MELCOR Model Research Articles

Development of a MELCOR Model for LVR-15 Severe Accidents Assessment

LVR-15 is a light-water-tank-type research reactor placed in a stainless-steel vessel under a shielding cover located in the Research Centre Rez (CVR) near Prague. It is operated at a steady-state power of up to 10 MWt under atmospheric pressure and is cooled by forced circulation. In 2011, the fuel was replaced, going from high-enriched uranium (HEU) to low-enriched uranium (LEU). After 2017, the State Office for Nuclear Safety (SUJB) asked CVR to evaluate the LVR-15 under Design Extended Conditions B (DEC-B). For this reason, a new model was developed in the MELCOR code, which allows for modelling the progression of a severe accident (SA) in light-water nuclear power plants and estimating the behaviour of the reactor under SA conditions. The model was built by collecting information about the LVR-15. Since the research reactor can have different core configurations according to the location of the core components, the core configuration with the most fuel (hottest campaign K221) was selected. Then, to create the radial nodalisation, the details of the core components were obtained and grouped in five radial rings and 27 axial levels. The simulation was run with the boundary conditions collected from campaign K221, and the results were compared with the reference values of the campaign with a negligible percentage of error. For the coolant inlet and outlet temperature, the reference values were 318.18 K and 323.5 K, respectively, while for the simulation, the steady state reached 319 K for the inlet temperature and 324 K for the outlet temperature. Additionally, the cladding temperature of the hottest assembly was compared with the reference value (353.72 K) and the steady-state simulation results (362 K). In future work, different transients leading to severe accidents will be simulated. When simulating the LVR-15 reactor with MELCOR, specific attention is required for the aluminium-cladded fuel assemblies, as the model requires some assumptions to cope with the phenomenological limitations.

Severe accident analysis induced by secondary pipeline break in a small modular PWR

The small modular PWR (SMPWR) usually adopts integral design. Under severe accident, the system responses are different from those large PWRs. It is necessary to study the severe accident behavior of the SMPWR. A MELCOR model is developed for SMPWR and its steady-state results are in good agreement with the design values. Severe accidents induced by secondary pipeline break accidents are simulated, and no pressure relief measures are taken to keep the primary loop under high pressure. The mitigation effects of passive containment air cooling system (PAS) and passive cavity injection system (PCIS) are evaluated under different cases. The results show that under high pressure conditions, PCIS can effectively cool the lower head. The earlier the PCIS operates, the more significant the mitigation effect can be. In addition, PAS can effectively reduce the peak pressure and temperature in the containment. This study can provide a reference for the formulation of severe accident management guidelines on SMPWRs.

Towards a more realistic MELCOR model for a dry cask for spent nuclear fuel. Part II: application.

Nowadays, a great deal of attention is devoted to the development of best-estimate models able to produce more realistic outcomes. This is also the case for system codes, such as MELCOR, that are being mostly used in a conservative way especially when dealing with the licensing process. The above-mentioned need for more realistic results is at the core of this two-paper series related to the creation of a more accurate MELCOR model for the HI-STORM 100S dry cask. The findings obtained from the sensitivity studies carried out in the Part I are leveraged to set up an improved MELCOR model, the characteristics of which are consistent with the typical features of Spent Nuclear Fuel (SNF), and with geometrical and material properties of the cask itself. The addition of an axial power profile in the Fuel Assembly (FA), the better characterization of the flow losses in the air gap between internal metallic canister and external concrete-based overpack, and the choice of an appropriate value for the concrete thermal conductivity, are taken into account conjointly in this Part II. The outcomes from the improved MELCOR simulation are reported mainly in terms of the Peak Cladding Temperature (PCT), being the variable under regulatory surveillance. However, in addition to PCT, calculated temperature profiles are displayed and compared against the ones resulting from the previous model.

Towards a more realistic MELCOR model for a dry cask for spent nuclear fuel. Part I: sensitivity studies.

The United States (US) Department Of Energy (DOE) has addressed the thermal analysis of the Spent Nuclear Fuel (SNF) stored within a dry cask system as a matter of high priority. In this regard, it is of utmost importance that simulation tools effectively reproduce the general thermal behavior of the modelled cask, including heat exchange and removal. Temperature distribution in the different components of the system is usually the focus of performed thermal analyses. In particular, attention is paid to the maximum temperature reached in the fuel cladding, namely the Peak Cladding Temperature (PCT). Within this framework, the present paper is the first of a two-paper series aimed at developing a more accurate model for the HI-STORM 100S cask. The dry cask in question is modelled and its behavior is simulated by means of the MELCOR code (version 2.2.18019). Stressing the need for a more realistic model rather than a conservative one, this paper reports the efforts undertaken to evaluate the influence of some specific modelling choices on the PCT. The study of the cask performance is therefore conducted taking into consideration three main factors: the axial power distribution in the Fuel Assembly (FA), the flow losses in the air gap between the internal canister and the external overpack, and the conductivity of the overpack concrete.

Strategic analysis on sizing of flooding valve for successful accident management of small modular reactor

In contrast to all-time flooded small modular reactor (SMR) systems, an in-kind flooding safety system (FSS) has been proposed as a passive safety system applicable to small modular reactors (SMRs) that adopt a metal containment vessel (MCV). Under transient conditions, the FSS can provide emergency cooling to dry reactor cavities and sustain long-term coolability using re-acquired evaporated steam in the reactor building on demand. When designing an FSS, the effect of the flooding flow area is vital as it affects the overall accident sequence and safety. Therefore, in this study, a MELCOR model of a reference SMR is developed and numerical analysis is performed under postulated accident scenarios. Without flooding, the MCV pressure of the reactor module exceeds the design pressure before core damage. To prevent core damage, an emergency flooding strategy is devised using various flow path parameters and requirements to ensure an adequate emergency coolant supply before the core damage is investigated. The results indicate that a flow area exceeding 0.02 m2 is required in the FSS to prevent MCV overpressure and core damage. This study is the first to report a strategic analysis for appropriately sizing an FSS flooding valve applicable to innovative SMRs.

Investigation of aerosol transportation phenomena in pool scrubbing by combining bubble plume measurements and single-bubble mass transfer analysis

The fission product filtration capability of pool scrubbing is an uncertain factor in severe accident risk analysis. Limited discussions on aerosol transportation phenomena exist due to insufficient understanding of aerosol mass transfer from single bubbles and bubble plume behavior. This study aims to investigate aerosol transportation during pool scrubbing by integrating individual bubble mass transfer into the overall bubble plume.Decontamination factor (DF) measurements were conducted for various nozzle submergences, followed by analysis of bubble diameter distributions in the corresponding bubble plume. Findings revealed that the MELCOR model overestimates bubble break-up, contrary to the expected positive impact on DF. To address this discrepancy, a computational fluid dynamics simulation was employed to analyze mass transfer coefficients considering different bubble diameters and aerosol properties. By integrating the mass transfer of single bubbles, the discrepancy in the DF tendencies was attributed to the underestimation of aerosol mass transfer coefficients in individual bubbles.

Validation study of sodium pool fire modeling efforts in MELCOR and SPHINCS codes

Discharge of sodium coolant into containment from a sodium-cooled fast reactor vessel can occur in the event of a pipe leak or break. In this situation, some of the liquid sodium droplets discharged from the coolant system will react with oxygen in the air before reaching the containment. This phase of the event is normally termed the sodium spray fire phase. Unreacted sodium droplets pool on the containment floor where continued reaction with containment atmospheric oxygen occurs. This phase of the event is normally termed the sodium pool fire phase. Both phases of these sodium-oxygen reactions (or fires) are important to model because of the heat addition and aerosol generation that occur. Any fission products trapped in the sodium coolant may also be released during this progression of events, which if released from containment could pose a health risk to workers and the public. The paper describes progress of an international collaborative research in the area of the sodium fire modeling in the sodium-cooled fast reactors between the United States and Japan under the framework of the Civil Nuclear Energy Research and Development Working Group. In this collaboration between Sandia National Laboratories and Japan Atomic Energy Agency, the validation basis for and modeling capabilities of sodium spray and pool fires in MELCOR of Sandia National Laboratories and SPHINCS of Japan Atomic Energy Agency are being enhanced.This study documents MELCOR and SPHINCS sodium pool fire model validation exercises against the JAEA’s sodium pool fire experiments, F7-1 and F7-2. The proposed enhancement of the sodium pool fire models in MELCOR through addition of thermal hydraulic and sodium spreading models that enable a better representation of experimental results is also described.

Pool scrubbing in unit 1 of fukushima daiichi

Under the frame of OECD/BSAF and ARC-F projects, CIEMAT estimated that around 40% of the Cesium initial inventory in the reactor core and nearly 60% of the Iodine one of the Unit 1 of Fukushima Daiichi, got absorbed in the suppression pool. CIEMAT has critically reviewed the bases of the MELCOR pool scrubbing module, including the major assumptions and approximations underneath, under the prevailing conditions in Unit 1 of Fukushima Daiichi during fission product scrubbing. The results have shown that most fission product mass is retained in the water nearby the injection point as a consequence of diffusiophoresis and inertial mechanisms. However, some of the approximations underneath the corresponding models have been found to be questionable and some new formulations have been considered and implemented within the stand-alone SPARC-90 code, which is the basis of the MELCOR models for pool scrubbing. In particular, a model capable of describing the scrubbing effect of jet injection has been implemented. This work presents better-fitted models for the scenario and, after implementing them into the SPARC-90 code, the resulting moderate quantitative impact on the accident scenario of Unit1 in Fukushima Daiichi, has been discussed to be well within the uncertainty range associated to the original code estimates. However, updating of MELCOR concerned models for fission products trapping in pools is highly recommended for physical consistency.

NPP Krško Station Blackout Analysis after Safety Upgrade Using MELCOR Code

The analysis of a Station blackout (SBO) accident in the NPP Krško including thermalhydraulic behaviour of the primary system and the containment, as well as the simulation of the core degradation process, release of molten materials and production of hydrogen and other incondensable gases will be presented in the paper. The calculation model includes the latest plant safety upgrade with addition of Passive Autocatalytic Recombiners (PAR) and the Passive Containment Filter Venting (PCFV) system. The code used is MELCOR, version 1.8.6. MELCOR is an integral severe accident code which means that it can simulate both the primary reactor system, including the core, and the containment. The code is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. The analysis is conducted in two steps. First, the steady state calculation is performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step is the calculation of the SBO accident with the leakage of the coolant through the damaged reactor coolant pump seals. Without any active safety systems, the reactor pressure vessel will fail after few hours. The mass and energy releases from the primary system cause the containment pressurization and rise of the temperature. The newly added safety systems, PAR and PCFV, prevent the damage of the containment building by keeping the thermalhydraulic conditions below the design limits. The analysis results confirm the capability of the safety systems to effectively control the containment conditions. Results of the analysis are given with respect to the results of the MAAP 4.0.7 analysis for the same accident scenario. The MAAP and MELCOR codes are the most popular severe accident codes and, therefore, it is reasonable to compare their results. In addition, sensitivity calculations performed by varying most influential parameters, such as the hot leg creep failure, blockage of a pipe connecting the cavity and the sump, inclusion of a radionuclide package in the MELCOR, etc. are done in order to demonstrate correct physical behaviour and the accuracy of the developed NPP Krško MELCOR model.

Assessment of SCALE and MELCOR for a generic pebble bed fluoride high-temperature reactor

This paper presents the development and application of SCALE and MELCOR models for a fluoride salt-cooled high-temperature reactor (FHR) based on publicly available specifications. SCALE version 6.3beta15 was used to generate power distributions and decay heat curves, and MELCOR version 2.2.18019 was used to calculate the thermal hydraulic response of an assumed FHR primary system. An approach was developed for determining the equilibrium state of the core using depletion of a core slice model and blending of fuel compositions at different burnups to provide three-dimensional fuel composition in the core. Results compared between a core with entirely fresh fuel and one with an equilibrium fuel composition revealed that the equilibrium core led to lower steady-state temperatures but slower cooldown during a loss of flow accident (LOFA). A sensitivity study was then conducted to explore the transient response of our FHR system to variations in thermal hydraulic parameters of the system using the equilibrium core. The inlet temperature, graphite thermal conductivity, and SCRAM time all provided significant control over peak fuel and coolant temperatures. Uncertainties in radionuclide decay data in SCALE were used to perform a decay heat sensitivity study, and we found that uncertainties in decay heat led to negligible impact on peak temperatures during the course of the transient. In all cases evaluated, the observed peak fuel temperatures remained approximately 700 K below anticipated failure limits for the LOFA.

Experimental study of interfacial area of bubble plume based on bubble tracking by wire-mesh sensor

The fundamental information of bubble plume such as interfacial area, three-dimensional velocity, and bubble diameter are necessary to develop a kinematic pool scrubbing model for uncertainty reduction of severe accident analysis. In this study, an algorithm was introduced to polygonize the cross-sectional void fraction of the wire-mesh sensor and reconstruct it in 3D to track the individual bubbles detected. Based on the result, the interfacial area and three-dimensional velocity of individual bubbles in a bubble plume were obtained. The results show that the interfacial area from MELCOR model tends to overestimate the analyzed results due to remaining of large bubbles in the downstream. In addition, the effect of subcooling temperature and steam fraction on bubble formation was evaluated by applying the developed method into condensable gas experiments. The interfacial area increased significantly under high subcooling temperature and steam fraction due to atomization of bubbles caused by steam condensation.

Supporting analysis for WCLL test blanket system safety

The Water-Cooled Lithium Lead Breeding Blanket is one of the most promising concepts to be used as a key component in fusion power devices. It provides tritium breeding and nuclear power conversion and extraction. Before its integration in a DEMO plant, a testing phase is required to gain data on both thermal and neutronic performances. For this purpose, a Test Blanket System is currently under conceptual design and will be integrated into dedicated ITER equatorial ports. Early insight on the safety performance of the Test Blanket System remains an essential element for integrating into ITER, and accident analysis is one of its critical components. Preliminary safety studies have been performed in the framework of Safety And Environment (SAE) work package of the EUROfusion consortium program to support the design and the integration of the Water-Cooled Lithium Lead Test Blanket System into ITER. At the present conceptual design phase, these studies are tailored to give insight on the effects of some parameters, such as postulated initiating event timing, plasma termination system or valves intervention, and possible safety provision implementation. A MELCOR model of the Water-Cooled Lithium Lead Test Blanket System has been developed, and two accident scenarios have been studied: an ex-vessel LOCA inside the port cell and a loss of flow accident because of pump seizure. The impact of mentioned parameters on accident evolution and consequences is investigated in support of the safety logic definition.

A Sensitivity Study of Critical Flow Modeling with MELCOR 2.2 Code Based on the Marviken CFT-21 Experiment

The paper presents a study of the critical flow phenomena modeling with MELCOR 2.2.15254 severe accident computer code. The Marviken Critical Flow Test number 21 (CFT-21) experiment was selected as a representative critical flow-focused Separate Effect Test (SET). Various modeling aspects were investigated, including the nodalization, model setup, parameters, and sensitivity coefficients. A local-type sensitivity study was performed to analytically identify the significant parameters and assess their impact on the modeling. A dedicated regression-based approach, using standard deviation, was developed to find the best-fit MELCOR modeling parameters. The primary purpose of this work was to determine the appropriate approach to model critical flow with MELCOR 2.2, investigate the model performance, assess the influence of nodalization choices, identify significant sensitivity parameters, and prepare recommendations with an emphasis on best-estimate modeling. An additional outcome was the benchmark of the recent code version with the Marviken test.

Preliminary uncertainty quantification of the core degradation models in predicting the Fukushima Daiichi unit 3 severe accident

A MELCOR model of a boiling water reactor has been developed by Sapienza University of Rome in the framework of the MUSA project aiming at identifying and quantifying uncertainty sources in severe accidents analyses. To develop the model, the Fukushima Unit 3 boiling water reactor has been taken as reference, and a preliminary sensitivity analysis with RAVEN software coupled with MELCOR has been performed to quantify the influence of core degradation parameters on selected key figures-of-merit, such as pressure vessel and containment pressures, core liquid level, hydrogen generation, lower head breach time and source term.Results of base case satisfactorily predict main pressure and liquid level data measurement from TEPCO. As well as, results from the uncertainty analysis envelope the majority of thermohydraulic TEPCO reported measured data. Furthermore, median of calculated core status is in likely agreement with recent TEPCO containment inspections and muon measurements, with about 25% of fuel rods remained intact in RPV and 65% of core masses ejected to the pedestal.

Plant-specific assessment of the natural circulation-induced creep rupture characteristics in the RCS pressure boundary during the SBO accident

In nuclear power plants (NPPs), one of the important yet uncertain phenomena during a station blackout (SBO) accident is the possibility for a counter-current natural circulation flow in the reactor coolant system (RCS) pressure boundary, which could divert the core melt progression and/or the in-vessel fission product (FP) distribution and, in turn, lead to a heatup and eventual creep rupture. From the risk and accident management points of view, one of key concerns is to predict more explicitly whether such a creep rupture can occur in the RCS pressure boundary, and if it indeed can, also predict the most probable time and the most likely location. This paper presents plant-specific analysis results on the potential characteristics of natural circulation-induced creep rupture of the RCS pressure boundary that could potentially take place during an SBO accident of a reference pressurized water reactor (APR1400). MELCOR-CFD (Computational Fluid Dynamics) coupled analysis was performed to determine more precisely the relevant physical parameter that could greatly influence on the behaviour of natural circulation and the ensuing creep rupture. Multiple sensitivity analyses were also carried out to investigate the influence that particular MELCOR modeling schemes and plant conditions have in estimating the RCS creep rupture characteristics. The analysis results and relevant insights were summarized in terms of particular points of interest.

Global uncertainty and sensitivity analysis of MELCOR and TRACE critical flow models against MARVIKEN tests

Uncertainties of numerical predictions play important role in assessment of safety margins in nuclear reactors. Critical Flow models are used in the different types of Nuclear Safety Analysis but these are especially important in Loss of Coolant Accidents (LOCA) analyses.In this paper validation of the input decks is performed of all of the Marviken critical flow tests. Next uncertainties of the prediction of these models implemented into TRACE and MELCOR computer codes are analysed and discussed. Results of the global uncertainty analysis are shown in terms of trends of uncertainty values with respect to the pressure, nozzle length, water level, discharge coefficient and hydraulic diameter.Application of the global sensitivity analysis methods allowed ranking of the sources of uncertainties over large spectrum of experimental conditions. This was achieved thanks to usage of Bayesian Integrated Uncertainty and Sensitivity Analysis (BIGUSA) methodology. The analysis revealed that the uncertainties of critical mass flow predictions are predominantly resulting from uncertainties of the hydraulic diameter in both codes. In the TRACE code the hydraulic diameter, nozzle length and discharge coefficient were dominating factors. For MELCOR dominating factors were hydraulic diameter, discharge coefficient and pressure.

MELCOR simulation of the SBLOCA induced severe accident for the SMR in a floating nuclear power plant

A floating nuclear power plant (FNPP) is a kind of nuclear power plant on a barge moored specifically in an area of the sea. This study focuses on a severe accident induced by cold leg small break loss of coolant accident (SBLOCA) occurring in a small modular reactor (SMR) in a FNPP of 100 MWt 2-loop pressurized water reactor (PWR) type. The severe accident was modeled using MELCOR computer code. A MELCOR model was developed for the primary and secondary systems, the engineered safety features, and the containment of the SMR. Three severe accident mitigation measures were analyzed using the model, namely the pressurizer (PZR) relief extension, the external vessel reactor cooling (EVRC), and the passive autocatalytic recombiners (PARs). The progression of the severe accident and the effects of three mitigation measures against the severe accident were analyzed. The simulation results indicated that the reactor core was exposed and heat up due to the coolant leakage and finally melted down due to high-pressure and lower-pressure injection failure. However, with the mitigation measures against the severe accident, the reactor pressure vessel (RPV) and containment integrity were maintained, thus preventing the large release of radioactive products into other cabins and the environment. Comparative analysis results also showed that all the three measures performed their scheduled functions and were significant in mitigating the accident. The study is useful in gaining insights into the processes of the severe accident induced by cold leg SBLOCA and can be used to identify appropriate mitigation strategies and countermeasures to deal with the severe accident in the FNPP.

Sensitivity analysis of core slumping and debris quenching behavior of Fukushima Daiichi Unit-3 accident

The Great East Japan earthquake and the subsequent tsunami which occurred on March 11th, 2011 put the operating Units 1–3 at Fukushima Daiichi Nuclear Power Plant (NPP) in severe accident conditions and core meltdown due to station blackout. Although research efforts have been made by various parties to study the accident scenarios since the Fukushima accident, there remain unresolved issues regarding the core degradation behavior suggested by measurement data such as water level, Reactor Pressure Vessel (RPV) pressure and Primary Containment Vessel (PCV) pressure. To analyze and resolve such issues would be helpful to promote further understanding of the severe accident scenario at Fukushima units as well as the decommissioning work undergoing.The current study focuses on a detailed analysis of the RPV pressure peak event that occurred in Unit-3 at 12:00 on March 13th 2011. Sensitivity analysis cases were carried out with MELCOR 2.2 codeiThe MELCOR code version applied in the current study is MELCOR 2.2.9541. The results used in this study were obtained before April 2018 under research collaboration with Waseda University.i with sensitivity parameters that can influence the RPV pressure behavior, such as the debris quenching heat transfer coefficient, the number of opening SRVs during the RPV pressure peak event, amount of core slumping and particulate debris diameter. The cases that could reproduce the RPV pressure peak were further discussed to show likely debris bed energy history and the water mass history in the lower plenum during the RPV pressure peak event. The current study suggests that 1) Opening of SRVs equivalent to the total area of 4–6 fully-open SRVs (or equivalent leak area) could have occurred during the pressurization phase of the RPV accompanied by heavy debris quenching effect, while the opening of SRVs equivalent to a total area of at least 2 fully-open SRVs (or equivalent leak area) could have occurred during the depressurization phase of the RPV accompanied by moderate debris quenching effect. 2) The particulate debris diameter is not a very sensitive parameter when evaluating the debris quenching effect of Unit-3 in the current MELCOR modeling. 3) The current modeling suggests that around 70–110 GJ of energy can be removed by coolant during the debris quenching period with 30 tons of water reduction from the lower plenum.

Analysis of the Effect of Vessel Failure and Melt Release on Risk of Containment Failure Due to Ex-Vessel Steam Explosion in Nordic Boiling Water Reactor Using ROAAM+ Framework

Abstract Nordic boiling water reactor (BWR) design employs ex-vessel debris coolability in a deep pool of water as a severe accident management (SAM) strategy. Depending on melt release conditions from the vessel and core–melt coolant interactions, containment integrity can be threatened by: (i) formation of noncoolable debris bed or (ii) energetic steam explosion. Melt is released from the vessel affect ex-vessel phenomena and is recognized as the major source of uncertainty. The risk-oriented accident analysis methodology (ROAAM+) is used for quantification of the risk of containment failure in Nordic BWR where melt ejection mode surrogate model (MEM SM) provides initial conditions for the analysis of debris agglomeration and ex-vessel steam explosion which determine the respective loads on the containment. Melt ejection SM is based on the system analysis code methods for estimation of leakages and consequences of releases (computer code) (MELCOR). Modeling of vessel failure and melt release from the vessel in MELCOR is based on parametric models, allowing a user to select different assumptions that effectively control lower head (LH) behavior and melt release. The work addresses the effect of epistemic uncertain parameters and modeling assumptions in MEM SM on the containment loads due to ex-vessel steam explosion in Nordic BWR. Sensitivity and uncertainty analysis performed to identify the most influential parameters and uncertainty in the risk of containment failure due to ex-vessel steam explosion. The results of the analysis provide valuable insights regarding the effect of MELCOR models, modeling parameters, and sensitivity coefficients on melt release conditions and predictions of ex-vessel steam explosion loads on the containment structures.

Challenges and sensitivities in the modelling of Fukushima Daiichi Unit 1 unfolding with MELCOR 2.2

The accident occurred in Japan on 11 March 2011 has revived the interest for the analysis of severe accidents. The scarce and sometimes unreliable data concerning boundary conditions, effectiveness of accident management measures and equipment performance, pose a tough challenge in modelling the accident scenarios.Throughout an analysis of the challenges posed by the Fukushima Daiichi Unit 1 data recorded, this paper describes the major postulates proposed by CIEMAT concerning the equipment and component responses, the effectiveness of accident management actions and the MELCOR model applied. Among most influencing assumptions are those related to the reactor pressure vessel (RPV) leaking pathways and failure mode, the water flow rate entering the reactor, the potential leaking pathways and failure mode and location from the primary containment vessel (PCV) to the reactor building, the corium relocation from RPV to the cavity and its distribution in the PCV, the potential stratification of the suppression pool and the hypotheses made a priori concerning fission product release and transport. Based on the postulated scenario and model, a remarkable agreement of the thermal footprints in terms of RPV and PCV pressures during 500 h has been achieved, in which the RPV and PCV leaks/failures as well as venting played a determining role in the short run of the accident and water injection heavily conditioned the long one. As for the scarce data related to fission products (FP), a consistent agreement is found in the suppression chamber, but estimates in the Dry-Well are about an order of magnitude below measurements despite showing the observed trend. A number of factors might affect FP comparisons to data, from the approximate method to derive dose rates (measurements) from FP masses (MELCOR results) to the RPV and PCV postulated failures. Anyway, based on the data available the set of hypotheses and approximations made seem to make up a defensible scenario for Fukushima Daiichi Unit 1.The studies and results presented in this paper have been achieved under the frame of the OECD/BSAF projects through the CSN-CIEMAT collaboration agreement on severe accidents research.