Debris Bed Research Articles

Numerical simulation of CS28-1 experiment by using CANDU severe accident analysis code, CAISER

Due to the demands for an accurate and realistic severe accident code in the PHWR safety analysis, a new severe accident code called CAISER is being developed at Korea Atomic Energy Research Institute (KAERI). The CAISER code simulates core degradation phenomena occurring in a calandria tank, and it consists of two main modules: A fuel rod degradation module and a fuel channel degradation module. The fuel rod degradation module simulates the severe accident phenomena in a fuel channel, which includes a core uncovery, fuel rods heatup, hydrogen generation due to steam-Zr oxidation, fuel rods slumping, fuel rods melting and relocation, and thermal interaction of relocated molten mass with a pressure tube or a calandria tube. The fuel channel degradation module simulates the overall severe accident phenomena in a calandria tank, including the sagging of a fuel channel, debris bed formation caused by a fuel channel failure, the molten pool formation, and the calandria tank failure. Each module is tracking mass and energy changes for the main components, which are fuel, cladding, pressure tube, calandria tube in a fuel rod degradation module, and debris bed, metallic and oxidic corium pools in a fuel channel degradation module. These two modules are closely interconnected to simulate phenomena both in a fuel channel and in a calandria tank at the same time. In this paper, the numerical modelling in the CAISER code is described, including the node system and the governing equations for each main component. For the purpose of CAISER code validation, numerical simulations were conducted for the CS28-1 experiment. The simulation results show that the temperature distributions calculated by CAISER are in line with the experimental data.

Sensitivity analysis of core slumping and debris quenching behavior of Fukushima Daiichi Unit-3 accident

The Great East Japan earthquake and the subsequent tsunami which occurred on March 11th, 2011 put the operating Units 1–3 at Fukushima Daiichi Nuclear Power Plant (NPP) in severe accident conditions and core meltdown due to station blackout. Although research efforts have been made by various parties to study the accident scenarios since the Fukushima accident, there remain unresolved issues regarding the core degradation behavior suggested by measurement data such as water level, Reactor Pressure Vessel (RPV) pressure and Primary Containment Vessel (PCV) pressure. To analyze and resolve such issues would be helpful to promote further understanding of the severe accident scenario at Fukushima units as well as the decommissioning work undergoing.The current study focuses on a detailed analysis of the RPV pressure peak event that occurred in Unit-3 at 12:00 on March 13th 2011. Sensitivity analysis cases were carried out with MELCOR 2.2 codeiThe MELCOR code version applied in the current study is MELCOR 2.2.9541. The results used in this study were obtained before April 2018 under research collaboration with Waseda University.i with sensitivity parameters that can influence the RPV pressure behavior, such as the debris quenching heat transfer coefficient, the number of opening SRVs during the RPV pressure peak event, amount of core slumping and particulate debris diameter. The cases that could reproduce the RPV pressure peak were further discussed to show likely debris bed energy history and the water mass history in the lower plenum during the RPV pressure peak event. The current study suggests that 1) Opening of SRVs equivalent to the total area of 4–6 fully-open SRVs (or equivalent leak area) could have occurred during the pressurization phase of the RPV accompanied by heavy debris quenching effect, while the opening of SRVs equivalent to a total area of at least 2 fully-open SRVs (or equivalent leak area) could have occurred during the depressurization phase of the RPV accompanied by moderate debris quenching effect. 2) The particulate debris diameter is not a very sensitive parameter when evaluating the debris quenching effect of Unit-3 in the current MELCOR modeling. 3) The current modeling suggests that around 70–110 GJ of energy can be removed by coolant during the debris quenching period with 30 tons of water reduction from the lower plenum.

Criticality Characteristics and Sensitivity Analysis of Fukushima Debris Beds Containing MCCI Products

Abstract This study investigates the criticality characteristics of debris beds that may have been formed through the molten–core–concrete-interaction (MCCI) at the pedestal floor of the damaged reactors in Fukushima Daiichi Nuclear Power Station. These were modeled as UO2-concrete systems submerged in water. First, a conservative model was used to evaluate the impact that the presence of concrete has on the neutron multiplication factor (keff) of debris beds. The good moderation capacities of concrete were proved, and it was found that recriticality would be possible under the considered conservative assumptions. Second, a more realistic model was used to perform an uncertainty and sensitivity analysis of a wide range of debris parameters (debris porosity, core meltdown grade, debris size, debris composition, concrete erosion factor, etc.). In this case, the results indicate that the probability of a recriticality event is very remote. It was also found that the presence of boron (B4C) from the control rods within debris has by far the highest influence on keff.

Analysis of the Effect of Vessel Failure and Melt Release on Risk of Containment Failure Due to Ex-Vessel Steam Explosion in Nordic Boiling Water Reactor Using ROAAM+ Framework

Abstract Nordic boiling water reactor (BWR) design employs ex-vessel debris coolability in a deep pool of water as a severe accident management (SAM) strategy. Depending on melt release conditions from the vessel and core–melt coolant interactions, containment integrity can be threatened by: (i) formation of noncoolable debris bed or (ii) energetic steam explosion. Melt is released from the vessel affect ex-vessel phenomena and is recognized as the major source of uncertainty. The risk-oriented accident analysis methodology (ROAAM+) is used for quantification of the risk of containment failure in Nordic BWR where melt ejection mode surrogate model (MEM SM) provides initial conditions for the analysis of debris agglomeration and ex-vessel steam explosion which determine the respective loads on the containment. Melt ejection SM is based on the system analysis code methods for estimation of leakages and consequences of releases (computer code) (MELCOR). Modeling of vessel failure and melt release from the vessel in MELCOR is based on parametric models, allowing a user to select different assumptions that effectively control lower head (LH) behavior and melt release. The work addresses the effect of epistemic uncertain parameters and modeling assumptions in MEM SM on the containment loads due to ex-vessel steam explosion in Nordic BWR. Sensitivity and uncertainty analysis performed to identify the most influential parameters and uncertainty in the risk of containment failure due to ex-vessel steam explosion. The results of the analysis provide valuable insights regarding the effect of MELCOR models, modeling parameters, and sensitivity coefficients on melt release conditions and predictions of ex-vessel steam explosion loads on the containment structures.

Development of relocation model for debris particles in core disruptive accident analysis of sodium-cooled fast reactor

During a core disruptive accident in a sodium-cooled fast reactor, the molten fuel or steel is solidified into debris particles that form a debris bed in the lower plenum. When the debris bed starts boiling, the flow of the coolant and vapor together relocates and flattens the debris particles. This process is called debris relocation. The thickness of the debris bed and its porosity have significant influence on the cooling ability of the fuel debris in the lower plenum. Therefore, it is necessary to evaluate the transient changes in the shape and thickness of the debris bed in the relocation behavior for accident analysis. To simulate the relocation behavior, a debris-relocation model based on the COMMEN code was developed in this study. The debris-relocation model is based on the shear strength mechanism, which is widely applied in soil mechanics. Shear strength is a function of the density, position, and shape characteristics of particles. The debris bed is relocated only when the shear stress of the global debris particles is greater than the shear strength. By incorporating the relocation model into the COMMEN code, the relocation transient processes of different particle beds in a nitrogen-injection experiment were simulated, and the results were compared with experimental results. The analysis indicates that the proposed debris-relocation model can effectively predict the occurrence of debris-bed relocation and reasonably simulate its transient process, although future improvement is necessary.

Pressure dependence of dryout in a heat-generating porous debris bed

The primary objective of this work is to numerically study the effect of system pressure on the occurrence of dryout phenomena in heap-shaped, heat-generating porous debris beds. This is achieved using a multiphase computational fluid dynamics (CFD) model that is implemented in the framework of the commercial CFD platform ANSYS FLUENT. The model is extensively validated with available experimental data on two different aspects: pressure drop in two-phase flow through porous media and prediction of dryout in typical porous debris beds. A wide range of system pressure, relevant to severe accident conditions, is considered in this analysis in order to obtain a thorough understanding of its impact on multiphase flow and dryout occurrence in porous debris beds. This analysis is performed for different subcoolings of flooding water, which gives additional knowledge on the effects of coolant subcooling on dryout occurrence. Results indicate that dryout in debris beds occurs at progressively higher power densities as the system pressure is raised. Similar effects are observed with increase in liquid subcooling as well.

Study on relocation behavior of debris bed by improved bottom gas-injection experimental method

During the core disruptive accident (CDA) of sodium-cooled fast reactor (SFR), the molten fuel and steel are solidified into debris particles, which form debris bed in the lower plenum. When the boiling occurs inside debris bed, the flow of coolant and vapor makes the debris particles relocated and the bed flattened, which called debris bed relocation. Because the thickness of debris bed has great influence on the cooling ability of fuel debris in low plenum, it’s very necessary to evaluate the transient changes of the shape and thickness in relocation behavior for CDA simulation analysis. To simulate relocation behavior, a large number of debris bed relocation experiments were carried out by improved bottom gas-injection experimental method in this paper. The effects of different experimental factors on the relocation process were studied from the experiments. The experimental data were also used to further evaluate a semi-empirical onset model for predicting relocation.

Experimental study on melt-jet behavior during SFR core disruptive accidents using simulant materials

Investigations on melt-jet behavior during Core Disruptive Accidents (CDAs) of Sodium-cooled Fast Reactors (SFR) are crucial to the design of In-Vessel Retention (IVR) strategy. In present work, aimed to achieve a comprehensive understanding on this behavior, a number of simulated experiments are conducted by discharging a molten low-melting-point alloy jet into a water pool. Different experimental parameters such as melt temperature, water temperature, melt penetration velocity, nozzle diameter and water pool depth are taken. Based on visual observation as well as the quantified particle and debris-bed properties measured (e.g. average size and size distribution of fragments, particle sphericity and bed porosity), the effect of different releasing conditions on the melt-jet behavior mode, particle characteristics and debris-bed morphology are clarified. Dominant factors that influence the melt-jet breakup phenomenon are identified over present range of conditions. The performed analyses also suggest that the fragment mass median diameter can be reasonably interpreted by the Kelvin-Helmholtz instability model which integrates solidification effect and the Weber number theory. Besides, map diagrams distinguishing hydrodynamic effect and thermal effect with regard to melt-jet behavior mode, debris bed morphology (porosity) and particle morphology (sphericity) are explored. Insight and experimental data obtained in this work will be utilized for the future improved validations of fast reactor safety analysis codes in China, such as the thermal–hydraulic models for the prediction of fragmentation and debris formation phenomena.

Modelling adhesive wear extension in fretting interfaces: An advection-dispersion-reaction contact oxygenation approach

This paper aims at modelling contact oxygenation concept which describes the effects of fretting loading conditions on di-oxygen flow within the interface and the associated partition between abrasive and adhesive wear domains in fretting wear scars. This is achieved by modelling oxygen transport within fretting interface using an advection-dispersion-reaction approach (ADR) by considering debris bed as a compact porous medium traversed by atmospheric gases. ADR model is calibrated using three fretting tests of flat-on-flat 34NiCrMo16 interface. Results show that ADR approach can predict the partition of abrasion and adhesion for wide range of sliding frequencies, normal forces and contact sizes. Besides, it can capture the transition of wear mechanisms from pure abrasive to mixed abrasive-adhesive wear at different loading conditions.

Implementation of framework for assessment of severe accident management effectiveness in Nordic BWR

Nordic Boiling Water Reactor (BWR) design employs ex-vessel debris coolability in a deep pool of water as a severe accident management (SAM) strategy. Depending on melt release conditions from the vessel and core-melt coolant interactions, containment integrity can be threatened by (i) formation of non-coolable debris bed, or (ii) energetic steam explosion.In order to assess the effectiveness of SAM the Risk Oriented Accident Analysis Methodology framework (ROAAM+) has been developed. The framework is further extension of the approach originally developed and applied by Prof. Theofanous and co-workers.This paper presents the implementation of ROAAM+ probabilistic framework for uncertainty quantification and risk analysis.We further apply ROAAM+ to the analysis of steam explosion risk in Nordic BWR assuming different state-of-knowledge situations and different containment fragilities. We employ an iterative processes of knowledge refinement using risk analysis as a guiding tool in identification of the major sources of uncertainty. We estimate failure domains and discuss ROAAM+ results in terms of possibility vs necessity of containment failure.

First fossil true water bugs (Heteroptera, Nepomorpha) from Upper Jurassic strata of North America (Morrison Formation, southeastern Utah)

ABSTRACT The first fossil hemipteran from the Late Jurassic of North America, Morrisonnepa jurassica n. gen. et n. sp., is reported and described from the Morrison Formation, Jurassic Salad Bar locality, San Juan County, Utah, USA. The new specimen is characterised and illustrated, showing morphological characters similar to nepomorphs such as forewing well-developed and folded flat on the abdomen, oval abdomen shape, and the presence of a short pair of appendages. The taxonomic allocation close to members of the Nepomorpha is discussed. Morrisonnepa jurassica n. gen. et n. sp. was collected with abundant plants, spinicaudatan carapaces, and a small amphibian from a finely laminated shale that overlies a coarser plant debris bed, supporting the presence of a possible oxbow lake or pond, environments developed within the greater ecosystem of the Morrison Formation during the Late Jurassic. In this context, we analyse the taphonomic and palaeoecological implications of the presence of aquatic insects. Besides providing morphological information on Jurassic nepomorphs, the new fossil helps illustrate how the aquatic insect assemblage was integrated during the Jurassic in North America.

In-situ X-ray tomography of wear – A feasibility study

Wear experiments using annular contacts were carried out to investigate the possibility of imaging wear scars and wear debris beds in-situ. Samples made of 6082T6 aluminium alloy were subjected to unidirectional wear under different conditions with the sample elements then locked in contact to preserve the end-of-wear condition. Post-wear 3D morphology was successfully imaged in-situ using X-ray tomography, with total depth of the wear scar, surface primary-profile roughness and real area of contact also obtained. This study is a proof-of-concept, illustrating the utility of tomography for studying wear, without the need to disassemble the contact or disturb the debris bed and serves as a starting point for a new approach to the investigation of wear phenomena.

Minimum diameter limit of particle size distribution and its effect on coolability of debris bed

The particle size distribution (PSD) of the debris bed is a major parameter influencing its coolability with respect to the two-phase pressure drop of the porous media. Most studies have focused on representative particle diameter as opposed to size distribution. Based on a severe accident situation (strong two-phase convection), we introduced a rearrangement of small particles and developed a corresponding size distribution formula by modifying the Rosin-Rammler distribution.A minimum particle diameter model was proposed as a function of dryout heat flux (DHF) based on a minimum fluidization velocity concept. In addition, a truncated Rosin-Rammler distribution was developed and a Sauter Mean Diameter (SMD) was formulated in the form of the gamma function with a new parameter: the minimum particle diameter (Dmin). Sensitivity analysis of DHF in a 1-D top flooded debris bed showed two notable results for the effect of Dmin on coolability. First, Dmin significantly reduced DHF uncertainty due to the distribution constant. Moreover, the calculated DHF increased by introducing the concept of a minimum particle diameter. In conclusion, treatment of small particles in the debris bed is essential for determining debris bed coolability. Additionally, the truncated Rosin-Rammler distribution could be utilized to provide the precise SMD in a porous media.

Experimental investigation of debris entrapment in annular contacts

In wearing contacts, a part of the wear debris is trapped between the surfaces, creating a wear debris bed. Many researchers report such beds providing a degree of protection against further wear. It was hypothesized, that in an annular contact, the level of debris entrapment increases with the width of the annulus resulting in decrease in wear. To test this hypothesis, two series of experiments using EN1A steel and 6082T6 aluminium alloy annuli, measuring 1, 2 and 4 mm wide were carried out. When steel was used, there was no clear correlation between ring width and total mass loss or steady-state wear rate. When aluminium was used, wear decreased with the ring width according to both metrics. This behaviour appears to correlate with worn surface roughness, suggesting worn surface roughness to be a parameter controlling debris entrapment.

A comparative assessment of various single tray core catcher options for future sodium cooled fast reactors

Various single tray in-vessel core catcher options to handle whole core decay heat for a 1500 MWt capacity sodium cooled Indian fast reactor (SFR) have been analyzed towards their efficiency in rendering acceptable core catcher temperatures. For this purpose, a one dimensional transient multiphase heat transfer model that considers sodium boiling, fuel particles re-melting and slumping has been adopted. The boundary conditions for the one dimensional model are obtained from computational fluid dynamics (CFD) simulations of turbulent natural convective heat transfer within the lower sodium plenum of the SFR. It is found that the top surface of the debris bed dissipates more heat than the bottom surface of the core catcher. A single plate core catcher is seen to experience melting of the core catcher plate (CCP) for a short duration in case of CCP thickness greater than 80 mm. The heat shield plate (HSP) of the double plate core catcher is observed to undergo complete melting due to vapor locking between the HSP and CCP. The phenomena of melting and slumping of debris particles is seen in double plate core catchers with a refractory lining. However, safe temperature levels of HSP and CCP were observed in the case of double plate core catchers with refractory lining. This study suggests that a double plate core catcher with a delay bed is the potential option for future sodium cooled fast reactors, which are to be designed to handle whole core meltdown scenarios.

Study on high-speed calculation of debris bed coolability for sodium-cooled fast reactors

Study on high-speed calculation of debris bed coolability for sodium-cooled fast reactors

Study on application of artificial neural network to debris bed coolability calculations for sodium-cooled fast reactors

Study on application of artificial neural network to debris bed coolability calculations for sodium-cooled fast reactors

Study on melt stratification and migration in debris bed using the moving particle semi-implicit method

In the severe accident of nuclear reactors, the melt stratification and migration in debris bed directly affect the melt heat transfer characteristics and the heat flux distribution on the lower head wall, which in turn affect the follow-up accident progressions including the pressure vessel failure mode, melt leakage, and molten corium concrete interaction. They are the key issues in nuclear reactor severe accident research. In the present study, the MPS method was applied to analyze melt stratification and migration in debris bed by including the models of radiation heat transfer among debris, melt phase change and surface tension. To validate the numerical method, the experiment was performed using the metal tin and ternary nitrate as the core melt simulant. Then, the stratification behavior of oxide and metallic core melt was investigated under varied parameters and conditions. The results indicated that the crust formed at the oxide and metal melt interface could block the stratification, which had important influence to the layered configuration. The melt contact area increased in the conditions of crust formation prior to the stratification completion. The high melt and debris temperatures and surface tension could contribute to the stratification, but the high viscous melt when undergoing phase transition needed a long time. Moreover, the comparison of melt stratification under adiabatic and isothermal boundary conditions showed that the oxide melt solidification trapped some metal melt in the voids. The penetration distance of metal melt in the oxide debris bed increased rapidly in the initial moment and gradually in the following moment, and finally the migration was terminated by melt crust. The melt penetration distance increased as the increase of debris bed porosity and melt and debris temperatures.

Design of core catchers for sodium cooled FBRs – Challenges

The whole core meltdown scenario in sodium cooled fast reactors is considered under design extension criteria and has a very low probability of occurrence. To mitigate such a hypothetical severe accident in fast breeder reactors, a core catcher has been provided to accommodate the core debris within the primary containment boundary. In sodium cooled fast breeder reactor (FBR), an in-vessel core catcher, is provided to receive and disperse the fuel debris arising out of core meltdown during Hypothetical Core Disruptive Accident (HCDA). The core catcher prevents hot debris from reaching and damaging the main vessel. It also enables adequate heat transfer by natural convection to keep debris in stable conditions. The present study focuses on design of core catcher with sacrificial barriers and techniques to improve natural cooling of debris on the core catcher for FBRs. Studies are carried out towards the selection of suitable sacrificial material for core catcher, numerical analysis on multilayer and thermal-hydraulic analysis on multi jets core catcher concepts for selection of best suitable technique for debris bed coolability. The core catcher concept with sacrificial barrier and cooling pipes is found to possess enhanced coolability of debris bed. Experimental studies on compatibility of sacrificial ceramic material with sodium indicate strong dependence on microstructure. It has also been found that the multilayer core catcher with MgO as delay bed of 50 mm thickness is adequate to reduce the bottom temperature of core catcher below the design safety limit. Judicious mix of all these findings is expected to have unique design to accommodate whole core meltdown in future fast reactors.

In-Vessel Retention of PHWRs: Experiments at Prototypic Temperatures

Abstract In pressurized heavy water reactors (PHWRs), multiple failures of engineered safety features may cause a failure of core cooling eventually leading to core collapse. The failed fuel and fuel channels relocate to the bottom of the calandria vessel (CV) and form a terminal debris bed, which generates decay heat. With time, the moderator evaporates and the terminal debris bed ultimately melts and forms a molten pool of corium. If corium breaches the CV and enters the calandria vault, large amounts of hydrogen and other fission gases may be generated due to molten core concrete interaction, which may pressurize the containment leading to containment failure. In addition, the passive catalytic recombiner devices may be incapable of managing such large amounts of hydrogen. Hence, in-vessel retention of corium is the only option to the avert progression of the accident. The heat removal capability of the CV needs to be demonstrated in order to attain the goal of in-vessel retention, to contain the corium during severe accidents. A lot of numerical analysis of heat removal capability of the CV has been done. However, experimental demonstration of in-vessel retention has been rarely presented in the literature, especially for PHWRs. In this paper, in-vessel retention at prototypic temperatures has been presented. Experiments have been carried out in scaled CVs. Different corium simulants have been used at elevated temperatures and corium coolability has been demonstrated.