REACTOR CORE AND PRIMARY SYSTEM BEHAVIOUR AND CONSEQUENCES IN SEVERE
ESTIMATION OF COREMELT EVENT TREES
H. Plank, Dr. H. Weisshäupl
Siemens AG, Nuclear Engineering Department, Mail Box 3220, D-91050
Although in commercial western nuclear power plants a very high safety standard
has been reached, in future nuclear power plants preventive measures should be
complemented by mitigative measures to meet the overall objective, that even in
the worst case there will be no necessity for evacuation or relocation of the
population from the vicinity of the damaged plant.
Any device for mitigation of coremelt accidents must be suitable for a wide
range of possible scenarios. Frequently the design of mitigative measures is
based on a few enveloping scenarios. This work presents two alternative methods
with respect to characteristic in-vessel scenarios. The first one is
comprehensive MAAP3.0B study, where the selection of the scenarios is based on
the results of a probabilistic safety analysis. In this way on the one hand
more than 80% of all coremelt scenarios could be analyzed by a limited number
of calculations and on the other hand likelihood distributions of some key
results (e.g. begin of melting, slumping, RPV-failure, hydrogen mass, mass and
temperature of melt at slumping) could be generated. The second method uses the
accident progression event tree (APET) method to generate characteristic plant
damage states and their estimated likelihoods.
The MAAP study showed that core degradation begins before 3 hours in about 35%
and after 19 hours in 65% of all coremelt accidents. The duration of core
degradation (begin melting til slumping) amounts to 40-80 minutes. At the time
of slumping 50-70% of UO2 relocate to the lower plenum in 25% and
70-76% in 75% of all scenarios. The total hydrogen mass was in 75% less than
600 kg (38% Zr-oxidation), the maximum mass was 820 kg (51% Zr-oxidation).
With the APET method a more detailed melt mass - likelihood distribution at the
time of vessel failure has been obtained. The most likely (72%) melt group CM3
contains 70-90% of total UO2 followed by group CM2 (21%) with 50-70% UO2 and group CM4 (3%) with 90-95% UO2. The likelihood of
scenarios where vessel failure already occurs when the fraction of relocated
melt is low (<50% UO2) has been estimated to 2%.
CORE MELT PROGRESSION: STATUS OF CURRENT UNDERSTANDING AND
R. W. Wright
This paper will emphasize late phase (ceramic) melt progression and the more
recent developments in melt progression research. Topics to be addressed will
include the difference between wet core and dry core behavior, the major branch
point of metallic melt drainage from the core or core blockage, the threshold
for ceramic pool meltthrough from the core that determines the mass of ceramic
melt released from the core into the lower plenum, the characteristics of the
released melt, other major phenomenological effects, and the effects of core
reflooding. For completeness, the principal results on lower head integrity
under melt attack will be mentioned. The paper will include results and
interpretation from experiments, from the TMI-2 accident, and also from
CODE VERIFICATION AGAINST EXPERIMENTS
STATUS OF ATHLET-CD DEVELOPMENT SHOWN BY THE
LOFT-FP-2-ANALYSIS AS AN EXAMPLE
J. Bestele, K. Trambauer
Gesellschaft für Anlagen und Reaktorsicherheit (GRS) mbH,
Forschungsgelände, D-84748 Garching
The Gesellschaft für Anlagen- und Reaktorsicherheit is developing in
cooperation with the Institut für Kernenergetik und Energiesysteme of the
University of Stuttgart the system code ATHLET-CD (Analysis of
Thermalhydraulics of Leaks and Transients with Core Degradation). The code
consists of detailed models of the thermalhydraulics of the reactor coolant
system. It is based on a lumped parameter approach for the five equation system
of the two phase flow with drift-flux model. The thermofluiddynamics module is
coupled with modules describing the early phase of the core degradation, like
cladding deformation, oxidation and melt relocation, and the release and
transport of fission products. The assessment of the code is being done by the
analysis of separate effect tests, integral tests and plant events. The code
will be applied to the verification of severe accident management
The test LOFT-FP-2 was conducted under the auspices of the Organization for
Economic Cooperation and Development at the LOFT (Loss-of-Fluid) facility, a
1:50 volumetrically scaled pressurized water test reactor, of the Idaho
National Engineering Laboratory on July 9, 1985. The objective of the test was
the examination of fission product release and transport phenomena during a
small break loss of coolant accident. The temperature in a special designed
fuel element in the center of the nuclear core exceeded 2100 K for several
The high temperature caused the degradation of the fuel element and the release
of fission products and aerosols. The test was terminated by flooding the
reactor pressure vessel with water. At the beginning of the flooding phase the
temperature in the core increased again due to the exothermal
metal-water-reaction of the rod cladding and caused further damage of the core
with extensive hydrogen generation and fission product release.
The test was calculated from scram until the end of the quenching of the hot
core with the system code ATHLET-CD in order to verify the code for reactor
case. All important phenomena which occurred during the transient, like loss of
coolant, core uncovery, heat-up and degradation, release and transport of
fission products and aerosols and the final flooding were calculated well. The
approach of coupling the modules which describe core degradation, fission
product release and transport with the modules which describe thermalhydraulics
and neutron kinetics turned out to be successful.
VERIFICATION OF MODIFIED VERSION OF SFD INTEGRAL CODE
ICARE2 AGAINST CORA-W2 (ISP-36) EXPERIMENTAL DATA
A. Kisselev*, A. Volcheck*, V. Strijov*, A.
Porracchia**, F. Jacq**
*Institute of Nuclear Safety Institute of Russian Academy of Science (NSI RAS),
**Institut de Protection et de Surete Nucleare (IPSN), France
Recently in various countries a lot of efforts are applied for development of
integral code for analysis of NPP core severe accidents. Such kind of the
development requires an association of efforts in international co-operation.
An example of such co-operation may be the activities, connected with
modernisation and validation of physical models of a code ICARE2.
The code ICARE2 has been developed in IPSN (France) since 1987. The code was
intended for simulation of thermal hydraulic processes during severe accidents
of PWR NPP type core. A number of modules for different physical phenomena were
included in the code: mechanical destruction of fuel rods, radiation heat
transfer, main chemical reactions, etc. However the majority of models used in
the code was based on the simplified approaches, which can be applied only for
definite conditions. Complexity of described object required generalisation the
code in a part of development more physically reasonable models. Such
modernisation was conducted in 1992-1994 in co-operation NSI RAS-IPSN.
Following modules were implemented in the code ICARE2 during this period
(models were developed in NSI RAS in the frames of a detailed mechanistic code
Fitting of the data base for new physical models was done on the basis of the
analysis of a wide spectrum of separate-effect tests: P. Hofmann (Germany),
Olander (USA), Sagat (Canada), etc.
- UZRO (chemical behaviour during UO2ZrH2O
interactions) - the solution of system of equations for oxygen diffusion in
- DROP (melt relocation) - the solution of one dimensional equations for heat
and mass transfer for liquid masses on the surface of the cladding.
- CROX (fuel-cladding deformation) - the analysis of mechanical behaviour of
multilayer structure ().
Finally the code with new modules were tested on international experiments
PHEBUS-B9+, CORA-13, PBF1.4, CORA-W1 and CORA-W2 (ISP-36 - international
standard problem as result of common efforts of KfK Germany and NSI "Kurchatov
In this report the results of ICARE2 code validation on CORA-W2 experimental
data are presented.
The main objectives of CORA-W2 test were investigation of physical processes
(onset of temperature escalation, H2-generation, cladding failure, melting,
oxidation, relocation, absorber rod behaviour) as separate effects, evaluation
of their integral contribution and examination the reliability of severe
accident computer codes for experiment with WWER-1000 bundle type.
The code ICARE2 with new physical models was applied to that experiment and it
- The new modules being implemented in ICARE2 codes show adequate
operationallity in frame work of integral SFD code (non significant increase of
computing time, well correlation with other set of models and modules of
standard ICARE2 code).
- Validation results are in good agreement with the integral test CORA-W2 for
following key quantities: hydrogen generation, temperature behaviour of the
bundle on different elevations, destruction of fuel rod cladding, dynamic of
melt relocation and oxidation.
- R. Gonzales, P. Chatelard, F. Jacq. ICARE2 - a computer program for severe
core damage analysis in LWRs. Note technique SEMAR 93/33. Cadarache, le
- Code Package SVECHA. Modelling of Core Degradation Phenomena at Severe
Accidents. Preprint NSI-RAS No NSI-18-94, Moscow, 1994.
STATUS OF THE INTERPRETATION OF THE PHEBUS FPT0 TEST WITH
ICARE 2: BUNDLE DEGRADATION
C. Jamond, B. Adroguer, S. Bourdon, S.
Ederli*, G. Repetto
IPSN/CEA, CE-Cadarache France
The aim of the PHEBUS FP project is to study in an in pile test facility, under
prototycal conditions, phenomena governing the release, transport, retention
and chemistry of fission products under Light Water Reactor severe accident
conditions. Phenomena to be investigated are those taking place in the core
region, the circuit and the containment building.
FPT0 was the first test of the program and went much farther into the so called
late phase than any other integral test performed so far; it fulfilled all the
following objectives: high cladding oxidation without starvation conditions,
fuel dissolution, FP release and significant bundle degradation with loss a
rod-like geometry using a bundle of 20 fresh fuel rods plus one SIC rod
pre-irradiated for 9 days.
In this paper are presented the comparisons between the calculated evolution of
the bundle using ICARE2 code and the experimental results. In spite of the fact
that the current V2 mod1 version of the code is only validated for the
prediction of the early phase of core degradation, this code is currently used
as a tool for the understanding of the late phase degradation behaviour of the
It is shown in a first time that the calculations performed with ICARE2 code,
after adjustment of the Zr-rich melt relocation criteria (cladding
embrittlement), simulate the experimental fuel temperatures at all measurement
levels, up to the end of the oxidation escalation. During this phase,
relocations of small amounts of materials occured, inducing partial blockages
in the bundle between the two space grids, and causing locally limitation of
the clad oxidation.
During the heat up phase of the test it is shown how these local material
relocations are one of possible cause that initiated the loss of rode-like
geometry in the bundle (calculated temperatures in the shroud near mid-phase
during this final phase are overestimated); other possible processes are
These phenomena led to a formation of a molten pool in the lower part of the
In these conditions it is presently difficult to reproduce the degradation
scenario of the bundle during the heat-up phase and the final state of the
bundle. Nevertheless the flexibility of the code and the possibility to use
different modelling options enabled some observed or expected degradation
processes to be imposed: upper grid catcher effect, candling of upper rod
plugs, inner rod slumping...
Finally phenomena which are modelled in ICARE2 code, and which are important to
calculate FPT0 test have been identified. Main deficiencies concern the non
prediction of the loss of rod-like geometry and the lack of transition toward a
core debris configuration. In these areas, new development efforts are
ICARE2 LATE PHASE DEGRADATION MODELS: APPLICATION TO TMI-2
F. Fichot, R. Gonzales, P. Chatelard, B.
Lefèvre and N. Garnier
Institut de Protection et de Sûreté Nucléaire (IPSN) du CEA
- Cadarache (France)
The ICARE2 code is designed to calculate in a mechanistic way reactor core
damage in LWRs. It is developed at the Institute for Nuclear Protection and
Safety (IPSN) of CEA in France, as an analytical support of the in-pile
experimental PHEBUS SFD and FP programs. Important progress has been made in
developing numerical codes such as ICARE2 to model the early phase of severe
accidents. The late stages of core degradation involve substantial melting and
material relocation. Models have been developed in ICARE2 to describe the
formation of solid debris as a result of cladding degradation. A mechanistic
model, using classical porous media results, is implemented to calculate the
heat and mass transfers within this debris bed. In case of a molten pool
formation, free convection effects are modeled to estimate the heat transfers.
A brief description of the main models for early and late phase degradation is
provided, and a calculation of the phase 2 of TMI-2 accident is presented.
After cladding oxidation and embrittlement, a large debris bed forms in the
center of the core. The rapid melting of solid particles in the debris bed
leads to the formation of a large molten pool. This pool is supported by a
stable lower crust but expands radially and finally liquid corium flows along
the sides of the reactor core. The material relocation and thermal behaviour
calculated in this simulation seem reasonable and confirm some of the assumed
scenarios of the accident.
SOURCE TERMS - FISSION PRODUCT TRANSPORT
CEC REINFORCED CONCERTED ACTION ON REACTOR SAFETY: SOURCE
W. Balz*, B. R. Bowsher**, C. G. Benson** and
E. D. Loggia
* CEC, DGXII-F, RTD Energy, Rue de la Loi 200, B-1049 Brussels, Belgium
** AEA Technology, Winfrith Technology Centre, Dorchester, Dorset DT2 8DH,
The CEC is supporting a major programme on reactor safety research. This
programme is organized through a Reinforced Concerted Action, whereby effort
and data from organizations within the European Union are contributed to the
action. The total funding for this work is 62 MEcu with the CEC contributing
The programme is divided into eight areas including the Source Term Project.
The objective of this project, which involves eighteen European organizations,
is to establish a consensus on the state of knowledge and the key areas of
uncertainty affecting the source term under severe accidents. Discussions to
define the Source Term Project concluded that the optimum use of resources
would be gained by dividing the programme into two phases. The first year
(1993) was based on a series of assessments, sensitivity studies and
state-of-the-art reviews to identify uncertainties in the database and areas of
consensus. Work in the second phase (1994/95) is benchmark studies. The work
encompasses fission product release from fuel, transport in the reactor coolant
circuit and behaviour in the containment. Benchmark activities in support of
the Phebus-FP project are also undertaken.
The main results of the assessment activities are described in this paper,
together with preliminary results from the experimental work conducted in the
THE RELEASE AND TRANSPORT OF LOW VOLATILITY FISSION
PRODUCTS UNDER SEVERE ACCIDENT CONDITIONS
R. R. Hobbins*, D. J. Osetek**
* RRH Consulting
** Los Alamos Technical Associates
The impetus to understand the release and transport of low volatility fission
products under severe accident conditions is the potentially significant
contribution of these radionuclides to the dose consequences calculated from
severe reactor accidents. The relative dose contribution of these radionuclides
to the total effective dose equivalent for a severe accident is documented.
Experimental data and theoretical and empirical models on the release and
transport of low volatility fission products are reviewed, and directions of
new work to enhance understanding in this area of severe accident technology
are identified. Current information indicates that ceramic crusts readily from
around molten fuel and limit the releases of low volatility fission products.
Additionally, temperature gradients in the release pathway cause strong
deposition of the low volatility fission products released from the fuel.
Further research is suggested to evaluate these phenomena and quantify their
effects on low volatile fission product release. The research should include
in-pile and out-of-pile experiments using prototype conditions to permit the
formation of crusts around accumulations of molten fuel and provision to
measure fission product release, timing of release, and deposition at high
LWR SEVERE ACCIDENT SOURCE TERM: PART 1: FISSION PRODUCT
RELEASE, TRANSPORT AND BEHAVIOR IN CORE AND PRIMARY SYSTEMS
T. Kress, D. Power, R. Lee and L.
U. S. Nuclear Regulatory Commission, Washington, DC 20555
Fission product releases to the environment, or source terms, arise as a result
of a highly diverse group of phenomena involved in any particular severe
accident sequence. For light water reactors (LWRs), these phenomena include
fission product release, transport and behavior in core and primary system and
in the containment. They include core heatup, fuel element degradation and
melting, pressure vessel attack and failure, possibly high pressure melt
ejection, interaction of core debris with concrete, retention of fission
products within the reactor coolant system, effects of hydrogen burns or
detonations, retention of fission products by suppression pools or ice beds,
late revolatilization of fission products from surfaces, and, clearly, the
effect of containment integrity or containment bypass and time and location of
containment failure, if it occurs. Because of the multiplicity of accident
sequences that can occur for a given plant as well as the diversity of the, as
yet, imperfectly understood severe accident phenomena, it is not surprising
that probabilistic risk assessments such as, for example, those documented in
NUREG-1150 have indicated large uncertainties in source terms which represent a
significant contribution to the uncertainty in the absolute value of risk.
Because of the difficulty and expense involved in performing prototypic
experiments, substantial reliance has been placed on the development and
validation of detailed mechanistic computer codes for analyzing severe accident
phenomena and the source terms associated with them. This paper discusses the
extensive research and other efforts that have taken place over the last decade
to address the technical issues which bear on being able to describe
quantitatively the source term(s) and its characteristics. It also summarizes
our present state of knowledge and points out areas where additional research
will add further to our understanding. Finally, this paper discusses the NRC's
efforts in revising the licensing source term (TID-14844) and the implications
of this revision, especially for siting and design of future power plants.
A STUDY OF DIFFUSIOPHORETIC PARTICLE DEPOSITION IN A STEAM
GENERATOR TUBE IN PWR PRIMARY CIRCUIT USING THE SOPHAEROS CODE
M. Missirlian, G. Lajtha*, M. Cranga
Institut de Protection et de Sûreté Nucléaire (IPSN) du
* Institut for Electric Power Research Co. Budapest (Hungary)
First a brief description is given of recent progress in connection with the
diffusiophoretic mechanism modelling in the PWR primary circuit fission
products transport code SOPHAEROS V1.1. This is followed by some calculations
for a steam generator tube of Pressurised Water Reactor geometry showing that
steam condensation increases the retention of fission products aerosols under
the influence of the diffusiophoretic mechanism, giving a clear indication of
the importance of this phenomenon in accident sequences.
MODELLING OF LATE-PHASE PHENOMENA
THERMALHYDRAULIC BEHAVIOR OF A MOLTEN CORE WITHIN A
STRUCTURE, SIMULATED WITH THE TOLBIAC CODE
B. Spindler, G.-M. Moreau, S. Pigny
Commissariat à l'Energie Atomique, FRANCE, Service de Thermohydraulique
des Réacteurs, Centre d'Etudes Nucléaires de Grenoble, 17 rue des
Martyrs, F38054 GRENOBLE CEDEX 9
The purpose of the TOLBIAC code is the simulation of the behavior of a molten
core within a structure (vessel lower head or external core catcher), taking
into account the wall ablation and the relative position of the molten core
Three components are described. The liquid metal phase corresponds to the
metals issued from the molten core and the possible ablation of a metallic wall
(pressure vessel). The liquid oxide phase is divided into heavy oxides issued
from the core and ligth oxides generated by the wall ablation. Finally the gas
phase represents the gas volume above the molten core, and those possibly
generated by the wall ablation.
Four mass balance equations (metals, oxides, light oxides, gas), two energy
balance equations (metals and oxides), and three momentum balance equations
(metals, oxides and gases) are written. The discretization uses either a
rectangular or a cylindrical 3D meshing. The equations are solved with a
two-step quasi-implicit method.
The constitutive laws of TOLBIAC (interfacial heat and mass transfer, wall heat
transfers) are taken from the literature, or very simple models with current
trends. The molten core configuration is mainly a stratified one: heavy oxides
at the bottom, above it a metal layer and finally a layer of the light oxides
generated in the liquid metal in contact with the wall. Other configurations
may also occur, depending on the wall position (vertical or horizontal) and
nature (metal, oxide with or without gas generation). A more or less high
mixing of the phases then takes place near the walls, and the constitutive laws
have to represent it.
However, the physical properties of the components (mixing of several
materials) are not well known, and the constitutive laws may therefore be not
The crusts formation at the wall, as well as at the upper surface, is taken
into account, as a thermal resistance and a variation of the liquid volume.
The ablation of the solid structure is calculated: the geometry of the cells in
contact with a solid wall changes with time. Liquid and gas are generated,
depending on the nature of the wall material. The ablation velocity depends on
the wall heat transfer coefficient and on the ratio between the conduction heat
flux in the wall and the total wall heat flux. It is calculated by solving the
thermal equation in the wall with a fine meshing near the ablation front.
A 1D conduction calculation is performed in order to determine the wall
temperature profile. A wall typically consists in a concrete layer in contact
with the molten core, followed by a steel layer, with imposed external
temperature and heat transfer coefficient.
The fission energy source is modeled and also the heat losses at the upper
surface (radiation or exchange with a water layer).
The qualification of the code is under progress. For a better physical
representation, the constitutive laws have to be improved.
Some results are presented in order to illustrate the code potentialities: the
behavior of a molten core in a vessel lower head on one hand, and in an
external core catcher on the other hand.
MOLTEN POOL MODELLING IN PWR SEVERE ACCIDENT SCENARIO
B. Spindler, J. P. Van Dorsselaere
DER/SERA, Cadarache, France
During severe accidents in PWR, the formation and behaviour of a pool of molten
materials ("corium") are important phenomena which may modify the accident
scenario. We propose here a simplified transient model of a liquid molten pool,
which can be integrated into scenario codes. This model yields the pool
temperature distribution and the wall fluxes, for a homogeneous corium in any
axisymetrical geometry varying with time. Two natural convection flow regimes
are studied: laminar and turbulent. The model used as place dicretization grid,
adapted at each time step to represent the flow in the pool. The heat transfer
coefficients between pool and walls are issued from correlations. The energy
equation is discretized by the Finite Volume Method, implicitly with respect to
time. This model was programmed first in an independent module, which
calculates the pool transient behaviour until the steady-state regime. It was
validated on two experiments: in water, then in UO2. Then, the
coupling of this independent module with the ICARE2 code was performed. This
coupling showed the model is satisfactorily running when integrated in a core
REVIEW OF MODELS AND CODES DESCRIBING WWER AND RBMK
REACTORS FIRST OUTLINE SYSTEMS BEHAVIOR UNDER HEAVY ACCIDENTS
A. A. Tutnov
Russian Research Centre "Kurchatov Institute", Moscow, Russia
Prediction of elements and nods of WWER and RBMK nuclear power plants behavior
under heavy accidents is necessary for NPP safety analysis. Complex of models
and program cods intended for assessment of pipelines, reactor vessels, steam
generator collectors and other elements of first outline equipment under normal
conditions and under accidental situations like different diameter pipeline
rupture was produced in RRC "Kurchatov Institute". Analysis of different
accidental situations appearance but without taking into account of dynamic
effects which are possible under pipeline rupture is conducted with the help of
"MAVR" and "CORPUS" cods. The last circumstance demand the solving of complex
mechanics problems connecting with moving of systems, consisting of elastic
bodies and liquids transferring along them, in particular hydraulic shock in
pipeline. Calculation of unstationary SSS appearing in the pipeline under
flowing stream perturbation is connected by finite cladding elements method,
each of them dynamic behavior is described by wave equations system of
thinwalled cladd theory. Mathematical simulation of quick running hydrodynamic
processes in liquid under pipeline depressurization is based on the
hydrodynamic equations for compressible liquid and also on thermodynamics of
phase conversions considerations. There are calculation schemes and some
calculation results in the paper.
NUMERICAL DECISION OF STEADY AND UNSTEADY PROBLEMS OF
THERMAL CONDUCTIVITY AND THERMAL ELASTICITY FOR REACTOR PRESSURE VESSEL
V. D. Loktionov, N. Yu. Medvedeva, S. E.
Krivulya, N. I. Yaroshenko
Electrogorsk Research & Engineering Center of NPP Safety
Code CORPUS simulating thermomechanical behaviour of PWR bottom during melting
interaction with Reactor Pressure Vessel (RPV) bottom is developed. Temperature
fields, stress and strain in the body reactor arising during melt interaction
under severe accident are investigated. Dependence of thermal characteristics
of reactor body material from temperature, complex behaviour of heat transfer
(convection, radiation), phase transformations on the "melt-body" boundary
under body material melting are considered. The problems are solved in
three-dimensional statement on the base of Finite Element Method (FEM).
Calculation methods, algorithms and programming code were developed.
Object-Oriented Design is used for numerical modelling of reactor body
behaviour under severe accident.
CORIUM-WATER INTERACTION STUDIES IN FRANCE
G. Berthoud, C. Brayer, M. Valette
COMMISSARIAT A L'ENERGIE ATOMIQUE, FRANCE, Direction des Réacteurs
Nucléaires, Département de Thermohydraulique et de Physique,
Service de Thermohydraulique des Réacteurs, Centre d'Etudes
Nucléaires de Grenoble, 17, rue des Martyrs - F 38054 GRENOBLE CEDEX 9
For some years, a multidimensional multiphase code (MC3D) is developed in
France in order to describe the different phases of an interaction. Up to now,
most of the work is devoted to the premixing phase and the code is validated
against an analytical experiment using solid balls heated up to
2200o K (BILLEAU) and more global experiments
using hot liquid melt: FARO (JRC ISPRA) using 150 kg of
UO2-ZrO2 and PREMIX (KfK) using 10 kg of
Apart from this, modelling of the propagation phase -the explosion itself- is
underway and will be presented.
ESCALATING AND PROPAGATING MELT/COOLANT INTERACTIONS IN
THE KROTOS EXPERIMENTS: STATUS OF KNOWLEDGE
H. Hohmann, D. Magalion, A. Yerkess, I.
Commission of the European Communities; Ispra, Italy
M. Corradini, J. Tang, B. Shamoun
Nuclear Engineering and Engineering Physics, University of Wisconsin-Madison,
M. Burger, M. Buck, E. V. Berg
Institut für Kernenergetik und Energiesysteme, University of Stuttgart
The experimental program KROTOS at JRC Ispra is quite unique because it
investigates the energetic FCI in a well-controlled one- or two-dimensional
geometries with the advantage of being able to use fuel simulants (Sn,
Al2O2) or more recently fule melt
(UO2-ZrO2). The KROTOS experiments /1/3/ have become one
of the main reference experiments for investigating the escalation and
propagation (of thermal detonation) behavior of interactions in melt/coolant
systems (vapor explosions).
An additional feature has been introduced in the last experiments by increasing
the diameter of the test tube to 20 cm, in order to facilitate penetration of
the melt against strong steaming for
UO2-ZrO2 (KROTOS-37) and checking the effect against Al2O3 (KROTOS-38). However, the substantial difference in
behavior remained. Increasing the tube diameter while not changing the
conditions of melt introduction yields in principal a step towards 2D behavior.
While 1D conditions favor specific checks of thermal detonation modeling at a
first level, stepwise introduction of 2D conditions enables specific checks of
2D effects such as weakening by lateral expansion (or "explosion
- The results from experimental series with Sn, Al2O3
and UO2-ZrO2) melts demonstrate a strongly different
behavior: only mild, barely sustained propagation for Sn (main reference
experiment: KROTOS-21), strong escalations to high, supercritical pressures of
~100 MPa with Al2O3 (main reference experiment:
KROTOS-28), applying the same trigger, and strong steam production and fine
fragmentation already during premixing with UO2-ZrO2,
currently preventing triggering and energetic interactions (KROTOS 32 to 37).
- The experiments can be taken as essentially one dimensional (test tube of
9.5 cm inner diameter and height of 1.25 m, upwards wave propagation after
triggering from below).
- Indications on the integral premixture conditions are at least given by the
level-meter measuring the level swell from which the total steam content can be
estimated, the thermocouples indicating the arrival of melt at certain
- A series of pressure transducers along the test tube allow detailed
measurements of the behavior after trigerring. These serve as a basis for
validating thermal detonation codes. From comparisons with resulting shock
wave velocities furhter indications can directly be obtained on the composition
of the premixture. In addition pressurization of the expansion vessel is also
Since in KROTOS-21 and KROTOS-28 the triggering and escalation behavior is best
characterized, these experiments give the basis for checking thermal detonation
codes. At present, calculations of KROTOS-21 with IDEMO/1,5/, ESPROSE/4,6,7/
and TEXAS-IV/8,9/, calculations on KROTOS-28 with IDEMO/3,11/ and ESPROSE/4,7/
have been published. Calculations on another Al2O3-experiment (KROTOS-26) have
also been presented in /4,8/; although early triggering makes comparisons
The pressure development in the tests can be predicted with the different codes
reasonably well. Essential features of the processes can be reproduced and thus
the codes appear to be in general appropriate to describe the phenomena,
although details of the experimental results have not yet been reproduced.
However, different assumptions on the premixture and the dominant processes
have been used, yielding different interpretations of the experimental results.
In /1,5/ it has been shown with IDEMO that hydrodynamic fragmentation is at all
not sufficient to explain the sustained pressure waves and short-time peaks
with the Sn-experiments. Thus, an additional, strong contribution of thermal
fragmentation has been assumed as a possible explanation suggesting also other
possibilities, such as heterogeneous water heating. This view is in principle
supported by TEXAS-IV calculations in /8/, however with a much smaller thermal
fragmentation which was sufficient due to emphasizing steam production. On the
other hand, calculations from ESPROSE/6/ emphasize heterogeneous water heating
as well as steam production showing hydrodynamic fragmentation to be sufficient
under these assumptions.
Nevertheless, the task remains to understand the importance of the respective
contributions and effects and, finally, to come to a general description which
can be extrapolated to a wide range of conditions. This can also be supported
by attempts to get more information on the premixture conditions, e.g., by
calculations with premixing codes. Such calculations have already been
performed with TEXAS/8/, however further confirmation is needed as well.
The situation with the Al2O2 experiments can be
characterized similarly. Here, agreement between the calculations with
IDEMO/5,11/ and ESPROSE/4,7/ exists that the hydrodynamic fragmentation is
strong enough to explain the rapid escalation to very high pressures obtained
experimentally. A stronger escalation from ESPROSE may be attributed again to
the assumption of nonhomogeneous heating of water in contrast to the IDEMO
calculation. Again, there remains a need to clarify the details of behavior and
thus to come to a generalized description of the exchange processes.
In the present contribution the above status of analysis of the Sn and
Al2O2 experiments, especially KROTOS-21 and KROTOS-28
will be discussed, considering and evaluating in detail the models applied in
the different codes and the main questions posed from this investigations and
the experimental results. Based on this information, new calculations with
IDEMO and TEXAS IV will be performed and discussed in direct, detailed
comparison. Concerning the different behavior in the
UO2-ZrO2 experiments, the main questions will also be
discussed and elaborated. Since these are obviously related to the premixing
behavior and more specifically to the breakup of the melt stream, first
attempts to clarify the differences will be undertaken based on film boiling
and jet breakup considerations/8,10,12/.
The final intention of this study is to determine the various features which
characterize the mixing, pressure escalation and propagation phases which occur
during melt/coolant interactions the thus discover why these characteristics
may be substantially different for Sn,
Al2O2 and UO2-ZrO2 melts. This
raises the concept of scaling of the explosion phenomena and what must be
preserved between various tests.
- Burger, M. et al., (IKE), Schins, H., et al., (JRC Ispra): Examination of
Thermal Detonation Codes and Induced Fragmentation Models by Means of Triggered
Propagation Experiments in a Tin/Water Mixture, Nucl Eng Design 131 (1991)
- Hohmann, H., Hegallon, D., Schins, H., Yerkess, A.: FCI Experiments in
Aluminum Oxide/Water System. CSNI Specialists Mtg. on Fuel-Coolant
Interactions, Santa Barbara CA, Jan 5-8, 1993.
- Hohmann, H., et al.: Advance in the FARO/KROTOS Melt Quenching Test Series,
22nd Water Reactor Safety Information Mtg., Bethesda, Maryland, October 23-26,
- Yuen, W. W., Theofanous, T. G.: The Prediction of 2D Thermal Detonations
and Resulting Damage Potential, CSNI Specialist Mtg on Fuel-Coolant
Interactions, Santa Barbara CA, Jan 5-8, 1993.
- Burger, M., Buck, M., Muller, K., Scahtz, A.: Stepwise Verification of
Thermal Detonation Models: Experimentation by Means of the KROTOS Experiments.
CSNI Specialist Mtg. on Fuel-Coolant Interactions, Santa Barbara CA, Jan. 5-8,
- Yuen, W. W., Chen X., Theofanous, T. G.: On the Fundamental
Microinteractions that Support the Propagation of Steam Explosions, Proc.
NURETH-5, Salt Lake City UT, Sept. 22-24, 1992, Vol II, 627-636.
- Theofanous, T. G., Yuen, W. W.: The Prediction of Dynamic Loads from
Ex-vessel Steam Explosions, Int. Conf. on “New Trends in Nuclear System
Thermohydraulics,” May 30-June 2, 1994, Pisa Italy.
- Tang, J., Corradini, M. L.: Modeling of the Complete Process of
One-Dimensional Vapor Explosions. CSNI Specialist Mtg. on Fuel-Coolant
Interactions, Santa Barbara CA, Jan 5-8, 1993.
- Bang, K. H.: The Role of Fragmentation Rate in Vapor Explosion Propagation:
Comparison of Models, Proc. Int. Seminar on the Physics of Vapor Explosions,
Tomakomai, Japan, Oct. 25-29, 1993, pp 229-233.
- Burger, M., Cho, S. H., v Berg, E., Schatz, A.: Breakup of Melt Jets as
Precondition for Premixing, Modeling and Experimental Verification. CSNI
Specialist Mtg. on Fuel-Coolant Interactions, Santa Barbara CA, Jan 5-8, 1993.
- Burger, M., et al.: Analysis of the Thermal Detonation Experiment KROTOS-28
with the IDEMO Code. Intl “Energic-Technic-Unmelt,” U. Brochmeier (Ed),
Ruhu-Univ Bochum, Inst. fur Energietechnik, Buchum, Germany, 1994.
- Burger, M., v Berg, E., Cho, S. H., Schatz, A.: Modeling of Jet Breakup on
a Key Process in Premixing Proc. Int. Seminar on the Physics of Vapor
Explosions, Tomakomai, Japan, Oct. 25-29, 1993, pp 229-233.
SEVERE ACCIDENT MITIGATION CONCEPT FOR THE EPR AND
The defense-in-depth concept of safety employed in commercial nuclear power
plants has led to a very high safety standard. Nevertheless for future designs
in addition to an improved accident preventional level measures for mitigation
of severe accidents with core melt down are generally envisaged.
To cope with the consequences of a severe accident means to deal with different
phenomena which may threaten the integrity of the containment or may lead to an
enhanced fission product release into the environment:
For the European Pressurized Water Reactor (EPR), jointly developed by French
and German industry, dedicated measures are planned to cope with these
- high pressure reactor pressure vessel failure
- energetic molten fuel-coolant interaction (in-vessel and ex-vessel)
- direct containment heating
- molten core concrete interaction (basemat meltthrough)
- hydrogen combustion
- long term pressure and temperature increase in the containment.
In addition special design features are foreseen, as e.g. a limitation of an
energetic water contact with ex-vessel melt, and a strong reduction of fission
product release into the environment by a double containment with collection of
the fission products in the annulus, vented by an emergency standby filter.
- for primary side depressurization
- for hydrogen control
- for stabilization and cooling of the melt
- for containment heat removal.
To show the adequacy and feasibility of these new features quite a lot of R&D
work has to be relied on and still to be performed, starting with generic
questions which help to get more insight into the phenomena involved going to
more design related questions to proof and demonstrate the adequacy of chosen
Research and development cooperations have recently been intensified to achieve
this ambitious goal. The German Forschungszentrum Karlsruhe and the French
research institutions of the CEA are working closely together with the industry
having between themselves a strong cooperation in performing experimental and
analytical investigations, complementing and supporting each other. Generic
investigations of different institutes, institutions and universities partly
sponsored by the federal government are necessary and helpful in understanding,
calculating and verifying the phenomena involved. Within a greater frame
parties from the European Countries have found together withing the nuclear
fission safety programmes of the European Commission to reach a common
understanding and to find solutions to open questions. Agreements and an open
information exchange between the worldwide parties involved in nuclear energy
are rounding up the picture.
EFFECTIVENESS OF THE MODERATOR AS A HEAT SINK DURING A
LOSS-OF-COOLANT ACCIDENT IN A CANDU-PHW REACTOR
D. B. Sanderson, R. G. Moyer and R. Dutton
AEL Research, Whiteshell Laboratories, Pinawa, Manitoba ROE 1LO
In a CANDU* pressurized heavy-water (PHW) reactor, the fuel and coolant are
separated from the heavy-water neutron moderator by horizontal fuel channels.
The fuel channel consists of a Zircaloy-2.5 Nb pressure tube and a Zircaloy-2
calandria tube, separated by a gas-filled annulus. The separation between the
primary heat transport fluid and the moderator is a distinctive feature of
CANDU reactors. In particular, the large volume (~250,000 litres) of cool
moderator provides a backup heat sink for the reactor core, should both the
normal and the emergency cooling system fail. This in-situ heat sink (about 10
mm away from the hot pressure tube) provides an inherently passive heat removal
mechanism in the event of loss-of-coolant accident (LOCA) scenarios.
Several events which are postulated in the licensing and safety assessment of
CANDU reactors result in a degradation of the normal heat removal mechanisms
from the fuel. In the more severe of these incidents, the pressure-tube
temperatures may increase as a result of convective heat transfer from the
hot coolant or steam in the channel, and due to thermal radiation from the
overheated fuel bundles. If the pressure tube becomes hot enough, it will
deform and contact its surrounding calandria tube. Upon contact, stored heat in
the pressure tube is transferred across the interface to the calandria tube,
conducted through its wall, and transferred from its outside surface into the
surrounding moderator. This process significantly improves the transfer of heat
from the fuel bundle to the moderator, helping to reduce fuel temperature and
fission product releases.
Verification of the effectiveness of the moderator as a heat sink during a LOCA
has been the focus of an extensive and ongoing research program at
AECL-Research Whiteshell Laboratories. The general methodology used in the
research program has been to perform small-scale separate-effects experiments
to develop and validate mathematical models which describe the phenomena. These
validated models are then integrated into a code (CATHENA) linking the various
phenomena to characterize the fuel channel response to a LOCA. Full-scale
integrated experiments are then performed to validate the code.
This paper describes the operation of the moderator as a passive heat sink in a
CANDU reactor and recent experiments that verify its operation in certain
severe accident situations. Recent code development and validation work, aimed
at modelling this phenomenon, are also reviewed.
* CANada Deuterium Uranium, registered trademark of AECL.
COOLABILITY OF SEVERELY DEGRADED CANDU CORES
D. A. Meneley1, C.
Blahnik1, J. T. Rogers2, V. G.
Snell1 and S. Nijhawan3
1 Atomic Energy of Canada Ltd.
2 Dept. of Mechanical and Aerospace Engineering, Carleton
University, Ottawa, Ontario, Canada
3 MIR Consulting, Toronto, Ontario, Canada
Analytical and experimental studies have shown that the separately cooled
moderator in a CANDU reactor provides an effective heat sink in the event of a
loss-of-coolant accident (LOCA) accompanied by total failure of the emergency
core cooling system (ECCS). The moderator heat sink prevents fuel melting and
maintains the moderator heat sink prevents fuel melting and maintains the
integrity of the fuel channels, therefor terminating this severe accident short
of severe core damage.
Nevertheless, there is a probability, however low, that the moderator heat sink
could fail in such an accident. The pioneering work of Rogers (1984) for such a
severe accident using simplified models show that the fuel channels would fail
and a bed of dry, solid debris would be formed at the bottom of the calandria
which would heat up and eventually melt. However, the molten pool of core
material would be retained in the calandria vessel, cooled by the independently
cooled shield-tank water, and would eventually re-solidify. Thus, the calandria
vessel would act as an inherent "core-cathcer".
The present paper reviews subsequent work on the damage to a CANDU core under
severe accident conditions and describes an empirically based mechanistic model
of this process. It is shown that, for such severe accident sequences in a
CANDU reactor, the end state following core disintegration consists of a porous
bed of dry solid, coarse debris, irrespective of the initiating event and the
core disassembly process.
The paper describes an improved model for the subsequent heat-up of the debris
bed which includes all the significant mechanisms that may affect the thermal
behaviour of the debris. The paper also describes a detailed model for the
thermal behaviour of the molten pool formed by eventual melting of the solid
debris bed which takes into account internal heat generation and
buoyancy-driven circulation in the pool, the formation of solidified crusts on
the upper and lower surfaces of the pool, heat generation in the crusts,
radiation from the upper crust and radiation absorption, emmision and
transmission in the steam over the pool, as well as the natural convection and
nucleate boiling processes of the shield-tank water on the outer surfaces of
the calandria vessel.
Application of these models to a dominant-frequency severe accident sequence in
a CANDU-6 reactor is described. Results for reference conditions for the
thermal transient of the debris bed show that its heat-up is relatively
slow, with melting in the interior of the bed beginning about two hours after
heat-up begins. However, the upper and lower surfaces of the debris remain well
below the melting point and heat fluxes to the shield-tank water are well below
critical heat fluxes for the existing conditions. The calandria vessel remains
well-cooled and retains its integrity throughout the transient. Sensitivity
studies in which important parameters are varied over wide ranges yield the
same conclusions, with the results indicating that, for larger pore sizes,
melting of the debris may not even occur.
Results for the thermal behavior of the subsequent molten pool show that the
calandria vessel wall will be protected by a thick solid crust below the pool
which will grow with time. Calandria wall temperatures and heat fluxes to the
shield-tank water ensure that the calandria vessel will maintain its integrity
in this phase of the transient also. Again, sensitivity studies varying the
important parameters confirm the general validity of this result. Should the
shield-tank water eventually boil off, the calandria vessel will fail, but thsi
will not occur in less than a day, giving the operators adequate time to
provide water from emergency supplies to the shield tank, which will then
retain the re-soldifying core indefinitely.
Contributors to the predicted effectiveness of cooling of a degraded core in a
CANDU, in addition to the inherent heat sink provided by the separate moderator
and the shield tank water are the low power density, about 15.6 MW/Mg of fuel
in a CANDU-6 based on the design power, and the extensive dispersion of the
debris bed in the calandria resulting in a shallow molten pool depth of about 1
metre maximum and about 0.65 metre average for the reference case for a
CANDU-6. These factors help to explain the different predicted behaviour of a
degraded CANDU core relative to that of an LWR in typical melt-progression
The results of this work confirm and provide more confidence in the conclusions
of the early studies that the calandria vessel will retain its integrity in
severe accidents in a CANDU reactor and will contain a disintegrated or molten
core for a long period without operator intervention, thus acting as an