Hypothetical Core Disruptive Accident Research Articles

Development of a numerical model for disassembly phase of hypothetical core disruptive accident and application to a medium-sized mixed oxide fuelled fast reactor

Safety assessment of fast reactors involves analyzing Hypothetical Core Disruptive Accidents (HCDA) that could result in reactor core disassembly. HCDA is generally analyzed in three distinct phases that include pre-disassembly, disassembly and post-disassembly phases. In the present work, a new coupled neutronics-hydrodynamics computer code called FAst Reactor DISassembly (FARDIS-1) is developed in 2-D cylindrical coordinate geometry to model the super-prompt critical power excursion during the disassembly phase of HCDA in a sodium-cooled fast reactor (SFR). The modelling approach followed is essentially similar to the one used in VENUS-II code but with a few improvements. The code is benchmarked against the predictions of VENUS-II for a sample case of HCDA in the FFTF reactor model. This is followed by an analysis of the disassembly phase in a medium-sized (500 MWe) mixed oxide fuelled fast reactor. Two cases are assumed at the beginning of the disassembly phase: a sodium voided core (conservative) and a partially voided core (realistic) and the influence of several core neutronics parameters on the transients is elucidated through parametric studies. The peak power, thermal energy release, peak pressure, maximum temperature and work-potential are estimated and their trends are discussed.

On physics of a hypothetical core disruptive accident in Multipurpose hYbrid Research Reactor for High-tech Applications – MYRRHA

The sensitivity of the reactivity of a fast reactor core to changes in its geometry and/or fuel relocation calls for particular attention with regard to criticality events. A category of these events, the so-called Core Disruptive Accidents (CDAs), are intensively studied in the safety assessment of Sodium-cooled Fast Reactors (SFRs), and more recently also in the case of other systems. Differences between SFRs and Heavy Liquid Metal Fast Reactors (HLMFRs) are significant and therefore warrant an understanding of phenomena and the development of models specific to HLMFRs. This paper provides a qualitative overview of the physics relevant to the investigation of a CDA in HLMFR, with a particular application to the Multipurpose hYbrid Research Reactor for High-tech Applications – MYRRHA. At first, a core compaction mechanism viable for an HLMFR has been postulated. In what follows, simulation by an already existing severe accidents code, as well as modelling based on fundamental physics and engineering, have been performed. It is demonstrated that, for a linear insertion of reactivity due to hypothetical core compaction, the reversal of reactivity evolution happens due to the Doppler effect and the thermal expansion of core materials. Subsequent expansion by fuel melting terminates the prompt-critical event and makes the system delayed-supercritical. Successive fuel and/or coolant boiling is responsible for the hydrodynamic disassembly of the core and it therefore effectively terminates the transient.

Numerical evaluation of hypothetical core disruptive accident in full-scale model of sodium-cooled fast reactor

A hypothetical core destructive accident (HCDA) has received widespread attention as one of the most serious accidents in sodium-cooled fast reactors. This study combined recent advantages in numerical methods to realize realistic modeling of the complex fluid–structure interactions during HCDAs in a full-scale sodium-cooled fast reactor. The multi-material arbitrary Lagrangian–Eulerian method is used to describe the fluid–structure interactions inside the container. Both the structural deformations and plug rises occurring during HCDAs are evaluated. Two levels of expansion energy are considered with two different reactor models. The simulation results show that the container remains intact during an accident with small deformations. The plug on the top of the container rises to an acceptable level after the sealing between the it and its support is destroyed. The methodology established in this study provides a reliable approach for evaluating the safety feature of a container design.

Development of a simple model for estimating the design limit of core void reactivity to prevent re-criticality of MOX-fueled cores in liquid metal-cooled fast reactors

The mixed oxide (MOX)-fueled core for liquid metal-cooled fast reactors has such high core performance and favorable neutron economy as to achieve an average fuel burnup of over 90 GWD/t with high power density. However, traditional MOX-fueled cores which are the ones without specific design measures to reduce the positive void reactivity worth has the potential to exceed prompt criticality during hypothetical core disruptive accidents (CDAs). Should the prompt criticality be exceeded, the core materials would heat up until their temperatures would exceed their boiling points. As a result, a release of significant mechanical energy cannot be ruled out, imposing serious damage on the reactor vessel. The released mechanical energy, termed “energetics,” is dominated by a rapid insertion of positive reactivity due to a fuel–coolant interaction (FCI) during the initiating phase of an unprotected loss-of-flow (ULOF) event. Considering the mechanism of energetics generation, a simple model was developed to estimate the design limit of positive component of core void reactivity against a ULOF event for traditional MOX-fueled cores. This simple model is constructed to evaluate the design limit based on the relation between the allowable positive reactivity insertion rate defined by the core characteristics and that assumed by the effect of FCI phenomena. The validity of the model is discussed by comparison with SAS4A which is the CDA analysis code devoted to calculate the initiating phase.

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.

Experimental study on fragmentation characteristics of molten aluminum/copper jet penetrating in sodium pool

Under Hypothetical Core Disruptive Accidents (HCDA) in Sodium-cooled Fast Reactor (SFR), Molten Fuel-Coolant Interaction (MFCI) will cause intensive fragmentation of molten fuel materials and violent sodium boiling, possibly resulting in recriticality due to sodium void effect and reactor core compaction. And the formation and sufficient cooling of core debris bed are significantly affected by the fragmentation of molten fuel materials in sodium pool. Due to approximate ambient Weber Number (Wea) with molten metallic fuel, kilogram-level molten copper jet is dropped into sodium pool to study the fragmentation characteristics under varied experimental parameters in the present experiments. The nucleate boiling of liquid sodium is estimated to occur at the instantaneous contact interface between molten copper and liquid sodium. The molten copper jet suffers from intensive fragmentation and solidification to produce many fragments under thermodynamic and hydrodynamic effects. Besides, experiment with molten aluminum jet injected into sodium pool is conducted without occurrence of sodium boiling to independently study the effect of hydrodynamic interaction on interphase penetration and fragmentation. During the experiments, temperature variations and dynamic pressure change in sodium pool are monitored to evaluate energy release. The resultant fragments are recorded and measured for morphology analysis and size distribution to examine the fragmentation characteristics. Based on experimental results, the fragmentation mechanism is discussed against hydraulic and thermal factors. The present findings contribute to further understanding the fragmentation characteristics of molten corium in sodium pool and optimization of mitigation systems against MFCI for SFR.

Cooling of heat generating core debris using multiple passive jets in a liquid metal pool of LMFBR

A 3D numerical study on coolability of heat generating core debris by multiple passive jets in a liquid metal pool has been carried out. The computational domain considered is a cylindrical enclosure, which resembles the lower plenum of a typical 500 MW LMFBR. The plenum has provisions to ensure the coolability of the core debris resulting from the hypothetical core disruptive accident. The arrangement avoids settlement of core debris on the main vessel and ensures safe retention of destroyed core on core catcher within the main vessel. The enclosure is filled with liquid sodium and is considered as an incompressible fluid. The transient form of governing equations for flow and heat transfer are solved numerically. The k−ε model is used for closing RANS equations while the pressure velocity coupling is done by PISO algorithm to obtain smooth convergence. The developed model is validated against benchmark results on natural convection in low Prandtl fluid available in the literature and also using in-house experimental data. The Rayleigh number based on heat generation rates is varied from 1012 to 1015. Different orientations and angles of passive cooling pipes with time dependent heat generation rate are analyzed. Three different bend angles viz 90°, 135° and 180° with bend orientations of inward, outward and opposite are considered. Also, the influence of central chimney on heat removal is brought out clearly. The fluid flow and heat transfer characteristics are described using isotherms, velocity vectors and in terms of Nusselt number. It is found that cooling pipes with 1800 bend having inward orientation provides favorable cooling capability over all other configurations. The central chimney is found to have profound influence in the heat removal from heat source. A useful correlation for Nusselt number is also proposed for thermal design purposes.

Probabilistic evaluation of the energetics upper bound during the transition phase of an unprotected loss of flow accident for a sodium cooled fast reactor by using a Phenomenological Relationship Diagram

Abstract One of the main research goals of the GEN-IV systems is enhancing their safety compared with the former Sodium-Cooled Fast Reactor (SFR) designs. A key issue is the capability of accidents prevention as well as of demonstrating that their consequences do not violate the safety criteria. In order to fulfill such requirements, risk analyses of severe core disruptive accidents are performed. Since the beginning of the SFR development, Hypothetical Core Disruptive Accidents (HCDAs) have played an outstanding role. Numerous safety analyses have been performed for developing and licensing past SFR designs and nowadays a large database of results is available. In particular, a large amount of results of the mechanistic SIMMER-II and SIMMER-III/IV analyses for various core designs and different power classes is available at the Karlsruhe Institute of Technology (KIT). The current paper describes the probabilistic approach based on the Phenomenological Relationship Diagram (PRD), which is used to evaluate the Probability Distribution Function (PDF) of the thermal energy release during the transition phase of an unprotected loss of flow accident scenario for a SFR. The technique allows taking into account the mechanistic nature of the accident scenario. In fact, the available results of the mechanistic analyses of HCDAs in SFRs are used to assess the PDFs of the dominant phenomena affecting the thermal energy release, which are propagated in the PRD by employing a Monte Carlo method.

Fragmentation of continuous molten copper droplets penetrating into sodium pool

ABSTRACT The progression of hypothetical core disruptive accidents (CDAs) in metal fuel cores is strongly affected by exclusion of molten metal fuel from the core region due to molten fuel–coolant interaction (FCI). As a basic study of FCI, the present paper focuses on the fragmentation characteristics of continuous molten copper droplets with a total mass from 20 to 50 g penetrating into a sodium pool. The results show that the fragmentation of the continuous molten copper droplets is sensitive to the change of the hydrodynamic and thermal conditions when the instantaneous contact interface temperature (Ti ) is lower than the turning point (Ttp ) and insensitive at Ti > Ttp . Compared with the fragmentation of a single droplet, the fragmentation of continuous droplets is accelerated and enhanced due to the collision between the droplets and the upward microjets. The present mass median diameter (Dm ) or dimensionless mass median diameter (Dm /D 0) of continuous copper droplets shows a distribution with smaller values than those of single copper droplet, and larger values than those of copper jets under similar thermal and hydrodynamic conditions. These results are promising to assure the termination of accidents in CDAs and useful to the core design with enhanced safety in FBRs.

New developments in modeling sodium spray fires with ASTEC-Na

One of the most significant containment-related safety issues during a severe accident at a sodium fast reactor is a spray fire, as the high energy ejection and combustion of sodium can cause a sudden, large spike in containment pressure, and can possibly cause the containment building’s early failure. This paper discusses the incorporation sodium spray fire models into ASTEC-Na, a simulation tool currently under development by IRSN for sodium fast reactor severe accidents. An extensive comparison to past experimental results is included, and a new empirical model that describes the combustion phenomena for large flow sodium spray fires is developed. This paper also extends ASTECNa to a hypothetical core disruptive accident, and discusses a fire confinement suppression technique.

Investigation on upper bounds of recriticality energetics of hypothetical core disruptive accidents in sodium cooled fast reactors

One key research goal for GEN-IV systems is an enhanced safety compared to the former Sodium Cooled Fast Reactor concepts. A key issue is built-in safety and the capability to prevent accidents and to demonstrate that their consequences do not violate aimed-at safety criteria. From the beginning of SFR development the Core Disruptive Accident (CDA) has played an outstanding role in the safety assessment. Under core disruptive accident conditions with core melting the fuel might compact, prompt criticality might be achieved and a severe nuclear power excursion with mechanical energy release might be the consequence. Numerous safety analyses accompanied the development and the licensing procedures of past fast reactor projects. A central issue of all analyses was the assessment of a realistic upper bound of energetics especially related to recriticalities in disrupted core configurations. Striving for an even higher safety level for next generation reactors a new strategy focused on the development and introduction of preventive and mitigative measures both to reduce the chance for a severe accident development and to mitigate its energetics. For assessing the effectiveness of these measures the knowledge of the CDA behavior is essential. In this context and on basis of new code developments, new experimental insights and extended studies for many reactor types of different power classes over the recent years, the issue of a realistic upper bound of energetics of the late core melt phases is again of relevance. Of special interest is the identification of natural and intrinsic mechanisms that limit the escalation of energetics. The current paper deals with these issues and tries to add supportive facts on the limits of CDA energetics. The evaluation of results of mechanistic SIMMER-II and SIMMER-III/IV analyses performed for various core designs and power classes and specific model case studies in 2D and 3D geometry indeed supports the idea of a limit of recriticality energetics. Intrinsic mechanisms exist, which limit the escalation energetics even in case of a strong blockage confinement suppressing any fuel discharge and allowing on-going sloshing recriticalities. In the light of the available information and taking into account relevant scientific publications and studies by the international community on the subject, one could conclude that an upper bound for energetics in the range given in the paper can be deduced.

Probabilistic Evaluation of the Post Disassembly Energetics of a Hypothetical Core Disruptive Accident in a Sodium-Cooled SMR by using a Phenomenological Relationship Diagram

An innovative probabilistic approach based on the Phenomenological Relations Diagram is developed to perform risk analyses of Hypothetical Core Disruptive Accidents in Sodium Fast Reactors (SFRs). The novel method is applied to evaluate the probability distribution of the thermal-to-mechanical energy conversion ratio in an ULOF/Post-Disassembly Expansion Phase scenario in a small- to medium- size SFR. The results provide a comprehensive picture of the fuel energy distribution in the system.

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in the relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. It could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling.

Visual study of ex-pin phenomena for SFR with metal fuel under initial phase of severe accidents by using simulants

ABSTRACTIn the present Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metal fuel within a subchannel is suggested as an inherent safety strategy in the initiating phase of a hypothetical core disruptive accident (HCDA). This safety strategy provides a negative reactivity driven by the melt dispersion; therefore, it could reduce the possibility of occurrence of a severe recriticality event. In the initiating phase, the melt could be injected into the subchannel horizontally by the internal pressure of the fuel pin. Complex phenomena occur during intermixing of the melt with the coolant after the horizontal injection of the melt. It is rather difficult to understand the several combined mechanisms that occur that are related to the dispersion and fragmentation of the melt. Thus, it seems worthwhile to study the horizontal injection of melt at lower temperatures, which could help to observe the dispersion phenomenon and understand the fragmentation mechanism. In this work, for a parametric study, tests were performed under structural conditions, coolant void conditions, and boiling conditions. As a result, in some cases, the injected molten materials were stuck around the injection hole. On the other hand, the molten materials were dispersed upward sufficiently well under the boiling condition when R123 was used as the coolant. The built-up vapor pressure was found to be one of the driving forces for the upward dispersion of the molten materials.

Evaluation of recriticality behavior in the material-relocation phase for Japan sodium-cooled fast reactor

As the most promising concept of sodium-cooled fast reactors, the Japan Atomic Energy Agency has selected the advanced loop-type fast reactor, so-called Japan sodium-cooled fast reactor (JSFR). Through the evaluation of event progressions during hypothetical core-disruptive accident (CDA) under the design extension condition, a CDA scenario for JSFR has been evaluated. It has already been demonstrated that in-vessel retention (IVR) against CDA could be achieved by taking adequate design measures under best estimate conditions.The whole sequence of CDA scenario for JSFR was categorized into four phases according to the progress of core-disruption status. In the third phase, so-called material-relocation phase, the accident events would progress in the subcritical state. However, if the uncertainties about the molten state of core remnant and their discharge behavior outward from core are conservatively superposed, the disrupted core may lead up to recriticality.In the present study, the factors leading to recriticality in the material-relocation phase were investigated using the phenomenological diagrams, and the recriticality behaviors were evaluated through parametric analyses using SIMMER-III/IV codes. The results of parametric analyses suggested that a significant mechanical energy leading to the boundary failure of reactor vessel would not be released even assuming recriticality due to the uncertainties about molten state and discharge behavior. Through the present evaluation of the hypothetical recriticality event, the CDA scenario for JSFR could obtain further robustness from the viewpoint of achieving IVR.

On the Multiple-Pin Modeling of the Fuel Bundle for the Simulation of the Initiating Phase of a Severe Accident in a Sodium Fast Reactor

Numerical simulations of the primary phase of a hypothetical core disruptive accident are currently based on a multiple-channel approach, which requires that subassemblies or groups of subassemblies be represented together as independent channels. Generally, a single-pin treatment is used to model the channel fuel pins. The limitation of this simplified approach should be assessed because it can affect voiding and melting patterns that in turn may influence reactivity insertions and power history. In the same manner, the single-pin hypothesis may introduce important biases in the prediction of can-wall thermal ablation. Radial propagation of the degradation and subsequent accident consequences may thus be affected. To improve the safety assessment of sodium fast reactors, two-dimensional effects are investigated using a multiple-pin model. Numerical results for a severe accident transient show that the current methodology is nonconservative and predicts the onset of sodium boiling with a delay. A two-node radial meshing of the subassembly is preferred for treating the peripheral ring of fuel pins separately from the rest of the pins. This treatment would allow overcoming the previous issue and give more accurate initiating phase simulations.

An Assessment of Corewide Coherency Effects in the Multichannel Modeling of the Initiating Phase of a Severe Accident in a Sodium Fast Reactor

To consolidate the safety assessment for liquid-metal fast breeder reactors (LMFBRs), hypothetical core disruptive accident (HCDA) sequences have been extensively studied over the past decades. Numerous analyses of the so-called initiating phase (or primary phase) of a HCDA have been made with the safety analysis system code SAS4A. The SAS4A accident analysis code requires that subassemblies or groups of subassemblies be represented together as independent channels. For simulating a severe accident sequence, a subassembly-to-channel assignment procedure has to be implemented to produce the consistent SAS4A input decks. Generally, one uses imposed criteria over relevant reactor parameters to determine the subassembly-to-channel arrangement. The multiple-assembly-per-channel approach introduces corewide coherency effects, which can affect the reactivity balance and therefore the overall accident development. In this paper, a subassembly-to-channel assignment procedure based on the subassembly power-to-flow ratio is presented and implemented to generate the SAS4A input decks over a range of parameter values. The corresponding SAS4A calculations have been performed on a large LMFBR. The purpose of the present series of calculations is to investigate the magnitude of errors encountered in the analysis of the initiating phase related to the subassembly-to-channel arrangement selection, by comparison with a one-subassembly-per-channel reference solution. It appears that a refinement in the channel arrangement substantially reduces corewide coherency effects. Analysis of the calculations also suggests that an accurate representation of the scenario requires the number of channels to be on approximately the same order of magnitude as the total number of subassemblies. Numerical results are examined to provide the reader with quantitative measurements of bias related to subassembly-to-channel arrangement.

Critical trajectories for aerosol particles: Their determination for impaction in fibrous filters and in oscillating bubbles

Critical trajectories for aerosol particles in a gas flow are the ones which divide an aerosol flux into different parts, for example aerosol which is, and is not deposited. They can exist in all gas flows in which aerosol motion is governed by gas velocity rather than by diffusion and we describe two mathematical methods for their calculation. For deposition by impaction on a filter fibre it is necessary to solve the differential equations for particle motion and an efficient iterative procedure is used to obtain the critical trajectories.Jonas and Schütz (1988) have shown that aerosol impaction is an important mechanism for the removal of aerosol from an oscillating sodium vapour bubble formed during a hypothetical core disruptive accident in a fast reactor. For these one-dimensional oscillations, when the gas velocity within a bubble is a linear function of position, we extend their work by calculating critical trajectories directly from the integral equation describing a depositing particle for two models with different initial conditions. With initially entrained uniform aerosol, the percentage impacted is independent of the inclusion of gravity in the calculations as long as regions empty of aerosol do not appear in the bubbles. Numerical results are obtained for a wide range of amplitudes of bubble oscillations and aerosol in the size range 1-30μm. In agreement with Jonas and Schütz, we find that a considerable fraction of the aerosol at larger sizes is removed by impaction. The theory is also shown to apply to other types of bubble oscillation including those of a spherical bubble.

Evaluation of debris bed self-leveling behavior: A simple empirical approach and its validations

During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of core material pool. However, coolant boiling may lead ultimately to leveling of the debris bed that is crucial to the relocation of molten core and heat-removal capability of debris bed. To clarify the mechanisms underlying this self-leveling behavior, several series of experiments were elaborately designed and conducted in recent years under the collaboration between Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). Based on the experimental observations and quantitative data obtained (mainly the time variation of bed inclination angle), a simple empirical approach to predict the self-leveling development depending on particle size, particle density and gas velocity was proposed. To confirm the rationality and wide applicability of this approach, over the past few years extensive efforts have been made by performing modeling investigations against a large number of experimental data covering various conditions, including difference in bubbling mode, bed geometry and range of experimental parameters. The present contribution synthesizes these efforts and gives detailed comparative analyses of the performed validations, thus, providing some insight for a better understanding of CDAs and improved verifications of computer models developed in advanced fast reactor safety analysis codes.

CHARACTERISTICS OF SELF-LEVELING BEHAVIOR OF DEBRIS BEDS IN A SERIES OF EXPERIMENTS

During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process, such as boiling mode, particle size, particle density, particle shape, bubbling rate, water depth and column geometry, were investigated, thus giving a large palette of favorable data for the better understanding of CDAs, and improved verifications of computer models developed in advanced fast reactor safety analysis codes.