Severe Accident Management Research Articles

ELSMOR - towards European Licensing of Small Modular Reactors:Methodology recommendations for light-water small modular reactors safety assessment.

Decarbonization of energy production is key in today's societies and nuclear energy holds an essential place in this prospect. Besides heavy-duty electricity production, other industrial and communal needs could be served by integrating novel nuclear energy production systems, among which are low-power nuclear devices, like small modular reactors (SMRs). The ELSMOR (towards European Licensing of Small Modular Reactors) European project addresses this topic as an answer to the Horizon 2020 Euratom NFRP-2018-3 call. The consortium includes 15 partners from eight European countries, involving research institutes, major European nuclear companies and technical support organizations. The 3.5-year project, launched in September 2019, investigates selected safety features of light-water (LW) SMRs with focus on licensing aspects. Providing a comprehensive compliance framework that regulators can adopt and operate, the licensing process of such SMRs could be optimized, helping their deployment. In this prospect, as a result of ELSMOR's work, this article gives an overview of the specific issues that LW-SMRs may bring about in the different domains of nuclear safety, in terms of: •Methodological standpoints: safety goals, safety requirements, safety principles (defence-in-depth implementation);•Main safety functions of reactivity control, decay heat removal and confinement management;•Severe accident management;•Other safety issues particular to SMRs: use of shared systems; performing of multi-unit probabilistic safety assessment (PSA); spent fuel management, transport and disposal management. In this article, adequate methodologies are developed to deal with these issues and to help assess the safety of LW-SMRs. This work gives a precious synthesis of the safety assessment issues of LW-SMRs and of the associated methodologies developed in the context of the ELSMOR European project.

A condition-informed dynamic Bayesian network framework to support severe accident management in nuclear power plants

In the nuclear industry, Severe Accident Management Guidelines (SAMGs) are developed with the objective of providing prescriptive actions for mitigating the consequences of Nuclear Power Plant (NPP) accidents. These guidelines are defined with respect to a set of expert-based prototypical severe accident scenarios and are symptom-based, i.e., they relate to the values of specific monitored plant parameters. Expert-based SAMGs might fail to deal with non-prototypical scenarios, i.e., scenarios not considered a priori by the experts. To address this issue, in this paper we propose a methodology based on Dynamic Bayesian Networks (DBNs) to support operators decision-making for avoiding the escalation of severe accidents to non-prototypical scenarios. The methodology is based on the integration of condition monitoring data into a DBN to allow it to i) infer the current Plant Operating Condition (POC), Initiating Event (IE), and NPP safety barriers Damage State (DS), ii) predict the grace time (GT) available to carry out mitigative actions, and iii) evaluate the effectiveness of the alternative candidate mitigative actions taking into account the safety barriers estimated DS and the predicted available GT. The proposed methodology is exemplified on a Loss of Coolant Accident (LOCA) in a WWER-1000.

Three-dimensional CFD analysis of PHWR exposed core under postulated severe accident condition

A three-dimensional steady-state thermal–hydraulic analyses of a large sized Pressurized Heavy Water Reactors (PHWRs) with varying levels of exposure is carried out for degraded core heat transfer assessment and presented in this paper. Low-frequency postulated severe accident without mitigating actions leads to exposure of fuel channels in PHWRs. Such exposure of fuel channels leads to their heat-up, governed by interdependent heat transfer mechanisms such as radiation, convection and conduction heat transfer and thermo-chemical interactions among the core constituents. In order to understand such complex behavior and plan the severe accident management guidelines, numerical analyses of the exposed core are of necessary. This paper illustrates the influence of the temperature and velocity profiles vis-à-vis different regions of the exposed core on the fuel channel heat-up. The interesting interdependency of convection and radiation heat transfer on fuel channel heat-up is reported. The dominance of convection heat transfer is observed for initial exposures of the fuel channels. However, the advanced stages of fuel channel exposures are observed to be dominated by radiation heat transfer. Intriguingly, it is observed that the convective cooling of exposed fuel channels by steam flow is counter-acted by exothermic heat generation from Zircaloy-made calandria tube surface and steam interaction. This leads to supplemental power addition for fuel channel heat-up. Moreover, it is interesting to note that the heat-up of fuel channels is restricted to central regions of the exposed core due to degraded steam cooling and higher-powered fuel channels. It is worth noting that fuel channel disassembly is observed in the advanced levels of fuel channel exposures. Moreover, fuel channel length equivalent to nearly 7–8 fuel bundle lengths undergo disassembly depending upon the fuel channel elevation and location in the core. The calandria vault water is observed to undergo film boiling at the onset of fuel channel disassembly.

Numerical simulation of in-vessel steam explosion for China third-generation PWR

In this study, in order to evaluate the Reactor Pressure Vessel (RPV) integrity under in-vessel steam explosion loads for China third-generation Pressurized Water Reactor (PWR), 2D and 3D numerical simulation of in-vessel steam explosion is conducted using MC3D code. The sensitivity analyses for in-vessel steam explosion are performed, including break area, water temperature, water level, melt temperature, and break height. The results show that the premixing stage cannot cause RPV failure and the possibility of the steam explosion blowing the RPV upper head off is almost non-existent. When the break area increases, the maximum pressure and impulse of the RPV inner wall increase obviously first and then generally decrease slightly. With the increase of water temperature, the maximum pressure and impulse first increase and then decrease. When the water level is higher, the maximum pressure variation is small. When the water level is very low, the maximum pressure decreases. The maximum impulse generally decreases with the decrease of the water level. When the melt temperature exceeds 2850 K, the steam explosion is not sensitive to the melt temperature. With the increase of the break height, the maximum pressure and maximum impulse first increase and then decrease. Furthermore, based on the results of sensitivity analysis, two cases of conservative conditions are selected, and the RPV integrity under in-vessel steam explosion loads for China third-generation PWR is considered. The result shows that the integrity of the RPV can be maintained under the in-vessel steam explosion loads for China third-generation PWR with a large margin. These studies provide some reference to improve the severe accident management strategy in China.

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.

Validation of EPG/SAG Rev. 4 of the Kuosheng Nuclear Power Plant Using the MAAP5 Code

The Boiling Water Reactor Owner’s Group released Emergency Procedure Guidelines and Severe Accident Guidelines Revision 4 (EPG/SAG Rev. 4) in 2018. The major improvement to EPG/SAG Rev. 4 was Contingency 1 (Alternate Level/ Pressure Control). Contingency 1 coordinates the reactor pressure vessel (RPV) water level and RPV pressure control action to prolong the availability of steam-driven injections and optimize the transfer to motor-driven systems. In this study, the effectiveness of the EPG/SAG Rev. 4 Contingency 1 strategy was compared with those of EPG/SAG Revision 2 Contingency 1 and Specific Major Incident Guidelines (SMI) using the Modular Accident Analysis Program, Version 5 (MAAP5). SMI was developed by the Taiwan Power Company to mitigate a Fukushima-like accident. The surrogate plant that analyzed is the Kuosheng Nuclear Power Plant (NPP). The Kuosheng NPP is BWR-6 Mark-III containment. MAAP5 is an integral severe accident analysis program that simulates the responses of a light water reactor power plant during a severe accident. This program has been used extensively for probabilistic risk assessments and for verification and validation of mitigation actions specified in severe accident management guidelines. The simulation scenarios were extended loss of alternative-current power and loss of ultimate heat sink. The low-capacity, motor-driven portable pump was the only available system for RPV injection in the first hour of the accident. In this time period, the RPV water level and pressure were controlled by reactor core isolation cooling and safety relief valves. After this study, the strategy of EPG/SAG Rev. 4 Contingency 1 was successfully validated, and the effectiveness of minimum pre-depressurization RPV water level and the low-capacity, motor-driven portable pump were also demonstrated in this study.

Influence of different thermodynamic parameters on variation across the aerosol size spectrum

During a severe accident in a nuclear power plant, fission products can be released into the containment in the form of radioactive aerosols. The behavior of these aerosols is influenced by various thermodynamic parameters, including temperature, pressure, and relative humidity. Changes in these parameters can impact the aerosol size spectrum, leading to variation in aerosol transport and deposition. Having a comprehensive grasp of how thermodynamic parameters impact change in the aerosol size spectrum is crucial for managing and mitigating the consequences of a severe accident, as it allows for better prediction of aerosol behavior and appropriate interventions. A series of experimental works on in-/soluble aerosols and their mixtures has been carried out using the IN–EX facility located at the Forschungszentrum Juelich the Germany. The changes in the aerosol size spectrum of SnO2 , CsI and two different mass composition mixtures (SnO2 : CsI = 7:3 or 4:6) under various thermodynamic conditions are discussed and analyzed. The results show that increasing the temperature leads to a broader aerosol spectrum for SnO2 , but not for CsI aerosols. With the increasing pressure, the aerosol size spectrum shifts towards smaller particles for CsI, but not for SnO2 . Furthermore, increasing the relative humidity causes the aerosol size spectrum of both SnO2 and CsI particles to shift towards larger particles, whereas CsI particles are more pronounced. The behavior of such mixtures relies on the composition of the primary substance within it. These findings provide robust information for improving the strategies of severe accident management.

An analysis method to assess in-Calandria corium retention time in PHWRs

In Pressurized Heavy Water Reactors (PHWRs), during a postulated severe accident involving the core collapse, the core internals would fall onto the Calandria bottom and form a debris bed. Under such conditions, the core debris bed would be contained by the Calandria, which also facilitates its cooling by the Calandria vault water. In absence of any accident management action, the Calandria temperatures would rise, causing loss of material strength that may also threaten the structural integrity of Calandria. The gross structural failure of Calandria would limit its ability to further retain the core debris and would jeopardize the in-vessel accident management strategies. The knowledge of structural degradation of Calandria with time till its ultimate failure is important for formulating and assessing the severe accident management guidelines. Assessment of Calandria failure under accident conditions involves several important aspects such as complex thermo-mechanical loadings, temperature dependent elastic-plastic and creep modeling, flexibility of different parts and boundary constraints etc. In view of this, a systematic analysis method based on the sequentially coupled thermo- mechanical analysis has been devised and used for the structural assessment of Calandria. The failure assessment has been carried out as per the specified failure criteria for prominent failure modes including large unbounded deformations and plastic instability, large inelastic strains/ ductility exhaustion and creep-stress rupture. The paper covers various considerations taken into account during the analyses along with few case studies on Indian PHWRs.

CFD analysis of PHWR exposed core under postulated severe accident condition

A two-dimensional steady-state thermal-hydraulic analyses of the exposed core with varying levels of exposure is carried out for heat transfer assessment. Low-frequency postulated severe accident without mitigating actions leads to exposure of fuel channels in Pressurised Heavy Water Reactors (PHWRs). Such exposure of fuel channels leads to their heat-up which is governed by interdependent heat transfer mechanisms such as radiation, convection and conduction heat transfer. In order to understand such a complex behaviour as well as to plan the severe accident management guidelines, numerical analyses of the exposed core are required. In this paper, one-of-a-kind thermal-hydraulic analyses of the exposed reactor-scale core geometry are presented for row-wise exposures of fuel channels. These numerical analyses are terminated at the onset of fuel channel disassembly i.e., loss of geometrical intactness of the exposed core. The interesting interdependency of convection and radiation heat transfer on fuel channel heat-up is brought out in this study. The dominance of convection heat transfer is observed for initial exposures of the fuel channels. However, for the advanced stages of fuel channel exposures, the heat transfer is observed to be dominated by radiation heat transfer. Intriguingly, it is observed that the convective cooling of exposed fuel channels by steam flow is counter-acted by exothermic heat generation arising from Zircaloy and steam interaction. Moreover, it is interesting to note that the disassembly of fuel channels is restricted to central regions of the exposed core.

Optimization of prevention and mitigation of severe accidents through Research: Major outcomes and outlook of the Spanish programme

Severe accidents research activities in Spain have grown substantially since the Fukushima-Daiichi accident. Most of these research activities have been conducting by a few organizations (national research center and academia) with the support of the Regulatory Authority, utilities and engineering companies. A great number of these activities were conducted in the analytical field and their main outcomes are contributing to a better understanding of severe accident phenomena (i.e., pool scrubbing; combustible gas distribution and combustion; fission products transport; etc.) and the efficiency of severe accident management actions (i.e., FLEX; PARs; activation of safety systems, long-term measures; etc.). The national framework set between researchers and end-users (regulators and utilities) is improving the safety of the Spanish nuclear power plants and allows for the assimilation of the research results by the nuclear industry.

Scope of conceptual development of Resilient supercritical Light water-cooled reactor (SCWR-R)

The research and development (R&D) of Supercritical Light Water-cooled Reactor (SCWR) has been based on the matured Generation III light water-cooled reactor (LWR) technology and the supercritical fossil fuel-fired power plant (SC-FPP) technology. The plant concept has unique features such as the simple and compact plant system and the high thermal efficiency. Currently, the Japanese concepts are being further developed at Waseda University with focuses on improving resilience of SCWR. This study proposes and presents the overall scope of the new Resilient SCWR (SCWR-R) concept with focuses on two viewpoints: (1) SCWR can utilize variety of different core designs ranging from thermal to fast spectrum for multi purposes (e.g., Pu breeding, Minor Actinide burning) with the same plant system to respond to different fuel cycle needs; (2) severe accident management of SCWR-R can be developed with the combination of IVR and PCV with S/C to minimize the PCV internal contamination by radioactive materials, which should greatly reduce post-severe accident management. To establish the SCWR-R concept, the multi-level physics simulation methodology is currently being developed to evaluate the debris criticality, cooling and radioactive aerosol management during the severe accident.

Optimization strategy for SAM in nuclear power plants based on NSGA-II

Abstract The Severe Accident Management Guide (SAMG) is an important component of nuclear safety regulations. Many studies are being conducted to optimize severe accident management (SAM) strategies. To ensure the safety of nuclear power plants, decision makers need to monitor multiple parameters with security threats. Therefore, it is particularly important to search optimal SAM strategies under different numbers of mitigation targets. The Non-dominated Sorting Genetic Algorithm-II (NSGA-II) is an evolutionary algorithm that does not require derivative differentiation and is capable of population search. In this study, a nuclear power plant accident optimization strategy is developed using the Modular Accident Analysis Program (MAAP) in conjunction with NSGA-II. The strategy enables decision makers to consider multiple mitigation objectives in a complex decision environment. Focusing on the CPR1000, this study applies the optimization strategy to automatically search for optimal mitigation strategies for small break loss of coolant accident (SBLOCA) and station blackout hot leg creep rupture accidents (SBOHLCR). Comparing the optimization results with the basic accident sequence, it is found that the reactor pressure vessel (RPV) failure time is delayed from 72,702 s to 128,730 s under SBLOCA and from 23,828 s to 28,363 s under SBOHLCR. This study has also verified that the optimal SAM strategy obtained by the strategy through dual objective optimization has better mitigation effects than a strategy that only considers one objective. This optimization strategy has the potential to be applied to other types of severe accident management studies in the future.

Main outcomes of the Phebus FPT1 uncertainty and sensitivity analysis in the EU-MUSA project

The Management and Uncertainties of Severe Accidents (MUSA) project was funded in HORIZON 2020 and is coordinated by CIEMAT (Spain). The project aims at consolidating a harmonized approach for the analysis of uncertainties and sensitivities associated with Severe Accidents (SAs) analysis, focusing on source term figures of merit. The Application of Uncertainty Quantification (UQ) Methods against Integral Experiments (AUQMIE – Work Package 4 (WP4)), led by ENEA (Italy), was devoted to apply and test UQ methodologies adopting the internationally recognized PHEBUS FPT1 test. FPT1 was chosen to test UQ methodologies because, even though it is a simplified SA scenario, it was representative of the in-vessel phase of a severe accident initiated by a break in the cold leg of a PWR primary circuit.WP4 served as a platform to identify and discuss the issues encountered in the application of UQ methodologies to SA analyses (e.g. discuss the UQ methodology, perform the coupling between the SA codes and the UQ tools, define the results post-processing methods, etc.). The purpose of this paper is to describe the MUSA PHEBUS FPT1 uncertainty application exercise with the related specifications and the methodologies used by the partners to perform the UQ exercise. The main outcomes and lessons learned of the analysis are: scripting was in general needed for the SA code and uncertainty tool coupling and to have more flexibility; particular attention should be devoted to the proper choice of the input uncertain parameters; outlier values of figures of merit should be carefully analyzed; the computational time is a key element to perform UQ in SA; the large number of uncertain input parameters may complicate the interpretation of correlation or sensitivity analysis; there is the need for a statistically solid handling of failed calculations.

Case studies on radioactivity and decay heat contribution from coating materials of irradiated accident tolerant fuel for a typical PHWR

Seven coating materials such as Al2O3, NbC, Si, SiC, TiN, Y2O3, and ZrC have been considered as a possible coating material towards development of accident tolerant fuel for a typical pressurized heavy water reactor. For an effective severe accident management strategy in addition to spent fuel handling, storage and transport, it is required to know the contribution to the radioactivity burden and decay heat load from the irradiated coating materials of accident tolerant fuel. Under irradiation of coating material in PHWR flux condition, the decay chain of each activated radionuclide has been simulated using ISOGEN code using one group homogenized point cross section. In this paper, radioactivity and decay heat contributions from different irradiated coating materials have been quantified and compared. The minimum contribution to radioactivity and decay heat is observed in SiC after irradiation of 445 days of normal operation. Out of all the coating materials considered in the study, most radioactivity and decay heat were added due to irradiation of coating materials Y2O3. From reactivity load, source term, and decay heat perspectives, silicon-based coating materials are found to be the most suitable.

ELSMOR – towards European Licensing of Small Modular Reactors: Methodology recommendations for light-water small modular reactors safety assessment

Decarbonization of energy production is key in today’s societies and nuclear energy holds an essential place in this prospect. Besides heavy-duty electricity production, other industrial and communal needs could be served by integrating novel nuclear energy production systems, among which are low-power nuclear devices, like small modular reactors (SMRs). The ELSMOR (towards European Licensing of Small Modular Reactors) European project addresses this topic as an answer to the Horizon 2020 Euratom NFRP-2018-3 call. The consortium includes 15 partners from eight European countries, involving research institutes, major European nuclear companies and technical support organizations. The 3.5-year project, launched in September 2019, investigates selected safety features of light-water (LW) SMRs with focus on licensing aspects. Providing a comprehensive compliance framework that regulators can adopt and operate, the licensing process of such SMRs could be optimized, helping their deployment. In this prospect, as a result of ELSMOR’s work, this article gives an overview of the specific issues that LW-SMRs may bring about in the different domains of nuclear safety, in terms of: Methodological standpoints: safety goals, safety requirements, safety principles (defence-in-depth implementation);Main safety functions of reactivity control, decay heat removal and confinement management;Severe accident management;Other safety issues particular to SMRs: use of shared systems; performing of multi-unit probabilistic safety assessment (PSA); refuelling, spent fuel management, transport and disposal management. In this article, adequate methodologies are developed to deal with these issues and to help assess the safety of LW-SMRs. This work gives a precious synthesis of the safety assessment issues of LW-SMRs and of the associated methodologies developed in the context of the ELSMOR European project.

In-depth investigation of natural convection thermal characteristics of BALI experiment through Eulerian computational fluid dynamics code and comparison with Lagrangian code

In-vessel retention through external reactor vessel cooling (IVR-ERVC) is a severe accident management (SAM) strategy that has been adopted and used in many nuclear reactors such as AP1000, APR1400, and light water reactor etc. Some reactor accidents have raised concerns about nuclear reactors among residents, leading to a decrease in residents' acceptability and many studies on SAM are being conducted. Experiments on IVR-ERVC are almost impossible due to its specificity, so fluid characteristics are analyzed through BALI experiments with similar condition. In this study, computational fluid dynamics (CFD) via Reynolds-averaged Navier-Stokes (RANS) and large eddy simulation (LES) for BALI experiments were performed. Steady-state CFD analysis was performed on three turbulence models, and SST k-ω model was in good agreement with the experimental measurement temperature within the maximum error range of 1.9%. LES CFD analysis was performed based on the RANS analysis results and it was confirmed that the temperature and wall heat flux for depth was consistent within an error range of 1.0% with BALI experiment. The LES CFD analysis results were compared with those of the Lagrangian-based solver. LES matched the temperature distribution better than SOPHIA, but SOPHIA calculated the position of boundary between stratified layer and convective layer more accurately. On the other hand, Lagrangian-based solver predicted several small eddy behaviors of the convective layer and LES predicted large vortex behavior. The vibration characteristics near the cooling part of the BALI experimental device were confirmed through Fast Fourier Transform (FFT) investigation. It was found that the power spectral density for pressure at least 10 times higher near the side cooling than near the top cooling.

Critical heat flux model for heated horizontally oriented cylindrical vessel and its application to PHWR calandria under severe accident scenario

In-vessel retention (IVR) of molten corium is a key severe accident management strategy adopted in pressurized heavy water reactors (PHWRs) and is recognised as the only option to manage a complete core melt scenario. Since the PHWR calandria vessel is completely surrounded by a large pool of water, the vessel itself is expected to act as a ‘core catcher’. The success of IVR strategy depends up on whether the actual heat flux imparted by molten corium is less than the critical heat flux (CHF). In the present work, a hydrodynamics model is developed and applied to obtain the variation of CHF along outer surface of 220 and 700 MWe Indian PHWR calandria vessel and tube. The thermal margin available for success of IVR strategy is evaluated. The governing equations for external buoyancy driven boundary layer flow are derived and solved in non-dimensional form. The details of the mathematical model, benchmarking exercises, validation aspects and application of the model to PHWR calandria vessel and tube are discussed. Parametric studies are presented to bring out the influence of diameter, system pressure, subcooling of water, depth of submergence and slip ratio on the variation of CHF along the cylindrical vessel surface.

Numerical simulation technologies for safety evaluation in plant lifecycle optimization method, ARKADIA for advanced reactors

This paper describes the development of Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA) which provides best possible solutions for the design, safety measures, maintenance, and decommissioning of advanced nuclear reactors. ARKADIA consists of state-of-the-art numerical simulation methods, the artificial intelligence technologies, and a knowledge base that includes experiences from the development and operation of the fast reactors and the related research and development. Two subsystems of ARKADIA: ARKADIA-Design and ARKADIA-Safety, are currently being developed individually as the first target of sodium-cooled fast reactors. In a subsequent development phase, the two subsystems will be integrated into a single system, ARKADIA, that is applicable to advanced reactors with a variety of concepts, coolants, configurations, and output levels. ARKADIA-Design offers the functions to support design optimization both in normal operating conditions and design basis events mainly during a conceptual design stage in the fields of core design, plant structure design including thermal-hydraulics analysis, and maintenance plan optimization. The main technology of ARKADIA-Design is based on the multi-level simulation approach by the coupled analysis with numerical analysis codes according to the user’s requirements. For instance, the coupled simulation of neutronics, core deformation, and core thermal hydraulics could evaluate the negative feedback effect of the core deformation reactivity during the plant transient. ARKADIA-Safety is developed for the automatic optimization of severe accident management and their feedback to a plant design. ARKADIA-Safety includes the thermal-hydraulics and multi-physics code, SPECTRA, for in- and ex-vessel integrated severe accident simulation. The basic capability to evaluate severe accident progress was demonstrated through the analysis of a hypothetical loss of reactor level event in a sodium-cooled fast reactor.

Use of supporting software tool for decision-making during low-probability severe accident management at nuclear power plants

In the project NARSIS – New Approach to Reactor Safety ImprovementS – possible advances in safety assessment of nuclear power plants (NPPs) were considered, which also included possible improvements in the field of management of low probability accident scenarios. As a part of it, a supporting software tool for making decisions under severe accident management was developed.
 The mentioned tool, named Severa, is a prototype demonstration-level decision supporting system, aimed for the use by the technical support center (TSC) while managing a severe accident, or for the training purposes. Severa interprets, stores and monitors key physical measurements during accident sequence progression. It assesses the current state of physical barriers: core, reactor coolant system, reactor pressure vessel and containment. The tool gives predictions regarding accident progression in the case that no action is taken by the TSC. It provides a list of possible recovery strategies and courses of action. The applicability and feasibility of possible action courses in the given situation are addressed. For each action course, Severa assesses consequences in terms of probability of the containment failure and estimated time window for failure. At the end, Severa evaluates and ranks the feasible actions, providing recommendations for the TSC. The verification and validation of Severa has been performed in the project and is also described in this paper. Although largely simplified in its current state, Severa successfully demonstrated its potential for supporting accident management and pointed toward the next steps needed with regard to further advancements in this field.

Uncertainty quantification for severe-accident reactor modelling: Set-up and first results of the Horizon-2020 project MUSA

The current Horizon-2020 project on “Management and Uncertainties of Severe Accidents (MUSA)” aims at applying Uncertainty Quantification (UQ) in the modelling of Severe Accidents (SA), particularly in predicting the radiological source term of mitigated and unmitigated accident reactor scenarios. A selected number of severe accident sequences of different nuclear power plant designs (e.g. PWR, VVER, and BWR) are addressed.The application of the Best Estimate Plus Uncertainty (BEPU) methodology to reactor accident scenarios requires a number of key steps: (i) the selection of severe accident sequences for each reactor design; (ii) the development of a reference input model for the specific design and SA-code; (iii) the definition of the figures of merit for the UQ-analysis; (iv) the selection of a list of uncertain model parameters to be investigated; (v) the choice of a statistical tool to propagate input deck uncertainties; (vi) the selection of a feasible approach (i.e., Monte Carlo versus order statistics) to address UQ by using a statistical software (i.e., UQ-tools DAKOTA, SUSA, URANIE, etc.); (vii) the running phase to achieve a high number of successful realizations with the SA codes; and, (viii) the statistical evaluation of the results (i.e., sensitivity analysis).This paper describes each of these steps such as settled in the reactor applications work package of the EU MUSA project and pays particular attention to the choices made by partners. It presents preliminary results also with an emphasis on the major challenges posed by BEPU application in the field of SA analysis.