Severe Accident Code Research Articles

Uncertainty and sensitivity analysis of the fission product behaviour in the Phébus FPT1 test with the system code AC2

The Phèbus FP (Fission Product) research program, conducted by IRSN and CEA, aimed to improve the understanding of the phenomena governing core melt down, fission product release and transport, as well as the fission product behaviour in the containment. The test facility represents a 900 MWe PWR at a 1/5000 scale. The second test, FPT1, was carried out in 1996 and utilised an Ag-In-Cd alloy control rod. There is a growing interest in Best Estimate Plus Uncertainty analysis both in the research, as well as in the licensing. It has a long-standing history in case of thermal–hydraulic system codes focusing on design basis accidents, but its application is limited on the severe accident domain. In recent times however an increasing attention has been paid to evaluating severe accident codes’ uncertainties. This study presents an uncertainty and sensitivity analysis (UaSA) of the FPT1 test, focusing on fission product behavior modeling within the severe accident code system AC2. The detailed application of the coupled codes ATHLET-CD and COCOSYS in this context is novelty. Detailed description of the methodological and practical approach is provided. The uncertain input parameters chosen and quantified are directly related to the phenomena describing the fission product transport and retention phenomena both in the primary circuit, as well as in the containment, while thermal–hydraulic conditions within the primary circuit were kept as much as possible constant. Finally, the study has been complemented by a sensitivity analysis deriving the Spearman's rank correlation coefficient for each uncertain input parameter.

Numerical simulation of corium flow through rod bundle and/or debris bed geometries with a model based on Lattice Boltzmann method

A new model is proposed to investigate the relocation and the distribution of hot corium flows in different configurations (rod bundle, porous debris bed) representative of a severe accident in a Light Water Reactor (LWR). Our model relies on the coupling between a modified Lattice Boltzmann Method (LBM), called Free-Surface LBM, that solves hydrodynamics of unsaturated corium and a Finite Volume Method (FVM) that solves heat transfers. Corium solidification and melting are addressed by implementing a correlation between the temperature and the viscosity. Several simulations on representative elementary volumes were performed, varying configurations (debris bed, rod bundle with and without grid). From the results, it is possible to capture important details of the flow at a scale lower than the pore scale and, at the same time, it is possible to take into account the average effects at the scale of several pores. Presented as a proof of concept these preliminary studies show the interest of this kind of CFD approach to identify which parameters at microstructure scale can potentially govern the corium relocation kinetics at macroscopic scale. It will provide useful information that might improve core degradation models in severe accident codes, such as ASTEC.

Analysis of The Influence of Models of Individual Physical Processes and Phenomena on the Calculation Time of the Source Term in Severe Accidents

Dependence of the CPU time needs for calculation of the source term during a severe accident at nuclear power plants with WWER reactors from the individual physical processes and phenomena used as part of the calculation tools similar to the SOСRAT/V3 severe accident code is investigated. This analysis allows revealing the most expensive models in terms of runtime. The simplification of these models can ensure the greatest acceleration of the calculation. The relevance of the task draws from the need to develop new or adapt existing calculation tools for assessing the intensity of radioactive emission sources for the tasks of emergency preparedness and response considering the specific requirements for the accuracy of numerical estimates and time to obtain them. The paper demonstrates the possibility of reducing CPU time without significant loss of accuracy of the numerical estimates by simplifying the spectrum of aerosols sizes. The efficiency of the proposed approach is demonstrated by the example of modeling the Phebus FPT1 experiment.

New correlations for focusing effect evaluation of the light metal layer in the lower head of a nuclear reactor in case of severe accident

In case of a severe accident (SA) in a nuclear reactor, the core heats up and materials melt and relocate to the lower head of the vessel. In such configuration, stratification occurs between the oxide and metal phases present in the melt. When the metal phase is lighter than the oxide phase, the maximum heat flux applied to the vessel wall is concentrated at the location of this top metal layer, creating the so-called “focusing effect”. The modelling of the focusing effect is essential for SA codes to properly evaluate the vessel failure time or to demonstrate the vessel integrity, when In Vessel Retention (IVR) strategy is implemented. However, existing correlations used in SA codes for focusing effect modelling have some limitations: if the thickness of the metal layer becomes very small, they predict an extremely high heat flux. Moreover, they were validated with tests using water, whereas the Prandtl number was found to play a significant role in this heat transfer.In this paper, the focusing effect is studied based on 3D Direct Numerical Simulations (DNS) of a cylindrical metal layer, covering all possible metal layer characteristics in terms of geometry and boundary conditions, especially for its top surface where the intensity of radiative heat transfer may vary. The analysis of the fluid behavior and the quantitative results obtained allow to derive new correlations and a new model based on the radial profile of the mean temperature of the system and on the radial profile at the top surface. Results at reactor scale confirm the overestimation of existing 0D model used in SA codes for a thin metal layer and highlight the relevance of the proposed model for focusing effect evaluation during transient situations.

GOTHIC and MELCOR benchmarking of the CNL strong condensation containment apparatus experiments

Understanding severe accident behaviour in water-cooled small modular reactors is crucial for their deployment, as well as having qualified software tools for safety analysis. A model benchmarking exercise was conducted where predictions from the software tools GOTHIC and MELCOR were compared against past experiments from the Strong Condensation Containment Apparatus (SCCA). Key phenomena that can be expected in small modular reactor (SMR) containments, given their small size and external passive water cooling, are explored, including how well the codes replicate stratification between a hot upper layer and a cooler under layer. This overall thermalhydraulics behaviour impacts other important phenomena, like heat transfer and aerosol deposition. MELCOR, an integrated severe accident code, and GOTHIC, a thermalhydraulic and containment code, were used in benchmarking activities, wherein the experimental results were compared to their simulation results. The findings demonstrated MELCOR’s capabilities to simulate the SCCA experiments and GOTHIC’s strength in predicting multiphase flow and thermalhydraulics in containment and cooling systems. In addition, the benchmark comparisons provided new insights toward modelling experiments with these codes as well as the behaviour observed in the SCCA experiment. These findings will guide accident management and emergency preparedness requirements for SMRs.

Experimental determination of deposition velocity of CsOH aerosols on CaCO3 at temperature range 170 - 290 °C

A field assessment at the Fukushima-Daiichi Nuclear Power Station revealed high radioactivity on the concrete shield plugs, which is estimated above 20 PBq for Cs-137 at units 2 and 3. This leads to significant interest in the retention of Cs on concrete during severe accidents (SA). However, the interaction of CsOH, as one of the main Cs forms released in SA, with concrete surfaces at elevated temperatures remains poorly researched. In this study, we have experimentally investigated the deposition behavior of CsOH on CaCO3, which is the primary phase existing on the surface of concrete, under humid atmosphere. As a result, the chemical reaction enhanced deposition rate (N), and increased linearly with CsOH concentration (Cg), as following expression:N(µg/cm2.s)=vdCg, where vd is temperature-dependent deposition velocity as given by lnvd(cm/s) = -3785.8/T + 3.766, for T in the range of 170 and 290 °C.This empirical model can be integrated into severe accident codes to quantify the chemical trapping of cesium on concrete surfaces during ex-vessel release. Moreover, it can contribute to understanding the reasons behind the high dose rate on concrete shield plugs at the Fukushima Daiichi Nuclear power stations and aid in developing effective decommissioning practices for concrete structures.

Application of reinforcement learning to deduce nuclear power plant severe accident scenario

Severe accident scenarios for nuclear power plants are determined through probabilistic safety analysis (PSA). In this process, it is possible to identify the failure sequence of specific components, but assessing the impact of component failure time on the severity of accident remains a challenge. In this study, a novel approach is presented that utilizes machine learning methodologies such as reinforcement learning (RL) to complement traditional PSA. The proposed process was validated by comparing whether the most severe accident scenarios obtained through critical accident simulations can be reproduced by a RL, since this is a novel use of machine learning techniques. The comparison shows feasibility of exploring critical accident scenarios using RL. To implement the reinforcement learning methodology based on the existing system code, supervised learning model that can predict the remaining time of reactor vessel failure was implemented in this study. Based on this prediction model and data from existing catastrophic accident simulation, RL was implemented. The results obtained from RL were subsequently validated with the results of severe accident code simulation. In summary, new methodology for applying machine learning techniques to the nuclear accident analysis process was presented, and the feasibility and potential of the proposed methodology were discussed.

ACE surrogate Model–Based uncertainty and sensitivity analysis methods for severe accident codes

This paper explores the alternating conditional expectation (ACE) algorithm-based surrogate model to advance the state-of-practice in uncertainty and sensitivity analysis methodologies for severe accident codes. For engineering purposes, the ACE algorithm has been used as an alternative means to find the optimal functional forms of the multiple input variables and response variables of interest. Analysis results here demonstrate that compared with the reference cases the proposed surrogate model provides much higher performance in terms of the coefficient of determination (R2) and normalized root mean square error (NRMSE), thus giving more robust insights into the relationship and correlation between the input parameters and figures of merit (FOMs) of interest. Relevant results and insights are summarized in terms of points of interest.

Methods for the radioactive release estimation under DEC-A conditions

After the Fukushima Daiichi Nuclear Power Plant accident, the importance was raised to strengthen the assessment of the safety level of Nuclear Power Plants (NPP) by considering situations more severe than those integrated during the design of the plants. Design Extension Conditions (DEC) term was introduced by IAEA and WENRA. One of the most important subjects in the analysis of these conditions is the evaluation of radiological consequences.The Reduction of Radiological Accident Consequences (R2CA) collaborative project started in 2019 in the frame of the Horizon-2020 Program of the European Commission. The project addresses a broad scope of LWR designs (Gen II, III and III +) through the analyses of bounding scenarios of Loss Of Coolant Accidents (LOCA) and Steam Generator Tube Rupture (SGTR) transients and explores DBA and DEC-A conditions. The Lithuanian Energy Institute is taking part in this project as well.The part of work provided in the frame of R2CA project is presented in this article. The generic BWR-4 type reactor with MARK-I containment was analyzed in the case of DEC-A conditions. In this analysis the main thermal hydraulic processes, fission product release and its transport from the broken loop through containment to the environment were investigated. The number of ruptured fuel assemblies is the key parameter determining the fission product release from the core. The methodology which would help more precisely evaluate the number of failed fuel assemblies is presented in this work. The methodology consists of the application of integral severe accident code ASTEC and fuel performance code TRANSURANUS. Analysis using ASTEC code showed the importance of the reactor core nodalisation scheme. Taking iterative calculations, the relative power to which fuel assemblies remain intact at the selected DEC-A condition scenario was obtained. These results were verified with TRANSURANUS code calculations considering the uncertainties of the most relevant parameters. Based on TRANSURANUS calculation results, the ASTEC core nodalisation was improved. Improved ASTEC model nodalisation allowed to calculate more realistic rates of fission products releases from the fuel to the coolant, from the coolant to the containment and as well from the containment to the environment.

Qualification Analysis of Experimental Data for High-Temperature Oxidation of Russian Zirconium Alloys in Steam

The paper presents the results of qualification analysis of experimental data on high-temperature oxidation of the Russian zirconium alloys in steam, which include an assessment of the consistency of measurement results in tests, and numerical modeling of experiments with SOCRAT-V1/V2 code. Conclusions are provided on the possibility of using the SSC RIAR experimental programs for validation of severe accident codes, and the applicability of the SOCRAT-V1/V2 code for prediction of the high-temperature oxidation of the Russian zirconium claddings of fuel rods.

Nitriding model for zirconium based fuel cladding in severe accident codes

A model has been developed to describe the nitriding of partially oxidized zirconium based cladding during an air ingress sequence when the reaction has become starved of oxidant (oxygen and/or steam), and the subsequent re-oxidation of nitride following of restoration of coolant. Key aspects of the model are the estimation of oxygen-stabilised alpha zirconium, α-Zr(O), formed during pre-oxidation and its reaction with the nitrogen. Nitriding of metallic Zr is much slower than α-Zr(O), and plays a comparatively minor role. The model is based on data from separate-effects tests comprised pre-oxidation, nitriding in the absence of oxidant, and re-oxidation in the absence of nitrogen, which were used to derive the kinetic parameters for the main reaction processes. Developmental assessment was performed using the test results, demonstrating favourable agreement for the main reaction signatures. Independent assessment against Integral Test data is underway.

Uncertainties impact on the major FOMs for severe accidents in CANDU 6 nuclear power plant

In the nuclear safety studies, a new trend refers to the evaluation of uncertainties as a mandatory component of best-estimate safety analysis which is a modern and technically consistent approach being known as BEPU (Best Estimate Plus Uncertainty).The major objectives of this study consist in performing a study of uncertainties/sensitivities of the major analysis results for a generic CANDU 6 Nuclear Power Plant during Station Blackout (SBO) progression to understand and characterize the sources of uncertainties and their effects on the key figure-of-merits (FOMs) predictions in severe accidents (SA).The FOMs of interest are hydrogen mass generation and event timings such as the first fuel channel failure time, beginning of the core disassembly time, core collapse time and calandria vessel failure time. The outcomes of the study, will allow an improvement of capabilities and expertise to perform uncertainty and sensitivity analysis with severe accident codes for CANDU 6 Nuclear Power Plant.

Uncertainty quantification for a severe accident sequence in a SFP in the frame of the H-2020 project MUSA: First outcomes

The Management and Uncertainties of Severe Accidents (MUSA) project, funded in HORIZON 2020 and coordinated by CIEMAT (Spain), aims at consolidating a harmonized approach for the analysis of uncertainties and sensitivities associated with Severe Accidents (SAs) focusing on Source Term (ST). In this framework, the objectives of the Innovative Management of Spent Fuel Pool Accidents (IMSFP – WP6), led by IRSN (France), are to quantify and rank the uncertainties affecting accident analyses in a Spent Fuel Pool (SFP), to review existing and contemplated SA management measures and systems and to assess their possible benefits in terms of reduction of radiological consequences.To quantify the propagation of the uncertainties of the input parameters to the output uncertainties of severe accident codes (ASTEC, MELCOR, RELAP/SCDAP), a diverse set of uncertainty quantification (UQ) tools (DAKOTA, RAVEN, SUNSET, SUSA) are used. The statistical framework used by the different UQ-tools is similar e.g. pure random (Monte Carlo) and Latin hypercube sampling (LHS).Fourteen partners from three different world regions are involved in the WP6 activities. The target of this paper is to describe the achievements during the first three years of the project. In a first part, a description is given of the SFP accidental scenario, of the key target variables and radionuclides chosen as ST Figures of Merit (FoM) and of the identified uncertainty sources in models and input parameters. A key element when defining the SFP scenario has been the consideration (or not) of the reactor building, as it is expected to significantly affect analyses. In a second part, the first insights coming out from the calculation phase of the project are presented. The review of existing SA management measures is also exposed, as well as systems whose benefits will be assessed in the second phase of the project. Finally, challenges that arise from such an exercise are discussed, as well as major difficulties found when applying UQ methodologies to SFP scenarios and solutions adopted.

Parametric MELCOR 2.2 sensitivity and uncertainty study with a focus on aerosols based on Phébus test FPT1

The Phébus test FPT-1 was carried out to investigate the fission product release, transportation, and distribution in the reactor coolant system and the simulated containment during a severe accident. The test was divided into three phases: the core degradation, aerosol transport, and containment phenomena. In this work, a parametric sensitivity and uncertainty analysis of the two first phases, i.e., core degradation and aerosol transport, is conducted with the focus on aerosol parameters.The study was done in a frame of WP4 of the European MUSA project whose main objective is to develop a methodology for severe accident uncertainty analysis.Due to the wide scope of the MUSA project, this study focuses not only on the UQ analysis but also on the parametric dependency and data interpretation. The main subject was investigating the uncertainty of the MELCOR 2.2 aerosol models and the relation between the chosen parameters as well as evaluating the usability of essential analytical tools available for such analyses.The calculations were carried out using the integral severe accident code MELCOR2.2 and the generic input deck of the FPT-1 experiment. The uncertainty quantification tool DAKOTA was used as a statistical tool. It was coupled with MELCOR within the SNAP environment.Fifteen uncertain parameters related with the aerosol release and transport were investigated. Two figures of merit were chosen to be the amount of deposition of Cs in the reactor coolant circuit, and the mass of air-borne aerosol in the containment. Preliminary analysis showed a significant correlation between the aerosol dynamic and agglomeration shape factors, density and minimum aerosol diameter with both FOMs. In addition to the standard uncertainty analysis, five MELCOR/DAKOTA UQ sensitivity calculations with modified parameters were performed and analyzed. Additionally, the impact of the parameters was investigated by excluding the most correlated ones from the analysis. The results confirm the strong correlation of shape factors and the importance of their ranges for overall UQ results.The authors show that by coupling the chosen analytical tools together, it is possible to carry out uncertainty quantification of severe accident analysis.Based on the study performed, the authors underline the need for further investigation of parameter distributions, the effect of the chosen number of samples on the results, and a global time-dependent UQ analysis.

PANDA experimental database and further needs for containment analyses

PANDA is a large-scale, multi-compartment, thermal-hydraulics facility located at the Paul Scherrer Institute (PSI) in Switzerland and is operated for nuclear safety projects. PANDA is continuously upgraded with new components and instruments to meet the requirements of each new project. The modularity of the facility and its instrumentation provide unique opportunities to perform experiments ranging from basic containment phenomena studied mainly in a single vessel to integral behavior of Passive Containment Cooling Systems (PCCS) requiring up to six PANDA vessels and four passive condensers (PCCs, IC) submerged in four water pools. Experimental data analyses are performed using system codes, lumped parameter (LP) severe accident codes and Computational Fluid Dynamics (CFD) codes. In this paper, we provide an overview of the PANDA experimental database and identify further data needs for containment analyses, considering the main results of recently completed PANDA projects.

THAI experiments on iodine behavior in a room chain representing flow conditions in large containments

Iodine which can be released into the containment during a severe accident, is considered one of the most important nuclides with potential for public dose. Significantly different iodine concentrations can be expected in the individual compartments and rooms of large containments, due to geometrical constraints and the strong dependence of many iodine reactions on local thermohydraulic conditions. Such inhomogeneity may be relevant for organic iodide formation which depends on the concentration of deposited I2 and the dose rate.The objective of the two tests Iod-33 and Iod-35 was to measure the distribution of iodine along a convective flown-through room chain. This arrangement shall simulate the pathway of iodine from the locations of release (primary loop rupture) or formation (sump) along the plant rooms up to the dome of the containment and back. With respect to iodine, the aims were to measure the distribution and mixing of gaseous iodine in different rooms, the decrease of gaseous iodine by its deposition on painted surfaces, and the distribution of deposited iodine on the different painted surfaces.The two tests were performed in the two vessel THAI+ facility with additionally installed painted surfaces (painted with Gehopon epoxy paint) under superheated and condensing conditions. The main part of the painted surface area was provided in the form of painted steel sheets and arranged to form two intermediate floors within the THAI test vessel including small flow openings. Gaseous iodine was injected over several hours into the room chain in which a slow convection was established by means of differential wall heating.The gaseous I2 concentration inside the THAI+ facility increased continuously until the I2 injection was terminated. Simultaneously, the iodine deposition onto the painted surface increased. The gaseous I2 concentration decreased rapidly after stopping the I2 injection, in favor of a continuation of the I2 deposition on painted surfaces. The multi-compartment effect along the room chain has been demonstrated by the test results and the data allows for further development and validation of the I2/paint reaction model of severe accident codes.

MELCOR integrated severe accident code application to safety assessment of high-temperature gas-cooled reactors

MELCOR is an integrated thermal hydraulics, accident progression and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (U.S. NRC) since the early 1980 s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts over the past decades have expanded its application space to non-LWR reactor concepts. Current MELCOR development efforts have been focused on providing the U.S. NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC’s near-term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enhancements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. In this paper, we describe the generic plant model developed to demonstrate MELCOR capabilities to perform high-temperature gas reactor (HTGR) safety evaluations. This model represents a TRi-structural ISOtropic particle fuel (TRISO) pebble bed HTGR with a primary system rejecting heat to a recuperative heat exchanger. Surrounding the reactor vessel is a reactor cavity contained within a confinement volume cooled by the Reactor Cavity Cooling System (RCCS). A range of demonstration calculations are performed to evaluate the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing reactivity control, primary system heat removal and reactor vessel decay heat removal, and confinement cooling.

Analysis of BDBA sequences in a generic IRIS reactor using ASTEC code

Integral Severe Accident (SA) codes are aimed at providing an exhaustive coverage of all the main phenomena taking place in a core melt accident. Today, these deterministic codes have reached a high level of maturity for the simulation of operating reactors and the nuclear technical community is starting to extend their applicability to advanced reactor designs, as Small Modular Reactor (SMR). In the framework of the NUGENIA TA-2 ASCOM (ASTEC COMmunity) collaborative project, a generic input-deck based on the IRIS design has been developed for the ASTEC code. The generic SMR ASTEC model has been already proved able to simulate the main thermal–hydraulic phenomena driving the passive mitigation of a SBLOCA in Design Basis Accident (DBA) conditions. The same initiator event, regarding the guillotine break of a Direct Vessel Injection (DVI) line, will be assumed for the simulation of beyond-design scenarios by considering the unavailability of selected passive safety systems. The results of the ASTEC simulations of four Beyond Design Basis Accidents (BDBAs) (study carried out with ASTEC V2.2, IRSN all rights reserved, [2021]) will be analyzed and discussed against the reference DBA sequence in the present paper. This study is aimed at proving the first insights about the capability of the ASTEC model of a generic IRIS reactor to be used in BDBA and in SA analyses, if significant core degradation takes place. In addition, it characterizes the role played by each safety system in SMR passive mitigation strategy and give the possibility to characterize the phenomenologies specific of SMR designs.

Steady-state thermal hydraulic modelling and turbine trip transient simulation of the NuScale integral pressurised water reactor using the ASTEC code

A steady state model of the NuScale reactor has been built using the thermal hydraulic module CESAR of the severe accident code ASTEC. Construction of the model was achieved through iterations based on convergence towards data provided by NuScale pertaining to important operating properties—core temperatures, coolant mass flow rate and secondary steam temperature—at 100% power (160 MW) and also at lower power levels. Good agreement with reference values was achieved to within 3% deviation. With a similar methodology, the previous model was adjusted to give a second steady state model meant to provide conservative initial conditions from which a turbine trip transient simulation would be performed. Sound agreement with literature was achieved, with maximum deviations of less than 1.8% for monitored parameters. Lastly, a turbine trip transient was simulated and compared to reference data from analyses performed with NRELAP5 and RELAP5/SCDAPSIM. The most important system behaviors and event timings are well captured and reproduced, with discrepancies arising from inevitable estimates made due to lack of information.

Dynamic Event Tree Analysis as a Tool for Risk Assessment in Nuclear Fusion Plants Using RAVEN and MELCOR

In the broad framework of the nuclear power plants (NPPs) industry, the dynamic probabilistic risk assessment could answer the time dependence deficiency of the event tree and fault tree analysis. The basic event tree approach relies on experts’ pre-constructed accident sequences without exploring the time-dependent nature of an accident scenario, which could strongly affect the accident sequence. Conversely, the effects of events timing can be studied by adopting a dynamic event tree (DET) approach. Developing a DET methodology requires integrating a system code capable of replicating an accident scenario and a logic-driver code able to generate the event tree sequence, trigger plant safety systems, and manage other relevant events throughout the simulation. For this purpose, Methods for Estimation of Leakages and Consequences of Release (MELCOR) and reactor analysis and virtual control environment (RAVEN) have been coupled through a Python script developed by the Sapienza University of Rome, Rome, Italy, to perform DET studies during accident transients in fusion and fission reactors. RAVEN is a software tool developed at the Idaho National Laboratory (INL), Idaho Falls, ID, USA, to act as a control logic driver and post-processing tool for different applications. MELCOR for fusion is a fully integrated design basis and severe accident code that simulates thermal-hydraulic behavior and self-consistently accounting for aerosol transport in nuclear facilities and reactor cooling systems for the evaluation of the source term in fusion reactors. The coupling between these codes will provide a wide range of NPP risk assessment analyses, establishing new best practices. In this work, a preliminary DET study has been performed, selecting as initiating event an ex-vessel loss of coolant accident (LOCA) in the water-cooled lithium lead test blanket system (WCLL TBS) to be tested in International Thermonuclear Experimental Reactor (ITER). Time-dependent parameters such as the intervention of the plasma shutdown system and the closure of the main system isolation valves have been sampled to study evolving system scenarios.