Radioactivity and Environment

Research, Assessment,Transmission of Knowledge

Safety of the Geological Storage of Radioactive Waste

Safety of Installations,Accident ScenariosIonising Radiation

and Human Health

Severe Accidents and CrisisAnticipation

5 5 Analysis, prevention and management 2

5.1 Molten corium pools in the vessel bottom 4head of a pressurised water reactor (PWR) during a severe accident

5.2 Spreading of corium - Modelling and safety 9analyses for the EPR reactor corium recovery system

5.3 Fission products in the containment 13Behaviour of organic iodides

5.4 The hydrogen risk during a severe PWR accident 185.5 ASTRID - Assessing a light water reactor accident in real time 22

5.6 RECI experiments - Iodine chemistry in catalytic 28hydrogen recombiners

5.7 SOURCE TERM programme 29

5.8 EPICUR - A new installation to study the volatility of radionuclides in the containment in severe accident situations

5.9 SARNET - A European network of excellence for severe 30accidents

5.10 IRSN/FzK cooperation on direct heating of containment

5.11 Initial results of PHÉBUS FPT3 experiment 31

5.12 Key dates - Theses vivaed 32Position of authors in the IRSN organisational chart

Severe Accidents andCrisis Anticipation

newsflashnewsflashnewsflashnewsflashnewsflashHead office77-83, avenue du Général-de-Gaulle92140 Clamart - FranceRegistered under Nanterre RCS B440 546 018

Telephone+33 (0)1 58 35 88 88

Postal addressB.P. 1792262 Fontenay-aux-Roses Cedex - France

Web sitewww.irsn.org

200Scientific and Technical Report

5

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The IRSN 2005 Scientific and Technical Report offers a summary of the scientific projects from 2004-2005 which reached a milestone in their work. This documentcould not have been prepared without the collaboration of the Editorial Committee and the researchers and specialists who devoted their time and energy to draftingand finalising these articles. We would like to thank each and everyone of them for their contribution.

The 2005 Scientific and Technical Report comprises 6 fascicules and is printed on 100% recyclable and biodegradable, chlorine-free coated paper using vegetable-based ink.

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IRSN - 2005 Scientific and Technical Report 1

5Severe Accidents and Crisis Anticipation

Analysis, prevention and management

he probability of a severe accident leading to reactor core meltdown is low because the

occurrence of such an accident would require the combination of a number of failures, in

particular safety system failures. However, this low probability is offset by the gravity of the

consequences if a release of radioactive substances into the environment were to occur. That

is why the IRSN placed a great deal of emphasis on the study of this type of accident.

Prevention firstly depends on the design of the reactor and on the existence of safety

systems. Accident management systems are implemented by the operator in order to limit the

consequences of accidents: this is in-depth defence. In this way, the new-generation European

pressurised water reactor (EPR) will be equipped with a recovery system designed to contain and

cool the corium(1) in order to preserve the integrity of the containment.

In addition to these steps, emergency plans and medium and long-term strategies for

the management of post-accident situations are defined to cover any accident which could

lead to radioactive contamination of the environment.

In the context of its research and expert survey work, IRSN studies the phenomenology of severe

accidents, the preventive measures that are or may be implemented and the management of crisis

situations which may be caused by such accidents. This work is described in the five articles in this

chapter. Two of these chapters deal with the knowledge acquired and applications regarding the

behaviour of corium when it reaches an accident-stricken reactor’s vessel bottom head and when

it spills into a recovery system of the type that will be used on the EPR reactor. A third article gives

the results of a research programme to study the transfers of mass and heat occurring in the

5

Severe Accidents and Crisis Anticipation2

T(1) Mixture of molten materials in the

reactor core.

(2) Emergency containment safeguardprocedure in case of accidentalpressure build-up.

(3) Gesellschaft für Reaktorsicherheit(German institute for reactorsafety).

5containment and their consequences for the

distribution of the hydrogen emitted during

core degradation. The underlying source of

preoccupation is the risk of loss of containment

integrity following violent combustion of the

hydrogen. The fourth article looks at iodine as

its 131I isotope constitutes the main short-term

radiation risk for populations. The behaviour of

organic iodines are examined, in particular.

These iodines are not stopped by the filtering

systems used in French containments in case

of venting of the containment(2).The knowledge

acquired through research programmes into

severe accidents has been integrated into

numerical simulation tools such as the ASTEC

software, jointly developed by the IRSN and

the GRS(3) since 1995. It has multiple fields of

application, covering:

physical analyses of scenarios to better

understand the phenomenology and assess

possible radiological releases outside the

containment;

accident management studies concerning

measures for the prevention and limitation of the

consequences of severe accidents and support

for the preparation of crisis management;

support for experimental programmes.

The implementation of this software and its

relative slowness are not, however, compatible

with the decision-making process in crisis situa-

tions. Simpler and faster tools are therefore

required. For this purpose, IRSN is developing,

on the basis of European cooperation, the

ASTRID code, which is described in the fifth and

last article in this chapter.

3IRSN - 2005 Scientific and Technical Report

Jean-Claude MICAELLI,Assistant Director

Department for the Prevention ofMajor Accidents

5.1 Molten corium pools in the vessel bottom headof a pressurised waterreactor (PWR) during a severe accident

The formation of a corium(1) pool in the vessel bottom head is a critical phase of a core meltdown accident in a pressurised waterreactor (PWR). Indeed, the heat flux required at the interface

between the molten pool and the vessel in order to remove the residualheat can exceed 1 MW/m2 in this situation, so directly threatening thevessel’s integrity even when there is cooling from the outside.It should be remembered, however, that some safety analyses have shownthat a flux such as this cannot be reached in reactors with a power of lessthan 600 MWe (AP600, VVER 440) as it is then possible to cool the vesselfrom the outside.

The risks of loss of containment integrity become high after failure of the vessel: ablation of the reactor

pit concrete by the corium, rapid heating of the atmosphere and dispersion of radioactive elements.

Molten corium pools have been the subject of research and studies since the end of the 1970s and

they remain topical in spite of the remarkable progress made in understanding and modelling the

physical and chemical phenomena that come into play. The processes which lead to the formation

of a molten pool and damage to the vessel are rather complex and are only partially understood at

the present time. They remain as difficult to study experimentally as they are to model. This difficulty

is due, in particular, to the large size of the pool (in a very turbulent state that cannot be reproduced

at reduced scale), phase changes (solidification or melting, separation into immiscible liquids) and

transfers of mass caused by the existence of thermochemical imbalances in the pool.

Owing to these gaps in our understanding of the phenomena, the assessment of the safety of some

new reactor concepts (AP1000, VVER640) in order to ensure that the corium is retained inside the

vessel(2) are subject to a great deal of uncertainty.

In the rest of this article, current knowledge of the behaviour of corium in a vessel is briefly

reviewed, with a description of the arrival of the corium in the vessel bottom head and the

formation of the molten pool, and a closer look at heat and mass transfers in the pool, in particular.

The main points are summarised at the end of this article.

Arrival of corium and formation of the molten pool

It is generally assumed that, when the corium reaches the area of the core near the vessel bottom

head, the latter is filled with liquid water. In all likelihood, the corium contains a large fraction of

non-oxidised zircaloy (cladding material). It is described as sub-stoichiometric as long as its composition

Severe Accidents and Crisis Anticipation4

(1) The word “corium” refers to themixture of molten materials (UO2,ZrO2, Zr, steel) which is formed bythe meltdown of fuel rods andcontrol rods.

(2) This is not the case for the PWRreactor.

5

is not (U-Zr)O2, and it is therefore prone to oxidation. The interac-

tion of corium at over 2,500 K with water at saturation level causes

fairly fine fragmentation of the corium jet in solid particles and heavy

production of steam which may lead to a considerable pressure

increase in the primary system. This phenomenon has been quantified

by means of experimental programmes such as FARO [1] (ISPRA).

The interaction produces particles of various sizes between 1 and 5

millimetres. One of the important questions raised, apart from the

size of the particles formed, is knowing whether the corium is also

oxidised when it is fragmented. This defines the production of

hydrogen associated with the oxidation reaction and governs what

happens to the corium, as will be seen in the fourth part of this

chapter. Only the ZREX experiments [2] (SNL) have been able to

provide a partial answer to this question. It appears that, if there is

no steam explosion, the fragmentation is not fine enough to lead to

significant oxidation of the debris. However, too few experiments

have been conducted to correctly quantify this phenomenon, and

this type of study has not been repeated owing to the excessively

high risks it involves.

Modelling of the fragmentation process is very complex and remains

relatively empirical. This is, therefore, a rather unpredictable stage of

the accident. When partially fragmented corium has accumulated in

the vessel bottom head, it gradually vaporises the residual water.

Assuming that no additional water is supplied or that the configuration

of the debris is such that it cannot be cooled, the temperature of the

materials gradually increases to the vicinity of 1,700 K, which is the

melting temperature of steel. A large quantity of molten steel from

the many structures (plates or tubes) fitted in the vessel bottom

head is gradually incorporated in the corium. Although the interaction

between liquid steel and (U-Zr)O2 corium have been the subject of

studies for a long time, it is only recently, in the context of the OECD

MASCA project [3] in particular, that the impact of these interactions

on the evolution of the corium in the vessel bottom head has been

realised.This aspect will be described in detail in part 4 of this chapter.

Florian FICHOTSevere Accident Study and Simulation Laboratory

IRSN - 2005 Scientific and Technical Report

5.1

Figure 1: Results of the RASPLAV AW-200-4 experiment with the ICARE code: formation of a pool from debris (the scale represents the volume fraction of molten materials, from blue –debris– to red –pool).

0.01Elevation (m)

-0.40-0.5 0.5

R (m)

RASPLAV AW200-4 liquid fraction

0

0.3

0.6

1

Moltenfraction

Severe Accidents and Crisis Anticipation6

Zircaloy and then metal oxides melt as the temperature rises and

accumulate to form a pool. Bearing in mind the results of tests,

ACRR-MP [4], PHEBUS FPT4 [5] and RASPLAV AW-200 [6], this

phenomenon is now fairly well known and modelled. This is illus-

trated by figure 1 (page 5) which shows the result of the calculation,

performed using the ICARE2 code [7], of the formation of a pool

from debris (UO2, ZrO2, Zr). The comparison with observations and

experimental measurements is highly satisfactory.

Flow regimes and heat transfers in a molten pool

As the pool is cooled by contact with the vessel and through its

upper surface by radiation or contact with water, natural convection

movements take place. Initial studies on molten pools mainly

concerned the characterisation of the convective regimes and the

assessment of heat exchanges with the vessel. Many experimental

devices, up to nearly full scale but on a semi-cylindrical section, have

correlated the heat flux with the Rayleigh number(3) inside the pool

for all the configurations of interest for the study of severe accidents.

Experiments with simulating materials have proved the most useful

as they make it possible to reach high Rayleigh numbers with complete,

reliable instrumentation.This has also made it possible to experimentally

characterise the turbulent transfer processes occurring in large pools.

The structure of the flow in a corium pool is characterised by three

characteristic areas which are shown diagrammatically in figure 2.

A boundary layer along the vessel is the seat of a downward flow.

This layer, which is very thin at the top of the pool, gradually widens

towards the bottom. It is characterised by a high temperature gradient

in the direction perpendicular to the flow. The central region of the

pool is the seat of an upward flow of very low velocity (on average)

for which the predominant transport mode is small-scale turbulent

diffusion. The temperature range is axially stratified in this region.

This causes an axial variation of the heat flux on the vessel, with the

maximum being located at the top of the region. Due to the effect of

the cooling of the upper surface, a layer several tens of centimetres

thick at the top of the pool is the seat of large-scale instabilities of

the Rayleigh-Benard type. In this layer, there is a lot of transport by

convection cells and it is often considered to be isothermal except

for a thin boundary layer on the upper surface. The heat flow at the

edge of this region is thus relatively uniform.

Many experimental programmes have been conducted on this subject:

Asfia & Dihr [8], ACOPO [9] on hemispherical geometry and COPO

[10], SIMECO [11], RASPLAV [6], BALI [12] on a semi-cylindrical section.

It should be noted that there are divergences in the various heat transfer

correlations and these become fairly significant for high Rayleigh

numbers. A great deal of care must therefore be taken when choosing

correlations: preference should be given to recent correlations proposed

by some authors [13] who have attempted to summarise large-scale

experimental results. In spite of this, exchange coefficients still differ

by a factor of between 2 and 4. However, most safety analyses have to

be based on these correlations which give the right order of magnitude

for heat fluxes. It will be seen, in the next section, that there may be a

number of stratified layers of material instead of a single homogeneous

pool. It should be noted however that, apart from SIMECO, there are

practically no experiments on stratified pools with two or three layers

and that the initial results do not appear to justify the application of

the correlations obtained for homogeneous pools. This aspect therefore

requires further experimental study.

Mass transfers and chemical interactionsin a molten pool

The complexity of the behaviour of molten pools is increased by the

numerous chemical interactions between materials at high temperature,

which can radically change the state and transfers of those materials.

As a consequence, it is extremely difficult to predict the movements of

materials in the reactor vessel and, therefore, the power distribution.

The effect of the actual materials was recently studied during the OECD

RASPLAV [6] and MASCA [14] programmes. It was then noted that, in

some conditions, substantial transfers take place in the pool, resulting

in stratified configurations which are rarely taken into account by the

conventional “envelope” approach. This will be discussed hereinafter.

Unstablezone

Stratifiedzone

Figure 2: Diagram of convective movements in a turbulent pool.

5.1

(3) The Rayleigh number is defined by the formula

RaR =

where R is the diameter of the pool, ql is the power density,βl is the thermal expansion coefficient of the mixture, λl andαl are the thermal conductivity and diffusivity respectively,and vl is the kinematic viscosity. This dimensionless numbercharacterises the magnitude of the work of the Archimedesforces compared with the dissipation of energy by viscosityand thermal diffusion.

g βl ql R5

αl vl λl

7

Among the important conclusions obtained by MASCA [14], it should

be noted that the addition of a small quantity of steel, in a proportion

of less than 20%, to a sub-stoichiometric corium (U-Zr-O) leads to

the formation of two immiscible liquids with a metallic phase that is

denser than the oxide phase. It has been shown that the metal layer

actually contains a non-negligible proportion of zircaloy and uranium,

with the latter significantly modifying the mixture’s density. Although

this result is perfectly explicable and even predictable, it has only rarely

been taken into consideration in previous studies on vessel resistance.

One of the parameters defining the density of the metallic phase

formed is the initial proportion of non-oxidised zirconium, i.e. the sub-

stoichiometry of the initial corium. The density of the metal formed is

proportional to a decreasing function of the initial sub-stoichiometry.

It is therefore important to note that “inverse” stratification phenomena

were only noted in experiments conducted in an inert atmosphere.

It is easy to assume that the sub-stoichiometry of the corium may

decrease in an oxidising atmosphere, which would have the effect of

reducing the density of the metallic phase and, perhaps, making it

lighter than the oxide. The MASCA-2 programme which is currently

in progress has the purpose of studying the behaviour of a corium

pool (as well as debris) in an oxidising atmosphere. However, there

are not yet enough properly analysed results to put forward any

conclusion on this point at present.

It should be noted that the MASCA programme [14] has played a part

in improving the NUCLEA thermochemical database [15] developed

on the initiative of the IRSN. We now have sufficiently accurate data

to be able to calculate the compositions of the oxide and metallic

phases in equilibrium in the immiscible domain. These compositions

have a key role in the modelling of a stratified pool when it must be

considered that the metal-oxide interface is in thermochemical

equilibrium. In previous modelling approaches, only thermal equilibrium

at the interface was taken into account. It may therefore be considered

that knowledge acquired recently in the course of experimental

programmes on the interactions of materials at high temperature have

led to significant development in the modelling of the molten pool.

Safety analyses and modelling

The usual trend for safety analysis consisted in simplifying the problem

to be studied by referring to a set of fairly simple envelope situations

which, it was believed, could be used to obtain a conservative estimate

of the thermal stresses placed on the reactor vessel. In this type of

study [16,17], it is generally accepted that, after fragmentation, the

corium is found in the vessel bottom head in the form of a bed of

debris and an unfragmented mass (pool) whose relative proportions

may vary within quite a wide range. Similarly, the rate of oxidation of

the debris and corium after fragmentation is assumed to vary over

quite a wide span (between 30% and 100%). On the basis of the

assumption that the greatest heat fluxes are obtained when all the

materials have melted, the transient period during which the actual

pool is formed is often modelled in a very simplified manner. This

choice appears reasonable at first sight but one must bear in mind

that it also means overlooking the transport of liquid steel in the

oxide debris and the interaction of metals with the debris when they

melt. We have seen, however, that it is precisely those interactions

which lead to stratified configurations that are rarely taken into

consideration in the conventional envelope approach, in which it is

nearly always considered that the metal layer floats over the oxide.

The latter case is the most conservative as it entails a local increase in

the heat flux (focusing effect) on the reactor vessel when the thickness

of the metal layer decreases. It is for this reason that in conventional

models of the vessel bottom head, the steel never mixes with the

corium and systematically forms a layer at the top of the pool of oxides.

Consequently, the mass transfers during formation of the pool and

during the melting of steel structures entail the possibility of transient

situations during which the heat fluxes on the reactor vessel could

be greater than those estimated for envelope situations. We can also

envisage scenarios in which the gradual addition of molten steel

(produced by the melting of the lower plenum structures) to the

pool may initially form a metal layer under the oxide pool. As the

proportion of steel increased, this layer would then become less

dense and would rise to the top of the pool. This type of transient

configuration is illustrated in figure 3 which shows a thin metal layer

that appears temporarily on the top of the oxide pool, constituting a

risk of local increase in the heat flux (focusing effect).

IRSN - 2005 Scientific and Technical Report

5.1

Metal layer

Oxide pool

Metal pool

Steam

Figure 3: Diagram of a transient configuration liable to be established duringthe development of a corium pool in a vessel bottom head.

Severe Accidents and Crisis Anticipation8

The conventional “envelope” approach is therefore now seriously

questioned. At the IRSN and other organisations, we are turning

towards more detailed modelling of the phenomena and simulta-

neously considering transfers of mass and heat in the pool, unlike the

conventional assumption where there is no interaction between

materials. Furthermore, models are being developed to deal with

changes in the materials over time rather than, as was previously the

case, considering a “final” steady state. This change of approach is

especially useful when studying current French pressurised water

reactors. Indeed, if the reactor vessel cannot be sufficiently cooled

from the outside it would be likely to rupture before the materials

melted completely and a steady state was established. The aim of

studies is then to predict the conditions in which reactor vessel rupture

would occur (time and location) as well as the mass, temperature

and composition of the corium that would come out.

If we want to achieve greater accuracy for the calculation of heat

fluxes, a 2D or 3D numerical simulation of the flow –CFD codes(4)–

taking turbulence into account must be performed. The existing

method of modelling turbulent flows in natural convection of a fluid

generating power internally is clearly inadequate, especially as it is

difficult, in conventional turbulence models (i.e. isothermal), to allow

for the effect of gravity terms on the generation or reduction of

turbulent kinetic energy. This is particularly true in the case of simple

models such as K-ε. At present, no existing model appears to be usable

in a mechanistic code for the simulation of severe accidents: models

which can give accurate results require excessively long computing

times. It may however be noted that some foreign organisations in

Russia and Japan are already using CFD codes for the study of reactors

by studying only the vessel bottom head, separately from the rest of

the reactor vessel. The accuracy and validity of these models has not

yet been proved.

Conclusion

After numerous experimental studies it was possible to understand

and to model, more or less satisfactorily, the formation of a pool

in the vessel bottom head, the pool’s thermo-hydraulics and the

interactions between materials. Now, the conventional envelope

approach has been seriously called into question and this has led

to the use of new models capable of calculating a pool’s transient

changes in order to tackle the problem. This is the case for models

currently being developed in the context of ASTEC [18] and

ICARE/CATHARE [19].These models also differ from earlier models

by taking into consideration the coupling of thermo-hydraulic

and thermochemical phenomena.This has become possible as a result

of a significant improvement in the quality of thermochemical

databases. Experimental research is still being conducted on this

subject in the context of projects such as MASCA [14] (second

phase), which aims to study the behaviour of a corium pool in an

oxidising atmosphere, and SIMECO [11] whose purpose is to study

the thermo-hydraulics of stratified pools.

5.1

References[1] D. Magallon. The FARO program recent results and synthesis. Proceedings of {CSARP} Meeting, Bethesda (USA), 1997.[2] D.H. Cho, D.R. Armstrong, W.H. Gunther, S. Basu. Experiments on interactions between zirconium-containing melt and water (ZREX): hydrogen generationand chemical augmentation of energetics. Proceedings of JAERI Conference, 97-011, (Japan), 1997.[3] F. Fichot, J.M. Seiler, V. Strizhov. Applications of the OECD MASCA project results to reactor safety analysis. MASCA Application Report,OECD-NEA, 2003.[4] B.D. Gasser, R.O. Gaunt, S. Bourcier. Late phase melt progression experiment MP-1. Results and Analyses, NUREG/CR-5874, SAND92-0804, 1992.[5] P. Chapelot, A.C. Grégoire, G Grégoire. Final FPT4 Report, DPAM-DIR 2004-0135, PH-PF IP-04-553, 2004.[6] V. Asmolov, at al. RASPLAV Application Report. OECD RASPLAV Seminar, Munich (Germany), 2000.[7] F. Fichot, V. Kobzar, Yu. Zvonarev, P. Bousquet-Mélou. The use of RASPLAV results in IPSN severe accident research program. RASPLAV Seminar 2000,Munich (Germany), 2000.[8] F. Asfia et V.K. Dihr. An experimental study of natural convection in a volumetrically heated spherical pool bounded on top with a rigid wall. NuclearEngineering and Design, 163, 1996.[9] T.G. Theofanous, M. Maguire, S. Angelini et T. Salmassi. The first results from the ACOPO experiment. Nuclear Engineering and Design, 169, 1997.[10] O. Kymäläinen, O. Hongisto et E. Pessa. COPO experiments on heat transfer from a volumetrically heated pool. IVO Process Laboratory,DLV1-G380-0377, 1993.[11] A.A. Gubaidullin, T.N. Dinh, B.R. Sehgal. Analysis of natural convection heat transfer and flows in internally stratified liquid. Proceedings of 33rd NationalHeat Transfer Conference, Albuquerque (USA), 1999.[12] J.M. Bonnet, J.M. Seiler. Thermalhydraulic phenomena in corium pools: the BALI experiments. Proc. of 7th Int. Conf. on Nuclear Engineering, Tokyo (Japan),1999.[13] J.M. Bonnet, J.M. Seiler. In-vessel corium pool thermalhydraulics for the bounding cases. NT SETEX/LTEM/01-247, CEA Grenoble, 2001.[14] V. Asmolov at al. MASCA Synthesis Report, OECD Report, 2005.[15] V. Chaud, P.Y. Chevalier, B. Cheynet, E. Fischer, P. Mason, M. Mignanelli. Contributions in final ENTHALPY report, ENTHA(03)-P018, EC 5th FrameworkProgram, 2003.[16] O. Kymäläinen, H. Tuomisto, T.G. Theofanous. In-vessel retention of corium at the Loviisa plant. Nuclear Engineering and Design, 169 pp. 109-130, 1997.[17] R.E. Henry, H.K. Fauske. External cooling of a reactor vessel under severe accident conditions. Nuclear Engineering and Design, 139, pp. 31-43, 1993.[18] S. Pignet, G. Guillard. Modeling of severe accident sequences with the new modules CESAR and DIVA of the ASTEC system code. NURETH-10, Seoul(South-Korea), 2003.[19] M. Salay, F. Fichot. Modelling of corium stratification in the lower plenum of a reactor vessel. MASCA Seminar, Aix-en-Provence (France), 2004.

(4) CFD: Computational Fluid Dynamics.

9IRSN - 2005 Scientific and Technical Report

5.2 Spreading ofcoriumModelling and safetyanalyses for the EPRreactor corium recoverysystem

The new-generation EPR pressurised water reactor is equippedwith an innovative safety feature: the corium recovery system.This compartment with a floor area of nearly 2 m2 is designed

to collect the corium in case of a perforation of the reactor vesselduring an accident involving core meltdown, in such a way that thecorium layer is thin enough for it to be cooled. In these conditions,it is clearly paramount to be able to reliably predict that the coriumwill spread in the recovery system under its own weight according to,in particular, its initial temperature and its flow rate. This question wasthe subject of a modelling programme conducted at the IRSN which ledto the development of the CROCO software, validated by comparingresults mainly obtained from the CORINE, KATS and VULCANOexperiments [1]. This programme is described in the first three parts ofthis article. The studies conducted at the IRSN to analyse the recoverysystem’s behaviour are briefly referred to in the fourth and last section.

5.2Fabrice BABIK, Michel CRANGA,

Jean-Claude LATCHE, Bénédicte MICHEL,Bruno PIAR, Didier VOLA

Applied Mathematics and Corium Physics Laboratory

Experiment VEU7: flow of corium on a ceramic substrate (left) and on a concrete substrate (right).

Severe Accidents and Crisis Anticipation10

Physical modelling of corium spreading

The spreading of the corium involves an extremely complex flow

which is the seat of coupled physical phenomena, the main ones being

solidification on the surface due to exchanges through radiation, the

erosion of the substrate (concrete), and chemical reactions between

the components of the corium and the concrete.The increase in tempe-

rature and then the erosion of the concrete also lead to the release of

water vapour and carbon dioxide which pass through the flow.

The method chosen to model this flow deals with the two phase assem-

blies separately. In the first, the corium liquid and solid phases are grouped

together in an equivalent continuum, and the dispersed gaseous phase is

isolated in the second. In the case of the latter, the balances of quantities

of motion and energy are simplified, and the inertia terms are ignored.

To develop the model describing the behaviour of the “corium phase”,

we must proceed in a number of steps [2], as follows.

The starting point is the definition of a local model for an incom-

pressible flow reacting to several components. Elementary components

with identical mass diffusivity are grouped in entities described as

macrocomponents, whose transport equations are processed. The

chemical reactions are then taken into account by applying the

assumption of thermodynamic equilibrium. The reaction terms in the

energy equation are then replaced by an assessment of the enthalpy

in relation to the composition. The latter operation is performed by

means of an external tool using a thermodynamic database (NUCLEA)

developed with CNRS (THERMA Laboratory in Grenoble) [3].

By using the averaging technique to formulate the balance equations

for the liquid-solid mixture and, in particular, assuming the phase

velocities to be equal, the stresses at macroscopic level can be written

in the usual form for a Newtonian liquid. An effective viscosity then

appears and this must be calculated by means of empirical correlations

drawn from literature. As a generalisation, effective non-Newtonian

laws of behaviour are also envisaged.

Then, two types of fluctuation on the microscopic level must be taken

into consideration: turbulence, in the usual sense of the word and

disturbances related to the gaseous bubbles passing through the flow.

These two phenomena are taken into account simply by supposing

that they result in additional diffusions, like mass, energy and quantity

of motion which are evaluated by means of empirical relations.We can

then refer to “algebraic closure” for the turbulence.

The model obtained in this way is complex to deal with on the nume-

rical level: indeed, it includes Navier-Stokes equations. Furthermore,

the geometric characteristic of the flow, i.e. its small vertical dimension

compared with the horizontal dimensions, results in very stretched

meshes. The latter characteristic is put to good use: an asymptotic

analysis, with respect to the ratio between the horizontal and vertical

dimensions, of the terms in each of the conservation equations,

shows that some of them are negligible. If they are eliminated, it is

possible to calculate the pressure directly (the dynamic pressure is

taken as being equal to the hydrostatic pressure). Then, the quantity

of motion equation can be decoupled, and this makes it considerably

simpler to reach a solution. The mass and energy balances remain

unchanged.The two equation systems –Navier-Stokes and the simplified

model– are processed with the simulation software developed for that

purpose, the CROCO code. Calculations show that the simplified

model gives results that are practically identical to the full model but

in much shorter computing times when the ratio between the horizontal

and vertical dimensions is less than 0.1.

5.2

Figure 1: Experiment VEU7: flow on ceramic substrate. Temperature distribution in the corium calculated with the CROCO code.

Height (m)

(m)VE_U7 on ceramic

Time (s)

1,100

1,200

1,300

1,400

1,500

1,600

1,700

1,800

1,900

2,000

2,100

2,200

2,300

2,400

2,500

0

0.01

0.02

0.03

0.04

0.05

0.75 0.85 0.95 1.05 1.15 1.25 1.35

11

Numerical solution

In general, numerically solving problems featuring a free surface is

difficult. For the simulation of the spreading of corium, this problem

is further complicated by large variations in the domain occupied by

the fluid and by the presence of boundary layers on the surface

owing to exchanges with the atmosphere through radiation. The first

aspect leads us to favour a spatial discretisation using a fixed mesh.

The scheme must then meet the two following requirements:

It must provide natural discretisation of the derivatives with

respect to time, especially in the areas of the mesh where the fluid

was absent at the previous time step and where, therefore, a standard

finite difference formula is not applicable.

It must maintain optimal precision in the vicinity of the free

surface.

This problem calls for the development of an original numerical

method [4], of the Galerkin transport type with the following principal

characteristics:

the free surface is discretised independently from the mesh and

transported by a Lagrangian method;

the time marching is based on decoupling the mathematical

operators into a transport operator and a diffusion operator;

transport is processed by the characteristics method;

the diffusion stage is discretised by a Galerkin technique, in which

the approximation spaces are the finite element spaces associated

with the mesh and the volume integrations are applied on the fluid

domain, i.e. limited by the discrete free surface.

The utilisation of non-Newtonian laws of behaviour in the diffusion

stage led to specific developments [5, 6].

Verification and validation

The first stage after developing a code or part of a code consists in

verifying the equations for the mathematical model. The solution is

known for a number of simplified cases, such as the problem of

conduction, advection-diffusion and Navier-Stokes equations. This

solution is established either analytically or using reference calculations

from international literature. In addition, approximate solutions are

available for spreading transients in extreme flow regimes:

For a null Reynolds number, like solutions that are asymptotically

valid for long times can be explicitly calculated for planar or axisym-

metrical flows with constant or variable total mass (constant input

flow, for example).

For planar flows of low viscosity, the evolution of the free surface

and of the flow rate through a vertical section can be obtained for

specific initial conditions by solving the Saint-Venant equations

analytically.

The CROCO code dealt successfully with all these test problems.

The validation stage then consists in verifying that the physical

model represents the actual situation correctly. For the spreading of

corium, the questions raised essentially concern the following:

the assumption of equality of the velocities of the liquid and solid

phases (without macrosegregation);

the empirical closing laws for the effective viscosity in partially

melted zones and turbulent diffusions, especially when due to a

stream of bubbles passing through the flow.

Quite logically, this leads to the following validation matrix:

Firstly, the code is compared with analytical experiments, featuring

solidification and without gas, where the physical variables of the

flow can be accurately characterised. This is the purpose of the first

series of experiments in the CORINE programme, conducted for IRSN

by CEA/DEN in Grenoble [7].The corium simulant solidifies on contact

with the substrate. Temperature variations with respect to time and

height (one measurement point per millimetre) are measured at

specific points in the flow. The KATS experiments, conducted at FzK(1)

[8], extend this study in the sense that the simulant used –thermite–

is only liquid at high temperature, so that solidification on the surface

due to exchanges through radiation can be observed. The comparison

of the temperature distributions obtained during the CORINE and

KATS experiments show that these flows correspond to two separate

regimes. In the CORINE experiments, the characteristic hydrodynamic

times are long compared with the thermal times and the isothermal

surfaces are almost vertical. These times are inverted in the KATS

experiments and cooling results in the appearance of thermal boundary

layers on the surface and in contact with the substrate.The first regime

occurs, in the case of the reactor, for spreading on a substrate of

IRSN - 2005 Scientific and Technical Report

5.2

Figure 2: Experiment VEU7: flow on ceramic and concrete substrates. Advance ofcorium with respect to time, measured and calculated by CROCO.

Position of front (m)

VE_U7 on ceramic Time (s)

0

0.1

0.2

0.3

0.4

0.5

0 5 10 15 20

On concrete (experiment)

On ceramic (experiment)

On concrete (calculation)

On concrete, degassing nottaken into account (calculation)

On ceramic (calculation)

(1) FzK: Forschungszentrum Karlsruhe – German institute ofnuclear energy techniques and Karlsruhe research centre.

Severe Accidents and Crisis Anticipation12

refractory material, while the second corresponds to spreading at

low flow rate on concrete with the diffusion of heat being increased

by bubbles.

Secondly, in a similar approach, correlations of turbulent diffusion

are validated by comparing them with the series of CORINE experi-

ments with bubbling, and with or without solidification, and then

with the KATS spreading tests on a concrete substrate.

In addition, the VULCANO experimental programme, in which the

material used is of the same composition as corium and the substrate

consists of either zirconium or concrete, comprises experiments

representing all the phenomena. It can therefore be used to obtain

overall confirmation of the modelling. Figures 1 and 2 (pages 10

and 11) show examples of experimentation and calculation.

The experiments confirm that there is no macrosegregation between

the liquid and solid phases and, allowing for the fact that the occurrence

of this phenomenon depends on thermophysical properties that are

often inadequately characterised (mass diffusions in a flow comprising

several components), this conclusion is strongly backed up by the

inclusion of experiments on real materials in the test matrix.

On completion of these verification and validation processes (see [7]

for a partial summary), the CROCO code is considered to have been

validated for safety analyses.

Studies concerning the EPR reactor’s corium recovery system

The confinement of corium to within the EPR reactor containment is

potentially obtained at the end of a scenario involving several stages.

At first, the corium is retained in the reactor pit. It erodes the

concrete floor and, then, the steel plate separating the reactor pit

from the spreading chamber. When this plate is perforated, the

corium flows into the recovery device. It then reaches detectors

which trigger the release of a store of water into the cooling system.

First of all, channels located under the floor of the recovery system

are flooded. The water then reaches the surface of the corium, a few

minutes after the corium comes into contact with the detectors.

As the spread of corium under water cannot be guaranteed, it is

essential that this should have happened before this moment. This

matter has been the subject of studies, the results of which are

briefly described herein.

Firstly, it is demonstrated that complete spreading is only possible if

the break opened by the corium in the steel plate is of a minimum

area. To do this, the relation between the size of that area and the

rate of spillage of the corium is ascertained by means of a simple

analytical model based on the Bernoulli equation. An initial series of

spreading calculations carried out using the CROCO V2 code then

shows that the detectors, wherever they are placed, can be reached

by the corium front in a much shorter time than the spillage time.

For example, in the case of a break equivalent to a diameter of only

20 cm, the spreading front can reach the end of the recovery device

in less than one minute. It is therefore impossible to count on a signi-

ficant delay between the beginning of the spillage and the moment

the detectors are reached. The spillage of corium into the recovery

device must be shorter in duration than the rising of water in the

cooling device, and this constraint governs the minimum area of the

break in the steel plate.

In a second series of calculations, it is verified that spreading of the

corium is confirmed for the minimum break area previously determined.

This study makes allowance for unfavourable assumptions. In particular,

it applies initial temperature values at which the corium is only partially

melted and consequently very viscous. In spite of this, the corium

front reaches the end of the spreading chamber in a few minutes.

This study concludes that the recovery system operates satisfactorily,

provided that the rate of spillage of the corium into the recovery

device is sufficient which is only ensured if the erosion, by the corium,

of the steel plate separating the reactor pit from the recovery system

results in a large enough opening.

5.2

References[1] C. Journeau, E. Boccaccio, C. Brayer, G. Cognet, J.F. Haquet, C. Jégou, P. Piluso, J. Monerris. Ex-vessel corium spreading: results from the VULCANO spreadingtests. Nuclear Engineering and Design, vol. 223, pp. 75-102, 2003.[2] B. Piar, B.D. Michel, F. Babik, J.C. Latché, G. Guillard, J.M. Ruggieri. CROCO: a computer code for corium spreading. 9th International Topical Meeting onNuclear Reactor Thermal Hydraulics (NURETH 9), San Francisco, 1999.[3] M. Barrachin, M. Salay, F. Fichot. The use of a thermochemical database to model stratification in a molten pool during a severe accident. STNM11,Karlsruhe, 2004.[4] D. Vola, F. Babik, J.C. Latché. On an numerical strategy to compute gravity currents of non-Newtonian fluids. Journal of Computational Physics, vol. 201,pp. 397-420, 2004.[5] D. Vola, L. Boscardin, J.C. Latché. Laminar unsteady flows of Bingham fluids: a numerical strategy and some benchmark results. Journal of ComputationalPhysics, vol. 187, pp. 441-456, 2003.[6] J.C. Latché, D. Vola. Analysis of the Brezzi-Pitkaranta stabilized scheme for creeping flows of Bingham fluids. SIAM Journal on Numerical Analysis, vol. 42,pp. 1208-1225, 2004.[7] B.D. Michel, B. Piar, F. Babik, J.C. Latché, G. Guillard, C. de Pascale. Synthesis of the validation of the CROCO V1 spreading code. OECD Workshop “Ex-VesselDebris Coolability”, Karlsruhe, Nov. 1999.[8] H. Alsmeyer, G. Albrecht, G. Fieg, U. Stegmaier, W. Tromm, H. Werle. Controlling and cooling core melts outside the pressure vessel. Nuclear Engineering andDesign, vol. 202, pp. 269-278, 2000.

13IRSN - 2005 Scientific and Technical Report

5.3 Fission products in the containmentBehaviour of organiciodides

During a severe accident with partial core meltdown in a pressurised water reactor, radioactive materials such as iodine,rare gases, strontium and caesium are released by the fuel and

carried in the form of aerosols or vapour into the primary cooling systemand into the containment.Among these materials, iodine, through its 131I isotope, poses the mainshort-term radiological risk for human populations. It is thereforeimportant to assess its behaviour as well as possible, by predicting itsvarious chemical forms(1) inside the reactor containment. The purpose is to obtain a realistic assessment of the possible releases into theenvironment and, thus, optimise the management of the accident’sconsequences.

The results of the international PHEBUS PF programme [1] show, in particular, that the combi-

nation of iodine with silver released during the melting of control rods can greatly reduce the

volatilisation of iodine from the sump water by forming the insoluble species AgI. On the other

hand, the presence of small but non-negligible quantities of gaseous iodine in the containment

as from the beginning of the accident, as was also shown by experiments in the PHEBUS PF

programme, lead to the production of volatile organic iodides (RI) in the short term. In case of

depressurisation of the containment, these iodides –unlike molecular iodine (I2)– are not

adsorbed by the paint covering the walls of the containment or stopped by the filter systems

used in French pressurised water reactors.

Many international nuclear safety organisations played a part in a set of experimental,

analytical or semi-integral programmes following the unexpected results of the first tests in

the PHEBUS PF programme. The aim was to better understand the behaviour of iodine and to

complete existing databases. This work was, naturally, focused on the study of organic iodides,

for which there are less well developed experimental databases and models than for molecular

iodine.

Organic iodides

In order to assess iodine releases, it is necessary to understand the mechanisms by which

organic iodides are formed and destroyed and to quantify their kinetics, especially as it has

been shown by large-scale experiments (such as those in the ACE [2] and PHEBUS PF

programmes) that they would quickly become the prevalent type of volatile iodine in case of

(1) Aerosols that are soluble/insoluble in the sump water in the containment, gaseous forms.

5.3

Nathalie GIRAULTLaboratory for the Study of

Fission Products

Severe Accidents and Crisis Anticipation14

a severe accident. It was for this reason that organic iodides were studied in a joint European effort.

This collaboration led to joint research programmes based on the performance of analytical tests

and their modelling, especially through the OIC [3] and ICHEMM [4] programmes of the 3rd and 4th

Framework Programmes for Research and Development (FPRD). This completed the results of

earlier programmes (CAIMAN [5], RTF [2], etc.). Their main aims were to identify and determine

the kinetics of the main mechanisms for the formation and destruction of organic iodides

under irradiation in the aqueous and gaseous phases over a wide range of experimental conditions.

Formation mechanisms and kinetics

The results of the OIC programmes made it possible to identify the main chemical reactions

involved in the formation of organic iodides, such as:

homogeneous reactions in the aqueous and gaseous phases, especially involving hydrocarbonated

substances released by paints or produced by the pyrolysis of cables or the oxidation of boron

carbide control rods;

heterogeneous reactions in the gaseous phase due to the adsorption of molecular iodine I2 on

painted surfaces.

In case of severe accident, the extent of each of these reactions will greatly depend on the thermo-

hydraulic and chemical conditions reigning in the containment. Homogeneous reactions in the

liquid phase are more likely if the sump water is neutral or slightly basic, i.e. when most of the

iodine is in solution. On the other hand, the formation of organic iodides by heterogeneous

reactions in the gaseous phase become predominant when conditions favour the volatilisation of

the iodine and its adsorption on painted surfaces, i.e. when the sump water is acidic, the silver

concentration is low compared with the iodine concentration and steam condenses on the walls of

the containment. The third formation mechanism, homogeneous reactions in the gaseous phase,

could be especially important in the event of severe accidents affecting some types of reactors,

such as boiling water reactors (BWR) which have control rods of boron carbide.

As far as heterogeneous reactions are concerned, in spite of the apparent wide scattering of the

experimental results which can be largely attributed to the varied experimental conditions and, in

particular, the different forms of iodine (I2 or I- ) considered, the measured percentages for conversion

from I2 to RI are fairly consistent. For all the programmes, the kinetic constants for the formation

of organic iodides vary from 10-5 to 10-3.h-1, and the type and ageing of the paint apparently has

little effect. For the interpretation of all the experimental data, two models were created and used

in the main European codes dealing with the chemistry of the iodine in containments.

The first model, which is rather empirical, known as the Funke’s model [6], developed by SIEMENS,

has been validated for the heterogeneous production of organic iodides by adsorption of I2 on

painted substrates (figure 1).

It mainly reflects the linear dependence of the kinetics of the formation of organic iodides under

radiation (krad) at the surface concentration of molecular iodine adsorbed on the paint [I] (mol.m -2)

and at dose rate D (kGy.h-1).

The second model, known as the Taylor’s model [7], developed by the IRSN, has been validated for

homogeneous production of organic iodides in the liquid phase by the reaction of I2 with organic

5.3

krad (h-1) = 5,18 10-3 exp -2 365 D[I]-0,57[ ]T

(1) AEAT: Atomic Energy AuthorityTechnology (United Kingdom).

15

of the work by Tang and Castleman [8]. This work was carried out in

dry air and at ambient temperature, i.e. in conditions that were not

representative of those in reactors. The kinetic law for breakdown of

ICH3 in humid air is of the first order and the associated kinetic

constant is little different from that previously measured in dry air, in

the region of 6.9 10-4 Gy-1. The ICH3 breakdown kinetics show little

change with temperature, as the activation energy is only 9 kJ.mol-1.

The two existing mechanistic models, covering the kinetics of

production of radiolysis products from the air and their reactions

with molecular iodine via IODAIR, developed by the IRSN [9] and the

AEAT(1) [10], were completed on the basis of this new experimental

data with reactions for the radiolytic breakdown of ICH3 into radicals

and their reactions with the primary and secondary radiolysis

products of air, (N°, e-, OH° and O°) and (NO3) respectively. This is

shown in figure 2 (page 16). This work allows to estimate the kinetic

constant for radiolytic breakdown of ICH3 which is in overall agreement

with the experimental results, especially at ambient temperature

(figure 3, page 16). According to these models, electrons and –to a

lesser degree– monoatomic nitrogen (N°) appear to be the main

agents responsible for the radiolytic destruction of ICH3 in the air,

within the dose rate range considered (> 500 Gy.h-1). This implies

destruction kinetics that are linearly dependent on the dose rate. The

law of dependence would be lower (1/2 power), however, for lower

dose rates. This is due to competition between the various species

produced by the radiolysis of air leading to the breakdown of ICH3.

As the principal primary products of the decomposition of ICH3 are

the radicals I° and CH3°, the models predict that the formation of

the iodine species I2O4 and IONO2 will be preponderant, as in the

radiolytic destruction of I2, along with carbonated species such as

CH3ONO and CH3ONO2. However, the validation of these models

is based only on comparison with overall breakdown kinetics, as the

speciation of the species produced has not been characterised

experimentally. These models are therefore used as a basis to draw

up a simplified mathematical law able to represent the variations in

the kinetic constant for radiolytic destruction of ICH3 with respect to

key parameters identified by means of sensitivity calculations, such

as dose rate D (kGy.h-1) and the degree of moisture (xH2O). This law

has the following form:

The current limitation of this model is that it does not allow for

variations in the composition of the mixture of products formed with

respect to different conditions such as the levels of humidity and

hydrogen in the air. Consequently, the presence of hydrogen would

considerably modify the composition of the products and favour the

IRSN - 2005 Scientific and Technical Report

5.3

species breakdown products released from paints due to the

combined effects of radiation, temperature and leaching by conden-

sates or sump water in the containment. It was constructed on the

basis of a more detailed reaction mechanism. In this case, expressing

the kinetics of the formation of organic iodides is complex and uses

a mathematical combination involving several kinetic constants to

describe the influence of a number of species such as molecular iodine,

iodides, reactive organic species from the paint ([R]) and oxygen.

Breakdown mechanisms and kinetics

Stationary organic concentrations measured in the containment,

especially during tests for the PHEBUS PF programme, suggests that,

in addition to sources for the formation of organic iodides in the

containment, there are important radiolytic or thermal destructive

reactions. The constants for adsorption of organic iodides on the

surfaces are low or null and, therefore, the surfaces do not trap those

iodides.

The new experimental data obtained in the context of the ICHEMM

project has made it possible to confirm the radiolytic destruction

kinetics of ICH3, which was previously established solely on the basis

Figure 1: Formation of organic iodides from painted surfaces – Comparison of the Funke empirical model with experimental results.

Coefficient k for speed of production of organic iodine (1/s.kGy/h)

Initial iodine charge in paint (DEP, mol.m-2)

10-10

10-9

10-8

10-7

10-6

1010-110-210-310-410-5

Production of organic iodides on painted surfacesexposed to non-condensable gaseous phase.

(20°C-25°C; 0.05-5 kGy/h)

k = 5.15 x 10-10 x DEP-0.5739

AEA - British paint

AEA - Finnish paint

Marchand/Funke - French paint

Correlation linefor all FIN data

AEA - French paint

k (s-1) = constant (1- 0,0013T) 1- 0,68 exp(-1,7xH2O) D

[ ]

[ ]

d[RI] (mol.dm-3.s-1) =

1,36 10-7 (0,68+0,0717T)109[R] [ I2]

dt [O2]

D1,22 - 0,11

5 104 + 109 [R]+1,1 1010 [I-]

Severe Accidents and Crisis Anticipation16

5.3

Figure 3: Radiolytic destruction of ICH3. Comparison of the IODAIR mechanistic model developed by the IRSN with experimental results.

[CH3I]/[CH3I]

Dose in kGy6 7542 311

0%

20%

40%

60%

80%

100%

Model calculation

TANG experiments

SIMS experiments

Figure 2: Mechanism of destruction, due to irradiation, of ICH3 diluted in moist air.

Predominantiodine

products

Predominantcarbon

products

CH3O2NO2

CH3O2CH3OOH

I2O4

CH3O

CH3CH3N

IO

CH3I

IONO2 ? I2

INO2 ?

I

CH3ONO2+NO2

+NO3

+H+O+NO3

+O+H+e-+NO3

+O2+H+e-

+NO+NO2

+HO2

+O2

CH3ONO

CH2I ICH2O2 ICH2OH

HI2 + CH2O

+O2

+OH+IO

+N

+NO2 +I

+O

+OH+O

+OH

Predominant iodine products are shown in bold. Those shown in red are volatile.

17

formation of volatile iodine species if the iodine oxides are the

products predominantly predicted by the models in the absence of

hydrogen with less than 0.5% of IONO2 formed for moisture levels

of 30%. In this case, defining the nature and preponderance of the

iodine species formed becomes fundamental for the final assessment

of iodine releases, as the outcome for each of these species may be

different. It is still not properly known what happens to the iodine

oxides whose non-volatile higher levels of oxidation (I2O5, I4O9) have

been revealed during analytical tests. They could be deposited in the

form of aerosols, so reducing the risks of gaseous iodine releases

outside the containment, or transformed into iodates (IO3-) by

dissolving in the droplets of water in suspension in the atmosphere

in supersaturated conditions or by hygroscopic effect. The iodine

nitroxides (of type IONO2), which are fairly volatile, could make a

more significant contribution to releases.

Conclusion

The combination of all the analytical and semi-integral experi-

mental programmes and the experiments in the PHEBUS PF

programme provided a large amount of data on the behaviour of

gaseous iodine and, especially, of organic iodides. These results

made it possible to validate or develop models for the formation

and destruction of organic iodides in conditions representative of

severe accident, especially in the presence of a radiation source.

The simplified models were input into the main European codes

concerning the chemistry of iodine in the containment. They can

be used to carry out reactor safety analyses in order to enhance

the prediction of iodine releases, especially by quantifying the

concentrations of organic iodides in the reactor containment

during an accident. This is particularly important with regard to

safety because these species, which are mostly very volatile, are

not adequately stopped by the filter systems in French reactors

in the event of depressurisation of the containment. It should,

nevertheless, be noted that the studies conducted are conserva-

tive in that they consider only methyl iodide, which is the most

volatile organic iodide. To obtain a more accurate quantification

of iodine releases, the existing databases would have to be

extended to allow for organic iodides at even higher dose rates

(tests are planned in the IRSN EPICUR installation [11]) and the

composition of the products resulting from the radiolytic

destruction of these species and of molecular iodine would have

to be determined. Although it is not indispensable to know the

exact composition of the mixture, it is important to be able to

predict the proportions of non-volatile species and volatile

species among the products resulting from radiolytic oxidation.

IRSN - 2005 Scientific and Technical Report

5.3

References[1] (a) P. von der Hardt, A. V. Jones, C. Lecomte, A. Tattegrain. Nuclear safety research, the PHEBUS-FP severe accident experimental programme. Nuclear Safety35, Vol. 2, 1994.(b) B. Clément, N. Girault, G. Repetto, D. Jacquemain, A.V. Jones, M.P. Kissane, P. von der Hardt. LWR severe accident simulation: synthesis of the results andinterpretation of the first Phebus FP experiment FPT0. Nuclear Engineering and Design 226, 2003, 5-82.[2] W.C.H. Kupferschmidt at al. The advanced containment experiment (ACE) radioiodine test facility experimental program. 3rd CSNI Workshop on IodineChemistry in Reactor Safety, Tokai-Mura, March1992, NEA/CSNI/R(91)15.[3] S. Dickinson, H.E. Sims, E. Belval-Haltier, D. Jacquemain, C. Poletiko, F. Funke, S. Hellmann, T. Karjunen, R. Zilliacus. Organic iodine chemistry. NuclearEngineering and Design 209, 2001, 193-200.[4] S. Dickinson, H.E. Sims, F. Funke, S. Guentay, H. Bruchertseifer, J.O. Liljenzin, H. Glänneskog, M.P. Kissane, L. Cantrel, E. Krausmann, A. Rydl. Iodine chemistryand mitigation mechanisms (ICHEMM). FISA Symposium, Luxembourg, November 2003.[5] (a) C. Leuthrot. Descriptif du dispositif expérimental CAIMAN. Rapport interne CEA/DEC/SECA/LTC 97/087,1997.(b) L. Cantrel. Analysis of the CAIMAN experiments and validation of the IODE code version 5.1. Rapport interne IRSN/DPAM/SEMIC 04/010, 2004.[6] F. Funke. Data analysis and modelling of organic iodide production at painted surfaces. OECD Workshop on Iodine Aspects of Severe Accident Management,Vantaa, Finland, May 1999.[7] P. Taylor. Transient formation of organic iodides: interpretation and modelling. Rapport interne IRSN/SEMAR 00/019, 2001.[8] I.N. Tang, A.W. Castleman kinetics of radiolytic decomposition of methyl iodide in air. J. Phys. Chem. 74 (22), 1970, 3933-3939.[9] F. Aubert, N. Girault. Qualification du modèle mécaniste de destruction radiolytique de l’iodure de méthyle – Élaboration d’un modèle simplifié, rapportinterne IRSN/DPAM/SEMIC 04/01, 2004.[10] D. Dickinson. The radiolysis of gaseous iodine species in air, in E. Krausmann (Ed). Data analysis and modelling of iodine chemistry and mitigationmechanism, EC Report SAM-ICHEMM-D010, 2002.[11] (a) B. Clément, R. Zeyen. The PHEBUS FP and international source term programmes, Proc. Nuclear Energy for New Europe 2005, Bled, Slovenia,September 5-8, 2005.(b) B. Clément. Towards reducing the uncertainties on source term evaluations: an IRSN/CEA/EDF R&D programme, Proc. Eurosafe Forum, Berlin, Germany,November 8-9, 2004.

5.4 The hydrogen riskduring a severePWR accident

In the context of studies on severe accidents affecting pressurised waterreactors, the “hydrogen risk” is defined as the possibility of loss ofintegrity of the reactor containment or of its safety systems following

the violent combustion of hydrogen released in the course of coredegradation. The steam that escapes through the break condenses on thecold walls of the containment during a severe accident started by a breakon the primary cooling system and causes convection loops inside it.The TOSQAN experimental installation was designed to study the principalphenomena concerned by this problem, especially those related to thedistribution of gaseous species in the containment.

Severe Accidents and Crisis Anticipation18

The TOSQAN facility (figure 1) consists of a closed

cylindrical vessel with a volume of 7 m3, a height of 4 m

and an inside diameter of 1.5 m. A central, vertical pipe

is used to inject superheated steam, air or helium. Being

a nonflammable gas, helium is used to simulate

hydrogen without any danger (as it has similar thermal

properties and density). The containment is equipped

with double walls, in which a heat-transport fluid is

circulated, its temperature being controlled by a heater-

cooler unit. In this way, an area of condensation, or

condensing wall, can be determined, while the other

warmer walls of the containment constitute non-

condensing walls. The experimental sequence for the

condensation experiments involves the injection of

steam into the containment, initially at atmospheric

pressure, with the wall temperatures being set.

Pressurisation leads to a state of thermodynamic equili-

brium, referred to as the constant condensation regime,

in which all the injected steam is condensed on the

containment walls (Malet et al., 2005). Some experiments involve the injection of helium when this

regime is reached. The TOSQAN facility is equipped with a wide range of instrumentation including

intrusive measurement techniques, such as thermocouples and pressure sensors, or techniques

requiring gas sampling (mass spectrometry, Auban et al., 2003, flow metering) and non-intrusive

measurement techniques based on optical diagnoses (laser velocity measurement with LDV and

PIV, Raman spectrometry, Porcheron et al., 2002, 2003). The experiments conducted in TOSQAN

constitute reference situations which can be used to validate the physical models used in contain-

ment thermo-hydraulic codes. The latter include TONUS, an application of the CAST3M code.

Figure 1: The TOSQAN facility.

Non-condensingwall

Non-condensingwall

Steaminjection

Condensingwall

19

Developed by the CEA for the IRSN, TONUS was validated in various

steam and helium distribution experiments, condensation experi-

ments in TOSQAN and in other French and German installations

such as MISTRA or ThAI. This was done in the context of a comparison

exercise or international benchmark (ISP47, Vendel et al. 2002)

supervised by the OECD. As regards the steady-state regimes in the

TOSQAN experiments, TONUS is a means of identifying the pressure

levels in the containment with a deviation of less than 0.3 bar, the gas

temperatures with a deviation of less than 3°C and concentrations

with a deviation less than 3% by volume. An example of qualification

is shown in figure 2.

Effect of condensation on air-steam flows

Condensation on walls in the presence of a non-condensable gas

such as air or hydrogen causes an accumulation of that non-conden-

sable gas in the vicinity of the wall. This mixture enriched with a

heavier gas (in the case of air) generates an inverted natural convec-

tion flow. This is due to the buoyancy forces descending the length of

the wall. The extent of this flow may vary according to the amount

of condensation or, in other words, according to the mass exchange

and the location of the condensation (in a central, vertical area of

the containment in TOSQAN) and it can then affect the overall flows

in the containment. The distribution of gaseous species may then be

modified. This is illustrated in figure 3, where the fields of concentra-

tion of the steam in the TOSQAN containment are represented for

two characteristic experiments: test 1 with homogeneous concentra-

tion and test 7 with stratified steam concentration.

This figure shows the following main characteristic areas:

area 1: injection,

area 2: recirculation,

area 3: flow descending the length of the wall,

area 4: described as “critical”, influenced by the entrainment in the

injection area and the flow in area 3.

In test 7, with low condensation flow on the wall, the downward flow

in area 3 is completely re-entrained by the injection area. The

injected steam is not distributed below the injection whereas, in test 1,

the downward flow in area 3 has a high condensation rate. This flow

Figure 2: Qualification of the TONUS code on the TOSQAN experiments: resultsobtained for gas temperature in test 7.

Figure 3: Steam concentration fields in test 1 at high condensationrate (left) and in test 7 at low condensation rate (right).

IRSN - 2005 Scientific and Technical Report

Gas temperature (°C)

Height in TOSQAN (m)0.40.0 0.8 1.2 1.6 2.0 2.4 2.8 3.2 3.6 4.0 4.4 4.8

50

60

70

80

100

90

110

120

CODE-EXPERIMENT vertical temperature profiles

Sumpzone

Condensingwall zone

Hot wallzone

Steaminjection

TONUS - At centre

TONUS - At mid-radius

TOSQAN - At mid-radius

TOSQAN - Near wall

TONUS - Near wall

TOSQAN - At centre

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.10

0.11

0.12

0.13

0.14

0.16

0.18

0.20

0.30

0.50

1.0

0.00

Area2

Area3

Area1

Area4

Test 1 Test 7

5.4

Jeanne MALETAirborne Pollutants and Containment Study and Research

Department

Emmanuel PORCHERON Laboratory for the Experimental Study of Containment,

Air cleaning and Ventilation

Severe Accidents and Crisis Anticipation20

can partly descend to below the condensing wall. This entrains steam

below the injection area and causes a mixing of gaseous species in the

containment. This is confirmed by the calculation of a characteristic

dimensionless number, the Grashof number –associated with the

wall (the ratio between buoyancy forces and viscous forces) which is

smaller for test 7 (3 109) than for test 1 (3 1010).

The mixture of gases in the containment is therefore greatly influenced

by the level of condensation on the wall. This result is also found by

“numerical testing”, referred to as test 1H (Malet et al., 2004). Test 1

is simulated with the same parietal temperature conditions and the

same injection flow by keeping constant the Richardson number for

injection, this number being characteristic of jet/plume type flows.

Test 1H differs from test 1 as the injected steam is replaced with a

non-condensable gas, namely helium. In this way, the buoyancy

effects caused by condensation are eliminated. The results are shown

in figure 4. It may be noted that in test 1H, the recirculation flow in

area 2 is dissipated along the wall in area 3 and is re-entrained in area 1.

The gas is not entrained towards the bottom of the containment, and

gas density stratification appears. This clearly illustrates the influence

of condensation, through the downward flow, on the distribution of

gases in the containment.

To corroborate these numerical results, figure 5 shows the typical

condensation flow by means of a velocity profile obtained experi-

mentally by PIV in the near-wall area for the condensation steady

state with an air-steam mixture (shown by t < 0 in figure 7). We also

see a clear acceleration of the parietal flow between the top and

bottom of the condensing wall, as shown by the comparison of

parietal profiles for the steady-state regime between the top

(figure 5) and the bottom (figure 6) of the condensing wall

(Porcheron et al., 2003).

Influence of a light non-condensable gas on air-steam-helium (hydrogen) flows

When helium is injected into the containment for between 0 and

600 s (figure 7), the light non-condensable

gas is initially distributed towards the

containment walls via the circulation loops

(figure 4). This has the effect of producing

a helium layer in the vicinity of the

condensing wall and insulating that wall

from the steam. Consequently, the inten-

sity of the condensation of steam

decreases in relation with the decrease in

the condensation rate observed (figure 7),

and the parietal flows slow down (figures

5 and 6). The decreases in the condensa-

tion rate and the velocity of parietal flows

are well correlated with respect to time.

This helium distribution phase in the

containment continues until a homoge-

neous air/steam/helium mixture is obtained

(around 2,500 s in figure 7).

A steady state for condensation with an

air/steam/helium mixture is reached. TheFigure 4: Injected gaseous species concentration field and velocity field in test 1 (left) and intest 1H (right).

5.4

Mixture

Steam volumefraction

Steaminjection

Flow due tocondensation

Containedflow

Heliuminjection

Helium volumefraction

Velocity

0.5 - 0.7 0.2 - 0.4 - 1

Same Richardson number

Figure 5: Parietal vertical velocity profiles measured at the top of the condensing wall.

Distance fromcondensing wall (mm)

-0.3

-0.25

-0.2

-0.15

-0.05

-0.1

0

0.05

Parietal vertical velocity profiles measuredat top level of condensing wall (m/s)

20 40 60 80 100 120 140 1600

Air / steamsteady stateAfter heliuminjection - t=800 sAir / steam / heliumsteady state

21

condensation rate is once again equal to the injected steam flow

rate, and the parietal flows speed up again.

Conclusions and outlook

Condensation can have a great effect on the distribution of gases

in the containment as it causes a parietal flow which may interact

with other flow areas. The TOSQAN experiments show, using the

TONUS code, that this flow related to condensation can be the

direct cause of homogenisation of the gaseous species. The flow

connected with condensation is ascertained experimentally by

measuring the velocity near the wall, which is a result rarely

obtained in a facility of this scale. Furthermore, the presence of

hydrogen (simulated by helium in TOSQAN) may also have an

effect on condensation. Indeed, it acts as insulation between the

condensing wall and the steam during the transient helium injection

phase, which brings about a decrease in the intensity of condensation

and, consequently, in the velocity of parietal flows connected with

the distribution of light non-condensable gas in the containment.

The presence of a light gas can therefore have the effect, in a transient

phase, of inhibiting the homogenisation of the gaseous species,

which is an unfavourable situation in the context of the hydrogen risk.

Other parameters, such as the activation of the spraying system in

the reactor containment, can also affect the distribution of gaseous

species: the spraying phenomenon is also the subject of a research

programme being conducted in TOSQAN.

IRSN - 2005 Scientific and Technical Report

5.4

ReferencesAuban (2003), J. Malet, P. Brun, J. Brinster, J.J. Quillico, E. Studer. Implementation of gas concentration measurement systems using mass spectrometry incontainment thermal-hydraulics test facilities: different approaches for calibration and measurement with steam/air/helium mixtures. 10th InternationalTopical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH -10), Seoul, Korea, October 5-9, 2003.Malet (2005), E. Porcheron, P. Cornet, J. Vendel. Experimental and numerical study of natural convective flows with filmwise condensation in a largecontainment vessel, to the 5th International Conference of Multiphase Flow (ICMF 2004), 30th May-4th June 2004, Yokohama, Japan, 2004.Malet (2005), E. Porcheron, J. Vendel. Filmwise condensation applied to containment studies: conclusions of the TOSQAN air-steam condensation tests,11th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-11), Avignon, France, October 2-6, 2005.Porcheron (2002), L. Thause, J. Malet, P. Cornet, P. Brun, J. Vendel. Simultaneous application of spontaneous Raman scattering and LDV/PIV for steam/air flowcharacterization.10th International Symposium on Flow Visualization, Kyoto, Japan.Porcheron (2003), L. Thause, J. Malet, P. Cornet, P. Brun, J. Vendel. Optical diagnoses applied for single and multi-phase flow characterization in the TOSQANfacility dedicated for thermal hydraulic containment studies. 10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH -10),Seoul, Korea, October 5-9, 2003.Vendel (2003), P. Cornet J. Malet, E. Porcheron. ISP 47, Containment thermal-hydraulics – Computer codes exercise based on experimental facilities. Nuclear,10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-10), Seoul, Korea, October 5-9, 2003.

Figure 7: Time evolution of relative pressure, steam condensation rate and volume fraction of helium measured at different heights in the containment.

Helium injection

Time (s)

0

0.05

0.1

0.15

0.25

0.2

0.3Helium volume fraction (xHe) Relative pressure (bar)

and condensation rate (g/s)

0 800 1,600 2,400 3,200 4,000 4,800 5,600-800

xHe - injectedxHe - top ofcondensing wallxHe - bottom ofcondensing wallRelative pressure

Condensation rate0

0.5

1

1.5

2.5

2

3

Figure 6: Parietal vertical velocity profiles measured at the bottom of the condensing wall.

Distance fromcondensing wall (mm)

-0.3

-0.25

-0.2

-0.15

-0.05

-0.1

0

0.05

Parietal vertical velocity profiles measuredat bottom level of condensing wall (m/s)

20 40 60 80 100 120 140 1600

Air / steamsteady stateAfter heliuminjection - t=800 sAir / steam / heliumsteady state

5.5 ASTRIDAssessing a lightwater reactor accident in realtime

In the event of an accident in a nuclear power plant, a large quantity ofradioactive material could be released into the environment. Dependingon the severity of the accident, it may be necessary to quickly take steps

to protect human populations and the environment. It may, however, takeseveral hours to implement such steps. It is for this reason that changes in the installation must be monitored and the foreseeable releases(1) must be assessed as soon as the alert is given. If releases occur, the foreseeableconsequences should be specified, in view of the actual development of the accident.The aim of the ASTRID project (Assessment of Source Term for EmergencyResponse Based on Installation Data(2)) was to develop a method and a computer tool enabling European countries to assess the foreseeablereleases in case of an accident affecting a light water reactor, with the goalof providing effective protection for populations and the environment.The estimation of the releases is then used to assess the actual or possibleconsequences in the environment in terms of doses and deposits, and todetermine the protective measures to be taken.

Severe Accidents and Crisis Anticipation22

Context and partners

The ASTRID system was developed as part of the European Commission’s 5th EURATOM Framework

Programme for Research and Development (FPRD). The project lasted from November 2001 to

October 2004. It was coordinated by the IRSN with the participation of technical and statutory

organisations from various countries:

Swedish Nuclear Power Inspectorate (SKI);

Radiation and Nuclear Safety Authority of Finland (STUK);

Hungarian Atomic Energy Authority (HAEA);

Gesellschaft für Anlagen und Reaktorsicherheit (GRS, German institute for the safety of nuclear

installations);

Forschungszentrum Karlsuhe (FzK, German institute for nuclear energy techniques and Karlsruhe

research centre);

Nuclear Power Plants Research Institute Trnava Inc (VUJE);

(1) Releases correspond to thereleases outside the installation.They are defined according to thenature of the radioactive productsand isotopes released, activities(Bq), emission rates (Bq/s), theduration of release of eachisotope, the origin of the releaseinto the environment (stack,building) and its height.

(2) Assessment of the source term incase of accident using the datafrom an installation.

installation in order to avoid or at least reduce the releases, and to take

action outside the installation to protect surrounding populations as

effectively as possible. This method uses a comprehensive approach to

evaluate the condition of the power plant and to organise the work of

crisis teams. The method is based on the barrier concept.

The ASTRID method features three barriers, as follows (figure 1):

The first barrier is provided by the fuel rod claddings and the fuel matrix.

The second consists of the reactor primary coolant boundary

(including effluent tanks and connected systems).

The third consists of the reactor building and, if required, the contain-

ment and its extensions, including the containment ventilation and

filtering systems.

We can define the safety functions that are intended to ensure the

integrity of the above-mentioned barriers by considering the failure

23

Vattenfall Power Consultant AB, formerly Swedpower AB (Sweden);

PCI Teknik & Information AB (Sweden).

The ASTRID system

The ASTRID system consists in a method for the appraisal of an

accident affecting a light water reactor and a computer tool used to

apply that method.

The methodThe aim of the ASTRID method is to identify and, as far as possible, to

foresee the occurrence of radioactive releases into the environment in

order to take the necessary control actions on the accident-affected

Figure 1: The three barriers.

IRSN - 2005 Scientific and Technical Report

5.5

ECCS

Fuel andcladding

Primary systemboundary inside andoutside containment

Reactor building and containment

AFWS RRS

M

FWS ECCS

RRS

N2

Karine HERVIOUIRSN Emergency Situations and Crisis Organisation Department

Ninos GARISSwedish Nuclear Power Inspectorate - SKI, Sweden

Bernd SCHWINGESGesellschaft für Anlagen und Reaktorsicherheit - GRS, Germany

Severe Accidents and Crisis Anticipation24

modes of each one. For example, the integrity of the first barrier is ensured provided that the

temperature of the fuel rod claddings does not exceed a given value. Two phenomena can lead to

an increase in the temperature of the claddings: an excessive increase in the energy produced by

the reactor core and degradation of the fuel cooling function. To avoid any excessive production of

energy in the reactor core, the core must be maintained in a critical or subcritical state. One of the

functions that must be controlled in order to ensure the integrity of the claddings and the fuel is

therefore the “reactivity” function. It must also be possible to remove power from the core via the

primary coolant to ensure cooling of the fuel, whether the reactor is in power operation or shut

down. The other function that is required to ensure cladding integrity is the “core cooling” function.

The assessment of reactor status is completed by checking the systems available to fill these

functions. The control actions taken by operators to manage the accident are also taken into

account in the assessment.

It is necessary to organise the implementation of the method used to evaluate the status of a

reactor in a crisis situation according to the approach proposed. A special schedule was drawn up

to define the status of the barriers (intact, degraded, faulty, unknown) and of the corresponding

safety functions (figure 2).

The ASTRID method entails periodically assessing the status of the three barriers (triple diagnosis)

and their foreseeable evolution (triple prognosis). The purpose is to characterise actual or possible

radioactive releases into the environment.

A diagnosis phase is followed by a prognosis phase:

The diagnosis consists in identifying the status of the reactor at the time of analysis. Signs of its

status, such as pressures, temperatures, levels, flow rates and activities which are transmitted by the

power plant, are used to evaluate the integrity of the barriers and the status of the safety functions.

The prognosis consists in studying what will happen in the course of time. It is based on the

status of the installation at a given time and assumes that no further event will occur, unless the

observed status of the installation inexorably leads to the loss of some systems.

5.5

Plant diagnosis/prognosis grid - No.

Diagnosis at: h Prognosis for date/hours/....:

Barrier state Critical safety functions Safety system Future state Future critical safety Future barrier statestate states of systems function state

Fuel matrix, Fuel matrix,cladding, cladding

Intact Sub-criticality Sub-criticality IntactDegraded Core Core Degraded to .... extentFailed cooling cooling Failure at hNot known Mass/Level Mass/Level Not known

Reactor Coolant Reactor Coolant Pressure Boundary (RCPB) Pressure Boundary (RCPB)

Intact RCPB integrity RCPB integrity IntactDegraded Degraded to .... extentFailed Failure at hNot known Mass/Level Mass/Level Not known

Heat sink Heat sink

Reactor containment Reactor containment

Intact Isolation Isolation IntactDegraded Heat sink Heat sink Degraded to ... extentFailed Atm. Cond. Atm. Cond. Failure at hNot known Not known

Ta b l e 1

Site: Plant unit: Date: Hour: Visa: From: To:

Inta

ctDe

grad

edFa

iled

Not

know

n

Inta

ctDe

grad

edFa

iled

Not

know

n

Figure 2: General chart to assist with implementation of method.

25

Aggravated prognosis consists in studying what could happen in

the event of an unforeseeable failure in view of the current accident

sequence or the control strategy implemented.

This diagnosis/prognosis approach must be applied continuously

throughout the accident in order to periodically update the assess-

ments, taking into account changes in the status of the installation.

It must be possible to make an accurate

assessment of the situation on the basis of

strict application of the method and good

knowledge of the possible accident deve-

lopment sequences.

The IT toolThe ASTRID IT tool was developed to help

experts assess the situation and its possible

developments by assisting in the application

of the method. It makes it easier to quantify

the key parameters required to manage the

crisis, such as the foreseeable moment of

occurrence of the releases and their severity.

More precisely, the purpose of the ASTRID

tool is to make it possible to follow up the

development of an accident, to foresee the

behaviour of the reactor and to estimate

the current or foreseeable releases. This relies on data concerning the

installation’s status that are regularly transmitted to crisis teams by

the nuclear power plant operator. The assessment is, in particular,

intended to provide information as a basis for decisions regarding the

protection of human populations. By using this tool, it is also possible

to assess certain parameters that are useful for the control of the

installation, such as the flow rate escaping through the break, the

available time before the start of releases in comparison with the

times required to return given items of equipment to service, etc.

The ASTRID tools has four main functions:

It helps the user with the diagnosis of the status of the installation.

It supplies precalculated release data in order to assess the order of

magnitude of the consequences for populations and quickly proposes

protective measures against them.

It simulates the development of the accident over time, provides

hypotheses on the availability of safety systems and allows the

releases to be assessed more accurately.

It enables the user to draw up expert assessment summaries.

Assistance for diagnosis of power plant status

The system assists the user in assessing the safety barriers and

functions (figure 3). The ASTRID tool displays important parameters

transmitted by the power plant so that the user has all the information

required to apply the method. These parameters, which are indicators

to characterise the power plant’s status, were identified during the

definition of the method for each type of reactor. For example, in

order to diagnose the status of the first barrier, the system must

examine changes, with respect to time, in the temperature of the

water on outlet from the core, the dose rate in the reactor building,

the hydrogen concentration and the stack activity.

The ASTRID system offers the possibility of automatically acquiring

data from the reactor calculator or a computer combining all the

data on the power plant. Furthermore, the user can implement

automatic monitoring of some important parameters with alarms

being triggered by predefined criteria such as drop to zero power or

the overshooting of a threshold.

Utilisation of precalculated releases

Crisis teams can propose actions to be taken to protect populations

before releases start, by means of a set of “base cases”, i.e. a set of

precalculated releases for various orders of magnitude. A set of base

cases is established on the basis of calculations performed by adop-

ting various probabilistic or deterministic approaches. With ASTRID,

the choice of the precalculated release suitable for the current accident

(whether real or simulated) is made in accordance with a test procedure

based on physical parameters (figure 4, page 26).

Simulation of the development of an accident situation

To determine the probable development of an accident situation, the

ASTRID tool can be used to perform simulations on the basis of the

diagnosis of power plant status and assumptions made by the user

regarding the availability and operating conditions of safety systems

with respect to time. Decision-makers must have access to an assess-

ment of the release risks and available times before the releases

occur in order to effectively protect the population. This information

is also useful to the operator in order to define the strategy to be

adopted to restore the installation to a satisfactory safety level to

IRSN - 2005 Scientific and Technical Report

5.5

Figure 3: Definition of the status of the first barrier of a pressurised water reactor (PWR).

Severe Accidents and Crisis Anticipation26

avoid releases as far as possible while ensuring the protection of personnel. The ASTRID tool helps

the user evaluate the installation’s thermo-hydraulic behaviour and transfers of fission products on

the basis of a simplified model of nuclear power plant systems. The code includes a model of the

primary system and the reactor cooling system (PROCESS), coupled with the German COCOSYS

code which assesses the thermo-hydraulic conditions in the reactor building and the behaviour of

fission products in the installation. The results must be available for urgent use: the computing time

of the simulation code installed on the ASTRID tool must therefore be in the region of a few

minutes. Users can then modify or test a great many assumptions and regularly reassess their

estimations within lead times compatible with the requirements of crisis management. The

release data calculated with software computing radiological consequences can then be automatically

transmitted. The results provided by the ASTRID tool were validated by comparison with results

obtained using the CATHARE code (figure 5) and the MAAP, MELCOR and ASTEC international

5.5

Figure 5: Change in mass of liquid water in primary system in case of a break on a system pipe (diameter: 7.62 cm)calculated by PROCESS and CATHARE (reference code for a reactor’s thermo-hydraulic behaviour).

Weight (t)

Time (s)

0

50

100

150

200

250

0 1,000 2,000 3,000 4,000 5,000 6,000

PROCESS: Weight of liquidwater in primary system

CATHARE: Weight of liquidwater in primary system

50 tonsCATHARE

50 tonsPROCESS

Figure 4: Part of a procedure to determine a precalculated release for a PWR reactor.

27

reference codes. The code’s field of applicability was defined in terms of accident sequences and

pressurised water and boiling water reactor types.

Summary reports

The various parties in a crisis situation must be regularly provided with summary reports on the

status of the installation and assessments performed. The ASTRID tool provides summary reports

in various formats to meet the requirements of various parties in the teams responsible for

technical assessment of an accident, such as authorities and decision-makers.

Conclusions and outlook

The ASTRID system, which includes an expert assessment method and a supporting computer

tool, constitutes a complete system to assess current or foreseeable releases in the event of

an accident situation affecting a light water reactor.

The method is fully applicable in case of crisis and could contribute to harmonising practices

in the field of accident assessment in Europe and facilitate communication between the parties

concerned, inside and outside a given country.

The computer tool has reached a relatively stable version. Some further work should, however,

be carried out so that it can be easily implemented in an emergency context. The GRS and the

IRSN are collaborating to achieve this objective. In the context of the EURANOS project,

supported by the European Commission in the framework of the 6th FPRD, the tool will be

tested in various emergency centres, including the IRSN Emergency Response Centre (CTC), to

obtain comments on its utilisation and to define possible improvements.

The ASTRID tool has some original features compared with the computer system installed at the

IRSN Emergency Response Centre for the assessment of releases in case of accident (SESAME(3)

system). Firstly, it is an integrated system. Its modelling scope is wide-ranging: it includes a

model to assess the thermo-hydraulic behaviour of the atmosphere in the containment and

can be used to model the main phenomena liable to occur during a severe accident and having

an impact on the amount of releases. Secondly, the project has demonstrated that it is now

possible, on the basis of available computing powers, to install scientific codes in a crisis tool

while at the same time ensuring computing times that are compatible with use in emergency

situations. This is bound to have an effect on the development of the next generation of

Emergency Response Centre (CTC) tools for accidents which may affect French pressurised

water reactors.

5.5

ReferencesASTRID/04.39 report – January 2005 – Final report for contract FIKR-CT-2001-00171.ASTRID/04.11 report – December 2004 – Description of the ASTRID system: method and tool for the assessment of power plant and of releases into theenvironment.ASTRID/04.15 – December 2004 – Validation report on the ASTRID computer tool validation report for French 900 MWe nuclear power plants.ASTRID/04.16 report – December 2004 – Validation report on the ASTRID computer tool validation report for VVER-440/V213 and VVER-1000/V320 reactors.ASTRID/04.22 report – December 2004 – Validation report on the ASTRID computer tool validation report for Finnish VVER reactors.

IRSN - 2005 Scientific and Technical Report

(3) Accident progression analysis andassessment methods.

Severe Accidents and Crisis Anticipation

newsflashnewsflashnewsflashnewsflashnewsflashnews

28

RECI experimentsIodine chemistry in catalytic hydrogenrecombiners

In the event of a severe accident occurring on a

pressurised water reactor, vapours and liquid or solid

particles comprising fission products and reactor

structure materials are produced during the core

degradation process and then transferred to the

containment via the break in the primary system.

These particles, in suspension in the containment

atmosphere, are then liable to be vaporised and to

react chemically while passing through the hydrogen

recombiners heated to a high temperature by the

exothermic reaction 2 H2 + O2g 2 H2O. In the case

of metal iodides (essentially CsI, AgI, InI and CdI2),

these additional reactions produce volatile iodine

forms (I + I2, HOI and HI) and ultrafine aerosols.

The volatile and insoluble forms of iodine are more

challenging, in terms of safety, than its solid forms,

as the trapping of gases is less effective than the

filtering of aerosol particles. Furthermore, molecular

iodine I2 is the precursor of methyl iodide ICH3,

the presence of which in the containment poses an

additional safety problem.

The general aim of the RECI (recombiner and

iodine) programme conducted by the IRSN on an

analytical test bench on the Saclay site was to

experimentally quantify the rate of conversion of

metal iodides into iodine according to temperature

and dwell time inside an electrically heated chemical

reactor, designed to simulate a catalytic recombiner.

The conversion rates measured on the RECI

bench with a CsI aerosol and then a CdI2 aerosol

gave high values, showing the thermal instability of

those iodides in oxidising conditions and in the

presence of steam. This thermochemical domain had

not previously been studied, unlike the case of the

primary system with no air intake.These experimental

results, obtained by heating particles of caesium or

cadmium iodide with a wide range of variation for

influencing parameters, such as temperature and

dwell time, and also the aerosol particle size, can in

all likelihood be extrapolated to silver and indium

iodides.

The IRSN carries on with its work in this field, on

the basis of iodine source term calculations(1) taking

into account the operation of hydrogen recombiners

(calculations performed with the IODE code in ASTEC),

and with the dimensioning of a comprehensive type

test bench, better representing the actual operation

of a recombiner than the RECI bench.

5.6Jean-Christophe SABROUX,

Florence DESCHAMPSAirborne Pollutant

Dispersion andContainment Study and

Research Department

(1) Maximum quantity ofradioactive iodineliable to escape fromthe reactor building incase of severeaccident.

29

newsflashnewsflashnewsflashnewsflashnewsflashnews

IRSN - 2005 Scientific and Technical Report

The IRSN inaugurated the EPICUR installation on

20th May 2005. The purpose of this installation is to

improve knowledge on the behaviour of volatile

radioactive products such as iodine and ruthenium

in a reactor containment in case of an accident

resulting in core meltdown. The installation

comprises an irradiator containing 925 TBq of 60Co

and supplying a dose rate of 10 kGy.h-1. The reactor

containment is represented on a very small scale by

a 5-litre irradiation tank containing a volume of

water and a volume of air simulating the sump and

the containment atmosphere, respectively, of a

reactor building.

40 tests are planned between 2005 and 2008.

The tests in progress, carried out in collaboration

with EDF, the CEA and the European Commission in

the context of the SOURCE TERM programme, are

aimed at improving the assessment of iodine

releases into the environment in the event of an

accident in order to better evaluate the means to

be implemented to minimise the consequences.

The first test, carried out on 12th May 2005,

consisted in irradiating an aqueous solution of stable

iodide with pH 5.0 and at a temperature of 120°C

traced with 37 MBq of iodine 131. The goal was to

quantify the release of gaseous iodine obtained under

the effect of radiation. This test made it possible to

qualify the device and the corresponding instrumen-

tation, and it has already provided useful data to

validate models for iodine radiolysis in the sump. This

makes it possible to predict the levels of gaseous

iodine concentration in the event of an accident.

5.8Séverine GUILBERT

Analytical Test Laboratory

EPICURA new installation to study thevolatility of radionuclides in thecontainment in severe accidentsituations

The SOURCE TERM programmeThe IRSN is launching a new research programme

entitled SOURCE TERM, in partnership with EDF,

the CEA and the European Commission to run for

the period 2005 to 2010. The objective of this

internationally open programme is to improve the

assessment of releases of radioactive products such

as iodine and ruthenium into the environment in

case of a core meltdown accident in a water reactor

so as to be able to better evaluate the means to be

implemented to minimise its consequences.

The first part of this programme concerns the

study of iodine chemistry in the primary system and

in the reactor containment in accident situations

with the aim of better quantifying potential releases

of iodine in various forms.

The second part is to focus on the study of the

degradation of boron carbide B4C control rods,

which are used in some European reactors (including

in France), and its impact on the evolution of a

severe accident and on the volatility of certain

radioactive products, especially iodine, in the

containment.

The third part is to examine the consequences of

degradation of fuel rods which may occur on contact

with air (especially as regards the release of ruthe-

nium) in case of an accident leading to perforation of

the reactor vessel or dewatering of an irradiated fuel

storage pit, the main point of these consequences

being the release of ruthenium.

The last part will entail the study of the release of

fission products from fuel at high temperature in

severe accident situations. Microanalyses of irradiated

fuels subjected to heat transients will be used to

help interpret experiments that are already available.

Experiments will also be conducted on highly irradiated

MOX and UO2 fuels.

5.7Didier JACQUEMAIN

Analytical Test Laboratory

Severe Accidents and Crisis Anticipation

newsflashnewsflashnewsflashnewsflashnewsflashnews

30

SARNETA European network of excellence for severe accidents

SARNET –the Severe Accident Research NETwork

of excellence, www.sar.net.org– which is coordinated

by the IRSN, gathers 200 researchers and 49 European

organisations committed to research on nuclear

reactor safety: organisations providing technical

support for safety authorities, universities and indus-

trial firms in 18 European countries. The goals of this

network are to augment and transmit scientific

knowledge on severe accidents for water reactors

and accidents involving core meltdown.

On 14th, 15th and 16th November 2005, the

network organised its first international seminar,

ERMSAR (European Review Meeting on Severe Accident

Research) in Aix-en-Provence (France) to report on

the main points of progress achieved in the 18 months

since it was set up. SARNET will organise a training

course, from 9th to 13th January 2006, for young

researchers on severe accident phenomenology and

probabilistic safety analyses for that type of accident.

This will form the basis for a book to be written for

the benefit of students and young researchers.

5.9Jean-Claude MICAELLI

Department for thePrevention of

Major Accidents

IRSN/FzK cooperation on direct heating of containment

The phenomenon of direct heating of a pressu-

rised water reactor’s containment is envisaged in

severe accident scenarios involving failure of the

vessel bottom head. In this case, the corium and hot

gases under pressure in the vessel are ejected into

the containment, so that its atmosphere heats up

and the pressure builds up.

A specific cooperation agreement has been esta-

blished between the IRSN and the FzK (Forschungs-

zentrum Karlsruhe) in order to better assess the

consequences of this phenomenon on contain-

ment resistance. The FzK has two complementary

experimental installations on a scale of 1:16, DISCO-C

and DISCO-H, which have been adapted to the

geometry of French reactors. The first of these is

intended for the study of the dynamic aspects of

dispersion and the second is used to study pressurising

of the containment, as well as dispersion.

In parallel, experiments are to be modelled using

the AFDM and MC3D mechanistic design codes will

allow the test to be interpreted with the aim of

validating or improving the models installed in the

ASTEC code system.

5.10Delphine PLASSART

Severe Accident Physics Bureau

An experiment in the DISCO-C facility

carbon monoxide and dioxide although the quantity

of methane (which was expected in small amounts

on the basis of pre-test calculations) was at the

lower detection threshold. Contrary to forecasts, the

hydrogen probes showed two cladding oxidation

phases –with nevertheless a proportion of steam in

the flow that was never zero. Additional measure-

ments were then carried out, after the FPT3 test,

on the fission product samples taken during the test

and on the test device by γ emission, radiography and

X-ray tomography. These measurements confirmed

moderate degradation of the bundle and no molten

pool, as was wanted. Several thousand γ spectra and

about 600 sensor signals were logged. These logs

must be analysed to compare the experimental

results with the results provided by ASTEC and

ICARE-CATHARE software, devoted to the calculation

of severe accidents in pressurised water reactors.

The FPT3 test was conducted on 18th November

2004 with a test bundle comprising 18 irradiated

fuel rods, two instrumented fresh fuel rods and a

boron carbide control rod. This was the 5th interna-

tional PHEBUS PF programme on core meltdown

accidents. The various test sequences conducted to

reproduce the phases of fuel degradation and the

release and transport of fission products were

conducted in accordance with the predefined protocol.

The test was continued until 23rd November in

thermo-hydraulic conditions in accordance with

forecasts for the phases of aerosol deposits and

iodine chemistry in the containment.

Sequential sampling operations were successfully

carried out and the instrumentation worked parti-

cularly well. The on-line γ spectrometry stations

detected the failure of claddings and the sensors,

designed to measure carbon gases, measured the

31

newsflashnewsflashnewsflashnewsflashnewsflashnews

IRSN - 2005 Scientific and Technical Report

Initial results of the PHEBUS FPT3experiment

5.11Thierry ALBIOL

Experimentation Project Section

Béatrice SIMONDI-TEISSEIRE

Laboratory forExperimentation and

Measurement ofAccidental Releases

FPT3 test bundle after degradation (tomographic images).

2005

Key dates

32

November

The FPT3 test, the 5th and last in the

PHEBUS PF international programme, was

carried out on 18th November 2004 in the

Phebus experimental reactor (see Flash on

page 31).The question of what use will be

made of the Phebus reactor after this test

remains open.

May

On 12th May 2005, the first test was

conducted in EPICUR, the new installation

for the study of the volatility of radionu-

clides in the containment in severe acci-

dent situations (Flash, page 29).

Severe Accidents and Crisis Anticipation

5.12

33IRSN - 2005 Scientific and Technical Report

Pascal Lemaître Effect of spraying on the hydrogen risk during a nuclear reactor accident,

study of changes in the characteristics of a set of water droplets in a gaseous air-steam mixture

Thesis prepared at the Laboratory of Aerosol Physics and Metrology vivaed on 12/14/04

at National institute of applied sciences in Rouen.

Fabrice Malet Experimental and numerical analysis of the propagation of turbulent premixed

flames in a heterogeneous, hydrogen-poor, moist atmosphere Thesis prepared at the Severe

Accident Physics Bureau, vivaed on 12/20/05 at Orléans University.

5.12

Theses vivaed

Position of authors in the IRSNorganisational chartArticle page

2 Department for the Prevention of Major Accidents (DPAM), central level.

4 Severe Accident Study and Simulation Laboratory (LESAG); Department for the Study and Modelling of

Fuel in Accident Situations (SEMCA); Department for the Prevention of Major Accidents (DPAM).

9 Applied Mathematics and Corium Physics Laboratory (LMPC); Department for the Study and Modelling

of Fire, Corium and Containment (SEMIC); Department for the Prevention of Major Accidents (DPAM).

13 Laboratory for the Study of Fission Products (LEPF); Department for the Study and Modelling of Fire,

Corium and Containment (SEMIC); Department for the Prevention of Major Accidents (DPAM).

18 Airborne Pollutants and Containment Study and Research Department (SERAC) ; Plants, Laboratories,

Transports and Waste Safety Division (DSU).

Laboratory for the Experimental Study of Containment, Air cleaning and Ventilation (LECEV); Airborne

Pollutants and Containment Study and Research Department (SERAC); Laboratories, Transports and

Waste Safety Division (DSU).

22 IRSN Emergency Situations and Crisis Organisation Department (SESUC); Environment and Response

Division (DEI).

28 Airborne Pollutant Dispersion and Containment Study and Research Department (SERAC); Department

for the Safety of Plants, Laboratories, Transportation and Waste (DSU).

29 Analytical Test Laboratory (LEA); Accident Experimental Study and Research Department (SEREA);

Department for the Prevention of Major Accidents (DPAM).

30 Department for the Prevention of Major Accidents (DPAM), central level.

Severe Accident Physics Bureau (BPHAG); Severe Accident and Radioactive Release Assessment

Department (SAGR); Department for Reactor Safety (DSR).

31 Experimentation Project Section (CPEX); Department for the Prevention of Major Accidents (DPAM).

Laboratory for Experimentation and Measurement of Accidental Releases (LEMRA); Accident Experimental

Study and Research Department (SEREA), Department for the Prevention of Major Accidents (DPAM).

34

5.12

Severe Accidents and Crisis Anticipation

35IRSN - 2005 Scientific and Technical Report

5.12

Cherbourg-Octeville■ Environment

Le Vésinet■ Environment and response

■ Human radiological protection

Saclay■ Safety of plants, laboratories,

transportation and waste

Clamart(Head office)Staff departments

Fontenay-aux-RosesOperational activities■ Defence nuclear expertise■ Environment and response■ Human radiological protection■ Reactor safety■ Safety of plants, laboratories,

transportation and waste

Orsay■ Environment

Mahina – Tahiti■ Environment

Agen■ Environment and response

Pierrelatte■ Response

■ Human radiological protection

Cadarache■ Environment■ Prevention of major accidents■ Human radiological protection■ Defence nuclear expertise

Les Angles – Avignon■ Response

■ Safety of plants, laboratories,transportation and waste

La Seyne-sur-Mer■ Environment

ClamartHead office77-83, av. du Général-de-Gaulle92140 ClamartTel: +33 (0)1 58 35 88 88

AgenB.P. 2747002 AgenTel: +33 (0)5 53 48 01 60

CadaracheB.P. 313115 Saint-Paul-lez-Durance CedexTel: +33 (0)4 42 19 91 00

Fontenay-aux-RosesB.P. 1792262 Fontenay-aux-Roses CedexTel: +33 (0)1 58 35 88 88

La Seyne-sur-MerCentre Ifremer de MéditerranéeB.P. 33083507 La Seyne-sur-Mer CedexTel: +33 (0)4 94 30 48 29

Le Vésinet31, rue de l’ÉcluseB.P. 3578116 Le VésinetTel: +33 (0)1 30 15 52 00

Les Angles – Avignon550, rue de la Tramontane – Les AnglesB.P. 7029530402 Villeneuve-Avignon CedexTel: +33 (0)4 90 26 11 00

Mahina – TahitiB.P. 519Tahiti Papeete, French PolynesiaTel: +689 54 00 25

Cherbourg-OctevilleRue Max-Paul FouchetB.P. 1050130 Cherbourg-OctevilleTel: +33 (0)2 33 01 41 00

OrsayBois-des-Rames (bât. 501)91400 OrsayTel: +33 (0)1 69 85 58 40

PierrelatteB.P. 16626702 Pierrelatte CedexTel: +33 (0)4 75 50 40 00

SaclayCentre CEA de Saclay91191 Gif-sur-Yvette CedexTelephone: +33 (0)1 69 08 60 00

IRSN sites

36 Severe Accidents and Crisis Anticipation

Editorial CoordinationScientific, Technical and Quality Assessment Division

Editorial CommitteeDESTQ: J. Lewi, M. Colin DSR: A. Dumas, G. Bruna

DSDRE: T. Bolognese, G. Monchaux DSU: P. Cousinou

DEI: N. Chaptal-Gradoz, D. Boulaud DEND: J. Jalouneix, D. Franquard

DCOM: M.L. de Heaulme, H. Fabre DPAM: B. Goudal

DRPH: J. Brenot, P. Monti

Articles written byIRSN

Coordination of ProductionCommunications Department, CPRP

Graphic Design and CoordinationLp active

Production AssistanceLAO Conseil, ré[craie]action, summer time

TranslationMic Assistance

PrintingIdéale Prod / JPA, Imprim Vert certified

Photo creditsIRSN IRSN/image-et-process (page 3)

IllustrationsStéphane Jungers

© Communication IRSN

ISSN no. pending

Radioactivity and Environment

Research, Assessment,Transmission of Knowledge

Safety of the Geological Storage of Radioactive Waste

Safety of Installations,Accident ScenariosIonising Radiation

and Human Health

Severe Accidents and CrisisAnticipation

5 5 Analysis, prevention and management 2

5.1 Molten corium pools in the vessel bottom 4head of a pressurised water reactor (PWR) during a severe accident

5.2 Spreading of corium - Modelling and safety 9analyses for the EPR reactor corium recovery system

5.3 Fission products in the containment 13Behaviour of organic iodides

5.4 The hydrogen risk during a severe PWR accident 185.5 ASTRID - Assessing a light water reactor accident in real time 22

5.6 RECI experiments - Iodine chemistry in catalytic 28hydrogen recombiners

5.7 SOURCE TERM programme 29

5.8 EPICUR - A new installation to study the volatility of radionuclides in the containment in severe accident situations

5.9 SARNET - A European network of excellence for severe 30accidents

5.10 IRSN/FzK cooperation on direct heating of containment

5.11 Initial results of PHÉBUS FPT3 experiment 31

5.12 Key dates - Theses vivaed 32Position of authors in the IRSN organisational chart

Severe Accidents andCrisis Anticipation

newsflashnewsflashnewsflashnewsflashnewsflashHead office77-83, avenue du Général-de-Gaulle92140 Clamart - FranceRegistered under Nanterre RCS B440 546 018

Telephone+33 (0)1 58 35 88 88

Postal addressB.P. 1792262 Fontenay-aux-Roses Cedex - France

Web sitewww.irsn.org

200Scientific and Technical Report

5

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