Design Of Sodium-cooled Fast Reactor Research Articles

Multi-physics code coupling methodology for the simulation of accidental transients in sodium cooled fast reactors

The French Alternative Energies and Atomic Energy Commission (CEA) has played a substantial role in advancing fourth-generation reactor technologies, including Sodium-cooled Fast Reactors (SFR), which are anticipated to offer enhanced efficiency, adaptability, and sustainability compared to existing nuclear technologies. In the context of safety assessment, it is crucial to precisely simulate accidental scenarios like Unprotected Loss Of Flow (ULOF) transients. These transients entail intricate multi-physics interactions within the reactor, particularly within the core, encompassing thermal-hydraulics, reactor physics, and fuel performance (e.g. changes in densities or temperatures will lead to neutronics effects, thus power, then causing feedbacks onto said densities and temperatures).In order to capture those complex phenomena, a multi-physics coupling approach for SFRs, which combines the CATHARE3 code for system scale thermal-hydraulics, the APOLLO3 code for reactor physics, and the GERMINAL V2 code for nuclear fuel performance has been developed at CEA in the recent years.This paper presents a comprehensive methodology of such coupling and its integration into a Verification, Validation and Uncertainty Quantification (VVUQ) process. The first part of the paper provides an introduction to the simulation tools used for the coupling. The following section explains the multi-physics coupling approach, facilitated by CEA’s Collaborative Code Coupling Platform (C3PO), which enables code integration through a Python interface to create a comprehensive simulation model. A detailed examination of the coupling algorithm and data transfer mechanisms is also provided.The generality and reliability of this approach are demonstrated by simulating two accidental transients: the Unprotected Loss Of Flow (ULOF) in the ASTRID reactor and the Loss Of Flow WithOut Scram (LOFWOS) in the Fast Flux Test Facility (FFTF), both corresponding to a primary pump trip without control rods falling. The results show that the proposed multi-physics coupling scheme is effective and robust for SFR simulations. The coupled model enables a more accurate and realistic representation of the complex phenomena occurring in the reactor during an accident.This research underscores the significance of multi-physics coupling in the analysis of SFRs and serves as a valuable point of reference for future SFR studies. The proposed approach offers an efficient way to investigate SFR responses during accidents or incidents, thereby advancing the development of dependable and precise simulation tools for SFR design and evaluation.

Analysis of the thermal-mechanical behavior of SFR fuel pins during fast unprotected transient overpower accidents using the GERMINAL fuel performance code

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered.Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up.The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

Experimental investigation on the heat transfer characteristics of argon space containing sodium vapor

A large quantity of argon is typically provided at the top of the main vessel of a sodium-cooled fast reactor (SFR) to serve as covering gas. The sodium coolant in an SFR evaporates and mixes with the argon covering gas during reactor operation. This mixed gas has a critical influence on the heat transfer at the top of the main vessel, particularly the temperature of the roof slab. The temperature distribution in the roof slab significantly affects reactor safety as it serves as the pressure boundary of the main vessel. Research on the behavior of sodium vapor has typically employed numerical simulations, and any experimental verifications have generally used simplified experimental models. To study the heat transfer characteristics of sodium vapor and eliminate the influence of simplifications on the results, the Heat transfer Experimental Facility for Argon space at the top of the main vessel (HEFA) was established based on a scaled-down top of the main vessel. The HEFA was employed to analyze influence of sodium vapor on the heat transfer characteristics of the reactor space by evaluating the effects of various heating temperatures with and without liquid sodium on the temperature distributions in the argon and roof slab. Furthermore, the emissivity of the liquid sodium surface was measured. The results show that the radiative absorption effect of sodium vapor primarily occurs near the liquid sodium surface. When the heating temperature was less than 400 °C, the presence of sodium vapor increased the temperature of the main vessel. When the heating temperature was 540 °C, the radiative absorption effect of the sodium vapor was enhanced, reducing the roof slab temperature. The observed phenomena can be considered to ensure reactor safety during the design of SFRs.

Safety-oriented SFR core design: Methodology and application to transient bifurcations and stability maps analyses

Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R&D program of CEA (French Alternative Energies and Atomic Energy Commission), a methodology is proposed to early consider safety requirement in the undergoing reactor design process. Before the use of mechanistic tools (CATHARE, SIMMER, EUROPLEXUS, etc.) whose input deck elaboration requires an advanced knowledge of the reactor design, the methodology proposed in this article involves several physical tools simulating phenomena likely to govern the choice of design parameters. These tools are mostly based on reduced order models (ROM), meaning that they mainly involve low-dimensional modelings (mostly 0D and 1D) that are validated versus experimental results. They are gathered in a platform that covers all kind of accidental phenomenology, from the initiator (pump trip, reactivity insertion, local flow blockages, etc.) until the reach of a stable and coolable state after corium relocation. Thus, enabling a large number of simulations in a reasonable computational time, this fast-running platform makes possible the characterization of some major accident transient bifurcations (such as boiling onset, boiling stabilization, primary power excursion, molten fuel vaporization, corium axial relocation in transfer tubes, etc.), in terms of probability of occurrence and of consequences on the transient evolution. It also enables to identify the main physical parameters causing the bifurcations in order to allow a straight feedback on the core design and to give some orientations for future R&D studies. In this article, a focus is firstly made on some scenario bifurcations to illustrate the platform capabilities. The boiling onset and possible reactor state stabilization are studied. The possibility of primary power excursion depending on the core design and the fuel vaporization possibilities are assessed. This platform is also of great help in order to rapidly compare several core designs when facing accidental transients. An example is given by comparing a CADOR core design to an ASTRID-like core design through flow stability maps in natural circulation conditions obtained by the fast-running tool platform. These applications demonstrate the efficiency of the presented methodology integrating safety at the very first design stages, and at facilitating the safety-oriented design of SFRs.

Experimental study on interface characteristics of liquid sodium-water vapor interaction in SWR

When the sodium-water boundary in the sodium-cooled fast reactor (SFR) fails, water vapor will enter the liquid sodium and cause the sodium water reaction (SWR) accident. SWR may lead to many serious consequences for power plant. Thus, it is a non-negligible factor to be considered in the design and operation of SFR. In this study, experiments on the interaction between liquid sodium and water vapor were carried out. The reaction phenomena, temperature of liquid sodium, morphology of reaction product and other characteristics of the chemical were summarized and analyzed in this paper. The experiment was conducted under different initial sodium temperatures (300–450 °C) and different mass fraction of water vapor (0.1–0.9 at 200 °C) conditions. Visualization scheme was used in this experiment and the reaction process was recorded by a high-speed camera. The results indicate that both mass fraction of water vapor and initial temperature of liquid sodium are important factors in the reaction process. The mass fraction of water vapor affects the actual reaction rate. The initial temperature determines the form of reaction products (liquid NaOH) on the surface of liquid sodium. For the experiments with initial temperature below 400 °C, the reaction products appear as bubbles and foams. While the initial experiment temperature exceeds 400 °C (including 400 °C), bubbles and foams gone, replaced by a thin film. By analyzing the conditions of bubble generation, the surface tension of liquid NaOH is an important cause, which can be affected by the temperature and bounding water molecules. Furthermore, according to the measurement results of sodium temperature, the reaction products in the form of a large concentration of foams have a strong blocking effect on mass transfer and greatly slow down the actual reaction rate. This research has important reference significance for the study of the mechanism of liquid sodium and water vapor reaction.

Numerical simulation of operational transients in sodium-cooled fast reactors

Reactor startup and shutdown are planned plant transients that are expected numerous times in the life of a nuclear reactor. These transients are desired to be carried out quickly for economic benefits and at the same time safely, which demands slower execution. The various plant design and safety constraints to be met during the execution of these transients in a typical medium-sized sodium-cooled fast reactor (SFR) were discussed. Operational procedures for the execution of these transients were reported in detail. A “wait and raise” mode for the startup and a “wait and lower” mode for the shutdown were adopted. During these procedures, primary and secondary loop flow rates were decided to be maintained at a nominal 100 % flow to avoid thermal stratification issues in sodium pools. Feedwater flow will be controlled based on sodium temperature at the steam generator (SG) outlet. The dynamics analysis for a reference SFR design was then carried out using plant dynamics code DYANA-P. The modeling details and approximations for simulating the startup and shutdown transients using the DYANA-P code were discussed. Two operational parameters, namely, control rod movement distance and waiting time between control rod movements, were optimized as 2.3 mm and 2.75 min, respectively, for the reactor's beginning of equilibrium cycle (BOEC) core configuration. The startup and shutdown durations were both estimated as ∼20.5 h each. The hot pool temperature change rate was estimated as ∼19 °C/h, which is less than the allowable limit of 20 °C/h. The SCRAM parameters, namely, reactor period and reactivity, were less than their alarm limits, thus, not interfering with the planned operations. The design and operational constraints of various plant systems were also satisfied. Parametric studies were also carried out for a few parameters: the fuel-clad interaction and the core burnup conditions. The assumptions of fuel being stuck to the clad (in feedback reactivity calculation) and the BOEC core configuration gives a higher power raising rate for a given control rod stroke and waiting time.

Versatile Test Reactor Conceptual Core Design

The VTR is a 300-MW(thermal) sodium-cooled fast reactor (SFR) designed for the specific purpose of delivering unique testing capabilities to enable the advancement of all reactor technologies. With its flux level, irradiation volume, and operational flexibility, the VTR will enable accelerated testing of materials, fuels, and various components needing irradiation testing. Proven SFR technologies and design approaches have been leveraged in designing the VTR core, ensuring the highest possible readiness level. This resulted in the VTR using ternary metallic fuel and delivering fast flux levels in excess of 4 × 1015 n/cm2·s over large useful volumes, corresponding to about 60 dpa/year in steel. As part of the design efforts, the VTR core performance has been determined for a representative configuration, ensuring that the reactivity control systems offer sufficient shutdown margins, that the core can be safely cooled in all situations, and that reactivity feedback coefficients are conducive to a favorable safety behavior. Furthermore, the incorporation of features such as fuel assembly storage in the shield region supports the flexible and reliable operation of the VTR. Additional design work has been ongoing as well. This includes thorough shielding performance evaluations to ensure safe operation of the VTR, verification and validation of the design tools used to achieve compliance with Nuclear Quality Assurance (NQA-1) requirements, early assessment of the impact of irradiation experiments on the core performance envelope and associated margins, and in-depth uncertainty quantification efforts to quantify the anticipated range of performance characteristics. An experimental program supporting the VTR core design has been set up, with the current focus being on thermal-hydraulic experiments. The purpose of this experimental program is to obtain confirmatory measurements to serve directly as part of the core design basis or as part of the validation cases supporting the simulation tools used.

Transient analysis of the main heat-transfer system of the pool-type sodium-cooled fast reactor

<p indent="0mm">Sodium-cooled fast reactors are the first choice of the fourth generation of advanced nuclear energy systems. They have attracted significant attention in the development of advanced nuclear energy technology owing to their advantages of proliferative nuclear fuel, transmutation of radioactive waste, and high thermal efficiency power generation. The design and construction of the secondary sodium-cooled fast reactor are ongoing in China. The physical characteristics of fast neutrons and the special thermal, physical, and chemical properties of liquid sodium metal cause special and complex scientific problems in the design of sodium-cooled fast reactors. In this study, taking the energy transfer process of sodium-cooled fast reactors as the object, we establish mathematical and physical models for analyzing large-scale transients in various energy transfer and conversion processes. The model is verified and validated by the natural circulation test of the Phoenix reactor and the international benchmark of the fast flux test facility loss in flow tests. The trend of the calculated values is consistent with that of the test data, and the results are in good agreement. For a typical pool-type sodium-cooled fast reactor, four types of transient events in the power plant are determined by engineering evaluation and deductive analysis. Among them, there are nine normal operation transients, 23 expected transients, 10 abnormal transients, and 7 emergency transients. Based on the established mathematical and physical models, the analysis results of each type of typical transient are given, which provide an important basis for the design of the system and equipment/components and lay a scientific and technological foundation for completing the design of the transient condition and operation mode of the sodium-cooled fast reactor as well as safety evaluation.

Study of Neutronic Performance of Sodium-Cooled Fast Reactor Design for Various Output Power Using Radial Fuel Shuffling Strategy

Study of neutronic performance analysis of Sodium-Cooled Fast Reactor (SFR) design based on the variations of output power using radial fuel suffling strategy has been performed. SFR is a type of the generation IV reactor that is widely researched for application on a commercial scale. The reactor design uses natural uranium as a fuel and Sodium as a coolant. The research has been carried out by using the SRAC code and JENDL-3.2 as a library with two dimensional R- Z suffling strategy of cylinder core for variations of power output 300, 350, 400, 450 and 500 MWTh. The neutronic parameters such as a multiplication factor (k-eff and k-inf) and burn-up analysis are observed. The SFR core is separated into 10 regions having the same volume fraction in the radial direction. At beginning of burn-up process, the reactor core is only filled with natural uranium fuel called fresh fuel and prepared for the one cycle of 10 years of burn-up time. The burn-up result in the first region is shifted into the second region, the burn-up result in the second region is shifted into the third region, and so on until burn-up result into the tenth region. The burn-up result in the tenth region is removed from the reactor core, then the first region can be filled with fresh natural uranium fuel and so on up to 10 times fuel cycles of reactor operation. The neutron calculation results indicate that the multiplication factors (k-eff and k-inf) in a critical condition are occurred on the 300 MWTh of output power. Overall, the output power of 300 MWTh has requirements and a greater chance of being operated for SFR designed.

Neutronic analysis of sodium‐cooled fast reactor design with different fuel types using modified CANDLE shuffling strategy in a radial direction

This study compares three fuel types using neutronic analysis for use in a sodium-cooled fast reactor (SFR) design with a modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) radial shuffling strategy. SFR is one type of generation IV reactor that is currently under investigation for commercial implementation. In this study, the SFR design utilizes natural uranium as the fuel input. The designed reactor core has a two-dimensional cylindrical geometry for each fuel mixed oxide (MOX), uranium-plutonium nitride ([U-Pu]N), and uranium-plutonium zirconium ([U-Pu]Zr). A radial shuffling strategy, using natural uranium as the fuel input, is applied to the SFR to manage the nuclear fuel burn-up process of the long-life reactor. This strategy is called the modified CANDLE burn-up scheme. The reactor core is divided into 10 regions with equal volume to represent the 10 years that the reactor operates without refueling. Initially, the first (innermost) region of the reactor core is filled with natural uranium fuel. The result of the burn-up from the first region is then shuffled into the second region. The third region is the result of the burn-up that is shuffled from the second region, and so on. This mechanism only requires natural uranium as the input for each 10-year fuel cycle. In this study, the fuel movement scheme is examined for three fuels. Global neutronic parameters, such as the multiplication factors and burn-up analyses, are observed and optimized. Overall, for an output power of 500 MWth and an active core radius of 110 cm and height of 210 cm, the study indicates that (U-Pu)N is the optimal fuel to be applied in the SFR with a modified CANDLE burn-up scheme. (U-Pu)Zr reaches a critical condition at an output power of 500 MWth with an active core radius of 130 cm and a height of 210 cm; whereas criticality for MOX is achieved at an output power of 1500 MWth with an active core radius of 210 cm and a height of 210 cm.

Analysis of gas entrainment phenomenon from free liquid surface for a sodium-cooled fast reactor design (Velocity profile and Strouhal number in a flow field)

Analysis of gas entrainment phenomenon from free liquid surface for a sodium-cooled fast reactor design (Velocity profile and Strouhal number in a flow field)

Analysis of the natural circulation test of PHENIX reactor by the THACS code

For the pool type sodium-cooled fast reactor (SFR), whether the primary circuit can establish a stable natural cycle after the reactor accident, and whether the primary natural cycle can take away the decay heat of the reactor core in the long-term, are the key issues in the design and safety analysis of SFR. In order to verify the natural circulation capacity of the primary circuit, a natural circulation test was carried out on pool type SFR Phenix by CEA before the reactor is decommissioned. One objective of this test is to verify the system codes used to simulate sodium natural convection phenomena.In this paper, the system code THACS is used to analyze Phenix natural circulation test, and the capacity of THACS to analyze the natural circulation of SFR is evaluated. Firstly, this paper introduces the components of Phenix reactor and the process of natural circulation test. Then, the models of program and the computational modeling of Phenix reactor are described. The calculation involves the whole process of the test: dry out of steam generators, scram, stop of pumps, two phases of natural circulation. Then, the results are compared with the test data, and the influence of inter-wrapper flow (IWF) on core outlet temperature is emphasized. The results show good applicability of THACS to analyze natural circulation in SFR, and the IWF can reduce the peak value of core outlet temperature at the beginning of natural cycle. Finally, the deviation between calculation and test results are analyzed, and the limitation of the models are evaluated.

Thermal-hydraulic analysis of a 7-pin sodium-cooled fast reactor wire-wrapped fuel bundle

The thermal-hydraulic characteristics of 7-pin sodium-cooled fast reactor (SFR) wire-wrapped fuel bundle were numerically investigated by commercial computational fluid dynamics (CFD) code ANSYS Fluent. The sensitivity analyses of mesh size, geometric simplification and Reynolds-Averaged Navier–Stokes (RANS) model were conducted. The turbulent Prandtl number model was adopted. The flow field and heat transfer were analyzed. The results show that the geometric simplification has great influences on friction factor and Nusselt number. The simplification should be taken into consideration seriously. The results of the standard k-ε model, realizable k-ε model and Reynolds stress model (RSM) model are close. The results of the Shear Stress Transport (SST) k-ω model has the maximum error compared with experiment correlation. The strong secondary flows and eddies were observed. The flow velocity near the contact of wire and rod is low and causes high temperature. At the cross section of z=50mm, the difference between the highest and lowest temperatures of the central rod is 13K, and the highest and lowest local Nusselt Numbers are 2.29 and 4.07 respectively, which indicates that the inhomogeneity of heat transfer is considerable. The velocity distribution of the boundary layer agrees well with the law-of-the-wall while the temperature boundary layer is much thicker than that of velocity. At least at y+<100, heat transfer is dominated by molecular heat conduction. At high Peclet number, the default value 0.85 of turbulent Prandtl number overestimates the heat transfer performance and the results of Kays model are closer to the experiment correlation. At very low Peclet number, Kays model has a quite different from experiment correlation. This study provides a reference for SFR design.

Evaluation of hot channel factor for sodium-cooled fast reactors with multi-physics toolkit

The evaluation of hot channel factor (HCF) is of great significance to the quantification of safety margins for reactor designs. In this paper, HCFs for a sodium-cooled fast reactor (SFR) are evaluated with the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit, which is developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of DOE for multi-physics reactor performance and safety simulations. The high-fidelity neutronics and thermal hydraulics solvers PROTEUS and Nek5000 in the SHARP toolkit are coupled to perform the multi-physics simulations for HCF evaluation. The HCFs induced by cladding manufacturing tolerance, fissile content mal-distribution, wire orientation and uncertainties on the cladding, coolant, and fuel properties are evaluated for a reference core SFR design (AFR-100). The HCFs calculated with the SHARP toolkit are compared to legacy HCFs for similar reactor types. The comparison demonstrates the reduction or elimination of modeling uncertainties in the calculation of HCFs using high fidelity advanced modeling and simulation tools without the need of expensive experiments. Moreover, the reduction of the uncertainties on HCFs evaluation allows an increase in nominal parameters and safety margin, which in turn improves the economic competitiveness of the SFR.

Numerical investigation of flow blockage accident in SFR fuel assembly

Sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declare. Due to the good thermo-physical properties of sodium, SFR has drawn a lot of attentions and there are considerable design experiences related to SFR. In the design of SFR, the pitch to diameter ratio (P/D) is kept small. Debris may accumulate in the flow channels and result in the flow blockage accident due to the corrosion of structural material. If single or several sub-channels are partially or totally occupied by debris, flow area will be suddenly reduced and a recirculation region can be expected downstream the blockage. Coolant temperature in this region can be greatly increased which may threaten the integrity of fuel assembly. In this paper, RANS method is employed to evaluate the impaired heat transfer process governed by blockages formed in a sodium-cooled 61-pin wire-wrapped rod bundle. The simulation results were firstly validated against the experimental data obtained from the literature. A good agreement demonstrated the capability of the employed approach to evaluate turbulent properties of sodium flow as well as the impaired heat transfer characteristics caused by blockages. With the validated method, the influence from blockages are systematically evaluated in which blockage size and its location are varied. Simulation results clearly indicate that a sharp increase of cladding and coolant temperature near the blockage can be expected even if the blockage area is not quite large. The local effect from blockage is also identified in the simulation. The variance of cladding and coolant temperature tends to vanish downstream the blockage and it is not feasible to detect the existence of blockage based on cladding temperature near the outlet even if the blockage area is sufficiently large.

Experimental and numerical investigation for the geometrical parameters effect on the labyrinth-seal flow characteristics of fast reactor fuel assembly

The labyrinth seal is widely used in the fast reactor fuel assembly to control leakage flow rate, adjust pressure drop and cool assembly outer wall. In the current investigation, a new non-rotating labyrinth seal with right-angle trapezoidal throttle tooth is designed and manufactured, and its flow characteristics are experimentally and numerically investigated. Reynolds number from 61,512 to 81,261 is considered and experiment data were generated from hydraulic experiment conducted at constant water temperature (84 °C) in the labyrinth-seal test section. Numerical results with different turbulence model, wall-normal grid spacing, inlet turbulence intensity and stream-wise grid length are compared with experiment results to demonstrate the effectiveness of numerical simulation. Geometrical parameters have a significant effect on labyrinth-seal flow characteristics. Increasing throttle-tooth number, spacing and throttle-tooth thickness can improve the labyrinth-seal sealing performance, while increasing gap width influences the flow velocity limitation and increases the leakage. Labyrinth-seal resistance coefficient is validated to be independent of Reynolds number (61,512–81,261). It is deduced that the fully rough regime still exists in an annular-tube turbulence flow with high wall roughness. The energy dissipation caused by throttling effect, jet, vortex and secondary transverse turbulence is the main sealing mechanism. Moreover, a leakage-coefficient empirical correlation is obtained by regression analysis of experiment data, which can be used for the structure design of sodium-cooled fast reactor and preliminary estimation of leakage flow rate in similar non-rotating seals.

Hydraulic experiment investigation on the plate-throttle entry tube flow characteristics of fast reactor fuel assembly

The entry tube of fast reactor fuel assembly plays an important role in controlling flow rate, adjusting pressure drop and limiting flow velocity, which has a significant effect on safety and economy of fast reactor. Due to some shortcomings of traditional straight-through entry tube, an optimized plate-throttle entry tube is designed and manufactured in the current investigation. Reynolds number from 8.99 × 104 to 1.65 × 106 is considered and experiment data are generated from hydraulic experiment conducted at constant water temperature (84 °C). The experiment results indicate that the flow characteristics of plate-throttle entry tube is excellent, i.e. it can significantly increase pressure drop with the large-diameter entry-tube holes to avoid excessive vibration. It is verified that the pressure drop of plate-throttle entry tube is more sensitive to the width of plate throttle rather than the diameter of entry-tube holes. In the experiments or practical engineering applications, the pressure drop should be adjusted by changing the width of plate throttle as much as possible to reduce cost. A dimensionless number called the throttling-area ratio is introduced to represent the geometrical parameters effect on flow characteristics of plate-throttle entry tube. The resistance-coefficient empirical correlations and an empirical criterion equation are obtained by regression analysis of experiment data, which can be used for structure design of sodium-cooled fast reactor and preliminary estimation of entry-tube flow characteristics.

Experimental and numerical investigation of sloshing behavior in annular region separated by several cylinders related to fast reactor design

Abstract The liquid sloshing behavior analysis of reactor vessel is an important topic of sodium-cooled fast reactor design. It is known that there are two pumps and four heat exchangers immersed in sodium liquid. Although lots of research work has been performed on the liquid sloshing inside simple cylindrical tank and coaxial circular cylinders, few about sloshing inside annular region separated by several cylinders (ARSSC), this paper deals with the analysis of the liquid sloshing inside the ARSSC geometry. A simplified cylindrical tank of fast reactor was designed, in which there are six internal cylinders used to model pumps and heat exchangers. By shaking table experiments, the number of internal cylinders was gradually increased, and the natural frequency and wave height of liquid sloshing were measured under each case. In numerical investigation, the VOF models were built to record wave height using the CFD simulation approach, the conditions were consistent with experiments. Much different with previous research, the characteristics of sloshing inside ARSSC geometry were strongly dependent on the number of internal cylinders. Experiment results showed that the natural frequency and wave height all decrease with the increase of internal cylinders gradually. The wave height response obtained by numerical simulation was in good agreement with the experimental results. In addition, as the increase of internal cylinders gradually, the maximum sloshing response occurred when the forcing frequency matched high order natural frequency of sloshing. Moreover, a correction factor was obtained by experiments for the natural frequency theoretical calculation formula. The conclusions can provide favorable reference for the seismic design of the sodium-cooled fast reactor.

A New Lumped Thermal Resistance Heat Transfer Model for Fuel Pin Structure

The lumped heat transfer methodology is simple, and the solution is very fast, so the lumped parameter approach has been widely used in thermal-hydraulic analysis for fuel pin heat transfer in nuclear reactors. In the conventional lumped thermal analysis of fuel pin structure, each component (such as a pellet, gap, cladding, etc.) is characterized by a concentrated bulk temperature (or averaged temperature), and a bulk thermal resistance. In contrast to this conventional lumped thermal resistance model, in this paper another kind of lumped thermal resistance heat transfer model for fuel pin structure has been developed. In this model, each fuel pin component is still represented by a concentrated lumped mean temperature node, while the location of the mean temperature position of each component is no longer set on the geometrical midpoint center; rather, it is assigned exactly onto the analytical temperature profile. Two thermal resistance elements are assigned for each component in this new model; between each component surface and its associated lumped mean temperature node a thermal resistance is assigned. Heat conduction in the radial direction between the mean temperature nodes of different components is purposely defined to take place at the in-between surface nodes. With this new arrangement, the location of the mean temperature positions for each component can be determined analytically, and all the thermal resistances are redefined, accordingly. The advantage of the presented method is that the temperature profile in the whole pin at any radial position can be reconstructed after a quite easy lumped heat transfer calculation. This advanced methodology can be used in nuclear reactor simulation studies where the fastness of the solution is of concern. It is of great advantage, e.g., for the early prediction of the formation of an internal molten fuel cavity within a fuel pin using this temperature profile, before the lumped pellet average temperature reaches the fuel melting point. This lumped thermal resistance model can be readily used for the sodium-cooled fast reactor design, especially for the optimized design of the pin structure. It can be also extended to the restructured fuel pin through the way that each restructured zone is treated as an individual component, e.g., for taking into account temperature-dependent thermophysical properties.

Preliminary Evaluation of Recriticality Potential for ESFR Core and Conditions for Its Elimination

In the case of a severe accident in a core resulting from unprotected loss of flow (ULOF) or unprotected transient overpower, damage can propagate from subassembly to subassembly and produce a whole-core–scale molten pool. Because the core is not in its most reactive configuration, a massive collapse of the molten material could result in a rapid supercritical condition with release of a large amount of energy. However, timely and sufficient fuel relocation outside the core by dedicated means could prevent any risk of recriticality and accident escalation. Based on a reference 1500-MW(electric) sodium-cooled fast reactor design, this paper describes the main results obtained in evaluating the recriticality potential of the European Sodium Fast Reactor (ESFR) core and conditions for its elimination during a ULOF-type transient. This study has been carried out in the frame of the Collaborative Project on European Sodium Fast Reactor of the 7th Framework Programme Euratom. The numerical analyses carried out in the present work allow one to estimate the amount of fuel mass that has to be removed from the core in order to maintain it in subcritical conditions, preventing the formation of a critical pool. Requirements for successful application of this approach, in terms of the negative reactivity insertion rate by fuel relocation and timing of discharge from the core, are derived.