In the severe accident management (SAM) strategy for Nordic boiling water reactors (BWRs), a flooded reactor cavity is conceived to receive corium in case of vessel failure, with the hope that the discharged corium will fragment and form a coolable particulate debris bed in the deep water pool. The so-formed debris bed on the cavity basement is supposed to be very hot at the beginning and therefore its quenching is a prerequisite for long-term coolability. In previous study the coupled MELCOR/COCOMO simulation was employed to simulate quench process of ex-vessel debris beds in severe accident scenarios. Although it successfully extended the MELCOR capability, the calculation was dramatically slowed down by explosive computational cost of COCOMO. To overcome the limitation, the present study is to develop a surrogate model (SM) which can replace the me-chanical code COCOMO and realize quick estimations of the quench process of ex-vessel debris beds. It was then coupled with MELCOR code for integral severe accident analyses of a Nordic BWR with cooling of ex-vessel debris beds. The SM was developed based on a database generated from COCOMO calculations of various one-dimension (1D) debris beds quenched in the reactor cavity, using artificial neural networks (ANNs). Finally, the coupled MELCOR/SM simulation was applied to safety analyses of postulated severe accident scenarios due to station blackout (SBO) in the BWR, where MELCOR performs integral analysis of accident progression while SM predicts the consequences (e.g. energy transfer) of debris bed quench. The simulation results show that the coupled MELCOR/SM simulation can predict the trends of containment pressure and pool temperature similar to those of the coupled MELCOR/COCOMO simulation. Compared with MELCOR standalone simulation, the coupled MELCOR/SM simulation predicted earlier pool saturation and containment venting.

The present study is intended to investigate the dryout and remelting phenomena of a debris bed during the late phase of an in-vessel severe accident progression. The SIMECO-2 facility at KTH is adapted to conduct the experimental investigation. For selection of an appropriate debris bed in the facility, pre-test simulations are performed by using the COCOMO code to determine: (i) simulant materials of debris particles; (ii) debris bed particle diameters; (iii) configuration and geometry of the debris bed (e.g., shape, layers, dimensions). Candidate particulate beds packed with different mixtures of particles are identified and simulated to obtain their thermal hydraulics in the hemispherical slice test section with radius of 500 mm and width of 120 mm. Based on the simulation results, a particulate bed is chosen and loaded in the SIMECO-2 facility for a scoping investigation. FBG probes with multiple measurement points of each probe are employed to acquire the temperature field of the particulate bed inductively heated. A video recording is applied to detect the dryout and remelting phenomena. In the scoping test, the dryout phenomenon occur first at the elevation of 5 cm from the bed surface under the induction heating power of 14.8 kW, which are comparable with the data predicted by the COCOMO code (6 cm from the bed surface under the heating power of 13.8 kW) in the pre-test simulations.

Different uncertainty analysis methods are applied to MELCOR simulation of two postulated severe accidents in a Nordic boiling water reactor (BWR): (i) station blackout (SBO) accident, and (ii) large break loss-of-coolant accident (LBLOCA) combined with SBO, with the objective to compare their performances in the estimates of 95/95 tolerance limits of two figures of merit (FOMs) - the hydrogen mass produced from core degradation and the timing of vessel failure. Given 17 uncertain input parameters of MELCOR with probability density functions (PDFs), the 95/95 estimates of the two FOMs are obtained through the uncertainty analysis. From the uncertainty analysis results, it is found that for the quantification of single FOM a larger sample size leads to a much more accurate and stable 95/95 estimate at a higher computational cost, and the three nonparametric methods (Wilks' method, Beran and Hall's linear interpolation method as well as Hutson fractional statistics method) behave similarly in both accidents, while the goodness-of-fit test method performs differently and tends to provide a more realistic 95/95 estimate in both accidents. For the quantification of multiple FOMs the bracketing method tends to provide a smaller 95/95 estimate than the Wald and Guba method does, in consistent with their mathematical definitions. The Wald and Guba method is more stringent than the bracketing method when all percentiles (coverage) are set as the same. The sensitivity analysis results show that the several most significant input parameters are ranked almost identically by Spearman rank correlation coefficient (SRCC) and Pearson correlation coefficient (PCC), but these coefficients are dependent on accident scenarios and output parameters. Among the 17 parameters chosen, molten cladding drainage rate is the most influential to the output parameters (timing of initial melt relocation, timing of vessel failure, residual heat, etc.) considered in the present study, probably due to its impacts on molten Zr exposure to steam and resulting oxidation.

The cornerstone of severe accident strategy of Nordic BWRs is to flood the reactor cavity for the long-termcoolability of an ex-vessel debris bed. As a prerequisite of the long-term coolability, the hot debris bedformed from fuel coolant interactions (FCI) should be quenched. In the present study, coupling of theMELCOR and COCOMO codes is realized with the aim to analyze the quench process of an ex-vessel debrisbed under prototypical condition of a Nordic BWR. In this coupled simulation, MELCOR performs an integralanalysis of accident progression, and COCOMO performs the thermal–hydraulic analysis of the debrisbed in the flooded cavity. The effective diameter of the particles is investigated. The discussion on thebed’s shape shows a significant effect on the propagation of the quench front, due to different flow patterns.Compared with MELCOR standalone simulation, the coupled simulation predicts earlier cavity poolsaturation and containment venting.

The CVH package in the MELCOR code is responsible for modelling the thermal-hydraulic behavior dur -ing severe accident. This work presents a sensitivity study of thermal-hydraulic nodalization in the core region for MELCOR simulations of postulated severe accidents in a pressurized water reactor (PWR). Three nodal schemes are developed with a good agreement of steady-state parameters. Two accident sce-narios: loss of coolant accident (LOCA) and station blackout (SBO) are simulated. The analysis is focused on the effect of the control volumes (CVs) in the CVH nodalization on the simulation results of in-vessel accident progression, including the core degradation, hydrogen generation and fission products release, etc. It is found that compared with the coarse CVH nodalization (1 CV), the radial refinement of CVH nodalization (7 CVs) leads to different impacts on accident progression in the two scenarios: faster core relocation and more hydrogen generation is predicted in the LOCA scenario, but it is opposite in the SBO scenario. The finest nodal scheme (49 CVs) with refinement in both radial and axial direction tends to predict an earlier occurrence of cladding rupture, RPV failure and faster core relocation, as well as more hydrogen generation. The CVH refinement makes little difference on the radioactive release.

In a postulated severe accident, the thermo-mechanical loads from the corium debris that has relocated to the lower head of the reactor pressure vessel (RPV) can pose a credible threat to the RPV's structural integrity. In case of a vessel breach, it is vital to predict the mode and timing of the vessel failure. This affects the ex-vessel accident progression and plays a critical role in the development of mitigation strategies. We propose a methodology to assess RPV failure based on MELCOR and ANSYS Mechanical APDL simulations. A Nordic-type boiling water reactor (BWR) is considered with two severe accident scenarios: i) SBO (Station Blackout) and ii) SBO + LOCA (Loss of Coolant Accident). In addition, the approach considers the dynamic ablation of the vessel wall due to a high-temperature debris bed with the use of the element kill function in ANSYS. The results indicate that the stress failure mechanism is the major cause of the RPV failure, compared to the strain failure mechanism. Moreover, the axial normal stress and circumferential normal stress make the dominant contributions to the equivalent stress sigma at the lower head of RPVs. As expected, the region with high ablation is most likely the failure location in both SBO and SBO + LOCA. In addition, comparisons of the failure mode and timing between SBO and SBO + LOCA are described in detail. A short discussion on RPV failure between ANSYS and MELCOR is also presented.

This paper presents a sensitivity study of nodal scheme in MELCOR simulation of severe accidents in a pressurized water reactor, with the objective to estimate the nodal effects on some in-vessel and exvessel processes and phenomena, including thermal-hydraulic response, core degradation and relocation, hydrogen generation, vessel failure, containment pressurization and venting, source term. For this purpose, three nodal schemes (i.e., coarse, medium and fine meshes) of the COR package of the MELCOR code are chosen to analyze two severe accident scenarios: small break loss of coolant accident (SBLOCA) and large break loss-of-coolant accident (LBLOCA), both combined with station blackout. The results show that the nodal schemes mainly affect the calculations of heat transfers from the core to coolant and heat structures, relatively affecting the core degradation and relocation to the lower head of the reactor pressure vessel. As for the consequences, the coarse mesh tends to predict slower core relocation progressions and a later failure of RPV lower head. Moreover, more hydrogen generation by cladding oxidation can be observed in the coarse mesh case. The nodal schemes have little impact on the estimation of in-containment source term. Meanwhile, the simulations with fine mesh may also provide more detailed distributions of corium masses and temperatures, as well as heat fluxes, affecting thermal and mechanic behavior of RPV lower head.

The severe accident management (SAM) strategy for a Nordic boiling water reactor (BWR) employs cavity flooding prior to vessel failure, so that the core melt (corium) discharged from the vessel could fragment and form a particulate debris bed. The key to the success of this SAM strategy is the coolability of ex-vessel debris beds.

The safety analysis involves knowledge about the reactor response to severe accidents under this SAM strategy, which requires the integral simulation of a system code such as MELCOR. Since currently the MELCOR code lacks the modeling of ex-vessel particulate debris beds, the present study aims to develop the capability of MELCOR for the simulation of debris bed coolability through the coupling of MELCOR with other codes, which are dedicated to this phenomenon.

The study is started from the qualification of a MELCOR model for severe accident analysis of a reference Nordic BWR, with the aim to help identify a proper core nodalization. For this purpose, three different core meshes (coarse, medium, and fine) are employed to obtain their impacts on corium release conditions. It is found the coarse mesh is sufficient in the present study, since it is not only computationally efficient, but also predicting earlier vessel failure and faster corium release, providing a more conservative condition for debris bed coolability analysis.

Two couplings are then adopted: (i) coupling of MELCOR with the COCOMO code, which is a mechanistic code for simulation of thermal hydraulics in debris beds; and (ii) coupling of MELCOR with a surrogate model developed in the present study. The first method can simulate the transient behavior of a debris bed during quench process. The second method can efficiently predict the coolability limit (dryout power) required in safety analysis. The surrogate model is developed based on the COCOMO prediction of two-dimensional debris beds.

The developed simulation tools, including the coupled codes and the surrogate model, are applied to the safety analysis of the reference Nordic BWR. The coupled MELCOR/COCOMO simulation is used to investigate the debris bed properties. The effective particle diameter is found as approximately 10% larger than the surface mean diameter of a debris bed with distributed sizes, quantified by the quench rate. For the effect of debris bed shape, it shows a faster quench process with a lower bed slope angle. The quench front propagation as well as the responses of local temperature and containment pressure are obtained.

The coupled MELCOR/surrogate model simulation is performed to estimate the coolability of ex-vessel vessel debris beds. The results show that debris beds are coolable under prototypical conditions with probable bed properties. The surrogate model is used to generate coolability maps, which show the debris bed coolability with the variation of bed properties. The sensitivity analysis indicates that the porosity and the geometry are the most influential to coolability limit. An uncertainty analysis methodology is proposed to obtain the probability of non-coolable debris beds.

Strategin för hantering av svåra haverier (SAM) från ett vattenfyllt nedre primärutrymme för nordiska kokvattenreaktorerna (BWR), så att härdsmältan (corium) som läckt ut från reaktortanken kan fragmentera och bilda en partikelformiggrusbädd. Nyckeln till framgången med denna SAM-strategi är kylbarheten av härdsmältan efter genomsmältning av reaktortanken (ex-vessel). Säkerhetsanalysen fordrar kunskap om reaktorns respons på svåra haverier enligt denna SAM-strategi, detta kräver en integrerad simulering av händelsen med en systemkod, som MELCOR. Eftersom MELCOR för närvarande saknar modellering av grusbäddar efter genomsmältning av reaktortanken, syftar denna studie på att utveckla MELCOR:s förmåga att simulera grusbäddars kylbarhet genom koppling av MELCOR till andra koder som är avsedda för detta fenomen. Studien utgår från utvärdering av en MELCOR-modell för analys av svåra haverier i en representativmodell för en nordisk kokarvattenreaktor. Syftet är att identifiera en korrekt härd nodalisering. Tre olika nodtyper (grovmaskig, medelmaskig och finmaskig) för nodalisering av härden används för att studera deras effekt på simuleringen av härdsmältans utsläpp. Den grovmaskiga nodaliseringen bedömdes lämpligast för den nuvarande studien, eftersom det inte bara är beräkningseffektivt, utan även förutspår tidigare reaktortankbrått och snabbare utsläpp av härdsmältan, vilket ger ett mer konservativt tillstånd för analys av kylbarhet av grusbädden. Två kopplingar antas sedan: i) koppling av MELCOR med den mekaniska koden COCOMO avsedd för simulering av termohydraulik i grusbäddar; och ii) koppling av MELCOR med en surrogatmodell utvecklad under denna studie. Den första metoden kan simulera störningar hos en grusbädd under kylningsprocessen. Den andra metoden kan effektivt förutsäga marginalerna mot torrkokning (dryout power) som krävs i säkerhetsanalysen. Surrogatmodellen är utvecklad baserat på COCOMOsförutsägelsen av tvådimensionella grusbäddar. IV De utvecklade simuleringsverktygen, inklusive de koppladekoderna och surrogatmodellen, tillämpas på säkerhetsanalysen av en referens nordisk kokarvattenreaktor BWR. Den kopplade MELCOR/COCOMO simuleringen används för att undersöka grusbäddens egenskaper. Den effektiva partikeldiametern är cirka 10% större än medeldiameter i för grusbäddsytaspartiklar med distribuerade partikelstorlekar som kvantifierats av snabbt nedkylningshastigheten. Grusbäddens form påverkar effekten av nedkylningsprocessen, en lägre lutningsvinkel ökar nedkylningen. Nedkylningsfrontens utbredning så som den lokala temperatur och inneslutningstrycket erhålls. Den kopplade MELCOR/surrogatmodellsimuleringen utförs för att uppskatta kylbarheten hos grusbäddar efter genomsmältning av reaktortanken. Resultaten visar att grusbäddar kan kylas under prototypiska förhållanden med sannolika grusbäddegenskaper. Surrogatmodellen används för att generera kylbarhetskartor, som beskriver systematiskt grusbäddens kylbarhet beroende på variation i grusbäddegenskaper. Känslighetsanalyser indikerar att porositet och geometrin är mest inflytelserika för marginalerna mot torrkokning. En metod för osäkerhetsanaly

The reactor pressure vessel (RPV) of a nuclear reactor is one of the key safety barriers preventing radioactive environmental releases during a severe accident. One of the promising strategies of severe accident management (SAM) is to retain the molten core having continuous decay heat inside the RPV by natural water cooling of the external vessel surface. The feasibility of such a strategy relies on complex safety analyses including accurateprediction of vessel thermo-mechanical behavior which can be assessed by mechanical stresses and strains. In this paper, we present the stress–strain response of an ablated RPV of a Nordic boiling water reactor (BWR) to dynamic thermomechanical loads set by expanding volumetrically heated molten pool inside the RPV cooled by water at the external surface. MELCOR 2.2.9541 severe accident code is used to simulate the in-vessel behavior and provides the input conditions for dedicated structural analysis of the RPV using ANSYS® Mechanical APDL 19.2. A creep model of the SA533B1 vessel steel is validated against uniaxial creep tests carried out by INEL (Idaho National Engineering Laboratory) and creep tests performed at CEA (French AlternativeEnergies and Atomic Energy Commission) as part of the OLHF (OECD Lower Head Failure) Project. Two generic severe accident scenarios are considered: (i) Station Blackout (SBO) and (ii) Station Black-out and Loss-of-coolant Accident (SBO + LOCA). In both scenarios, we found that the RPV has maintained structural integrity considering two failure criteria: stress-based and strain-based.