T. Kress, D. Powers, R. Lee and L. Soffer
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 bare 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 applications of this revision, especially for siting and design of future power plants.


V. Gustavsson
Vattenfall Energisystem AB, P.O. Box 528, S-16216 Stockholm, Sweden

In 1992-94 a PSA Level 2 Study was performed for the Ringhals 2 NPP - a three-loop W-PWR on the west coast of Sweden.

This paper gives an overview of the study with emphasis on phenomena and their importance for processes resulting in activity releases.

The following phenomena were included as possible causes of activity releases from the containment:

  • Hydrogen Deflagration/Detonation
  • Direct Containment Heating
  • In-vessel ans Ex-vessel Steam Explosions
  • Global Vessel Failure
  • Thermal Attack of Pentrations
  • Basemat Failure due to Core Concrete Interaction
The largest contribution to rapid overpressurization of the containment is derived from hydrogen deflagration.

The Level 2 Study also includes a sensitivity analysis where the impact on the results from uncertainties in the phenomena are estimated. This is shortly described in this paper together with the effect of human errors.

Finally the results of the study are summarized and discussed.


A. Jones
CEC, Ispra, Italy

Abstract not available


I. Shepherd, E. Hontanon, S. Gaillot, L. Herranz, E. Bonanni, K. Akakane
Safety Technology Institute Commission of the European Communities Joint Research Centre 21020 Ispra (Va), Italy fax +39 332 785053, phone +39 332 789489
CEA Cadarache, France, CIAMAT, Madrid, Spain, NUPEC, Japan

In the Phebus experiments fission products and structural materials from a realistic source will enter the containment vessel in the form of aerosol particles. One of the objectives of the test series is then to see whether steam condensation onto these aerosols will influence the rate at which they are removed from the atmosphere. The relative humidity is the most important parameter in determining this condensation rate so each Phebus-FP test has a different target humidity that should be reached.

The Phebus containment vessel is designed so that the temperatures of all the surfaces and the sump can be regulated: they can either be held constant or changed in a pre-determined manner. The experimenters then have to choose values of the surface temperatures that produce the desired humidity when steam and hydrogen are entering the vessel at a specified rate and afterwards when the aerosols are settling.

The first approach in determining the best surface temperature for the vessel was to run computer codes such as JERICHO and CONTAIN that are used for calculating the consequences of an accident in a full-sized reactor. The results of these calculations demonstrated a sensitivity to heat and mass transfer correlations. The codes have not been validated in geometries similar to the Phebus vessel so a priori there were few grounds for preferring one correlation to another.

In this paper we summarize blind pre-test calculations performed with the codes JERICHO, MELCOR, CONTAIN, CONTEMPT and CONT, show how the results compared with the thermalhydraulic experiments and go on to analyze the reasons for the discrepancies in the code results.

Armed with the knowledge of these tests another benchmark exercise was organized with boundary conditions corresponding to measured values from the experiments. Codes using the well-mixed hypothesis as well as multi-compartment and three dimensional codes were used. Comparing the results of the calculations with measurements contributed to further understanding of the thermalhydraulic behaviour of the Phebus-FP vessel and a better understanding of the uncertainties of the codes.


S. B. Dorofeev, A. A. Efimenko, A. S. Kochurko, V. P. Sidorov
Russian Research Centre "Kurchatov Institute", Moscow, 123182, Russia

A review of hydrogen combustion research at Kurchatov Institute is presented. Results on the spontaneous detonation scaling methodology and on the loads from different combustion and explosion modes are summarised.

Criterion for spontaneous detonation onset possibility and its application to severe accident in a nuclear power plant is discussed. Theoretical and experimental results on spontaneous detonation onset conditions are summarised. Three series of large scale turbulent jet initiation experiments have been carried out in KOPER facility (50 m3 and 150 m3). Series of jet initiation experiments in initially confined H2 - air mixtures have been carried out in KOPER facility (20 - 46 m3). Turbulent deflagration/DDT experiments were carried out in large scale confined volume of 480 m3 in RUT facility. Transition to detonation was observed at min. of 12.5% H2. Results showed, that the characteristic volume size should be used for conservative estimates in accident analysis. Series of experiments on detonation transition from one mixture to another of lower sensitivity has been carried in DRIVER facility. The experiments were aimed on the estimation of the minimum size of a detonation kernel. The received results are in a good agreement with the 7 criterion.

3D computer codes 3ET and b02 have been developed for description of loads from detonations. Series of large scale H2 detonation experiments have been carried out in RUT facility (16 - 25 %H2, two initiator locations). Experimental results form a database on detonation loads on reactor-relevant scale and complex 3D geometry. B02 code was evaluated against experimental data. Good agreement of main loading parameters was observed for fully developed detonations. Effect of DDT location on loads has been studied experimentally (UTR facility) and numerically in 1D geometry. It was found, those peak overpressures depend strongly on DDT location, and can be significantly higher than that from detonation. Impulses depend mainly on the mixture volume. Data of large scale detonation, deflagration and DDT experiments (RUT facility) confirm these observations. Experimental data on turbulent flame propogation and on resulting loads (including DDT events) were received in large-scale RUT experiments.

Recent results of combined hydrogen injection/ignition experiments are presented. The experiments are aimed on the investigation of possible consequences of deliberate ignition at dynamic conditions. Experiments include the large-scale tests on the effects of igniter location, ignition time, injection rate (0.1 - 1 kg/s) and injection point on the combustion mode. The possibility of initiation of local detonations due to ignition at dynamic conditions was observed in the tests. The experiments showed a good local mixing and large-scale hydrogen concentration nonuniformities. Multiple explosions at continuous injection were observed. The possibility of local detonations resulting from deliberate ignition should be taken into account in accident analysis.

* Research sponsored by the US NRC and KfK, Germany


C. C. Chan, W. A. Dewit, G. W. Koroll
AECL Research Whiteshell Laboratories Pinawa, Manitoba ROE 1LO CANADA

During a loss of cooling accident, hydrogen (or deuterium) can be formed due to metal-steam reaction. This hydrogen can leak into the containment building to form a combustible mixture. The pressure loading on the containment resulting from a hydrogen burn depends on whether the burn is a deflagration or a detonation. Direct initiation of detonation is very unlikely because it requires a high energy source such as solid explosive that is not present inside a reactor. However, a detonation is still possible by way of a Deflagration to Detonation Transition (DDT). For insensitive H2-air-steam mixtures, flame acceleration is the most probable mechanism for transition to detonation. This paper describes some recent results of a study on DTT resulting from flame acceleration and based on these results, establishes the criteria for DDT for these insensitive mixtures.

The propagation of a freely expanding flame is intrinsically unstable. Due to a feedback mechanism between the combustion induced flow and the combustion itself, a flame can accelerate very rapidly if obstructions are placed along its path. If appropriate conditions (in terms of the composition of the mixture, the flame speed and the configuration of the obstruction) are present, a DDT can occur. Presently, these conditions for H2-air-steam mixtures have not been determined. This paper describes a systematic study of flame propagation in a duct filled with obstacles to identify the transition limit (in terms of mixture composition), the transition distances and the critical flame speed leading to a DDT. Experiments were performed in a 28 cm diameter, 6-m-long combustion duct. Baffle type obstacles, having a blockage ratio (blocked area to cross-sectional area ratio) of 0.4 were mounted along the duct to induce turbulence in the unburned gas. Piezoelectric pressure transducers were mounted along the wall of the duct to monitor the location of a DDT as well as flame speed just prior to the transition process. Results showed that the range of mixture (in terms of the H2 concentration) for which DDT has been observed reduces as the steam concentration increases. DDT was not observed in any mixture containing more than 30% of steam. Results also showed that DDT did not occur if a flame had not accelerated to a speed corresponding to a flame Mach number greater than 1.5. The transition limits and the critical flame speed are necessary conditions for DDT; both of these criteria have to be satisfied before a DDT can occur.


G. W. Koroll1, W. A. Dewit1, W. R. C. Graham2
1Containment Analysis Branch AECL Research - Whiteshell Laboratories Pinawa, Manitoba ROE 1LO
2Chemical Engineering Branch AECL Research - Chalk River Laboratories Chalk River, Ontario KOJ 1JO

Catalytic recombination is a means of passive hydrogen removal from a post-accident containment atmosphere. It is attracting the attention of the utilities, designers and regulators as simple backfittable means of improving the margins of safety for hydrogen. The current generation of catalytic recombiners has demonstrated high capacity, resistance to fouling and generally good safety orientation under most fore seeable conditions. This paper describes progress in catalytic recombiner development, with emphasis on the activities at AECL.

The essential feature of the AECL recombiner is a novel catalyst material which is unmatched in wetproofing and thermodynamic range of operation. The catalyst is the product of the 20 years of engineering of high-quality catalysts to separate hydrogen isotopes in heavy water manufacture, a continuing key technology area in AECL. The catalyst is a nuclear product manufactured exclusively by AECL. It is successfully used to control hydrogen in other nuclear applications such as in the transport and storage of wet radioactive materials, and in fuel reprocessing operations. The active catalyst material is platinum, matrixed in an inorganic wetproofing support and bonded to a stainless steel substrate by a proprietary process. The formulation has extraordinary stability and robustness. Problems such as spalling and evaporation of wetproofing at high temperatures are eliminated. For this application, the catalyst is manufactured in thin, tough sheets with low mass for fast heat dissipation and low bulk for compact size per unit capacity.

Function and performance of the AECL recombiner were demonstrated in the 6.6 m3 and 10.7 m3 Containment Test Facility (CTF) vessels at AECL Whiteshell Laboratories (WL) using a 1/10-scale test model and a full-scale prototype recombiner. The AECL recombiner removes hydrogen (and carbon monoxide) at a high capacity and is resistant to foreseeable poisons and fouling agents in containment, during normal operation, or in an accident.

The AECL recombiner is self-starting in the range of temperatures 20 to 150oC at a low H2 concentrations (<1.0% V) and starts at low temperatures in a condensing atmosphere. Humid performance at low temperatures is considered a vital test of wetproofing effectiveness and is critically important in containments where pressure suppression systems are employed.

The nominal capacity for hydrogen recombination is 5.0 0.2 kg/h per m2 of inlet area to the recombiner in 5.0% H2 in air at 25oC. The capacity is increased by approximately 30% per metre of chimney length. The capacity is insensitive to the presence of the diluents such as steam, CO2, or N2 as long as the minimum stoichiometric amount of oxygen is available to recombiner the hydrogen. The capacity increases directly in proportion to the initial pressure in the range 1-2 atm. The capacity increases in proportion to the concentration of the limiting reactant. Finally, the capacity for hydrogen removal increases in proportion to the inlet area to the recombiner housing. This simple scaling parameter was verified in identical tests with the 1/10-scale test model and full-scale prototype recombiners.


B. Raj Sehgal, J. Andersson, N. Dinh, T. Okkonen
Nuclear Power Safety, Royal Institute of Technology
100 44 STOCKHOLM Sweden

A program of simulant material experiments is planned to investigate the physical processes that occur during the progression of a postulated melt-down accident in a light water reactor (LWR). The experiments will employ glass type melt materials at ~1000 K to 1700 K to represent the corium (UO2-ZrO2) melt. These simulant materials have the advantage that they are oxidic and their composition can be changed to widely vary the physical properties, e.g., viscosity, thermal conductivity etc. Additionally, because of their high temperature, they will model the radiation heat transfer and film boiling phenomena, which play a role in the corium melt interactions with structures and water. These materials can be electrically heated, form crusts and display slurry type viscous flow behaviour on cooling. Another advantage is that their cost is generally less than U.S.$ 1.00/kg.

Specific experiments proposed in the initial program are pertinent to (a) the interaction of corium melt and the lower head of the pressure vessel, (b) the interaction of corium melt with in-vessel and ex-vessel water pools. We expect that experiments on in-vessel ex-vessel melt and debris bed coolability will also become a part of this program of experiments on melt-structure water interactions.

The first set of experiments performed investigate the ablation process in a LWR vessel, as the melt discharges from the vessel to the containment through the failure location, be it a penetration, or the initial opening during through the creep-rupture process. The simulant material chosen is a mixture of Pbo and B2O3 which melts at ~1000 K and has very low viscosity at 1000 K and above. The vessel material chosen is lead, which melts at 600 K. The objective of these experiments is to determine if a crust formed retards the heat transfer to the vessel wall or is swept out by the flowing melt and is ineffective. The extent of the final size of the hole is affected significantly by the presence or absence of the crust.

Scaling of the hole ablation process has been performed. It appears that the experiments performed with a ? 10 to 100 l of melt will cover the values of the parameters for the prototypic accident conditions. A one dimensional code for melt flow and vessel interaction, called HAMISA (hole ablation in severe accidents) has been completed, and a two dimensional code is being prepared. The process is basically two dimensional, as has been demonstrated in the experiment with a thick plate, having an initial hole of 10 mm.


F. J. Erbacher
KfK, Karlsruhe, Germany

Abstract not available


A. Popkov1, V. Chudanov1, P. Vabushchevich1, V. Strizhov1, J. C. Lutche2, J. M. Veteau
1INS/RAS, Russia
2IPNS, France

Abstract not available


L. V. Benet, C. Caroli, P. Cornet, N. Coulon, M. Durin*, J. P. Magnaud, M. Petit
CE Saclay
91191 Gif Sur Yvette Cedex

In many countries, the safety requirements for future light water reactors include accounting for severe accidents in the design process.

As far as the containment is concerned, mitigation features allowing to limit pressure and temperature inside the building are to be assessed. There is also a need to accurately to estimate local hydrogen concentrations in order to evaluate the hydrogen risk.

For this purpose, codes have been developed in the past twenty years around the world. they generally include models for all the physical phenomena involved during severe accidents. Most of them are based on a multi-compartment approach for the spatial description. This approach have well known limitations when local concentrations must be evaluated in large volumes.

On the other hand, general purpose multi-dimensional thermal hydraulics computer codes are able to predict complex situations such as stratification occurrence, but they often lack some models which are necessary to be included in the analysis of the containment under a severe accident. The ability to treat long transients is also a concern with this type of codes.

For these reasons, CEA/DMT has undertaken the development of the GEYSER/TONUS code. This code will allow the coupling of parts of the containment described in a lumped parameter manner, together with meshed parts. Physical model of classical lumped parameter codes, such as condensation, will be adapted for the spatially described zones. The objective is to be able to treat complete scenarios and optimized numerical methods are developed for the transient problem. The code is built in the CASTEM 2000/TRIO EF system which allows, thanks to its modular conception, to construct sophisticated applications.

In this paper, the GEYSER/TONUS code is described, and some applications are shown.

* Presently at Commissariat à l'Energie Atomique, IPSN/DPEI, CE Fontenay, BP


S. R. Kinnersly
AEA Technology

Filtration is a means of reducing the release of fission products to the atmosphere during a severe accident if the containment function is bypassed. A number of Western LWRs have pre-installed filters to remove fission products if deliberate venting is needed to prevent containment failure. For UK gas cooled reactors, ad-hoc filtrations a potentially valuable measure for reducing releases if there is a leak during a severe accident and for extending the range of severe accident management options available. An ad-hoc filter would be rapidly assembled during a severe accident. The principles involved in ad-hoc filtration may be more generally applicable to other reactor designs, particularly those without a strong secondary containment or which are susceptible to containment bypass.

The design of an ad-hoc filter requires consideration of :

  • availability of materials;
  • ease of construction and operation;
  • stability of materials under heat and irradiation;
  • filter efficiency (aerosols and vapours);
  • filtration capacity;
  • decay heat removal;

The focus here will be on generic aspects of heat and mass transfer in ad-hoc filter design. Key filter parameters will be identified and discussed in terms of projected loadings on the filter. The advantages and disadvantages of alternative materials and designs will be considered.

Acknowledgement: This paper is based on work funded by the UK Health and Safety Executive and by Nuclear Electric plc.


W. C. H. Kupferschmidt, J. C. Wren, J. M. Ball
Research Chemistry Branch
AECL-Whiteshell Laboratories
Pinawa, Manitoba

Radioiodine is recognized as one of the most hazardous fission products that can be released from fuel during a reactor accident. This is due to the combination of its large inventory in fuel, short half-life and biological activity.

Furthermore, iodine has volatile chemical forms, which increases the probability of its release to the environment. However, iodine chemistry is quite complex, and the behaviour of iodine under accident conditions can be influenced by numerous variables. For example, the extent to which CsI, the most likely form of iodine to enter containment, would undergo reaction to form volatile I2 or organic iodides, is dependent on the dose rate, solution pH and impurities present within containment. Surfaces also play an important role in influencing iodine volatility, both as a sink and as a source of impurities that can increase iodine volatility.

This paper highlights recent findings in iodine research and summarizes the current state of understanding of iodine chemistry relevant to reactor safety. This paper also puts forward recommendations for minimizing iodine release from containment.

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