Hypothetical Core Meltdown Research Articles

Investigation of molten fuel coolant interaction with simulated corium in sodium

In case of a hypothetical core meltdown accident in Sodium cooled Fast Reactor (SFR), corium interacts with cold pool sodium and results in Molten Fuel Coolant Interaction (MFCI). During MFCI, the corium fragments and resultant debris settle on the core catcher as a bed and continue to generate decay heat. The size and state of fragmented debris play a crucial role in post-accident heat transfer. An experimental study has been undertaken to investigate the fragmentation of simulated corium in sodium using real-time X-ray imaging. Experiments were conducted by releasing ∼400 g of molten mixture of alumina and iron at ∼2400 °C into ∼4 kg of sodium at various temperatures. The fragmentation phenomenon was acquired using a digital flat panel detector and the captured images were subjected to image analysis to extract information on melt fragmentation. Post-experiment, fragmented debris was retrieved and subjected to sieve analysis to obtain characteristics towards assessing the dominant fragmentation mechanism. Experiments were also conducted with water for identifying key differences in the fragmentation and debris characteristics. Non-energetic fragmentation, generation of fine debris, and relatively minor sodium vaporization observed in the small-scale experiments present a reduced potential for energetic interaction in sodium than that of water.

Generation of simulated corium using thermite process

Nuclear reactors are built with many inherent and engineered safety features with defence in depth philosophy, to maintain integrity of core during normal as well as accidental conditions. However, a severe accident scenario such as hypothetical core meltdown has to be considered in safety analysis of the plant, towards demonstrating a safe and coolable geometry of degraded core. Progression of corium, its subsequent interaction with coolant and settlement in the reactor vessel determines the extent of core damage and needs thorough investigation for ensuring safe mitigation of the accident. A comprehensive experimental program has been taken up at IGCAR to investigate severe accident phenomena in Sodium cooled Fast Reactors (SFR). Highly exothermic thermite reaction is used to generate a high temperature melt consisting of ceramic and iron simulating the corium. A dedicated test facility ‘THEME’ has been setup and few small scale simulation experiments were conducted to study various phenomena like subassembly melting, corium interaction with coolant and settlement of debris on corecatcher.

Numerical simulation of Post Accident Heat Removal in a typical pool-type SFR under severe core relocation scenario

Under a hypothetical core meltdown and relocation scenario in Sodium cooled Fast Reactor (SFR), there may be a chance for reactor main vessel failure if the core debris are allowed to settle at the bottom of this vessel. To avoid this, an in-vessel core catcher assembly is installed in pool type SFRs just above the bottom wall of the main vessel. The function of this assembly is to collect, retain and passively cool the core debris by assisting natural convection decay heat removal from these debris. In this work, the passive decay heat removal from the core debris settled on single-tray-type design of core catcher has been numerically simulated under different amounts of core relocation using a validated 2-D axisymmetric CFD model. The maximum temperature experienced by the debris bed, core catcher plate and the welded joint between the core support structure and the main vessel (i.e., triple point) are evaluated in order to estimate the amount of core relocation that can be safely handled by the primary containment system under Post Accident Heat Removal (PAHR) scenario. The numerical study revealed that the single-tray-type design of core catcher can potentially assist the passive cooling even under the settling of debris material with heat generation equivalent to 60% of the full core on the core catcher. In addition to this, the effect of the size of opening created on grid plate during core relocation on Post Accident Heat Removal (PAHR) is studied for different amounts of core relocation. From this numerical study, the thermal capability of single-tray type in-vessel core catcher of SFR is assessed. The predictions are useful in determining the safe thermal margins available under core relocation scenarios and extending the functionality of single-tray-type core catcher for handling Whole Core Meltdown Accident (WCA) scenario in future SFRs.

Analysis of composition and temperature of debris bed based on settling characteristics of relocated core debris in pool type SFR

The hypothetical core meltdown accident in sodium cooled fast reactors (SFR) is a severe accident situation during which the energetic dispersion of core materials and their relocation may pose a threat to the integrity of reactor containment. During core meltdown accident, part of the molten core materials relocated downwards would be fragmented in the lower plenum sodium pool and settle over the in-vessel core catcher plate and forms a particulate debris bed. The recriticality potential and coolability of the debris bed are strongly influenced by the composition of fragmented particles of different core materials and their size distribution. In this paper, simplified models have been developed to evaluate the incipient conditions i.e., composition profile and initial temperature of the particulate debris bed formed over an in-vessel core catcher plate following the core meltdown accident in pool type SFR. The particle mixture settling process subsequent to the fragmentation of molten core materials in sodium pool has been analyzed to evaluate the composition profile of the debris bed. A 1-D model for particle motion through the stagnant fluid under the influence of gravity has been considered to evaluate the settling characteristics of core materials debris (mainly UO2 and stainless steel) in liquid sodium for a wide range of possible sizes. By extending this model, the differential settling behaviour of mixture of UO2 and stainless steel particles in liquid sodium pool has been investigated. The composition profile of the debris bed along its thickness has been evaluated for the material relocation scenario considered with various mass ratios of stainless steel and UO2 mixture using the published experimental results based particle size distributions. Further, a 1-D transient conduction model is implemented to estimate the temperature of debris particles when they settle on the core catcher plate. Simulations revealed that the particles will be quenched to near the temperature of the liquid sodium in lower pool while reaching the core catcher. The results are useful in assessing the coolability characteristics of debris bed formed on the core catcher and its recriticality potential in the analysis of the Post Accident Heat Removal.

Estimation of evaporation rate of cesium from hot sodium pool to inert cover gas of typical SFR

Estimation of evaporative release of Cesium-137 (Cs-137) from the hot pool of Sodium cooled Fast Reactor (SFR) under normal operating condition and hypothetical core meltdown accident scenario is important in safety analysis due to the radiological consequences associated with Cs release. A methodology has been developed to estimate the cesium evaporation rate from hot pool to inert cover gas region of typical SFR under fuel pin clad failure scenario based on the temperature and concentration dependent activity coefficient of cesium estimated using the Flory–Huggins solution theory. The predicted equilibrium partition coefficient (KD) values of cesium are validated with theoretical and experimental data available in literature. It has been observed that the KD of cesium is decreasing with increase in Cs concentration. A correlation has been developed for KD as a function of temperature in the infinite dilute solution regime. The present methodology predicted Retention Factor (RF) values of Cs based on the estimated KD values are closer to the experimental results compared to the published results of empirical models in the SFR hot pool temperature range. The methodology has been used for evaluating the evaporative release of cesium along with sodium from hot pool to cover gas region of typical SFR under fuel pin clad failure scenario. These estimations are useful for evaluating the radioactive dose due to Cs transport from SFR sodium pool to its inert argon cover gas region.

An estimate of the order of magnitude of the explosion during a core meltdown-compaction accident for heavy liquid metal fast reactors: A disquieting result updating the Bethe–Tait model

Criticality and recriticality considerations in heavy liquid metal fast reactors (HLMFRs) after a hypothetical core meltdown accident are discussed. Although many aspects of system behaviour in such scenarios can be deduced directly from the classical theory of sodium-cooled fast reactors (SFRs), certain ideas that have been accepted as true for SFRs cannot be extrapolated to HLMFRs without sufficiently careful thought. In this paper, we are concerned, as in SFRs, with fuel compaction, but with one important difference: there would be no boiling of the surrounding heavy liquid metal pool. Utilizing a Bethe–Tait model, it is shown that, due to the power flattening effect of the heavy liquid metal, explosive excursions at least an order of magnitude higher than for SFRs in similar situations are conceivable.

Fragmentation of molten debris during a molten fuel-coolant interaction

The potential for an energetic molten fuel-coolant interaction (MFCI) during a hypothetical core meltdown accident is of concern in nuclear safety analysis. An important aspect of a MFCI is the fine fragmentation and intermixing of molten core debris with the core coolant. The fragmentation characteristics of the molten debris (a mixture of UO 2 and zircaloy cladding) particles produced during a recent high-energy in-pile experiment are analyzed. The experimental results suggest that two mechanisms contributed to the fragmentation of the molten debris in this experiment, in which an MFCI occured. Phenomenological modelling of these two mechanisms and the effects of the governing parameters are presented.

Multiphase, multicomponent hydrodynamics in HCDA analysis: Present status and future trends

The scope and complexity of hypothetical core meltdown accident phenomenology is outlined. Approaches for assessing the recriticality characteristics and the tendencies for severe, containment challenging events is addressed also. Some results of initial analyses with SIMMER-II, which attempts to include the numerous multicomponent, multiphase effects on material movements and reactor neutronic response, are described. Perspectives are offered regarding the role of the internal structures in the termination of neutronic activity during a core meltdown accident sequence.

Release of Fission and Activation Products during Light Water Reactor Core Meltdown

The most relevant open questions combined with activity release during hypothetical core meltdown accidents refer to the chemical behavior of the highly reactive elements iodine, cesium, and tellur...

A Condensed Review of the Technology of Post-Accident Heat Removal for the Liquid-Metal Fast Breeder Reactor

The analyses performed to evaluate the capacity to contain the core materials after a hypothetical core meltdown accident are generally grouped under the area of post-accident heat removal (PAHR). ...

Response of liquid metal fast breeder reactor containment to a hypothetical core meltdown accident

This paper presents an analysis of the response of the containment building of a 2500-MWt liquid metal fast breeder reactor to a hypothetical reactor core meltdown. Although not mechanistically justifiable, this type of event is chosen for analysis as a basis for risk assessments. Containment space atmosphere compositions, temperatures, pressures, and structural temperatures are calculated, based on decay energy release and chemical reactions associated with the incident. The CACECO containment analysis code, which was used to make the calculations, is described in detail. Results of the study show that by utilizing the passive heat absorption capability of structures normally present in containment design, reactor plant containment integrity can be maintained for more than a day, even for extreme hypothetical events.

A Study of Post-Accident Molten Fuel Downward Streaming through the Axial Shield Structure in the Liquid-Metal Fast Breeder Reactor

A two-dimensional transient freezing model is developed to study the dynamics of solidification of a fluid flowing in a cylindrical channel, the walls of which are cooled below the freezing temperature of the fluid. The model is applied to the case of molten fuel flowing downward through the coolant channels in the lower shielding structure of current liquid-metal fast breeder reactor designs subsequent to a hypothetical core meltdown accident. The results indicate that under post-accident conditions, a high potential exists for rapid relocation of significant quantities of core debris across the shield structure.

Fast Reactor Fuel Interactions with Floor Material after a Hypothetical Core Meltdown

The hypothetical accident approach to analysis of fast reactors has been applied to the meltdown of an entire core and its interaction with containment floor materials of construction. The objectiv...