Severe Accident Conditions Research Articles

Fission product release from nuclear fuel II. Validation of ASTEC/ELSA on analytical and large scale experiments

This article is the second of two articles dedicated to the mechanisms of fission product release from a degraded core. The models of fission product release from nuclear fuel in the ASTEC code have been described in detail in the first part of this work (Brillant et al., this issue). In this contribution, the validation of ELSA, the module of ASTEC that deals with fission product and structural material release from a degraded core, is presented. A large range of experimental tests, with various temperature and conditions for the fuel surrounding atmosphere (oxidising and reducing), is thus simulated with the ASTEC code. The validation database includes several analytical experiments with both bare fuel (e.g. MCE1 experiments) and cladded fuel (e.g. HCE3, VERCORS). Furthermore, the PHEBUS large-scale experiments are used for the validation of ASTEC. The rather satisfactory comparison between ELSA calculations and experimental measurements demonstrates the efficiency of the analytical models to describe fission product release in severe accident conditions.

Distribution of fission product residual decay heat in stratified core melt of LWR and its influence on sidewall heat flux

The model of spatial distribution of fission product (FP) activity and decay heat in core melt of light water reactor under severe accident conditions is proposed. It is based on thermodynamical consideration of “U–Zr–O–Fe–FP” system as a two-phase (oxide and metallic) stratified melt. The thermochemistry of FP phase distribution, heat-mass transfer and radioactive transformations of ∼100 isotopes of Cs, Mo, Ru, Ba, Sr, Zr, La, Ce, Eu, Nd, Nb and Sb are considered self-consistently. The calculations carried out with specially designed code FPMC (fission products–molten core) showed that up to 15–25% of decay heat may be released in the metallic phase of the melt. Estimation of this value is fundamentally important in simulations of heat and mass transfer in stratified melt, such as in-vessel melt retention and reactor lower head failure. For numerical assessment of the influence of decay heat spatial distribution in stratified molten pool on the lower head failure time the multivariate simulation of the later stage of large break loss-of-coolant accident in VVER-1000 reactor was carried out by SOCRAT and HEFEST codes coupled with FPMC.

Advances in mechanistic understanding of iodine behaviour in PHEBUS-FP tests with the help of ab initio calculations

Analysis of the results of the Phébus experimental program identified phenomena related to fission product behaviour that are insufficiently described by available empirical models. The combining experimental and theoretical strategy implemented by IRSN to develop models that are able to predict any pressurised water reactor (PWR) severe accident transient is illustrated here for one of these phenomena, the presence of gaseous iodine at the break of the reactor coolant system (RCS). Kinetics modelling of the {I,O,H} chemical system in the RCS under severe accident conditions computed using quantum chemistry and validated against separate-effect experiments are used to re-interpret with a notable improvement the FPT1 experimental data contrary to the FPT3 one. After being implemented in the ASTEC (Accident Source Term Evaluation Code) software, these new models will contribute to a reduction in uncertainties in the prediction of the iodine source term.

Hydrogen hazard and mitigation analysis in PWR containment

This paper describes the analytical results for the estimation of hydrogen concentration in the containment atmosphere based on zirconium oxidation reaction following a severe accident. The analysis provides useful information about the potential challenge of local hydrogen accumulation in the containment, which may be used to reduce the hydrogen detonation risk and to design the capacity and arrangements of mitigation measures. The containment analysis has been performed using computer code COGAP which uses the scenario of loss of coolant accident. The behavior of pressure and hydrogen concentrations in containment as a function of time under the severe accident condition is presented in graphical form. The mitigation measures (recombiners) are essential to maintain containment atmosphere in the safe stable conditions. A hydrogen control system is to mitigate the hydrogen risk by comparing results from a reference accident sequence with and without recombiners. This comparison show that combustible gas occur in few local areas in the containment for a limited time span and hydrogen concentration is reduced significantly with the use of recombiners.

Restriction of Iodine Release under Severe Accident Conditions at NPP MIR.1200

Restriction of Iodine Release under Severe Accident Conditions at NPP MIR.1200

The COLIMA experiment on aerosol retention in containment leak paths under severe nuclear accidents

CEA and RSE managed an experimental research concerning the investigation of aerosol retention within concrete containment cracks under severe accident conditions. The main experiment was carried out in November 2008 with aerosol generated from the COLIMA facility and a sample of cracked concrete with defined geometric characteristics manufactured by RSE.The facility provided representative conditions of the aerosols suspended inside the containment of PWRs under a severe accident. Prototypical aerosol particles were generated with a thermite reaction and transported through the crack sample, where surface characteristics, temperature, pressure drop and gas leakage were properly reproduced.The paper describes the approach adopted for the preparation of the cracked concrete sample and the dimensioning of the experimental apparatus, the test procedure and the measured parameters. The preliminary results, obtained from this single test, are also discussed in the light of the present knowledge about aerosol phenomena and the theoretical analyses of particle behaviour with the crack path.

Prediction of Boiling-Induced Natural-Circulation Flow in Engineered Cooling Channels

Boiling-induced natural-circulation flow in various engineered cooling channels is modeled and solved by considering the conservation of mass, momentum, and energy in the two-phase mixture, along with the two-phase friction drop and void fraction. The model is applied to estimate the induced mass flow rates through a uniform annular gap and a nonuniform annular gap between the reactor vessel and insulation under the in-vessel corium retention-external reactor vessel cooling conditions, and in the engineered corium cooling system of an ex-vessel core catcher during a severe accident. Dependence of the induced flow rate on various system parameters including the channel gap size, inlet diameter, inlet subcooling, and wall heat flux has been identified numerically. Results of the present study provide useful information for enhancing the design of engineered cooling channels to assure long-term cooling and retention of corium under severe accident conditions.

The Analysis of Severe Accident Induced Steam Generator Tube Rupture and LERF Risk

This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with countercurrent natural circulating high temperature gas in the hot leg and SG tubes. The first step performs thermal-hydraulic analysis to predict the creep rupture parameter of the tubes in severe accident. The next step applies the creep rupture models to test the potential for the degraded SG to rupture before the hot leg. Then, the mean of the SG tube rupture probability was applied to estimate large early release frequency in LERF (Large and Early Release Frequency) model, and the overall LERF risk due to the Induced SGTR was calculated. In the final step, implementation of severe accident management guidance (SAMG), such as the RCS depressurization and refilling to SG, is evaluated using PSA approach. It can be found that strategy of RCS depressurization and refilling to SG can mitigate the result of induced SGTR and LERF effectively.

Leak behavior of steam generator tube-to-tubesheet joints under creep condition: Experimental study

To address concerns regarding excessive leakage from throughwall cracks in steam generator tube-to-tubesheet joints under severe accident conditions, leak rate testing was conducted using tube-to-collar joint specimens. The tube interior and the interface between tube and collar (crevice) were pressurized independently using nitrogen gas. The leak rate through the crevice was almost zero when the specimens were pressurized at ∼500 °C; this low leak rate is attributed to thermal mismatch effects preventing much leakage. The near zero leak rate was maintained until the onset of large leakage at higher temperatures. The leak rate behavior after the onset of the large leakage was not much affected by the crevice length or heat-to-heat variation of Alloy 600 tubes. This suggests that once the crevice gap opens, the creep rate of the low alloy steel collar becomes dominant. Specimens with different tube diameters behaved essentially the same way. To simulate a flawed steam generator tube in the tubesheet, the crevice region was pressurized through a hole in the tube. This simulation resulted in essentially the same behavior as those specimens whose tubes and crevices were pressurized independently. Oxidation of low alloy steel collars in air tests can increase the flow resistance, and thus tests using nitrogen gas would provide more conservative leak rate data.

중대사고시 노심용융물의 Rayleigh-Benard 자연대류 예비 실험

원자력발전소 중대사고시 노심용융물의 Rayleigh-Benard 자연대류 문제에 대한 예비실험으로 두 평판 사이의 거리, 측면벽의 유무 및 평판의 기하구조가 열전달에 미치는 영향에 대해 실험적 연구를 수행하였다. 열전달 실험을 대신하여 상사성의 원리를 이용한 황산-황산구리 수용액의 전기도금계를 물질전달계로 채택하였다. 실험은 <TEX>$Ra_s$</TEX>는 <TEX>$1.06{\times}10^7{\sim}2.91{\times}10^{10}$</TEX>의 범위에서 실험적 조건을 변화시켜가며 열전달을 측정하였다. 실험결과 단일 수평평판에서 측정한 열전달은 McAdams의 수평평판 자연대류 열전달 상관식과 일치하였고 두 평판에서 측정한 열전달은 Dropkin과 Somerscales, Globe와 Dropkin의 Rayleigh-Benard 자연대류 열전달 상관식과 매우 유사한 경향을 보였다. 두 평판 사이의 거리가 작을 경우 열전달이 높다가 거리가 증가하면 단일 수평평판에서의 자연대류 열전달과 같아졌다. 평판에 설치된 휜(Fin)은 열전달을 향상시켰다. 모든 경우에서 측면벽이 없는 경우의 열전달이 측면 벽이 있는 경우보다 항상 높았다. Rayleigh-Benard natural convection experiments were carried out as the preliminary experiment to simulate the natural convection of the core melt at the severe accident conditions. This work focused on the influences of plate separation distance(s), the existence of the side walls and crust geometries of upper and lower plates. Based upon the analogy concept, a cupric acid-copper sulfate electroplating system(<TEX>$H_2SO_4-CuSO_4$</TEX>) was employed as the mass transfer system and measurements were made for <TEX>$Ra_s$</TEX> ranging from <TEX>$1.06{\times}10^7$</TEX> to <TEX>$2.91{\times}10^{10}$</TEX>. The test results measured for a single horizontal plate were in good agreement with the correlation reported by McAdams and those for two horizontal plates showed the similar trend to the existing Rayleigh-Benard heat transfer correlations developed by Dropkin and Somerscales, Globe and Dropkin. The measured heat transfer rate decreased with the increasing separation distance between the two plates and became similar to those for a single horizontal plate. Fin arrays mounted on both upper and lower plates enhanced the heat transfer rates. For all cases, the heat transfer rates measured for open side walls are higher than those for closed ones.

Capability of the RELAP5 code to simulate natural circulation behavior in test facilities

Many advanced reactor designs incorporate passive systems mainly to enhance the operational safety and possible elimination of severe accident condition. Some reactors are even designed to remove the nominal fission heat passively by natural circulation without using mechanical pumps e.g. ESBWR, AHWR, CHTR, CAREM, etc. while in most other new reactor concepts, the decay heat is removed passively by natural circulation following the pump trip conditions. The design and safety analysis of these reactors are carried out using the best estimate codes such as RELAP5, TRAC and CATHARE, etc. These best estimate codes have been developed for pumped circulation systems and it is not proven about their adequacy or applicability for natural circulation systems wherein the driving mechanism is completely different. Some of the key phenomena which are difficult to model but are significantly important to assess the natural circulation system performances are – low flow natural circulation mainly because the flow is not fully developed and can be multi-dimensional in nature; flow instabilities; critical heat flux under oscillatory condition; flow stratification particularly in large diameter vessel; thermal stratification in large pools; effect of non-condensable gases on condensation, etc. Though, these best estimate codes use a six equation two-fluid model formulation for the thermal-hydraulic calculation which is considered to be the best representative of two-phase flows, but their accuracies depend on the accuracies of the models for interfacial relationships for mass, energy and momentum transfer which are semi-empirical in nature. The other problem with two-fluid models is the effect of ill-posedness which may cause numerical instability. Besides, the numerical diffusion associated due to truncation of higher order terms can affect the prediction of flow instabilities. All these effects may lead to inability to capture the important physical instability in natural circulation systems and instability characteristics i.e. amplitude and frequency of flow oscillation. In view of this, it is essential to test the capability of these codes to simulate natural circulation behavior under single and two-phase flow conditions before applying them to the future reactor concepts.In the present study, one of the extensively used best estimate code RELAP5 has been used for simulation of steady state, transient and stability behavior of natural circulation based experimental facilities, such as the High-Pressure Natural Circulation Loop (HPNCL) and the Parallel Channel Loop (PCL) installed and operating at BARC. The test data have been generated for a range of pressure, power and subcooling conditions. The computer code RELAP5/MOD3.2 was applied to predict the transient natural circulation characteristics under single-phase and two-phase conditions, thresholds of flow instability, amplitude and frequency of flow oscillations for different operating conditions of the loops. This paper presents the effect of nodalisation in prediction of natural circulation behavior in test facilities and a comparison of experimental data in with that of code predictions. The errors associated with the predictions are also characterized.

Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

A severe accident has inherently significant uncertainties due to the complex phenomena and wide range of conditions. Because of its high temperature and pressure, performing experimental validation and practical application are extremely difficult. With these difficulties, there has been few experimental researches performed and there is no plant-specific experimental data. Instead, computer codes have been developed to simulate the accident and have been used conservative assumptions and margins. This study is an effort to reduce the uncertainty in the probabilistic safety assessment and produce a realistic and physical-based failure probability. The methodology was developed and applied to the OPR1000. The creep rupture failure probabilities of reactor coolant system (RCS) components were evaluated under a station blackout severe accident with all powers lost and no recovery of steam generator auxiliary feed-water. The MELCOR 1.8.6 code was used to obtain the plant-specific pressure and temperature history of each part of the RCS and the creep rupture failure times were calculated by the rate-dependent creep rupture model with the plant-specific data.

Oxidation of fuel cladding candidate materials in steam environments at high temperature and pressure

Under certain severe accident conditions, the fuel rods of nuclear power plants are exposed to high temperature/pressure steam environments in which the Zr alloy cladding is rapidly oxidized. As alternative claddings, the oxidation resistances of SiC-based materials and stainless steels with high Cr and/or Al additions have been examined from 800-1200 °C in high-pressure steam environments. Very low reaction kinetics were observed with alumina-forming FeCrAl alloys at 1200 °C while Fe-Cr alloys with only 15-20% Cr were rapidly attacked.

Complementary safety assessments of the French nuclear facilities

EDF has conducted, after the Fukushima event, complementary safety assessments of its nuclear facilities. The aim of this in-depth review was to assess the resilience of each plant to extreme external hazards, situations that could lead to severe accident conditions. These analyses demonstrate a good level of safety for all of EDFʼs nuclear facilities. Supplementary measures post-Fukushima have been put forward to the ASN with the aim of continuing to improve the level of safety at the plants. Once the ASN position is issued, EDF will develop an action plan over several years, covering both supplementary studies and modifications that have been identified. Suite à lʼaccident de Fukushima, EDF a procédé à des évaluations complémentaires de sûreté pour lʼensemble de son parc de production. Ce réexamen approfondi a consisté à vérifier les marges de sûreté des installations face à des situations dʼagressions externes extrêmes pouvant conduire à un accident grave. Ces analyses ont permis de confirmer le bon niveau de sûreté de lʼensemble du parc nucléaire dʼEDF. Des mesures complémentaires post-Fukushima ont été proposées à lʼASN pour concourir à élever encore le niveau de sûreté des centrales. Après lʼavis de lʼASN, EDF élaborera un plan dʼaction sur plusieurs années, comprenant à la fois les études complémentaires et les modifications décidées.

The role of CFD combustion modeling in hydrogen safety management—Part I: Validation based on small scale experiments

Abstract The potential consequences of hydrogen release and combustion during a severe accident in a Light Water Reactor (LWR) have received considerable attention after the Fukushima accident. The pressure loads resulting from hydrogen combustion can be detrimental to the structural integrity of the nuclear reactor safety systems and the reactor containment. Therefore, accurate prediction of these pressure loads is important from a safety point of view. The considered pressure loads are determined by the turbulent flame acceleration. This flame acceleration is determined by the amount of turbulence generated locally during the combustion process. Lumped parameter codes can only assume, but cannot compute this local turbulence generation process due to their inherent limitations. Therefore, a three dimensional CFD based approach is needed. In this paper, it is demonstrated that it is of utmost importance to apply successive mesh and time step refinement systematically in CFD analyses for the considered application. This has been applied in this paper. Within the field of hydrogen safety, we have not found CFD analyses published earlier in the open literature that were based on successive mesh and time step refinement. In the light of the above, the lack of the demonstration that the numerical errors are negligible in the CFD analyses of hydrogen deflagrations must be considered as a serious shortcoming. Therefore, the authors are developing a CFD methodology that can be used for the analyses of hydrogen deflagrations in experimental facilities as well as in full scale reactor containments, with which it can be demonstrated efficiently that the mesh and time step requirements are fulfilled. Such a CFD methodology is deemed to be essential in order to be able to prove containment integrity under severe accident conditions. The authors are well underway with the development of this methodology, and the first part of the development will be presented in this paper. Furthermore, it is demonstrated that the turbulent flame acceleration, and therefore the resulting pressure loads, depend also strongly on the initial turbulence present in the ignition region. Therefore, it is recommended that for future validations, hydrogen deflagration experiments will be performed where this initial turbulence is accurately measured.

Numerical Study of Hydrogen Ignition by Passive Autocatalytic Recombiners

A detailed model is proposed for numerical simulation of hydrogen ignition inside box-type passive autocatalytic recombiners (PARs). The model is focused on the reactive channel flow between two catalytic sheets of a recombiner. It includes complex chemistry and multicomponent transport for homogeneous hydrogen combustion and complex surface chemistry for heterogeneous hydrogen recombination. First calculations are dedicated to H2/air mixtures without steam at atmospheric pressure and room temperature. The analysis of the total homogeneous and heterogeneous heat release rates according to the inlet hydrogen molar fraction reveals three possible operation regimes for the recombiners from pure catalytic conversion to pure gaseous combustion. A physical criterion is then proposed for the ignition of H2/air mixtures inside the recombiners. The numerical ignition threshold at 5.4% of hydrogen without steam is in good agreement with experimental data. The criterion is then applied to the ternary diagram including all representative H2/air/H2O mixtures for severe accident conditions in pressurized water reactors. It shows a sharper transition from the catalytic regime to the gaseous one for high hydrogen concentrations. A specific strategy finally allows defining an extended PAR hydrogen ignition limit in the entire ternary diagram, which is well corroborated by the available experimental database.

Response Analysis on Electrical Pulses under Severe Nuclear Accident Temperature Conditions Using an Abnormal Signal Simulation Analysis Module

Unlike design basis accidents, some inherent uncertainties of the reliability of instrumentations are expected while subjected to harsh environments (e.g., high temperature and pressure, high humidity, and high radioactivity) occurring in severe nuclear accident conditions. Even under such conditions, an electrical signal should be within its expected range so that some mitigating actions can be taken based on the signal in the control room. For example, an industrial process control standard requires that the normal signal level for pressure, flow, and resistance temperature detector sensors be in the range of 4~20 mA for most instruments. Whereas, in the case that an abnormal signal is expected from an instrument, such a signal should be refined through a signal validation process so that the refined signal could be available in the control room. For some abnormal signals expected under severe accident conditions, to date, diagnostics and response analysis have been evaluated with an equivalent circuit model of real instruments, which is regarded as the best method. The main objective of this paper is to introduce a program designed to implement a diagnostic and response analysis for equivalent circuit modeling. The program links signal analysis tool code to abnormal signal simulation engine code not only as a one body order system, but also as a part of functions of a PC-based ASSA (abnormal signal simulation analysis) module developed to obtain a varying range of theR-Ccircuit elements in high temperature conditions. As a result, a special function for abnormal pulse signal patterns can be obtained through the program, which in turn makes it possible to analyze the abnormal output pulse signals through a response characteristic of a 4~20 mA circuit model and a range of the elements changing with temperature under an accident condition.

Characterization of PWR vessel steel tearing under severe accident condition temperatures

Abstract In the event of a severe core meltdown accident in a pressurised water reactor (PWR), core material can relocate into the lower head of the vessel resulting in significant thermal and pressure loads being imposed on the vessel. In the event of reactor pressure vessel (RPV) failure there is the possibility of core material being released towards the containment. On the basis of the loading conditions and the temperature distribution, the determination of the mode, timing, and size of lower head failure is of prime importance in the assessment of core melt accidents. This is because they define the initial conditions for ex-vessel events such as core/basemat interactions, fuel/coolant interactions, and direct containment heating. When lower head failure occurs (i) the understanding of the mechanism of lower head creep deformation; (ii) breach stability and its kinetic of propagation leading to the failure; (iii) and developing predictive modelling capabilities to better assess the consequences of ex-vessel processes, are of equal importance. The objective of this paper is to present an original characterization programme of vessel steel tearing properties by carrying out high temperature tearing tests on Compact Tension (CT) specimens. The influence of metallurgical composition on the kinetics of tearing is investigated as previous work on different RPV steels has shown a possible loss of ductility at high temperatures depending on the initial chemical composition of the vessel material. Small changes in the composition can lead to different types of rupture behaviour at high temperatures. The experimental programme has been conducted on various French RPV 16MND5 steels for temperatures ranging from 900 °C to 1100 °C. Comparisons between the tests performed on these various 16MND5 steels show that this approach is appropriate to characterize the difference in ductility observed at high temperatures. The aim of this experimental study is also to contribute to the definition of a tearing criterion by identifying, on the basis of CT results, the related material parameters at temperatures representative of the real severe accident conditions. This experimental campaign has been carried out in partnership with IRSN in the framework of a research programme whose purpose is to complete the mechanical properties database of 16MND5 steel and to model tearing failure in French RPV lower head vessels under severe conditions ( Koundy et al., 2008 ).

Study of boron behaviour in the primary circuit of water reactors under severe accident conditions: A comparison of Phebus FPT3 results with other recent integral and separate-effects data

Abstract Boron carbide (B 4 C) is widely used as an absorber material in many commercial reactors, such as boiling water reactors, Russian VVERs, later French pressurised water reactors, and is planned to be used in Areva's EPR™. Under severe accident conditions, B 4 C reacts with its surrounding stainless steel cladding, producing eutectic melts above 1200 °C; remaining bare B 4 C and B 4 C/metal mixtures are then exposed to steam and oxidize highly exothermically. As well as hydrogen, gases and aerosols containing boron and carbon compounds are produced, which affect the transport and deposition of radiologically important fission products such as iodine and caesium in the circuit, and subsequent behaviour in the containment. The influence of a B 4 C control rod on fuel degradation and fission product release through to the late phase, material transport in the circuit and behaviour in the containment was studied in the nuclear-heated integral experiment Phebus FPT3 performed by IRSN at Cadarache. Carbonaceous gas production was monitored, and evidence for substantial deposition of B-containing compounds in the circuit was gathered. The electrically heated QUENCH-07 and -09 tests performed at Karlsruhe Institute for Technology (KIT) used similar bundle geometry and similar but less severe test conditions, with a fast cooling phase that arrested bundle degradation before the late phase. Extensive separate-effects tests on oxidation of B 4 C and its interaction with surrounding materials have also been performed by IRSN (BECARRE programme) and KIT (BOX, LAVA, QUENCH-SR). This paper compares the phenomena involving B 4 C observed in these experiments, taking the recently released data from Phebus FPT3 as a base, concentrating on transport and deposition behaviour in the circuit, with reference to degradation where necessary, e.g. as a source of carbonaceous gases. In particular, there is evidence concerning blockage formation in the hot leg of the FPT3 circuit, which affects the transport of fission products, and this is supported by similar observations in the separate-effect tests. The FPT3 data are now under quantitative interpretation, and work in progress to improve understanding in this area is indicated.

Ruthenium release modelling in air and steam atmospheres under severe accident conditions using the MAAP4 code

In a nuclear power plant (NPP), a severe accident is a low probability sequence that can lead to core fusion and fission product (FP) release to the environment (source term). For instance during a loss-of-coolant accident, water vaporization and core uncovery can occur due to decay heat. These phenomena enhance core degradation and, subsequently, molten materials can relocate to the lower head of the vessel. Heat exchange between the debris and the vessel may cause its rupture and air ingress. After lower head failure, steam and air entering in the vessel can lead to degradation and oxidation of materials that are still intact in the core. Indeed, Zircaloy-4 cladding oxidation is very exothermic and fuel interaction with the cladding material can decrease its melting temperature by several hundred of Kelvin. FP release can thus be increased, noticeably that of ruthenium under oxidizing conditions. Ruthenium is of particular interest because of its high radio-toxicity due to 103Ru and 106Ru isotopes and its ability to form highly volatile compounds, even at room temperature, such as gaseous ruthenium tetra-oxide (RuO4). It is consequently of great need to understand phenomena governing steam and air oxidation of the fuel and ruthenium release as prerequisites for the source term issues. A review of existing data on these phenomena shows relatively good understanding. In terms of oxygen affinity, the fuel is oxidized before ruthenium, from UO2 to UO2+x. Its oxidation is a rate-controlling surface exchange reaction with the atmosphere, so that the stoichiometric deviation and oxygen partial pressure increase. High temperatures combined with the presence of oxygen in the atmosphere lead to fuel expansion and formation of cracks. In these conditions, intra- and inter-granular diffusions of ruthenium in the fuel matrix are so enhanced that it is possible to consider an instantaneous volatilisation of ruthenium oxides at the fuel surface. Based on these considerations, a completely new model has been implemented in the EDF local version of the MAAP4.07 severe accident code (Modular Accident Analysis Program), owned by EPRI (Electric Power Research Institute). The fuel oxidation modelling takes into account many kinds of atmospheres (steam and/or air and/or hydrogen), the stoichiometric evolution and the oxygen partial pressure of the fuel matrix. The release of ruthenium oxides is calculated considering their particular reaction constants. The model was assessed by the simulation of different CEA-VERCORS experiments in air, steam and mixed atmospheres. These experiments are specifically designed to study FP release from fuel under different atmospheres and temperatures. This paper deals with the main results obtained with MAAP4.07 when simulating these tests.