This thesis deals with the severe accident of the gas-cooled fast reactor GFR. At the beginning of the study there is a review of the gas-cooled fast reactor subject. Next part is focused on description of possible solutions for severe accidents with emphasis on the solution applied in the Generation III+ reactors. Chapters that deal with material and thermal balance with severe accident of GFR demonstration unit, along with the chapter which analyses features of the corium, create a basis for the conceptual design of core catcher of GFR demonstration unit, which forms the final part of this thesis.

No description available.

No description available.

During the case of a hypothetical severe accident in a light water reactor, core damage may occur and molten fuel may interact with water resulting in explosive interactions. A Fuel-Coolant Interactions (FCI) consists of many complex phenomena whose characteristics determine the energetics of the interactions. The fuel melt initially undergoes fragmentation after contact with the coolant which subsequently increases the melt surface area exposed to coolant and causes rapid heat transfer. A substantial amount of research has been done to understand the phenomenology of FCI, still there are gaps to be filled in terms of the uncertainties in describing the processes such as breakup/fragmentation of melt and droplets. The objective of the present work is to substantiate the understanding in the premixing phase of the FCI process by studying the deformation/pre-fragmentation of melt droplets and also the mechanism of melt jet breakup. The focus of the work is to study the effect of various influential parameters during the premixing phase that determine the intensity of the energetics in terms of steam explosion. The study is based on numerical analysis starting from smaller scale and going to the large scale FCI. Efforts are also taken to evaluate the uncertainties in estimating the steam explosion loads on the reactor scale. The fragmented core is expected to form a porous debris bed. A part of the present work also deals with experimental investigations on the coolability of prototypical debris bed. Initially, the phenomenology of FCI and debris bed coolability is introduced. A review of the state of the art based on previous experimental and theoretical developments is also presented. The study starts with numerical investigation of molten droplet hydrodynamics in a water pool, carried out using the Volume Of Fluid (VOF) method in the CFD code ANSYS FLUENT. This fundamental study is related to single droplets in a preconditioning phase, i.e. deformation/pre-fragmentation prior to steam explosion. The droplet deformation is studied extensively also including the effect of the pressure pulse on its deformation behavior. The effect of material physical properties such as density, surface tension and viscosity are investigated. The work is then extended to 3D analysis as a part of high fidelity simulations, in order to overcome the possible limitations of 2D simulations. The investigation on FCI processes is then continued to the analysis on melt jet fragmentation in a water pool, since this is the crucial phenomenon which creates the melt-coolant pre-mixture, an initial condition for steam explosion. The calculations are carried out assuming non-boiling conditions and the properties of Wood’s metal. The jet fragmentation and breakup pattern are carefully observed at various Weber numbers. Moreover, the effect of physical and material properties such as diameter, velocity, density, surface tension and viscosity on jet breakup length, are investigated. After the fundamental studies, the work was extended to reactor scale FCI energetics. It is mainly oriented on the evaluation of uncertainties in estimating the explosion impact loads on the surrounding structures. The uncertainties include the influential parameters in the FCI process and also the code uncertainties in calculations. The FCI code MC3D is used for the simulations and the PIE (propagation of input errors) method is used for the uncertainty analysis. The last part of the work is about experimental investigations of debris coolability carried out using the POMECO-HT facility at KTH. The focus is on the determination of the effect of the bed’s prototypical characteristics on its coolability, in terms of inhomogeneity with heap like (triangular shape) bed and the radial stratified bed, and also the effect of its multi-dimensionality. For this purpose, four particle beds were constructed: two homogeneous, one with radial stratification and one with triangular shape, respectively. The effectiveness of coolability-enhanced measures such as bottom injection of water and a downcomer (used for natural circulation driven coolability, NCDC) was also investigated. The final chapter includes the summary of the whole work. / Under ett svårt haveri i en kärnkraftsreaktor kan en härdsmälta bildas och smältan växelverka på ett explosivt sätt med kylvattnet. En sådan FCI (Fuel-Coolant-Interaction) inbegriper flera fysikaliska processer vilkas förlopp bestämmer hur stor den frigjorda energin blir. Vid kontakt med vattnet fragmenteras först härdsmältan vilket i sin tur leder till att en större yta exponeras för kylvattnet och att värmeöverföringen från smältan snabbt ökar. Mycket forskning har ägnats åt att förstå vad som sker under en FCI men det finns fortfarande luckor att fylla vad beträffar t ex osäkerheter i beskrivningen av fragmentering av såväl smälta som enskilda droppar av smält material. Syftet med detta arbete är främst att underbygga en bättre förståelse av den inledande delen av en FCI genom att studera dels hur enskilda droppar av smält material deformeras och splittras och dels hur en stråle av smält material fragmenteras. Vi studerar särskilt vilka parametrar som mest påverkar den energi som frigörs vid ångexplosionen. Problemet studeras med numerisk analys med början i liten skala och sedan i full skala. Vi söker också uppskatta de laster som explosionen utsätter reaktorns komponenter för. En annan viktig fråga gäller kylbarheten hos den slaggansamling som bildas under reaktorhärden efter en FCI. Slagghögen förväntas ha en porös struktur och en del av avhandlingen redogör för experimentella försök som genomförts för att utvärdera kylbarheten i olika prototypiska slaggformationer. I avhandlingens inledning beskrivs de fysikaliska processerna under en FCI och kylningen av en slaggansamling. Det aktuella kunskapsläget på dessa områden presenteras också utgående från tidigare experimentella och teoretiska studier. Studierna i avhandlingen inleds med numerisk analys av hydrodynamiken för en enskild droppe smälta i en vattentank där VOF-metoden i CFD-programmet ANSYS FLUENT används. Denna grundläggande studie rör en enskild droppe under förstadiet till fragmentering och ångexplosion då droppen deformeras alltmer. Deformationen studeras ingående också med hänsyn tagen till inverkan av en tryckpuls. Inverkan av olika egenskaper hos materialet, som densitet, ytspänning och viskositet studeras också. Arbetet utvidgas sedan till en beskrivning i 3D för att undvika de begränsningar som finns i en 2D-simulering. Studierna av FCI utvidgas sedan till en analys av fragmentering av en stråle smälta i vatten. Detta är en kritisk del av förloppet då smälta och vatten blandas för att ge utgångstillståndet för ångexplosionen. Beräkningarna genomförs under antagande att kokning inte sker och med materialegenskaper som för Wood´s metall. Mönstret för fragmentering och uppsplittring studeras ingående för olika Weber-tal. Dessutom studeras effekten på strålens uppsplittringslängd av parametrar som diameter och hastighet för strålen samt densitet, ytspänning och viskositet hos materialet. Efter dessa grundläggande studier utvidgas arbetet till FCI-energier i reaktorskala. Här ligger tonvikten på utvärdering av osäkerheter i bestämningen av den inverkan explosionen har på omgivande konstruktioner och komponenter. Osäkerheterna inkluderar eventuell bristande noggrannhet hos såväl de viktiga parametrarna i FCI-processen som i själva beräkningarna. Den sista delen av arbetet handlar om experimentella undersökningar av slaggformationens kylbarhet som genomförts i uppställningen POMECO-HT vid avdelningen för kärnkraftsäkerhet på KTH. Vi vill bestämma effekten av formationens prototypiska egenskaper på kylbarheten. För detta ändamål konstruerades fyra olika formationer: två homogena, en med radiell variation i partikelstorlek och en med triangulär variation. Vi undersökte också hur förbättrad kylning kan uppnås genom att tillföra kylvatten underifrån respektive via ett fallrör (kylning genom naturlig cirkulation). I det avslutande kapitlet ges en sammanfattning av hela arbetet. / <p>QC 20150507</p>

The goal of the present study is to investigate the effect of different scenarios of core degradation in a Nordic-type BWR (boiling water reactor) on the reactor pressure vessel failure mode and timing. Specifically we consider the effects of (i) in-vessel pressure, (ii) melt properties. Control rod guide tube (CRGT) cooling and cooling of the debris from the top are considered as severe accident management (SAM) measures in this study. We also consider the question about minimal amount of debris that can be retained inside the reactor pressure vessel (RPV). Analysis is carried out with coupled (i) Phase-change Effective Convectivity (PECM) model implemented in Fluent for prediction of the debris and melt pool heat transfer, and (ii) structural model of the RPV lower head implemented in ANSYS for simulation of thermo-mechanical creep. The coupling is done through transient thermal load predicted by PECM and applied as a boundary condition in ANSYS analysis. Results of the analysis suggest that applying only CRGT and top cooling is insufficient for maintaining vessel integrity with 0.4 m deep (~12 tons) corium melt pool. The failure of the vessel by thermally induced creep can be expected starting from 5.3 h after the dryout of the debris bed in the lower plenum. However, earlier failure of the instrumentation guide tubes (IGTs) is possible due to melting of the nozzle welding. The internal pressure in the vessel in the range between 3 to 60 bars has no significant influence on the mode and location of the global RPV wall failure. However, depressurization of the vessel can delay RPV wall failure by 46 min for 0.7 m (~ 30 tons) and by 24 min for 1.9 m (~ 200 tons) debris bed. For 0.7 m pool case, changes in vessel pressure from 3 to 60 bars caused changes in liquid melt mass and superheat from ~18 tons at 180 K to ~13 tons at 100 K superheat, respectively. The same changes in pressure for 1.9 m case caused changes in liquid melt mass and superheat from ~40 tons at 42 K to ~10 tons at about 8 K superheat, respectively. Investigation of the influence of melt pool properties on the mode and timing of the vessel failure suggest that the thermo-mechanical creep behavior is most sensitive to the thermal conductivity of solid debris. Both vessel wall and IGT failure timing is strongly dependent on this parameter. For given thermal conductivity of solid debris, an increase in Tsolidus or Tliquidus generally leads to a decrease in liquid melt mass and superheat at the moment of vessel wall failure. Applying models for effective thermal conductivity of porous debris helps to further reduce uncertainty in assessment of the vessel failure and melt ejection mode and timing. Only in an extreme case with Tsolidus, Tliquidus range larger than 600 K, with thermal conductivity of solid 0.5 W∙m‑1∙K‑1 and thermal conductivity of liquid melt 20 W∙m‑1∙K‑1, a noticeable vessel wall ablation and melting of the crust on the wall surface was observed. However, the failure was still caused by creep strain and the location of the failure remained similar to other considered cases. / APRI-8

Syftet med studien är att beskriva forskningen som finns avseende informationsflöde, erfarenheter och upplevelser av anhöriga som grupp när en närstående drabbas av en allvarlig, livsomvälvande eller livshotande olycka eller katastrof. Studien har gjorts genom en scopingreview där fem databaser har genomsökts för att få en överblick av relevant forskning. Resultat påvisar vikten att information kring deras närstående når anhöriga i ett tidigt skede för att minska onödig stress. I arbetet med anhöriga visar det sig att socialarbetare bör ha en plats i krishanteringsarbetet då de ofta har utbildning och erfarenheter av att jobba emotionellt och stärkande. När en allvarlig olycka eller katastrof sker har det visat sig att även anhöriga påverkas. När anhöriga inte har förmågan att på egen hand ta sig igenom krisen bör det finnas krisinterventioner att sätta in. Interventionen bör vara individuellt anpassad och kan därmed behövas långsiktigt. / The purpose of this paper is to examine the needs close family and relatives have of information and support when a serious accident or disaster occurs. This was done through a scoping review. Five databases were searched to summarize relevant research. The results indicate how important it is with information to close family and relatives in an early stage to minimize their negative stress. Social workers should have a given place with the trauma team since they usually have the training and skills to work emotionally and supportive with people. When a severe accident or disaster occurs, it is proven that close family and relatives are also affected. A crisis intervention should be available when they don’t have the resources to cope with the crisis on their own. Crisis interventions should be individually adapted and might be needed fora long time.

Considering the late in-vessel phase of an unlikely core melt down scenario in a light water reactor (LWR) with the formation of a corium pool in the lower head of the re-actor pressure vessel (RPV) the possible failure modes of the RPV and the time to failure have to be investigated to assess the possible loadings on the containment. In this work, an integral model was developed to describe the processes in the lower plenum of the RPV. Two principal model domains have to be distinguished: The temperature field within the melt and RPV is calculated with a thermodynamic model, while a mechanical model is used for the structural analysis of the vessel wall. In the introducing chapters a description is given of the considered accident scenario and the relevant analytical, experimental, and numerical investigations are discussed which were performed worldwide during the last three decades. Following, the occur-ring physical phenomena are analysed and the scaling differences are evaluated between the FOREVER-experiments and a prototypical scenario. The thermodynamic and the mechanical model can be coupled recursively to take into account the mutual influence. This approach not only allows to consider the tem-perature dependence of the material parameters and the thermally induced stress in the mechanical model, it also takes into account the response of the temperature field itself upon the changing vessel geometry. New approaches are applied in this work for the simulation of creep and damage. Using a creep data base, the application of single creep laws could be avoided which is especially advantageous if large temperature, stress, and strain ranges have to be covered. Based on experimental investigations, the creep data base has been de-veloped for an RPV-steel and has been validated against creep tests with different scalings and geometries. It can be stated, that the coupled model is able to exactly describe and predict the vessel deformation in the scaled integral FOREVER-tests. There are uncertainties concerning the time to failure which are related to inexactly known material parame-ters and boundary conditions. The main results of this work can be summarised as follows: Due to the thermody-namic behaviour of the large melt pool with internal heat sources, the upper third of the lower RPV head is exposed to the highest thermo-mechanical loads. This region is called hot focus. Contrary to that, the pole part of the lower head has a higher strength and therefore relocates almost vertically downwards under the combined thermal, weight and internal pressure load of the RPV. On the one hand, it will be possible by external flooding to retain the corium within the RPV even at increased pressures and even in reactors with high power (as e.g. KONVOI). On the other hand, there is no chance for melt retention in the considered scenario if neither internal nor external flooding of the RPV can be achieved. Two patents have been derived from the gained insights. Both are related to pas-sively working devices for accident mitigation: The first one is a support of the RPV lower head pole part. It reduces the maximum mechanical load in the highly stressed area of the hot focus. In this way, it can prevent failure or at least extend the time to failure of the vessel. The second device implements a passive accident mitigation measure by making use of the downward movement of the lower head. Through this, a valve or a flap can be opened to flood the reactor pit with water from a storage res-ervoir located at a higher position in the reactor building. With regard to future plant designs it can be stated - differing from former presump-tions - that an In-Vessel-Retention (IVR) of a molten core is possible within the reac-tor pressure vessel even for reactors with higher power.

Considering the late in-vessel phase of an unlikely core melt down scenario in a light water reactor (LWR) with the formation of a corium pool in the lower head of the re-actor pressure vessel (RPV) the possible failure modes of the RPV and the time to failure have to be investigated to assess the possible loadings on the containment. In this work, an integral model was developed to describe the processes in the lower plenum of the RPV. Two principal model domains have to be distinguished: The temperature field within the melt and RPV is calculated with a thermodynamic model, while a mechanical model is used for the structural analysis of the vessel wall. In the introducing chapters a description is given of the considered accident scenario and the relevant analytical, experimental, and numerical investigations are discussed which were performed worldwide during the last three decades. Following, the occur-ring physical phenomena are analysed and the scaling differences are evaluated between the FOREVER-experiments and a prototypical scenario. The thermodynamic and the mechanical model can be coupled recursively to take into account the mutual influence. This approach not only allows to consider the tem-perature dependence of the material parameters and the thermally induced stress in the mechanical model, it also takes into account the response of the temperature field itself upon the changing vessel geometry. New approaches are applied in this work for the simulation of creep and damage. Using a creep data base, the application of single creep laws could be avoided which is especially advantageous if large temperature, stress, and strain ranges have to be covered. Based on experimental investigations, the creep data base has been de-veloped for an RPV-steel and has been validated against creep tests with different scalings and geometries. It can be stated, that the coupled model is able to exactly describe and predict the vessel deformation in the scaled integral FOREVER-tests. There are uncertainties concerning the time to failure which are related to inexactly known material parame-ters and boundary conditions. The main results of this work can be summarised as follows: Due to the thermody-namic behaviour of the large melt pool with internal heat sources, the upper third of the lower RPV head is exposed to the highest thermo-mechanical loads. This region is called hot focus. Contrary to that, the pole part of the lower head has a higher strength and therefore relocates almost vertically downwards under the combined thermal, weight and internal pressure load of the RPV. On the one hand, it will be possible by external flooding to retain the corium within the RPV even at increased pressures and even in reactors with high power (as e.g. KONVOI). On the other hand, there is no chance for melt retention in the considered scenario if neither internal nor external flooding of the RPV can be achieved. Two patents have been derived from the gained insights. Both are related to pas-sively working devices for accident mitigation: The first one is a support of the RPV lower head pole part. It reduces the maximum mechanical load in the highly stressed area of the hot focus. In this way, it can prevent failure or at least extend the time to failure of the vessel. The second device implements a passive accident mitigation measure by making use of the downward movement of the lower head. Through this, a valve or a flap can be opened to flood the reactor pit with water from a storage res-ervoir located at a higher position in the reactor building. With regard to future plant designs it can be stated - differing from former presump-tions - that an In-Vessel-Retention (IVR) of a molten core is possible within the reac-tor pressure vessel even for reactors with higher power.

La thèse s'inscrit dans le contexte des accidents graves dans les réacteurs nucléaires qui sont étudiés au laboratoire de physique et modélisation des accidents graves (LPMA) du CEA de Cadarache. Un accident grave survient lors de la perte du caloporteur au niveau du circuit primaire ce qui provoque une dégradation du combustible et la création d'un bain de corium. Celui-ci va ensuite se propager en cuve et fortement endommager les structures du réacteur. Pour la sûreté nucléaire, il est donc nécessaire de pouvoir prévoir la propagation de ce corium, d'où la création en 2013 de la plateforme PROCOR (Java) permettant aux travers d'applications industrielles de simuler cette propagation. Ces applications sont un ensemble de modèles physiques, couplés sur une macro boucle en temps, ayant chacun un ensemble d'équations algébriques et différentielles qui sont résolues en interne des modèles. Les modèles de la plateforme sont généralement des modèles OD dont la discrétisation spatiale est remplacée par des corrélations généralement issues de l'expérience. Chaque modèle a aussi un ensemble d'états et de règles de transition, et un changement d'état peut alors survenir à l'intérieur de la macro boucle en temps. Au début de la thèse, le couplage était simplement un chaînage des modèles sur la macro boucle en temps : chaque modèle est résolu l'un après l'autre, l'ordre étant défini par le créateur de l'application, et les modèles sont synchronisés à la fin de cette boucle. Les résultats des applications industrielles de la plateforme en modifiant simplement le pas de temps de la macro boucle en temps montrent une forte dépendance du schéma avec ce pas de temps. On a par exemple 10 % d'écart sur les flux imposés sur la cuve du réacteur en passant d'un pas de temps de 100 s à 50 s, ce qui a un fort impact sur les résultats de sûreté nucléaire. / This thesis focuses on solving coupled problems of models of interest for the simulation of severe accidents in nuclear reactors~: these coarse-grained models allow for fast calculations for statistical analysis used for risk assessment and solutions of large problems when considering the whole severe accident scenario. However, this modeling approach has several numerical flaws. Besides, in this industrial context, computational efficiency is of great importance leading to various numerical constraints. The objective of this research is to analyze the applicability of explicit coupling strategies to solve such coupled problems and to design implicit coupling schemes allowing stable and accurate computations. The proposed schemes are theoretically analyzed and tested within CEA's procor{} platform on a problem of heat conduction solved with coupled lumped parameter models and coupled 1D models. Numerical results are discussed and allow us to emphasize the benefits of using the designed coupling schemes instead of the usual explicit coupling schemes.

In Nordic design of boiling water reactors, a deep water pool under the reactor vessel is employed for the core melt fragmentation and the long term cooling of decay heated corium debris in case of a severe accident. To assess the effectiveness of such accident management strategy the Risk-Oriented Accident Analysis Methodology has been proposed. The present work contributes to the further development of the methodology and is focused on the issue of ex-vessel debris coolability. The height and shape of the porous debris bed are among the most important factors that determine if the debris can be cooled by natural circulation of water. The bed geometry is formed in the process of melt release, fragmentation, sedimentation and packing of the debris in the pool. Bed shape is affected by the coolant flow that induces movement of particles in the pool and after settling on top of the bed. The later one is called debris bed self-leveling phenomenon. In this study, the self-leveling was investigated experimentally and analytically. Experiments were carried out in order to collect data necessary for the development of a numerical model with an empirical closure. The self-leveling model was coupled to a model for prediction of the debris bed dryout. Such coupled code allows to calculate the time necessary to have a coolable configuration of the bed. The influence of input parameters was assessed through sensitivity analysis in order to screen out the less influential parameters. Results of the risk analysis are reported as complementary cumulative distribution functions of the conditional containment failure probability (CCFP). Sensitivity analyses identified: effective particle diameter and debris bed porosity as the parameters that provide the largest contribution to the CCFP uncertainty. It is found that the effect of the initial maximum height of the bed on the CCFP is reduced by the self-leveling. / Kokvattenreaktorer av nordisk typ har en djup vattenbassäng under reaktorkärlet som kan utnyttjas för att kyla härdsmältan och de fragmenterade härdresterna vid ett svårt reaktorhaveri. För att bedöma effektiviteten av en sådan haverihantering har man föreslagit användande av en riskorienterad metodik för haverianalysen (ROAAM, från engelska ”Risk-Oriented Accident Analysis Methodology”). Föreliggande projekt fokuserar på kylbarhet hos härdresterna utanför reaktortanken och bidrar till den pågående vidareutvecklingen av ROAAM till ROAAM+. Höjden på och formen för den porösa ansamlingen av härdrester (här också kallad partikelbädd) är bland de viktigaste faktorerna som avgör om resteffekten kan kylas bort med hjälp av naturlig cirkulation av vattnet i bassängen. Ansamlingens geometriska form skapas under hela processen från utsläpp av  härdsmältan via fragmentering och sedimentering i bassängens botten. Formen kan sedan förändras med tiden genom att partiklar rör sig och omfördelas i kylflödet. Detta fenomen kallas en självnivellerande process. I detta arbete studeras denna självnivellerande process experimentellt och analytiskt. Experimenten utfördes i en särskild experimentuppställning utformad för att att samla in data och parametrar som behövs för att simulera fenomenet och utveckla en beräkningsmodell som sluts empiriskt. Denna modell kopplades sedan till en modell för beräkning av dryout i partikelbädden. Genom denna koppling av de två beräkningsprogrammen är det är möjligt att beräkna tiden för partikelbädden att nå en kylbar konfiguration. Inverkan av variationer i modellens indata studeras med hjälp av känslighetsanalys. Härigenom identifierades de minst inflytelserika parametrarna såsom effektiv drifttid, partikeldensitet, experimentell ovisshet i de empiriska samband som används för att sluta modellen, samt omlokaliseringstid efter det att reaktorn snabbstoppats (SCRAM).  Dessa parametrar avfördes sedan från den fortsatta känslighetsanalysen. Ett artificiellt neuralt nätverk tränades för att användas i stället för den kopplade koden och möjliggöra den beräkningseffektivitet som krävs för att studera hur osäkerheter i indata förs vidare i riskanalysen. Resultaten är presenterade i form av komplementära, kumulativa fördelningsfunktioner för den betingade sannolikheten för brott på reaktorinneslutningen (CCFP, från engelska ”conditional containment failure probability”). Det visas att CCFP kan variera inom ett brett område beroende på de valda kombinationerna av frekvensfunktioner för ingångsparametrarna. Resultaten visar att effektiv partikeldiameter och hög porositet är de två parametrar som ger de största bidragen till osäkerheten i CCFP. Vi har också funnit att fenomenet självnivellering har en gynnsam inverkan på CCFP och leder till lägre utsläppsrisk. Det vore värdefullt att förfina de modeller som beskriver bildandet av den initiala partikelbädden. Detta är särskilt viktigt i de scenarier där det finns kort tid för självnivellering innan partikelbädden börjar smälta igen, dvs när man har relativt hög initial temperatur i partikelbädden och/eller hög specifik värmeeffekt. / <p>QC 20170315</p> / APRI