Incoherency effects in clad relocation dynamics for LMFBR CDA analyses

Reactivity feedbacks owing to clad relocation during a core disruptive accident (CDA) are of primary importance in LMFBR safety. A calculation of these feedbacks would require a knowledge of clad dynamics. An experimental and theoretical investigation of the flooding phenomenon (which is relevant to clad dynamics) was carried out. The experimental data indicate that flooding velocities agree well with Wallis's correlation. An extension of this correlation to LMFBR conditions gave rise to a set of flow regime maps, which allow a first order estimate of cladding velocities and could be utilized for evaluating reactivity feedbacks.     

On certain aspects of alternative core disruption accidents in LMFBRs

This paper examines the initial course of a postulated accident scenario in an LMFBR, involving the rupture of all piping connected to the reactor vessel, in the event of an earthquake (or an equivalent scenario involving both loss of heat removal and system rupture). The core is successfully shut down, but decay heat imposes a threat to core integrity.At the onset of pipe rupture, the fuel elements are cooled by natural circulation, followed by subcooling and nucleate boiling. Continuation of sodium evaporation leads to core dryout, clad melting, and subassembly wall failure. Clad melting is found to occur at a location close to the top of the core at a rate of 0.08 ft³/s. The sodium vapor velocity is not high enough to carry the molten steel to the upper blanket region; therefore, flow-channel blockage in the lower axial blanket region is expected.At the time of clad melting, the fuel temperature rises by approximately 2°F/s, while the temperature-rise rate at the can wall, due to heat radiation from the fuel pin, is 10°F/s. Failure of the can wall initiates gross fuel motion. Continuation of the heat generation in the fuel pellet leads to the melting of the control rod support. Both the fuel motion and the control rod failure mark the start of reactivity insertion.     

Analysis of hypothetical overpower accidents in the SNR-300 Mark 1A core with modeling of failure incoherence

Incoherency effects due to the statistical nature of fuel rod failure have been assessed for transient overpower (TOP) accidents with different initiating ramp rates. Characteristic single subassembly feedbacks from sodium voiding, fuel injection, and fuel sweep-out after failure were calculated and superimposed incoherently. Effective feedback functions were thus generated for characteristic subassembly groups (channels) and used as tabular reactivity input to corresponding multi-channel whole core TOP simulations using the HOPE computer model. Analyses were performed for the SNR-300 Mark 1A end of life (EOL) core with initiating ramp rates of 0.15, 0.5 and 5 $/sec. The results are compared with those from a calculation with coherent subassembly behavior in each channel. It is concluded for the mild transient that irrespective of the assumed axial failure location, negative fuel sweep-out reactivity from early failed subassemblies will prevent the accident from reaching a prompt critical stage and lead to a gradual shutdown if the assessed failure incoherencies are applied. For the higher ramp rates (0.5 $/sec) the assessed incoherencies will not always lead to an early shutdown, but they can considerably reduce the potential to accumulate high ramp rates in the initiating phase of TOP accidents.     

Hydrodynamics of post-disassembly fuel expansion

The early expansion of the fuel following disassembly in an LMFBR core disruptive accident is modeled. Spherical expansion in the sodium is assumed. A Lagrangian, finite-difference hydrodynamic code (FEXPAN) describes the motion. Disassembly employs VENUS-II, and a consistent equation of state for fuel was used throughout disassembly and FEXPAN. Time-dependent mechanical work and fuel vaporized without fuel mixing are obtained. FEXPAN is compared with time-independent expansion for the effect of fuel mixing. For example, for a particular accident analysed 75 MW sec mechanical work was calculated for expansion with no mixing versus 10 MW sec with complete core mixing.     

Fuel modelling in fast reactor whole core accident calculations

Transient fuel pin behaviour and fuel motion models are of major importance for the analysis of the initiation phase of unprotected whole-core LOF and TOP accidents in LMFBRs. The role played by these models is highlighted by discussing LOF and TOP accident sequences. This is followed by an overview of present whole-core fuel pin behaviour and fuel motion models and also a critical evaluation of these models.     

Analysis of reactivity induced accidents at Pakistan Research Reactor-1

Analysis of reactivity induced accidents in Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel, has been carried out using standard computer code PARET. The present core comprises of 29 standard and five control fuel elements. Various modes of reactivity insertions have been considered. The events studied include: start-up accident; accidental drop of a fuel element on the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results reveal that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is concluded that the reactor, which is operated safely at a steady-state power level of 10 MW, with coolant flow rate of 950 m³/h, will also be safe against any possible reactivity induced accident and will not result in a fuel failure.     

Energetics of simulated HCDA bubble expansions: Some potential attenuation mechanisms

Experiments conducted to increase our understanding of the dynamics and thermodynamics of expanding bubbles similar to the core disruptive accident (CDA) bubble in liquid metal fast breeder reactors (LMFBR) are described. The experiments were conducted in a transparent model of a typical demonstration-size loop-type LMFBR in which water at room temperature simulated the sodium coolant. Nitrogen gas (1450 psia) and flashing water (1160 psia) qualitatively simulated sodium vapor and molten fuel expansions. Three physical mechanisms that may result in attenuation of the work potential of a hypothetical CDA (HCDA) were revealed by the experiments: (1) the pressure gradient existing between the lower core and the bubble within the pool, (2) the hydrodynamic effects of vessel internal structures, and (3) the nonequilibrium flashing process occurring in the lower core. These three mechanisms combine to result in a coolant axial slug kinetic energy that is only 14% of the work potential of the ideal quasi-static nitrogen expansion and only 5% of the work potential of the ideal quasi-static flashing water expansion.     

CALIPSO—A computer code for the calculation of the thermodynamics, fluiddynamics and geometry changes following pin failure

The computer code CALIPSO calculates the thermodynamics and fluid-dynamics of fuel, fission gas and coolant as well as changes in geometry subsequent to pin failure in an anticipated liquid-metal fast breeder reactor (LMFBR) accident. In the documented version, CALIPSO is well suited for the analysis of the out-of pile SIMBATH experiments carried out at Kernforschungszentrum Karlsruhe (KfK) which simulate the above-mentioned accident with thermite technology. In two-dimensional geometry the fuel pin and its associated coolant channel, initially separated by the fuel cladding are treated as a single fluid domain. The conservation equations of mass, momentum and energy are solved separately for each component. The transient evolution of the temperature profile in the cladding is modeled in detail, thus permitting the analysis of various phase transition processes (melting, freezing and clad failure propagation). The coolant channel has a variable cross section and it is surrounded by an outer channel wall for single pin experiment analysis. Axially the coolant channel is connected to a simplified model of the whole sodium loop.     

Studies on Sodium Boiling Accidents in a Large Fast Breeder Reactor, (II) Radial Propagation of Sodium Boiling by Reactivity Effect

A computer code BORE was developed, with which analyses were performed on channel plugging accidents that would occur on a 1,500 MWe LMFBR. The BORE code calculated the dynamic characteristics of coolant boiling and fuel failure propagation radially in the core, and the requirements of core instrumentation systems were also analyzed. The results show that coolant boiling and/or fuel failure in a channel plugging accident are propagated only to a limited number of adjacent channels when sensors are installed that detect anomalies in channel flow, channel outlet temperature, boiling or reactivity. It is also concluded that the coolant void effect is not serious from the standpoint of safety when the time required for boiling propagation to adjacent channels can be made longer than 0.15 sec.     

Thermal-hydraulic modeling for the analysis of LMFBR disrupted core behavior

This paper presents the analytical models of thermal-hydraulic phenomena of major interest in the analysis of LMFBR hypothetical core disruptive accidents. These models have been incorporated in LEVITATE [1], a code for the analysis of fuel and cladding dynamics under loss of Flow (LOF) conditions. LEVITATE has recently become part of the SAS4A [2] code system, replacing the older, less sophisticated SLUMPY [3] model.The influence of different thermal-hydraulic models on fuel motion is illustrated by a comparison between the results calculated by LEVITATE, the data from the L7 and L6 TREAT experiments [4] and the results calculated by SLUMPY. The results calculated by LEVITATE are in good agreement with the experimentally observed early fuel dispersal.     

Thermal hydraulic and safety analysis for core conversion (HEU–LEU) of Syrian Miniature Neutron Source Reactor

The paper presents the behavior and properties analysis of the low enriched uranium fuel compared with the original high enriched uranium fuel. The MNSR reactor core was modeled with both fuel materials and the reactor behavior was studied during the steady state and abnormal conditions. The MERSAT code was used in the analysis. The steady state thermal hydraulic analysis results were compared with that obtained from the experimental results hold during commissioning the Syrian MNSR. Comparison with experimental data shows that the steady-state behavior of the HEU core was accurately predicted by the MERSAT code calculations. The validated model was then used to analyze LEU cores with two proposed UO₂ fuel pin designs. With each LEU core, the steady state and 3.77 mk rod withdrawal transient were run and the results were compared with the available published data in the literatures for the low enriched uranium fuel core. The results reveal that the low enriched uranium fuel showed a good behavior and the peak clad temperatures remain well below the clad melting temperature during reactivity insertion accident.     

A subchannel analysis code MATRA-LMR for wire wrapped LMR subassembly

In sodium cooled liquid metal reactors design limits are imposed on the maximum temperatures of the cladding and fuel pins. Thus an accurate prediction of the core coolant/fuel temperature distribution is essential to LMR core thermal hydraulic design. The detailed subchannel thermal hydraulic analysis code MATRA-LMR is being developed for LMFBR core design and analysis based on COBRA-IV-I and MATRA. The major modifications and improvements implemented in MATRA-LMR are as follows: sodium property calculation subprogram, sodium coolant heat transfer correlations, and most recent pressure drop correlations. To assess the development status of this code, benchmark calculations were performed with the ORNL 19 pin tests and EBR-II seven-assembly SLTHEN calculation results. The calculation results of MATRA-LMR were compared to the measurements and to the SABRE4 and SLTHEN code calculation results, respectively. Finally, the major technical results of the conceptual design for the KALIMER U-10%Zr binary alloy fueled core have been compared with the calculations of the MATRA-LMR, SABRE4 and SLTHEN codes.     

The coupling algorithm between fuel pin and coolant channel in the european accident code EAC-2

In the field of fast breeder reactors the Commission of the European Communities (CEC) is conducting coordination and harmonisation activities as well as its own research at the CEC's Joint Research Centre (JRC). The development of the modular European Accident Code (EAC) is a typical example of concerted action between EC Member States performed under the leadership of the JRC. This computer code analyzes the initiation phase of low-probability whole-core accidents in LMFBRs with the aim of predicting the rapidity of sodium voiding, the mode of pin failure, the subsequent fuel redistribution and the associated energy release.This paper gives a short overview on the development of the EAC-2 code with emphasis on the coupling mechanism between the fuel behaviour module TRANSURANUS and the thermohydraulics modules which can be either CFEM or BLOW3A. These modules will also be briefly described. In conclusion some numerical results of EAC-2 are given: they are recalculations of an unprotected LOF accident for the fictitious EUROPE fast breeder reactor which was earlier analysed in the frame of a comparative exercise performed in the early 80s and organised by the CEC.     

Analytical study on elimination of severe recriticalities in large scale LMFBRS with enhancement of fuel discharge

The possibility of severe recriticality could be excluded if the molten core materials are discharged from reactor core in the early stage of core disruptive accident (CDA). Based on this idea, several design measures for future commercial liquid metal-cooled fast breeder reactors (LMFBRs) have been proposed to enhance the molten fuel discharge from core in order to prevent formation of the core-wide molten pool with high mobility. One promising concept in these design candidates is modified-FAIDUS (Fuel subassembly with Inner DUct Structure). The event progression in unprotected loss of flow (ULOF) accident in a sodium-cooled large scale FBR with modified-FAIDUS was analyzed to assess the effectual performance of modified-FAIDUS in preventing severe recriticality using the SAS4A and SIMMER-III codes. Two parametric cases were performed covering the uncertainty of duct wall failure mechanism, one with stable fuel crust and another with unstable crust condition. The calculation showed that the final amount of discharged fuel from core in both cases was more than 20% of initial core inventory. The degraded core after fuel discharge is composed of the mixture of solidified fuel, swollen fuel chunks and molten steel, of which low mobility prevents massive fuel motion. The reactor power lowered to decay heat level and the reactivity lowered around −20 $, thus, the possibility of severe recriticality was eliminated.     

The steel behaviour in CABRI experiments: Interpretation of post test examinations

Non-destructive and destructive PTE of CABRI test sections provided a lot of experimental results allowing to elucidate clad/steel behaviour during the CABRI tests. It seems now well-stablished that the steel may play an important role during an unprotected accident in an LMFBR.

The most important phenomena discussed in this report are clad relocation, clad ballooning, and clad ablation, fuel/steel mixing and redistribution of metallic fission products toward the steel phase. Fuel bulk freezing caused by fuel/steel mixing was an important mechanism for blockage formation.     

Uncontrolled withdrawal of a control rod without scram

In the present work the thermal-hydraulics of reactivity-induced transients in low enriched uranium (LEU) core of a typical material test research reactor (MTR) are analyzed using the previous program developed by Khater et al. The analysis was done for uncontrolled withdrawal of a control rod with scram-disabled conditions. Initiating reactivity events with and without the influence of reactivity efficiency curve (“S” curve) were considered. The results of the proposed transients are analyzed and compared with each other. In transient without the “S” curve influence, a high primary peak power of 406.18 MW is attained and a clad melt down takes place after 1.85 s. In the transient with the “S” curve influence, a high super prompt-critical situation is produced (1.762$ at 0.895 s) with a very high primary peak power of 801.05 MW at 0.912 s. Also, a fast clad melt down is resulted in the hot channel at 1.088 s and a stable film boiling is established. This study indicates that, compared to the application of linear reactivity curve, the application of the reactivity efficiency curve results in the prediction of higher peaks in power and temperatures (fuel, clad and coolant) with a fast clad melt down.     

Development of a neural simulator for research reactor dynamics

A research reactor simulator was designed and developed by neural networks model. This simulator can predict the reactor power and temperatures (fuel, clad and coolant) in normal and accident condition considering reactivity feedbacks. The main advantage of this method, as compared with custom calculational methods (simulation with PARET and RELAP) is real time simulation without the need for much skilled or experienced setting. The response of benchmark reactor core predicted by neural simulator was compared to that obtained from PARET code and close agreement was observed.     

Simulation of the transient behaviour of LMEBR fuel pins under consideration of special burnup phenomena using the BREDA-II model

A new fuel pin model was developed to describe the influence of specific burnup phenomena on the behaviour of fuel pins under transient overpower conditions in a liquid metal fast breeder reactor (LMFBR). It has been used for transient fuel pin deformation analysis during hypothetical core disruptive accidents (HCDA) and for the purpose of interpreting fuel pin failure tests. The fuel pin model, designated as BREDA-II, is based on the equations of the quasi-static theory of thermal elasticity. The fuel is regarded as elastic and the cladding as elasto-plastic material. The equations for the stress-strain analysis are based on the plane strain approximation. A multiregion fuel pin model allows to simulate long-time and transient burnup phenomena. The long-time effects taken into account are the steady state swelling of fuel, the change in fuel porosity and the production and partial release of fission gases. During a power excursion transient fuel swelling and pressure increase due to transient fission gas behaviour are included in the deformation analysis. Potential fuel pin failure is indicated by the application of various criteria of failure. In subsequent model calculations the behaviour of an irradiated LMFBR fuel pin during an overpower transient corresponding to a reactivity ramp of $5/sec is simulated and interpreted from the point of view of reactor safety.     

The application of the simple fuel pin transient and melting model to thermal-hydraulics in LMFBR subassembly

A simple conduction model with phase change has been developed for the transient analysis of a reactor fuel pin based on lumped parameter techniques. The purpose of this analysis is to provide a simple useful tool to obtain the general information about fuel and clad leaning into the cooling transients and melting. Such a simple fuel and clad thermal transient model is particularly useful to multichannel analysis where conventional conduction computer codes require considerable computing time and storage space. At the present, this formulation is being employed for the analysis of sodium thermohydraulics, sodium voiding, and melting of cladding and fuel in a subassembly of a fast reactor core. A detailed analysis of the predicted coolant, fuel and cladding thermal transients leading into sodium voiding and fuel pin melting has been made in comparison with the results of various in-pile experiments and with the predictions from the existing more complicated codes.     

First 3-D calculation of core disruptive accident in a large-scale sodium-cooled fast reactor

The SIMMER-IV computer code is a three-dimensional fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. The present study has attempted the first application of SIMMER-IV to a core disruptive accident in a large-scale sodium-cooled fast reactor. A principal point of this study was to investigate reactivity effects with fuel relocation under three-dimensional core representation including control rods. The calculation has indicated that the fuel discharge from the core was disturbed by a significant flow resistance at the entrance nozzle in the current design. Additional static neutronic calculations have been performed to compare basic neutronic characteristics between different scale cores. The static neutronic calculations have clarified that the outward fuel compaction within the inner core increased the reactivity in the large-scale core unlike the small-scale core.