Lower Plenum Research Articles

Investigation of computational model for the natural circulation at dual channel facility

The current work investigates a computational model to study the thermal and hydraulic air behavior during the natural circulation at air ingression and accidents. This is done with the RHYS coupling ASYST VER 4 package. The test facility considered for the present study is a dual vertical channel facility comprised of two parallel channels connected to the upper and lower plenum. The flow fields in the heated and cooled channels were comprehensively characterized by analyzing axial temperature and velocity distributions using varied uniform iso-flux (100–1400 W/m2) and different outer surface temperatures (278, 288, 298, and 308 K). Temperature and velocity reversal recorded after maximal spots due to natural convection. The temperature rise from 278 to 308 K gave an average of 25.51 and 25.19° increase in air and inner wall temperatures, respectively, while air velocity increases at high cooling intensity (278 K) within the heated channel, in the cooled channel, low cooling intensity (308 K) resulted in higher velocity. The convective heat transfer is represented in terms of heat transfer coefficients, which are used to compute the Nusselt number. Additionally, the ASYST model was validated with data from literature sources, indicating strong agreement.

Simulation of late phase phenomena with the system code AC2

Simulation codes like the AC2 system code package developed by German Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) gGmbH are used in international reactor safety research to investigate hypothetical severe incidents and accidents in nuclear power plants. Valuable impetuses for the development of prevention and mitigation measures for such events can be derived from the analyses. This article presents selected work by PSS on the development of simulation models for the late phase of nuclear accidents. Topical issues as the formation and thermodynamic interactions of debris beds in the lower plenum of the reactor pressure vessel (RPV), the release of fission products as well as the interaction of the molten core material with concrete in the reactor cavity in case of RPV failure are discussed. The analyses carried out provide important findings regarding the validation of the developed models. In addition, future development needs are derived for the consideration of further relevant phenomena.

Quantitative evaluation of uncertain parameters for thermal-hydraulic experiments based on the COSINE code

Identifying important phenomena and parameters of LBLOCAs is an important step in nuclear power safety evaluation. Traditional identification processes are based on experience and incorporate important phenomena and parameters without quantifying specific aspects of the models. To accurately identify physical effects, this paper presents the use of multiphase field subchannel code for simulation analysis as applied to numerical examples and specific reactor thermal-hydraulic problems. The results indicate that the code can simulate the relevant phenomena, and the calculation band of the refill stage is narrow, while the calculation band of the reflood stage is wide. In addition, the key impact models are captured. The radiation model most significantly impacts the cladding temperature in the core, and the heat transfer model of transition boiling to the gas phase is most strongly influenced by the quenching time. The interfacial heat transfer model of the large bubble regime for the liquid plays a major role in influencing the water inventory of the lower plenum in the downcomer.

A comprehensive analysis of thermal–hydraulic signatures in neutron noise of WWER-type reactors

The neutron noise phenomenon, occurring in all types of nuclear reactors, provides valuable insights into the dynamic behavior of these reactors. In this work, a well-justified approach for modeling neutron noise induced by thermal–hydraulic sources in a hexagonal reactor core is presented in detail. These sources encompass fluctuations in key features of the inlet coolant, including flow rate, temperature, and boric acid concentration. Our proposed approach touches on various aspects of the distribution of thermal–hydraulic parameters among coolant loops, accounting for both uniform and non-uniform distributions. The latter allows for the asymmetric and asynchronous distribution of their fluctuating component among the coolant loops. To manage such complexity, it becomes imperative to determine the contribution of each fuel assembly based on each coolant loop. To achieve this, CFD simulations are conducted for the downcomer and lower plenum of the reactor pressure vessel. These simulations provide essential data to establish the coolant mixing matrix, which accounts for non-uniform distributions. The obtained mixing matrix is subsequently integrated into our neutron noise simulator, DYNOSIM, ensuring the faithful reproduction of neutron noise. Our study comprehensively analyzes both uniform and non-uniform scenarios. The radial and axial distributions of neutron noise amplitude and phase are examined for these scenarios. Their qualitative verifications are conducted by prior works in this field. Furthermore, we compare one of the uniform scenarios with reference results obtained using a frequency domain simulator. Notably, our CFD simulation results closely match the experimental data. Our findings demonstrate the effectiveness and efficiency of this methodology in modeling and understanding neutron noise in hexagonal reactor cores.

MAAP code analysis for the in-vessel phase of Fukushima-Daiichi Nuclear Power Station Unit 1 and comparison of the results among Units 1 to 3

The accident progression of the in-vessel phase of Fukushima Daiichi Nuclear Power Station Unit 1 was analyzed using the MAAP code. Although there is a large uncertainty in the initial stage of accident progression behavior in Unit 1 with little measurement data, it is presumed to have similarities to that of Unit 3. As a result, in Unit 1, since there was almost no alternative water injection during the in-vessel phase, cooling of the debris transferred to the lower plenum was small. It was likely that a large molten pool of metals had formed, and that the steam supply to the high-temperature core materials was suppressed and metal oxidation was relatively small. The analysis results for Unit 1 were compared with those for Units 2 and 3, and differences between units such as the thermal conditions of the debris that relocated to the pedestal and the degree of metal oxidation were shown.

Experimental study of turbulent flow in lower plenum with flow skirt of reactor pressure vessel of pressurized water reactor

The turbulent flow in lower plenum of reactor pressure vessel (RPV) of pressurized water reactor (PWR) is an important phenomenon for design of flow mixing device. A high-precision matched index of refraction (MIR) technique is requisite for time-resolved particle image velocimetry (TR-PIV) measurements in RPV with complicated internals. In this study, turbulent flow in a scaled-down RPV model was measured by TR-PIV under different Reynolds numbers (Re) and asymmetrical flow rate with aid of the high-precision MIR technique of sodium iodide (NaI) solution and acrylic. The time-averaged flow structures include flow separation at the corner of lower support plate, fluid bifurcation around the lower edge of flow skirt, flow turning upward below the vortex suppression plate, horizontal jets at holes of flow skirt, vertical jets at holes of vortex suppression plate, improving turbulent mixing and Reynolds stresses in the lower plenum. The coherent structures were studied by proper orthogonal decomposition (POD) analysis. The flow skirt and vortex suppression plate break down large vortices through them. With Re increasing, Reynolds stresses decrease fast, and the energy fractions of low-order modes transfer into high-order modes, because turbulent vortices break up. In the case of asymmetrical flow rate, the time-averaged velocity and Reynolds stress decrease slightly, and the cumulative energy fraction decreases at flow skirt and vortex suppression plate.

Optimal controller design for reactor core power stabilization in a pressurized water reactor: Applications of gold rush algorithm.

Nuclear energy (NE) is seen as a reliable choice for ensuring the security of the world's energy supply, and it has only lately begun to be advocated as a strategy for reducing climate change in order to meet low-carbon energy transition goals. To achieve flexible operation across a wide operating range when it participates in peak regulation in the power systems, the pressurised water reactor (PWR) NE systems must overcome the nonlinearity problem induced by the substantial variation. In light of this viewpoint, the objective of this work is to evaluate the reactor core (main component) of the NE system via different recent optimization techniques. The PWR, which is the most common form, is the reactor under investigation. For controlling the movement of control rods that correspond with reactivity for power regulation the PWR, PID controller is employed. This study presents a dynamic model of the PWR, which includes the reactor core, the upper and lower plenums, and the piping that connects the reactor core to the steam alternator is analyzed and investigated. The PWR dynamic model is controlled by a PID controller optimized by the gold rush optimizer (GRO) built on the integration of the time-weighted square error performance indicator. Additionally, to exhibit the efficacy of the presented GRO, the dragonfly approach, Arithmetic algorithm, and planet optimization algorithm are used to adjust the PID controller parameters. Furthermore, a comparison among the optimized PID gains with the applied algorithms shows great accuracy, efficacy, and effectiveness of the proposed GRO. MATLAB\ Simulink program is used to model and simulate the system components and the applied algorithms. The simulation findings demonstrate that the suggested optimized PID control strategy has superior efficiency and resilience in terms of less overshoot and settling time.

Nuclear Meltdown Relocation and Core Catcher Analysis

Nuclear meltdown with the potential human and environmental harm is one of the major accident hazard (MAH) faced by nuclear power plants. Limiting (or entirely avoiding) criticality events are the main design strategies for reactors of generations 3½ and 4 (Gen3½ and Gen4). These include ensuring negative void and negative temperature coefficients (for both moderator and fuel) regardless of operational conditions, which provide a self-regulating mechanism that helps preventing accidents occurrence (i.e., to address safety and reliability aspects of Gen4’s goals). However, in severe accident scenarios (e.g. during loss-of-coolant, LOCA, events) where failure to extract heat from the reactor may lead to core degradation, strategies to mitigate reactor meltdown and relocation are critical in the design of safety protocols. This work aims to numerically investigate core relocation as an integrated multi-fluid and heat dynamics problem in which flow of melted materials (UO2, Zircaloy and graphite) are modelled through interface capturing/tracking methods. Two interface tracking/capturing methods were compared, the level-set volume of fluid method (VOF) in Ansys Fluent, and the compressive advection method (CAM) in Fluidity/ICFERST. Both methods are in good agreement for the core relocation simulation. An in-vessel core catcher (IVCC) of tungsten alloy was also proposed to demonstrate core degradation control strategy through cooling of the melted multi-materials. The IVCC was simulated with a multifluid model in Ansys Fluent, in a specified applied heat flux model. The thickness of the IVCC is 0.20 m and the heat flux used is 600 kW m-2. The tungsten material used was able to withstand both thermal and mechanical loads on the lower plenum by extracting decay heat from the corium.

Experimental study on the influence of velocity on hydraulic characteristics in reactor lower plenum

Studies on the hydraulic models of reactors are generally based on geometric similarity and Euler number similarity. However, there is no unified criterion for selecting the modeling velocity. Furthermore, discrepancies remain in the conclusions regarding the influence of velocity on hydraulic characteristics of reactors. In this study, experiments were conducted on a flat-plate mock-up, which is 1:1 scaled and simplified from a pressurized water reactor. The effect of velocity on the characteristics of pressure drop, core inlet flow distribution, mixing and flow fields in the lower plenum was investigated in the velocity range of 0.2 Vp (velocity in prototype) to 1.0 Vp. The comparative results indicated that when the velocity increased from 0.2 Vp to 1.0 Vp, the Euler number decreased by more than 20 %, the flow distribution at the core inlet and mixing characteristics in the lower plenum were similar; the dimensionless flow fields in the lower plenum were also similar.

Integral simulation of the first 3 weeks of severe accident at Fukushima Daiichi Unit 1 with SOCRAT code. Part I – Qualification of model and simulation of the initial phase

This paper is the first part of the summary overview which presents the results of IBRAE's work over the last 11 years to study the severe accident at Unit 1 of Fukushima Daiichi Nuclear Power Station. The paper is dedicated to the description of the integral model of the Unit 1 with assumptions made, and to its qualification with the available data. The qualification includes the comparison of simulation results with measurements for the initial phase of the accident which starts with the initiating event (earthquake) and ends with tsunami arrival. The important issues discussed are the direct modeling of isolation condenser performance and verification of the possibility to use an integral approach to estimate the nuclide inventory in the core. These questions are of methodological importance for modeling the main phase of the accident.Direct modeling of isolation condenser as part of integral calculation allows verifying the hypothesis about the isolating valves left open inside containment after tsunami arrival and station blackout. The isolation condenser system could influence the development of severe accident by limiting the pressure peak in the reactor at the time of melt relocation into the lower plenum. Thus, verification of the isolation condenser model at the initial stage of the accident justifies the correctness of results obtained for the core degradation phase of the accident.The integral approach in nuclide inventory estimate presumes the need to mix the fuel with different burn-ups in the effective groups of fuel assemblies, which may bring an uncertainty in the nuclide inventory prediction. Quantification of this uncertainty is important for assessment of prediction accuracy for the source term.

Numerical study of debris bed formation behavior for severe accident in sodium-cooled fast reactor by using least square MPS-DEM method

During a Core Disruptive Accident (CDA) in a Sodium-cooled Fast Reactor (SFR), debris beds may form in the lower plenum of the reactor vessel as a result of the accumulation of discharged debris, which is formed from the solidification and fragmentation of core molten materials during their release due to interactions with the sodium coolant. Since the configuration of debris beds plays a vital role in their long-term coolability and neutronic subcriticality for achieving in-vessel retention, it is essential to understand and assess the Debris Bed Formation Behavior (DBFB). In this study, an advanced Moving Particle Semi-implicit (MPS) method, called least square MPS (LSMPS), is utilized and coupled with the Discrete Element Method (DEM) for the numerical investigations of DBFB characteristics in a two-dimensional liquid pool. The applicability and reliability of the proposed LSMPS-DEM method are validated by comparing simulation results with experiments using simulant materials. The LSMPS-DEM method is then employed to study DBFB in sodium. It is observed that debris properties significantly impact the sedimentation and accumulation processes of DBFB, leading to variations in the characteristics of the formed debris beds. The present study is expected to be useful for further investigations of DBFB under more realistic conditions and for the design improvement of safety devices to mitigate the consequences of CDA in SFR.

MAAP code analysis focusing on the fuel debris conditions in the lower head of the pressure vessel in Fukushima-Daiichi Nuclear Power Station Unit 3

Based on the updated knowledge from plant-internal investigations, experiments and computer-model simulations until now, the in-vessel phase of Fukushima-Daiichi Nuclear Power Station Unit 3 was analyzed using the MAAP code. In Unit 3, it is considered that ca. 40% of UO2 fuel was molten when core materials relocated to the lower plenum of the reactor pressure vessel. Initially relocated molten materials would have been fragmented by mixing with liquid water, while solid materials would have relocated later on. With this two-step relocation, debris in the lower plenum seems to have been permeable for coolant, thus debris seems to have been once cooled down effectively. Although the present MAAP analysis seems to slightly underestimate core-material oxidation during the relocation period, this probable underestimation was compensated for by an existing study that was considered more reliable, so that more realistic debris conditions in the lower plenum could be obtained. Probable debris reheat-up behavior was evaluated based on interpretation of the pressure data. This evaluation predicted that the fuel debris in the lower plenum was basically in solid-phase at the time when it relocated to the pedestal. With this study, basic validity of the former prediction of the Unit 3 accident progression behavior was confirmed, and detailed boundary conditions for future studies addressing the later phases were provided.

System Thermal-Hydraulics Model for Fluoride Salt-Cooled Reactor Based on Small Modular Advanced High Temperature Reactor (SmAHTR) Design Concept

Abstract A system thermal-hydraulics model for a fluoride-salt-cooled high-temperature reactor (FHR) based on the small modular advanced high-temperature reactor (SmAHTR) design concept is developed, using RELAP5-3D. The SmAHTR components modeled in the simulations include the reactor core, lower plenum, upper plenum, top plenum, three primary heat exchangers (PHXs) equipped with three primary pumps, and three director reactor auxiliary cooling system (DRACS) equipped with three fluid diodes. Flows through the reactor core are represented by 19 individual fuel channels, one reflector-hole channel, and one downcomer channel. In each of the 19 SmAHTR fuel block channels, the fuel elements are modeled in five groups using five heat structures, each with their corresponding power level. The total reactor power is 125 MWth. Using representative core power distributions for the SmAHTR at beginning-of-cycle (BOC) and at end-of-cycle (EOC), two steady-state system thermal-hydraulic model simulations with RELAP5-3D were performed using a nominal pressure drop loss factor of 1.5 for all nine junctions in each of the 19 fuel channels modeled. Exit coolant temperatures ranged from 688 °C to 739 °C (BOC) and from 696 °C to 721 °C (EOC), while peak fuel centerline temperatures in the highest power block were 1249 °C (BOC) and 1029 °C (EOC). By adjusting the loss factors to modify coolant flow rates in each channel, a more uniform exit coolant temperature was possible, ranging from 701 °C to 709 °C (BOC) and 698 °C to 714 °C (EOC), while peak fuel centerline temperatures were reduced to 1204 °C (BOC) and 988 °C (EOC), giving results more consistent with earlier studies of the SmAHTR.

Analysis of High-Pressure extended SBO accident in VVER-1000: A sensitivity study on lower plenum structures modeling and core region nodalization in MELCOR

The lower plenum structures of VVER-1000 reactors exhibit a unique shape. This paper aims to investigate the sensitivity of various important parameters related to safety using different models in the MELCOR 1.8.6 code's COR package. The study focuses on modeling the lower plenum structures and the core region using different CVH package nodalizations during a High Pressure Extended Station Black-Out accident. At first, MELCOR calculated parameters are investigated during the course of a severe accident for a basic model of the lower plenum structures and core region nodalization. Different important parameters from a safety point of view are also examined. These include the total amount of in-vessel hydrogen production, the lower head breach time and elevation, the synchronicity between lower head breach and debris ejection, the total amount of hydrogen produced during molten corium concrete interaction, the total amount of ejected debris to the cavity, and the total amount of radioactive material release. Then, by considering sixteen different simulations corresponding to four lower plenum structures modeling and four core region nodalizations, the sensitivity of the mentioned parameters is investigated. The obtained results show a sensitivity of about 12.7% for the timing of the accident (equivalent to), 103% for the hydrogen produced inside the vessel (equivalent to 417.9 to 847.8 kg), 143% for the hydrogen produced outside the vessel (equivalent to 445.7 to 1081.4 kg), 47% for the total hydrogen produced (equivalent to 1207.5 to 1780.7 kg), and 76% for total radioactive material release (equivalent to 625.311 to 1101.892 kg) up to 24 h after the accident.

An assessment of scaling methods for ECC behavior in downcomer of UPTF with RELAP5 code

Various experimental investigations in scaled-down test facilities have been conducted to observe the ECC behaviors during a Large-Break Loss of Coolant Accident (LBLOCA). The ability of the scaling method to describe the similarity between the prototype and the scaled models required validation. In this paper, a series of simulations were performed to investigate the characteristics of ECC penetration in UPTF and the models scaled by four scaling methods including three-level scaling method, linear method, modified linear method and Hierarchical Two-tiered Scaling (H2TS) method. The ECC behaviors, such as the inventory and pressure in lower plenum, steam mass flow at the break, have been analyzed. And the applicability of the scaling method has been evaluated based on the analysis of the ECC behaviors in both of the prototype and scaled models. The results showed that the transient process in the model scaled by H2TS method with aspect ratio of 1.87 and modified linear was in good agreement with the prototype during ECC injection. Moreover, in order to investigate the effect of the aspect ratio, different aspect ratio in H2TS method was applied to scale the model and the inventory in the lower plenum between the UPTF and the model has been compared. In addition, the modified Wallis correlation for CCFL accounting for the condensation effect has been performed.

Investigation of thermal hydraulic behavior of the High Temperature Test Facility's lower plenum via large eddy simulation

A high-fidelity computational fluid dynamics (CFD) analysis was performed using the Large Eddy Simulation (LES) model for the lower plenum of the High–Temperature Test Facility (HTTF), a ¼ scale test facility of the modular high temperature gas-cooled reactor (MHTGR) managed by Oregon State University. In most next–generation nuclear reactors, thermal stress due to thermal striping is one of the risks to be curiously considered. This is also true for HTGRs, especially since the exhaust helium gas temperature is high. In order to evaluate these risks and performance, organizations in the United States led by the OECD NEA are conducting a thermal hydraulic code benchmark for HTGR, and the test facility used for this benchmark is HTTF. HTTF can perform experiments in both normal and accident situations and provide high-quality experimental data. However, it is difficult to provide sufficient data for benchmarking through experiments, and there is a problem with the reliability of CFD analysis results based on Reynolds–averaged Navier–Stokes to analyze thermal hydraulic behavior without verification. To solve this problem, high-fidelity 3-D CFD analysis was performed using the LES model for HTTF. It was also verified that the LES model can properly simulate this jet mixing phenomenon via a unit cell test that provides experimental information. As a result of CFD analysis, the lower the dependency of the sub-grid scale model, the closer to the actual analysis result. In the case of unit cell test CFD analysis and HTTF CFD analysis, the volume-averaged sub-grid scale model dependency was calculated to be 13.0% and 9.16%, respectively. As a result of HTTF analysis, quantitative data of the fluid inside the HTTF lower plenum was provided in this paper. As a result of qualitative analysis, the temperature was highest at the center of the lower plenum, while the temperature fluctuation was highest near the edge of the lower plenum wall. The power spectral density of temperature was analyzed via fast Fourier transform (FFT) for specific points on the center and side of the lower plenum. FFT results did not reveal specific frequency-dominant temperature fluctuations in the center part. It was confirmed that the temperature power spectral density (PSD) at the top increased from the center to the wake. The vortex was visualized using the well-known scalar Q-criterion, and as a result, the closer to the outlet duct, the greater the influence of the mainstream, so that the inflow jet vortex was dissipated and mixed at the top of the lower plenum. Additionally, FFT analysis was performed on the support structure near the corner of the lower plenum with large temperature fluctuations, and as a result, it was confirmed that the temperature fluctuation of the flow did not have a significant effect near the corner wall. In addition, the vortices generated from the lower plenum to the outlet duct were identified in this paper. It is considered that the quantitative and qualitative results presented in this paper will serve as reference data for the benchmark.

Comparison of Melted Corium Relocation during Severe Accident of High Temperature Reactor using Moving Particle Semi-Implicit Method

System failure in nuclear reactors can cause degradation of a reactor core, allowing melting and relocation of the corium to the lower plenum in the nuclear reactor system. In this study, a severe accident simulation was carried out using the Moving Particle Semi-Implicit (MPS) method. In this method, we model the relocation of molten corium on the reactor core (support plate) to the lower plenum for several conditions with variations: corium material, lower plenum conditions, temperature, viscosity, and density. Those treatments were carried out in order to be able to compare and analyze the characteristics of the corium melt by reviewing the velocity profiles. The formation of a corium pool and debris bed can result in significant temperature differences and high heat flux against the walls of the reactor vessel, causing a decrease in the integrity of the reactor vessel and reactor failure.Keywords: Corium, Uranium Dioxide (UO2), Zirconium Dioxide (ZrO2), Fluid Relocation, Moving Particle Semi-Implicit (MPS).

Investigation of nuclear reactor core thermal-hydraulic characteristics after partial loss of flow accident

In normal operation conditions of nuclear power plants, the distribution of primary coolant between fuel channels would be considered almost uniform. When different number of Reactor Circulation Pumps (RCPs) are switched off, known as an abnormal condition, this uniform distribution is disturbed and different conditions occur for each channel depending on its position in the core. In this research, the normal and abnormal condition (with one or two tripped RCPs) for a VVER-1000/446 is investigated. For evaluation of the core neutronic calculations and thermal power distribution, USNRC’s PARCS system code is employed. Then a thermal-hydraulics module was developed for performing the T/H calculation of the core zone. The input velocity of each channel in abnormal condition was calculated based on developed CFD model in downcomer and lower plenum of Reactor Pressure Vessel (RPV) by ANSYS-CFX. The results show that, in normal operation, the hot channel is related to the central fuel assembly of the reactor core with the highest relative power equal to 1.29 and total power of 23.74 MW. In this case, the minimum inlet velocity, the maximum coolant outlet velocity, and the maximum fuel temperature are 5.6 (m/s), 330.96 (°C), and 1345.8 (°C), respectively. In the cases of operation with one and two tripped RCPs, the hot channel is related to the fuel assemblies with the lowest inlet velocities. The lowest velocities are 0.32 m/s, 0.24 m/s, and 0.22 m/s respectively for the condition with one tripped pump, two tripped pumps placed oppositely, and two tripped pumps placed contiguously. The hot channel numbers in these cases are 158, 102, and 90, respectively. In these channel, the condition of outlet flow would be superheated, but the fuel temperature (1006.3, 1050.9, and 987.9) do not reach the maximum allowable margin. The study confirms the necessity of the coolant distribution consideration in OLCs as well as events that may disturb the symmetry of the coolant flow. It also showed that the lateral fuel assemblies are more at risk in this situation because of a significant reduction in coolant flow. Likewise, the investigations proved the safe continuation of the operation in PLOFA conditions with the preventive algorithm of the emergency protection system and without any need for immediate mitigation actions or operator intervention.

Numerical investigation of the external emergency coolant transportation inside reactor annular down-comer

A numerical investigation of the ECC (Emergency Core Cooling) coolant transportation in the reactor annular down-comer is performed in this paper. The density difference of coolant is paid special concern and is quantificationally analyzed. If the ECC coolant with low density and high velocity is injected into the annular down-comer, the horizontal kinetic energy decays as the azimuthal angle of down-comer increases. As a result, the ECC coolant transports through the annular down-comer along the clockwise direction. If the ECC coolant with high density and low velocity is injected, the coolant falls vertically and arrives the lower plenum as it enters the annular down-comer. As the lower plenum is filled with the ECC coolant, part of the ECC coolant moves upward due to the mass force effect. With the density difference, remarkable velocity oscillation is generated in the lower space of the down-comer. The transportation of ECC coolant insider the down-comer is decided by Froud number. When the Froud number is larger than 1.2, the inertia force dominant mixing is formed and the ECC coolant transports in the clockwise direction. When the Froud number is lower than 0.8, the buoyancy driven mixing is formed and the ECC coolant transports in the anti-clockwise direction.

MAAP code analysis focusing on the fuel debris condition in the lower head of the pressure vessel in Fukushima-Daiichi Nuclear Power Station Unit 2

Based on the updated knowledge from plant-internal investigations, experiments and computer-model simulations until now, the in-vessel phase of Fukushima-Daiichi Nuclear Power Station Unit 2 was analyzed using the MAAP code. In Unit 2, it is considered that the enthalpy of the core materials was relatively low when they relocated to the lower plenum of the pressure vessel, then, cooled by the coolant and solidified there. Although the MAAP code tended to underestimate the degree of core-material oxidation during the relocation period, this probable underestimation was compensated for by an existing study that was considered more reliable, so that more realistic debris conditions in the lower plenum could be obtained. With this study, basic validity of the former prediction of the Unit 2 accident progression behavior was confirmed and detailed boundary condition for the later phase was provided. This boundary condition should be utilized for future studies addressing debris reheating process leading to lower head failure and debris relocation toward the pedestal.