Melt density as a material property is considered by some experts to be a key parameter in
explaining the material effect on explosion energetics in previous KROTOS experiments(prototypical corium, alumina). The high melt density of oxidic prototypic core melts yields
smaller droplets, thus faster solidification and higher voiding. This may explain the lower
energetics of the explosion with prototypic melts. Another explanation has been proposed (Piluso
et al., 2005): the difference of behavior between simulant materials and oxidic prototypicalmaterials could be linked to their ability to be supercooled (i.e. to remain liquid below themelting point). Alumina is known as being a material with a high supercooling (more than
600 K) whereas few data exist on ability to be supercooled for oxidic corium compositions.

Oxidation effects in codes needs development of modeling and complementary experimental
data for validation, especially for sub-stoichiometric corium melt (O/M<2.00).

II.2.3 Reactor application task
With regard to reactor applications, in the framework of SERENA-Phase 2, the scattering between FCI
codes was reduced as a result of better understanding and modelling of the premixing/fragmentation phase.
Nevertheless, the range of pressures and impulses due to steam explosion are still of the same order ofmagnitude which may induce some damage to the reactor cavity pit and could have impact on containment
integrity. While the voiding phenomena did not play an important role at the experimental scale, its
importance at the reactor scale cannot be ruled out. There is also a need to confirm the importance ofsolidification. If the effect is moderate at experimental scale, the calculations indicate a strong mitigating
effect at reactor scale (the proportion of solidified drops increases with the scale). Reactor case calculation
with FCI codes, performed for PWRs and BWRs ex-vessel generic configuration, have shown impulses of
150 kPa.s at bottom and 125 kPa.s at wall for PWRs and 110 kPa.s at bottom and 90 kPa.s at wall for
BWRs. For impulses to remain on the safe (not too conservative) side, the level of dynamic load to thecavity wall needs be lower than 100 kPa.s.
It is finally recognized that consistent improvements can be obtained by discretising the drop
population in a multi-size approach (codes MC3D, JEMI, JASMINE). The large size spectrum allows in
itself a reduction of the uncertainty.
Finally, there is still an outstanding issue concerning FCI code predictions at the reactor scale,stemming primarily from the basis of extrapolating the experimental results to reactor scale. Expert
judgments considering also experimental results are to be part of any safety demonstration for SE.
II.3

Ranking and knowledge of EVSE key phenomena

Based on the findings from the OECD-SERENA program (NEA, 2015) and also from the European
SARNET programs and other national programs (ongoing as well as recently completed), the experts haveranked, for this status document, the importance of key EVSE phenomena as well as assessed the status ofknowledge. Especially over the past few years (2012-2016), important collaborative efforts have beenachieved to improve a common understanding and modeling of SE (Meignen et al., 2014) in the frame of
EC-SARNET 1 (FP-6) and 2 (FP-7). For this section, it is important to note that all relevant phenomena are
identified in the list regardless of their importance ranking. However, it should be noted that the ranking ofimportance was not necessarily a consensus one for some phenomena, neither was the assessment ofknowledge for the experts involved in the EVSE status document.
Table 2 describes the status of the knowledge and the ranking on the main properties having a role
during FCI/SE phenomena. Table 3 describes the main reactor case scenario parameters having an impact
on ex-vessel steam explosion energetics and the status of knowledge. Table 4 is focused on the main

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phenomena occurring during the premixing/fragmentation and steam explosion steps and the status ofknowledge.
Table 2: Main properties having an impact on ex-vessel steam explosion phenomena
Properties

Rankinga

Status of knowledgeb

Liquidus temperature

H

M

Solidus temperature

H

M

Other thermodynamic data (cp, thermal diffusivity…)

H

I

DensityH

M

Melt undercooling ability

M

I

Surface tension

H

M

Viscosity

L

M

Emissivity

L

I

Three levels have been established to quantify the degree of relevance:a. Ranking: H: HighM: Medium

L: Lowb. Status of knowledge: S: Satisfactory M: Medium

I: Insufficient

For the properties involved in SE, the following can be added:The liquidus and solidus temperatures, and thermodynamic parameters are of first importancefor SE energetics evaluation because these data are used for solidification points or coolingmodelling. The level of knowledge is not the same according to the nature of corium, and, from ageneral point of view some essential thermodynamic data of the corium liquids are still missing.

The ability of supercooling/formation of metastable solids according to the corium compositions
and cooling conditions will play a role on SE. Taking into account the nature of corium (metallic,sub-stoichiometric or stoichiometric), different behaviors may exist; small analytical experiments
are useful to establish criterions of supercooling according to the nature of corium.



The thermophysical properties of the corium are partially known from general point of view. As
for thermodynamic properties, it depends on the nature of the corium. Experts note that surface
tension is the key parameter for fragmentation process. In contrast, viscosity (dynamic viscosityliquid state), and emissivity, are expected to be of limited importance.

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Table 3: Main scenario conditions having an impact on ex-vessel steam explosion energetics
and status of knowledge
Reactor scenario conditions

Rankinga

Status of knowledgeb

Jet velocity

HM

Jet diameter

H

I

Break location (orientation of corium jet)

H

I

Melt (corium properties)

M-H4

M

Melt superheat

H

S

Coolant (water) temperature

H

S

Coolant (water) impurities

M

M

Water level

H

M

Containment pressure

M

M

For the reactor scenario conditions, the following comments can be added:The melt jet fragmentation occurs mainly once it penetrates the water. Jet velocity at theentrance into the water impacts strongly the corium fragmentation process and so the SE
energetics. Experimental data on corium jets for ex-vessel configuration have been essentiallyobtained in vertical configurations with gravity driven velocity assuming scenarios for which thereactor vessel will fail through a central hole in the lower plenum. An important experimental
data base now exists for a given range of prototypical corium, especially thanks to
SERENA/Phase 2 program. Nevertheless, other conditions of failure of the reactor vessel have to
be considered for PWRs and BWRs modifying the corium fragmentation phenomena, as forexample: “Focusing effect” and lateral break: non-vertical jet, non-round jet, presence of vessel
thermal insulation and of concrete vertical wall for the first impact of the corium jet. Thismode of failure is mostly relevant to the scenarios where external vessel cooling is provided
 Instrumentation tube guide failure: turbulent conditions, formation of spray, non-verticaljet.
 Pressurized jets: velocity of the jet is driven by residual pressure difference between insideand outside the vessel. The effect is expected to be as non-negligible on premixingfragmentation phenomena, based on expert judgement. Unfortunately the knowledge iscurrently very limited on this topic due to the lack of experimental data and parametric laws
in this configuration.



The coolant temperature and the water sub-cooling play an important role on steam explosion
energetic. Some already existing experimental data have shown the importance of this parameteron void formation.



The nature of the water can also have an impact. Up to now, all the FCI/Steam explosion
experiments performed all over the world since 25 years used pure distilled water whereas in thereal reactor severe accident conditions, water will be a more complex liquid with paint4. Corium properties: ranking Medium to High, depending on specific properties considered (see Table 4).

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NEA/CSNI/R(2017)15
constituents, fission products, boric acid, sea water, like in Fukushima Daiichi. The main
properties of the water will be the same order of magnitude (e.g. with H3BO4 or NaCl), but theformed steam, and especially the interface properties between the steam and the water or the
corium droplet, could be different, having possible impacts on SE phenomena. Further
knowledge regarding the impact of realistic water composition on steam explosion energetic
might be obtained with small scale experiments. Previous work with non-nuclear materials andspecific conditions for the water (adding additives) has shown the suppression of steam explosion
triggering (Furuya and Kinoshita, 2002).


The level of water in the reactor pit plays an important role on steam explosion energetics for
PWRs and BWRs. Two families of scenarios can be considered for LWRs reactor pit: partiallywater filled pit (“shallow” water) or water full pit (“submerged” vessel). When the pit ispartially filled, the energetics are a function of the water level, depending on the initial corium jet
diameter. For the configuration with a low level of water (a few tenth centimeters- “shallow”
water), recent tests performed with simulant materials have shown possibly spontaneous andenergetics steam explosion (Grishchenko et al. 2013) with conversion ratios approaching that ofother simulant materials. These recent tests have been considered in the literature as “stratified”
SE, and generated an interest in the subject. For the submerged vessel configuration, there arefew experimental data (Na et al. 2016). Given that experimental results are very limited for bothof these reactor cases (“shallow” water or “submerged” vessel), new experimental data would benecessary to establish a criterion regarding SE energetics or triggering for these configurations.



The containment pressure (range from 1 to 4 bars): the impact of low ambient pressure is
expected to be moderate on the energetics.

Table 4: Main phenomena parameters having an impact on ex-vessel steam explosion energetics
and status of knowledge
Phenomenon

Rankinga

Status of knowledgeb

Jet fragmentation in gas space

LS

Jet fragmentation in water

H

M

Melt droplet characteristics in premixing (mean size, size distribution, nature of
the corium: metallic, sub-stoechiometric, oxidic)

H

M

Heat transfers

H

S

Melt solidification

H

M

Melt oxidation

H

I

Voiding in premixing

H

M

Triggering

H

M

Fine fragmentation in explosion

H

M

Stratified steam explosion

H

I

Mechanisms of pressurization

H

M

For the phenomena involved in SE, some comments can be added:

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II.4



The coarse fragmentation and the resulting drop size is of high importance, since it affects heattransfers and fine fragmentation rate during explosion. The knowledge on oxidic corium withsmall temperature interval of solidification is relatively well known, but much less is known for
metallic, sub-stoichiometric and large temperature interval of solidification of corium mixtures.
Concerning the jet fragmentation, there is a limited knowledge for the lateral vessel failure, forexample, in case of “focusing effect” and also for the pressurised jet.

The heat transfer characteristics are obviously of utmost importance in steam explosion forsteam production and melt solidification. Knowledge is estimated as relatively satisfactory,
although some aspects related to heat transfer around drops and fragments still need some
improvements.

The solidification and oxidation phenomena are both expected to be important and there isknowledge gaps remain. Both are strongly dependent on the corium composition and conditionsof cooling and steam removal at the surface of the droplet. Considering metallic, substoichiometric (O/M<2.00) or stoichiometric corium with a large solidification temperatureinterval, up to now, very few experimental data exist with no parametric laws.



The void in premixing by boiling and hydrogen production was considered to be of highimportance for mitigation of steam explosion. For hydrogen production, too few experimental
data exist to propose a model and evaluate the impact on SE energetics.



The triggering from thermal fragmentation. The triggering mechanisms for melt-bottom contact
and melt spreading need clarification and characterization. Predicting triggering in a deterministic
way seems out of reach. A statistical approach seems more adequate for the inherently chaotic
phenomena.

The fine fragmentation in the process of the explosion is of utmost importance. Knowledge isglobally limited, particularly considering the effect of partial solidification and possiblesupercooling for the corium.



The mechanisms of pressurisation (micro-interaction versus direct boiling), despite recent
clarifications, further improvements are needed in understanding and convergence of views.
EVSE containment building consequences

The methodology to estimate the consequences of ex-vessel steam explosions for LWRs and BWRs andthe maximum acceptable values on the containment is different according to country specific regulatory
requirements. It can be based on probabilistic or deterministic approaches, which are not mutually
exclusive.
Today, both one-dimensional and multidimensional FCI codes can be used in the deterministic
approach. For probabilistic studies, fast running simplified tools and surrogate models have beendeveloped based on the results of the full models of the FCI codes (Grishchenko et al., 2016). The
surrogate models allow for extensive sampling to address aleatory uncertainty from a large number ofpossible severe accident scenarios and epistemic uncertainty in the modeling due to various closures. One
important point concerning the nature of triggering in the reactor pit is that it can be assumed to bestochastic under severe accident conditions. Several methods have been proposed. No physical criterion or
empirical law has been justified to determine the spatial and time localization of the triggering for exvessel configurations. Some criteria have been used (e.g. melt bottom-contact), but they are based on a16

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heuristic approach and expert judgment. It is consequently difficult to establish general bounding case for
ex-vessel steam explosion. Alternatively a statistical treatment has been proposed for the effect of thetriggering time (Grishchenko et al., 2015), although it requires significant computational efforts for theestimation of explosion loads with different triggering time for the same melt release conditions. The
method provides a comprehensive treatment and can be used for both the bounding loads and moredetailed data on distribution of the impulses.
The mechanical consequences depend on the reactor pit design (PWRs, BWRs). Even among the LWR
reactor family, important differences on the design of the reactor pit and primary circuit can significantly
affect the mechanical response to a given load. The main risks, which are primarily plant specific are:damage of internal concrete structure and potential reactor vessel detachment under the effect of highpressure in the cavity, with a risk of missile formation which can strike the containment wall. In addition,with the displacement of internal structures, the floors can strike the containment wall, and the heavymaterials (such as steam generators) can be destabilized, leading to failure of some containment
penetrations with a risk of containment radiological by-pass. It is also strongly dependent on the scenario
leading to ex-vessel steam explosion, especially the break location of the reactor vessel, the melt jet
velocity, the water level in the pit and the corium melt composition. The current knowledge is limited forsome important phenomena (e.g. oxidation) and the capabilities of the tools to handle realistic 3D
situations.
It is now possible to propose two classes for consequences of ex-vessel SE:


Direct consequences: mechanical constrain on the containment concrete, on the access doors andon the reactor vessel.

Indirect consequences: motions of elements of the reactor circuit, that can affect leak tightnessof the containment (e.g. containment penetrations for the main steam line pipes) or equipment
responsible for long term cooling of the debris.
 For the direct consequences, the main risk is the failure of concrete containment or accessdoors though imposed mechanical loads, possibly by the displacement of walls due to theimpacts. The question is then the assessment of the fragility of concrete structures against thesteam explosion loads, noting that steel rebars are present in the concrete structures. For
preliminary evaluations, it is common to make reference to the steam explosion impulse of
the load, although some studies have shown that it cannot be the only factor since themaximum pressure also has an impact. The loss of integrity of the reactor pit for PWR’s
should occur after several tens to several ten of kPa.s of impulse. For large pit BWR’s, theloss of integrity of the reactor pit might be obtained with smaller loads due to thinnerwalls. Nevertheless, the loads themselves might be smaller in larger pits due to the venting
effect. The pressure wave following the inverse square law of the pit’s radius, the distance atwhich the explosion occurs from the wall will be one of the most important parameter
regarding the reactor pit integrity. For some BWR designs, access hatches have doors that canwithstand about 6-10 kPa.s and can be reinforced up to ~20-50 kPa.s.
 For the indirect consequences, the main risk concerns the displacement of heavy elements
present in the containment as: Reactor Pressure Vessel (RPV), concrete floors, cross pipes,
Steam Generator. Deformation/rupture and entrainment of Reactor Coolant System (RCS)
piping enters in this category. In the range of 100-200 kPa.s, it can be considered that motion
start to be significant, especially for the vessel, and can cause, for certain designs, failure of
the containment integrity in the penetrations.

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In this approach, it is also important to take into account the mid-term management of a severe
accident of the nuclear reactor: if the containment failure does not come right after the explosion, steam
explosion can cause irreversible damages on the operability of equipment needed for accident management(pipes, pumps... needed for continuous water supply, containment cooling and/or venting) that can render
the situation unrecoverable for Severe Accident long term management.
In the next paragraphs, some national examples of EVSE consequences evaluation are presented.
In the United States, an early study (Theofanous et al., 1994) provided an assessment of ex-vessel SE
energetics for operating PWRs and BWRs. The reference plants in this study included all BWR and PWR
containment designs. The study concluded the steam explosion induced containment failure is not aconcern with the exception of BWR Mark III containment, where structural failure could not be excluded.
An assessment of ex-vessel steam explosions in the AP600 advanced PWR was performed (Esmaili et al.
1996) in mid-1990s. The study looked at three different scenarios for the potential of steam explosion(unsubmerged RPV, partially submerged RPV, and fully submerged RPV) and a large number ofsensitivity parameters to calculate maximum impulse loading on the cavity wall. There was a large spreadin the calculated impulse load ranging from as low as 15 kPa.s to as high as 600 kPa.s. An assessment of
ex-vessel steam explosions in the AP1000 reactor was performed (Esmaili and Khatib-Rahbar, 2004) in2004. This study was similar to the one for AP600 and resulted in impulse loads in the range between
10 kPa.s to 300 kPa.s.
In Switzerland, an assessment has been completed for BWR (Leibstadt) (IAEA, 1997) and it has
shown the possibility of destroying the pedestal by an impulse of 150 kPa.s, with consequences on thereactor containment building.
In France, evaluations where made separately by IRSN (Cenerino et al., 2013) and EDF (Markovic
and Idoux, 2011), with different conclusions. IRSN produced a PSA-2 study on PWR's reactors with linked
calculations of loads and mechanical impact with simplified 2D calculations. The considered risk was thedisplacement of floors and impact on the containment. It was concluded that the risk of failure wasconditionally important and needed to be better assessed with, in particular, more accurate mechanistic
evaluations. This was done latter on by IRSN using 3-dimensional complex mechanical models. Althoughsome conservatisms of the previous IRSN study could be pointed out, the risk of containment failure could
not be ruled out, particularly considering the uncertainty of the load evaluations and their lack of identified
conservatism. For the EPR reactor, design provisions have been implemented in the reactor pit before thevessel failure to prevent any energetic corium/water interaction.
In Japan, researches on the containment failure probability due to ex-vessel steam explosions were
performed for Japanese BWR and PWR model plants based on the model validated by FARO andKROTOS 22. In these researches, a steam explosion was evaluated assuming the external trigger at thetime of maximum premixed mass in the pedestal area (or in the suppression pool of) a BWR model and in
the reactor cavity of a PWR model plant. After SERENA-2, the evaluation model was validated based on
KROTOS and TROI and major uncertainty parameters were identified in pre-mixing and fragmentation
phases (Hotta et al. forthcoming). Output of this study has been applied in a sample evaluation of actual
plants assuming cavity injection.
These mechanical evaluations show clearly that the values of impulses and pressure of steam explosion
are too close in some cases to mechanical or motion limits for PWRs and BWRs (Gen 2 and 3), having
consequences on reactor building containment integrity. Thus, it is necessary to have a better knowledge
and modeling of ex-vessel steam explosion. An improved understanding will decrease the uncertainties onconsequences of EVSE on reactor containment and enable development of mitigation measures for EVSE.
Another point concerns the medium and long term management of the severe accident in case of SE even if
the containment is still safe; this point has never been evaluated.

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II.5

EVSE national approach after Fukushima

Over the last decade, both experimental and analytical research on the topic of SE in nuclear reactors has
greatly improved our understanding of the subject. Notwithstanding that, there is a strong desire in nuclear
safety organisations of some OECD countries to achieve a better precision in their analytical tools for anassessment of SE risk. This desire is predicated, in part, by an anticipated need to make a more robust
safety case for the operating reactors, prompted by the accidents at Fukushima.
The safety analysis practice in most OECD countries included an assessment of SE risk, albeit, withvarying levels of complexities and sophistication in assessment methodologies. With the conclusion of theOECD SERENA program (Phases 1 and 2), the nuclear safety researchers in the OECD countries felt that
improvement in predictive capabilities of various analytical tools is needed to gain more confidence incalculated steam explosion risks and safety margins.
The members of the writing group discussed the subject at length during the development of the EVSE
status paper, and determined that the regulatory bodies of the participating countries be polled forregulatory requirements concerning steam explosion.
The main observations from the survey responses are:the regulatory requirements vary between countries;
most countries, if not all, consider EVSE still an open issue, due to remaining uncertainties in large
part;the detailed methodologies for SE risk assessment vary between countries in complexities andsophistication though most of them have same basic elements – a deterministic assessment of steam
explosion loads (impulse and/or pressure) with a structural mechanics assessment of containment fragility
and a probabilistic framework to evaluate the risk of containment failure associated with a steam
explosion; and finally;
the needs for and scope of continued research on the subject vary between the countries.
The writing group of the EVSE status paper note that the events at Fukushima, with the current
knowledge and interpretation of the accidents, have no particular bearings on the likelihood of SE or its
severity. At the same time, the writing group recognizes that changes in national regulatory requirements
post-Fukushima are dictated by considerations other than the above. Further, the experts group recognizes
that in light of new regulatory requirements in some countries and for specific reactor designs, there is aneed to have an improved precision in steam explosion risk assessment, despite SE code improvements
made during SERENA-2.
Concerning the main experimental SE programs developed after Fukushima severe Accidents, we canquote the following national programs:
In France, in the frame of the French national program RSNR-Post-Fukushima supported by the
French Research Ministry, a new program called ICE (Piluso, 2015) (Water Corium Interaction, RSNRPost-Fukushima action), led by IRSN has started in 2014 and will be finished in 2019: this program isdevoted to the studies of new FCI/SE configurations (main scientific topics: oxidation, solidification, jetfragmentation), closer to real configuration with an experimental part in CEA/KROTOS facility with
prototypical corium. MC3D FCI code (IRSN) is being developed to complement the experiments and willcapitalize the knowledge gained on FCI/SE.
In Korea, The Korea Atomic Energy Research Institute (KAERI) has conducted the TROI) programfor a study on fuel-coolant interaction (FCI) since 2001: more than 70 experiments using several prototypic
materials have been performed. KAERI, as a part of national program finished in February 2017, has
carried out tests to simulate a fuel coolant interaction using ZrO2 and corium under the conditions of thereactor vessel submerged in water (Na, et al., 2016)Error! Bookmark not defined..
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In Sweden, the recent experiments undertaken at KTH using simulant materials, called PULiMS(Pouring and Under-water Liquid Melt Spreading) and SES (Steam Explosion in Stratified configuration)
investigating the effects of melt spreading on the pit floor lead to regular spontaneous explosions at highmelt superheat and moderate water sub-cooling (5-25 K) (Grishchenko, 2013). This configuration is an
important issue because in some accident scenarios (large jet, small water height), the melt should
effectively spread as a liquid on the pit floor and this melt might participate in the explosion, or trigger it.
In the frame of European program “Severe Accident Facilities for European Safety Targets” (SAFEST)(Euratom, 2014-2018), complementary experiments are planned to have a better knowledge of thisenergetic steam explosion configuration.
Concerning FCI/Steam explosion activities in 2016, it is to be noted that in Europe, the SARNET
network is still continuing in the frame of NUGENIA program, providing in particular exchanges among
the experts and promoting access to some of the facilities in Europe in the frame of the European SAFEST
program (2014-2018).
II.6

EVSE improvements and needs for reactor application

In the frame of OECD/SERENA-Phase 2 program, valuable improvements to the understanding andmodeling for Ex-Vessel SE have been obtained on the important stages of the premixing/fragmentation andexplosion phenomena. Simplified generic configurations for BWRs and PWRs reactor case (compositionof corium, vertical jet, 2D configurations) have been studied. Consistent progress on the FCI/SE
phenomena knowledge and robustness in modelling, have been obtained as described in the previous
sections of the status paper.
Nevertheless, these EVSE generic configurations are representative of a limited number of scenarios
whereas there are other configurations like e.g. for the vessel failures (central hole or lateral failure). The
exact status on EVSE has been recently provided in a common paper (Meignen et al., 2014), stressing theimportant needs and phenomena to be studied in priority for FCI/SE for application to the reactor case.
Together with the above established tables, it is possible to highlight the major requirements to fulfill theneeds for more realistic reactor applications.
Some important topics are now summarized here:
Jet orientation and fragmentation: up to now, only strictly vertical, gravity driven jets have been
investigated in experiments and for FCI/SE reactor case studies (PWR or BWR, Gen2). Nevertheless, forreactor applications, other cases have also to be considered like lateral jet fragmentation. Indeed, in case of
ex-vessel cooling strategy by natural convection (for example, some Gen 3 reactors), a probable way ofvessel failure is a lateral breakthrough of the thin upper metallic layer caused by the “focusing effect”.
Another configuration to be also considered concern the difference of pressure between the reactor pit andthe vessel: even if it is planned to have depressurization procedure in case of severe accident, it isnecessary to evaluate conditions with low remaining pressure difference between the vessel and the reactor
pit (a few bars, ejection speeds of around 10 m/s). So complementary experiments with prototypicalcorium have to be performed respecting these two configurations to validate and improve the modelingcurrently used in FCI codes.
Melt solidification: this topic was a major issue of the OECD/SERENA-Phase 2 program, through theso-called “material effect”. It has been found difficult to provide precise models combining solidificationand fragmentation during the premixing. Concerning the impact of solidification on fine fragmentation, no
models exist and the impact is based on criteria discriminant drops that should fragment and those which
are solid. The models and criterions are however facing a difficulty with a lack of experimental data, withvariations of corium compositions. Three corium composition families can be identified representative ofreactor cases: metallic, sub-stoichiometric (O/M<2.00) and oxidic with large interval of solidification.
Based on explosion simulation results one can estimate the combined effect of the melt droplets
solidification and the void in mixture on the reduction of the explosion strength in the experiments. To
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facilitate the extraction of the influence of melt solidification on the explosion strength from the combined
influence of solidification and void, it would be advantageous to perform experiments with different
solidification levels respecting similar premixing conditions.
Oxidation: past experimental programs ZREX/ZRSS (NEA, 1994) have shown the importance ofoxidation phenomena in the case of a SE with a pure metallic melt, a case representative of some particularsituations for the vessel failure due to the focusing effect. The OECD/SERENA-Phase 2 program has
shown the importance of oxidation process for sub-stoichiometric or stoichiometric melts with otherconsequences on SE (mitigation by hydrogen release). One of the major issues regarding the modeling ofoxidation in the current FCI codes is that it is still largely parametric due to an important lack of analyticalexperimental data and models related to the oxidation of a high temperature single drop or fragment. The
ICE French program (Piluso et al., 2015) was built to answer partly to this point, but this program iscovering a limited range of corium composition for reactor case application. Furthermore, integral tests
with large masses of prototypical sub-stoichiometric and metallic alloys of corium (Zr, Fe, U) will benecessary to evaluate the impact of hydrogen and chemical energy releases at large scale, and to validate
oxidation models in FCI codes.
Melt fine fragmentation: the modelling of fine fragmentation in the FCI codes is currently based either
on isothermal single drop experiments, either on experiments using simulant drops at relatively low
temperatures (1 000-1 500 °C) in a water flow (DROPS, SIGMA, DROPSG). The hydrodynamic
fragmentation of a prototypical drop in reactor conditions was never studied experimentally. Thus, possible
effects which are specific of reactor conditions, such as very high temperatures, partial solidification, andvery high pressures are currently not well known. These effects are therefore either ignored by the codes,
either taken into account but with no possibility of validation. Therefore, in order to improve the current
modelling of fine fragmentation in FCI codes, it would be necessary to study experimentally thehydrodynamic fragmentation of a single prototypical drop in reactor conditions.
Voiding in premixing: voiding in premixing is the result of two physical phenomena: vapor (andpossibly hydrogen) production due to heat and mass transfer around the melt drops, and the two-phase flow
description of the coolant (heat and mass transfer around bubbles, bubble dynamics, etc.). Heat and masstransfer around melt drops is relatively well known, but some points require improvements, such as forexample the heat fluxes distribution between vapor production and liquid heating. Improvements may beobtained by developing more detailed descriptions (modeling or direct numerical simulation, DNS), and byadditional experimental data (for single spheres or cylinders with diameters comparable to the drops inpremixing, there is no experimental data on convective film boiling for temperatures above 1 000°C).
Regarding the two-phase flow description of the coolant, the physics are well known, but the mainquestion is which level of detail is actually necessary in FCI codes in order to be able to correctly predictvoiding in large scale experiments and reactor cases. At the end, in order to validate improvements in FCI
codes related to voiding, it will be very important to have large scale experiments with instrumentation
allowing measurement of spatial distribution of void with enough precision for FCI code validation.
Mechanism of pressurisation: some important experimental data related to this topic has beenobtained recently in the TREPAM (Berthoud, 2009) experiment, which simulated heat transfers between
fragments and coolant in conditions close to an explosion in reactor conditions. Some further
improvements are however needed in order to better evaluate the void produced by a fragment, and thus
the resulting pressurization. Additional analytical experiments at small scale (one droplet) could be
performed and also some DNS calculation, which could be a promising tool for this purpose.
Triggering: it is considered at the moment as a stochastic event. This conservative assumption is notsatisfactory because it has a huge impact on SE energetics evaluation. Some experts estimated that selftriggering through thermal fragmentation with small perturbations has a low probability of occurrence forcorium. Melt-bottom contact is for some other experts a preferential criterion. Recently, at KTH (Kudinov,2014), systematic spontaneous and energetic steam explosions have been observed during the spreading
phases of simulant materials in a limited height of water, while no explosion ever occurred in the past
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DEFOR tests done with exactly the same conditions except a much larger water height (complete jet
breakup before reaching the bottom), confirming thus the tendency for self-triggering during contact withthe bottom. This tendency needs however to be confirmed for prototypical corium. This configuration for
SE has to be taken into account to understand the phenomenology and improve, at least, criteria.
Nevertheless it is necessary first to understand clearly the triggering phenomena in KTH experiments
before any extrapolation to real materials and reactor case conditions.
Melt “stratification”: it has long been considered that the stratified melt could not participate actively
in a steam explosion. However, this assertion was based on idealized situations with simulant materials.
Considering the potential of the stratified melt to increase the load and recent tests performed at KTH(Kudinov, 2014) with simulant materials (2014-2016), confirmations of a weak impact for more realistic
conditions is necessary. Experimental corium spreading on concrete under water should be considered as apossible configuration for EVSE. Nevertheless it is necessary first to understand and model the SE
phenomena involved in KTH experiments before any extrapolation to real materials and reactor caseconditions.
These highlighted topics show that there are still needs in the future for both small-scale analyticalexperiments in order to enhance the models currently used in FCI codes, and large scale experiments for amore reliable qualification and validation of the codes and also to investigate new EVSE configurations(lateral or metallic jets, stratified configurations for example).

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