Sodium-cooled Fast Reactor Research Articles

Spatio-temporal evolution of sodium aerosol and temperature distribution during sodium pool fire in a confined environment

In case of a Core Disruptive Accident in sodium-cooled fast reactor, sodium will be ejected from the reactor vessel into the Reactor Containment Building (RCB) and results in sodium fire. Liquid sodium combustion generates aerosol, which gets dispersed and remains suspended in RCB based on geometry of the building and degree of turbulence. The Environment Source Term (EST) primarily depends on the dynamics of sodium aerosols inside the containment. Hence, it is essential to carry out a safety analysis of SFR using computational tools and models shall be validated with experimental data, particularly aerosol dynamics inside the containment along with thermal hydraulics. Besides, all codes assume uniform aerosol concentration in large containment buildings and the evolution of suspended concentration was predicted. However, in actual case a non-uniform concentration profile exists, which leads to stratification of aerosol concentration inside the containment. Towards assessment of aerosol dynamics in realistic scenario, spatial–temporal characteristics of sodium aerosol and temperature distribution were studied by conducting sodium pool fire experiments in a large containment.The sodium fire experiment was conducted by using ∼2.0 kg of sodium in a pool area of ∼0.25 m2. The measured maximum pool and gas temperatures are 685 °C and 126 °C respectively, and the temperature gradient at the wall boundary layer was 7.5 °C/cm. The estimated average sodium burning rate and aerosol release rate are 17.3 kg/m2hr and 0.24 g/s, respectively. The spatial variation of aerosol characteristics was measured for both horizontal and vertical planes. The measured maximum aerosol concentration in the horizontal plane ranges from 2.76 to 5.32, 2.16–3.30 and 2.09–4.21 g/m3 for the bottom, middle and top elevations respectively. For the vertical plane, aerosol concentration varied from 2.16 to 3.65 g/m3 for central positions and 2.53–5.32 g/m3 for off-center positions. The trend of average suspended aerosol concentration at each elevation showed that the concentration was higher at the top elevation than at the middle and bottom for up to 15 min i.e., during aerosol generation period and after that concentration was found to be higher at the bottom elevation for the next 25 min. Based on the measurements, it is observed that the aerosol distribution is not well mixed within the experimental period. A significant difference in aerosol mass concentration was observed between each elevation, between estimation and observations based on a zero-dimensional model. The evolutions of aerosol median size by theoretical simulation capture the overall trend reasonably.

Large leakage sodium-water reaction accident analysis in a sodium-cooled fast reactor with paralleling steam generators

In modern sodium-cooled fast reactor (SFR) designs, paralleling steam generators (SGs) are used to improve the economy. However, large leakage sodium-water reaction (SWR) models were developed only for designs using one SG. It is necessary to derive a pressure wave propagation model with parallel channels. Thus, a detailed pressure wave propagation model based on conversion equations was derived and verified with experimental data in this study. Using the developed model, a large leakage SWR accident in the SFR with four parallel SGs was simulated. The accident process was analyzed, and the design was verified to accommodate a large leakage SWR. Sensitivity analysis of critical parameters on large leakages was simulated and investigated. The SG pipe area, rupture disk bursting pressure, failure of the rupture disk, and release pipe area were found to have significant influences on the maximum pressure. Thus, the findings of this study are valuable for SG designs and protection from large leakage SWR accidents.

Modelling of a passive hydraulic lock concept for the mitigation transfer tubes in a SFR core

To improve the safety of the integrated Sodium-cooled Fast Reactor, recent safety complementary devices can prevent or mitigate the consequences of a hypothetical severe accident consecutive to a loss of reactor cooling sequence. In particular, the M−TT device is set by empty transfer tubes, in place of some fuel sub-assemblies present in the reactor core, using an innovative passive hydraulic lock. It is based on the concept of hydraulic diode controlled by the flowrate of primary pumps. This system is dedicated to provide an easy way to discharge corium towards the core catcher in the hypothetical case of severe accident. In bonus, concerning the prevention of severe accident, it opens a direct flow path between the hot pool, equipped with decay heat removal exchangers, and the cold pool. This can lead to an easier passage of colder sodium towards the cold pool to promote core cooling by natural convection in case of the loss of the primary flow. This paper is devoted to the evaluation of the assessment of the hydraulic lock function and of the small magnitudes of the perturbations induced by our device on the thermal–hydraulic behavior of the reactor during normal operations. Through an analytical analysis and an up-scaling numerical approach, ranging from the local scale (CFD) to the system one, we claim that the lock function is kept during the normal situations at various power regimes. The leak flow represents no more than 1% of the primary pump flow. In addition, no important reactor thermal–hydraulic perturbation is brought by adding transfer tubes in the core. This is true for normal steady states, whatever the power is, but also for a full-power loss of forced flow accidental transient leading to the natural convection cooling.

Development of a statistical evaluation method for core hot spot temperature in sodium-cooled fast reactor under natural circulation conditions

To enhance the safety of a sodium-cooled fast reactor (SFR), natural circulation must be able to remove the decay heat in the core, even if the AC power supply to the forced circulation equipment is lost. Under natural circulation conditions, deterministically evaluating the core hotspot temperature considering the effects of uncertainty on the sodium flow and heat is difficult because the flow is driven by buoyancy, and the flow rate and the temperature distribution influence each other. This study develops a statistical evaluation method for the core hot spot temperature by propagating the uncertainty through the input of the analysis to the output using the Monte Carlo sampling method. The core hot spot evaluation method is applied to three representative events during the natural circulation decay heat removal operations in a loop-type SFR.

Experimental and numerical analysis of the liquid metal mixing phenomenon in complex jets of Gen-IV nuclear system

Nuclear energy is essential in addressing the problem of climate change and energy crisis. Liquid metal mixing phenomenon at the core outlet of liquid metal fast reactors is an important issue in Gen-IV nuclear systems. In this paper, the experimental and numerical analysis of the liquid metal mixing process in complex jets were carried out based on the structure of fuel assembly heads in sodium-cooled fast reactors. The liquid metal mixing characteristics were obtained in the experiment. And a high fidelity LES model was developed and validated experimentally. The impact of different head structures on the mixing process was analyzed and an index of liquid metal temperature mixing degree was newly proposed and compared with previous formulas. Result shows that, the temperature mixing region can be divided into five zones along the flow direction. Compared to other three structures, the three-sided outlet head with offset hedge has the lowest fluctuations of velocity and temperature in the upstream of the mixing flow. Also, the three-sided outlet head with offset hedge has the best temperature mixing ability, and the index is two times larger than that of the prototype. The three-sided outlet head with offset hedge proved to be the optimal structure.

Characteristics and Mechanisms of Debris Bed Formation Behavior in Severe Accidents of Sodium-Cooled Fast Reactors: Experimental and Modeling Studies

A Sodium-cooled Fast Reactor (SFR) is one of the optimized candidates in Generation IV nuclear reactor systems, but safety is an essential issue for SFR development and application. Most knowledge was accumulated through SFR safety investigations, especially for Core Disruptive Accidents (CDAs). During the CDA of SFRs, the molten materials in the core region are likely to discharge into subcooled sodium and form a debris bed on the lower region of the reactor vessel. Noticing that elaboration on the characteristics and mechanisms of Debris Bed Formation (DBF) behavior should be essential for the subsequent analysis of debris bed coolability and accident progression through various experimental and modeling studies, much knowledge was obtained during the past decades. Motivated to promote future investigations on CDAs of SFRs, the previous experiments and modeling studies on DBF behavior are systematically reviewed and discussed in this paper. The experimental results showed that the flow-regime and accumulated-bed characteristics during DBF were influenced by varying parameters and realistic conditions. Through the modeling studies, several empirical models were proposed for predicting the flow regime and accumulated-bed characteristics in DBF. In addition, to promote further development of research, the future prospects concerning DBF behavior are also described.

Thermal Assessment of Repository Footprint Requirement for Advanced Fuel Cycles

Decay heat is an important constraint for repository size and design because it can drive processes that affect performance and compromise critical materials. This paper investigates how compliance with repository thermal limits is affected by three decay heat management strategies: waste package loading, waste package spacing, and surface storage time. In particular, this paper focuses on how repository area, a result of package spacing, is impacted by waste loading and surface storage time. A two-part analytical heat transfer model is presented and executed iteratively to determine the minimum allowable repository area. The analysis considers three generic close-contact repository designs along with the wastes generated from the 40 fuel cycle analysis examples used to generate metric data for the Fuel Cycle Evaluation and Screening study. Detailed results are presented for two fuel cycles: the once-through use of low-enriched uranium fuel in light water reactors and the continuous recycling of U and Pu in sodium fast reactors. Two limits for surface storage time are identified: the time required for disposal to be possible at all and the time at which further surface storage time yields no gains. The impact of waste loading is also diminished with increasing surface storage time. In general, the generic salt repository is most flexible to accept high-heat-generating wastes with less surface storage time than other repository environments. Limited-recycle fuel cycles are shown to pose a disposal challenge because of elevated, sustained decay heat generation in the waste.

Atomistic simulation study of the structure, segregation and stability of grain boundaries in the U-Zr metallic fuel

Uranium-zirconium (U-Zr) metallic fuel is considered as fuel candidate for sodium-cooled fast reactors due to its compatibility with sodium, intrinsic passive safety characteristics and higher burnup capability compared to the oxide systems. In the present study, we performed molecular dynamics (MD) and static simulations to investigate the stability, cohesive and structural properties of U-Zr grain boundaries, with 10 and 20 at.% of Zr content. The effect of the Zr composition, temperature and atomic configurations were investigated, and the fracture mechanism highlighted. The cohesive energies of special low energy GBs indicate that the intergranular fracture mechanism plays a significant role in the brittle regime. Moreover, we investigated the role of the atomic configurations in determining the GB properties of the alloys. Indeed, for a given misorientation angle, we found that different atomic configurations do not only yield different energies, but also different strain distribution, atomic coordination and different net expansions (free volume at the boundary). In general, we observed that Zr atoms tend to segregate in low energy configurations. We also observed a systematic increase in GB and surface energies with the Zr concentration, in which the surface energies are systematically higher than the corresponding GB energies. The reported findings highlight important effects of the atomic configurations on properties of investigated GB structures. Therefore, it clearly indicates that the atomic configurations of GBs must be correctly determined for the accurate modeling of the alloys properties.

Characterization of thermal stratification using validated liquid metal pool mixing models

Weakly stratified thermal fronts in the upper plena of liquid metal cooled reactors can pose prediction complexities for safety analyses. The strong thermal diffusivity of the coolant changes gravity’s influence on the flow by reducing horizontal density gradients at time scales faster than gravity can act on them. While damped out thermal behavior results in stratification, the underlying convection currents continue contributing to enhanced mixing. Experimental turbulence spectral information, captured by distributed velocity (ultrasonic) and temperature (fiber-optic) measurements, is used to construct microscale parameters and quantify the amount of enhanced vertical thermal diffusivity in a scaled gallium surrogate of a pool type sodium fast reactor upper plenum. These constructed microscale parameters are used to validate empirical models for use in liquid metal pools. This experimental data and, separately, the validated models, are fed into a simple 1-D transport model to quantify the impacts to reactor safety and allow for testing of the efficacy of using the more easily available bulk parameters (i.e., core barrel Reynolds and Richardson numbers) to estimate the enhanced diffusivity. The 1-D model represents the experimental fronts’ progressions most closely after accounting for these characteristics of mixed convection unique to liquid metals.

Droplet group combustion model improvement for sodium spray fire pre-combustion analysis

Sodium spray fire could happen in sodium-cooled fast reactor, which will cause the rapid rise of temperature and pressure in the vessel. In this paper, based on the droplet pre-combustion model, droplet motion model, multi-node spray fire combustion model and flow field model, a one-dimensional sodium fire combustion analysis code NAFCALX is developed. Its original droplet group combustion model and other inadequacies are improved, coupling the change of droplet with the movement of droplet during droplet combustion. The improved model is used to simulate the ABCOVE and KAERI sodium spray fire experiments. The predicted temperature and pressure peaks and the trend of temperature changing with time were in good agreement with the experimental values, which is much better than that from the original model. Through the comparison and analysis of the calculation results, the accuracy of the improved model is verified.

Flow induced vibration measurement of a full-length sodium fast reactor fuel bundle

Nuclear technology continues to be an integral part of the world’s integrated energy infrastructure to provide sustained carbon-free emissions form of energy. As existing nuclear power plants age, and newly conceived nuclear reactors are developed there are several phenomena that continue to limit the operations and lifetime of these systems including thermal fatigue, irradiation material degradation, and flow induced vibration. Of the new nuclear power plant companies, TerraPower is proactively addressing each of these phenomena through use of advanced design approaches and aggressive testing programs. This study details the outcomes resulting from the experimental testing program that supports design aspects of a sodium fast reactor fuel bundle and implications of flow induced vibration. Specifically, this study adds to the body of knowledge through three contributions: (1) The outcomes of this study result in the first-ever experimental data collected with sufficient temporal and spatial measurement resolution to validate multiphysics computational tools related to flow-induced vibration; (2) phenomenological comparisons of fluid–structure interactions are made between a straight ducted fuel bundle and a diluted duct fuel bundle; (3) pin-to-pin interactions are observed as they relate to different configurations within a single fuel bundle. The hydraulic characterization is compared against that of experimental data collected in previous studies to verify relevance and credibility of boundary conditions. The outcome of this work disseminates a new understanding of the vibrational response of a full-length SFR fuel bundle under representative hydraulic conditions to the community.

Experimental study on design optimization of the core catcher for the CEFR

ABSTRACT In severe accidents, sodium-cooled fast reactors (SFRs) are more concerned with core disruptive accidents (CDAs), and in-vessel retention (IVR) is an important mitigation measure to ensure the integrity of the reactor vessel. The distribution and accumulation uniformity of the melt in the lower chamber seriously affects the residual heat discharge capability. The installation of a core catcher is an effective mitigation method, and its structure and arrangement play a key role in mitigating severe accidents. In this study, the experimental facility is designed based on the core catcher of the China Experimental Fast Reactor (CEFR). An experimental platform is established to simulate the falling and relocation of melts in the core catcher under the condition of CDAs. At the same time, a quantitative evaluation index is established to analyze the influence of the number of chimneys, the inclination angle of the chimney canopy, the central chimney height, and the spacing of chimneys on relocation of melts in the core catcher. Finally, some suggestions for correlation optimization design are given.

Off-design performance evaluation of multistage axial gas turbines for a closed Brayton cycle of sodium-cooled fast reactor

In this study, the validity of reducing the number of gas turbine stages designed for a nitrogen Brayton cycle coupled to a sodium-cooled fast reactor was assessed. The turbine performance was evaluated through computational fluid dynamics (CFD) simulations under different off-design conditions controlled by a reduced flow rate and reduced rotational speed. Two different multistage gas turbines designed to extract almost the same specific work were selected: two- and three-stage turbines (mid-span stage loading coefficient: 1.23 and 1.0, respectively). Real gas properties were considered in the CFD simulation in accordance with the Peng–Robinson's equation of state. According to the CFD results, the off-design performance of the two-stage turbine is comparable to that of the three-stage turbine. Moreover, compared to the three-stage turbine, the two-stage turbine generates less entropy across the shock wave. The results indicate that under both design and off-design conditions, increasing the stage loading coefficient for a fewer number of turbine stages is effective in terms of performance and size. Furthermore, the Ellipse law can be used to assess off-design performance and increasing exponent of the expansion ratio term better predicts the off-design performance with a few stages (two or three).

Coolant leak rate through fatigue cracked pipe bend: An experimental and design standard based investigation

Leak-Before-Break (LBB) requirement plays a vital role in demonstrating the structural integrity of sodium-cooled Fast Breeder Reactor (FBR) piping system. Crack size at penetration (2CS), crack size that corresponds to detectable coolant leak rate (2CL), and unstable crack size (2CG) are the primary crack parameters associated with the LBB design of FBR piping system. The detectable leak rate through a piping system is directly related to the differential pressure (ΔP) and crack size. The design pressure of FBR piping systems is relatively low, resulting in a high value of CL. Considering the need for an optimum LBB design of the FBR piping system, an experimental approach is adopted to arrive at a realistic estimate of CL. Based on the experiments, a power-law type correlation can be derived to predict the value of CL as a function of ΔP. The evidence from this study points towards the conservatism present in the current FBR LBB design philosophy as per RCC-MRx A16. These observations have several implications for achieving an optimum LBB design criterion for an FBR piping system.

Validation study of sodium pool fire modeling efforts in MELCOR and SPHINCS codes

Discharge of sodium coolant into containment from a sodium-cooled fast reactor vessel can occur in the event of a pipe leak or break. In this situation, some of the liquid sodium droplets discharged from the coolant system will react with oxygen in the air before reaching the containment. This phase of the event is normally termed the sodium spray fire phase. Unreacted sodium droplets pool on the containment floor where continued reaction with containment atmospheric oxygen occurs. This phase of the event is normally termed the sodium pool fire phase. Both phases of these sodium-oxygen reactions (or fires) are important to model because of the heat addition and aerosol generation that occur. Any fission products trapped in the sodium coolant may also be released during this progression of events, which if released from containment could pose a health risk to workers and the public. The paper describes progress of an international collaborative research in the area of the sodium fire modeling in the sodium-cooled fast reactors between the United States and Japan under the framework of the Civil Nuclear Energy Research and Development Working Group. In this collaboration between Sandia National Laboratories and Japan Atomic Energy Agency, the validation basis for and modeling capabilities of sodium spray and pool fires in MELCOR of Sandia National Laboratories and SPHINCS of Japan Atomic Energy Agency are being enhanced.This study documents MELCOR and SPHINCS sodium pool fire model validation exercises against the JAEA’s sodium pool fire experiments, F7-1 and F7-2. The proposed enhancement of the sodium pool fire models in MELCOR through addition of thermal hydraulic and sodium spreading models that enable a better representation of experimental results is also described.

Stability and bifurcation analysis of Generation IV reactors via point reactor models with temperature reactivity feedback

While the stability of present-day reactors is well studied, stability analysis of Generation IV reactors is in its very beginning. In this paper, the stability of point nuclear reactor models with temperature reactivity feedbacks is analyzed. The examined models are referred to as G-L reactor models, which are point reactor models with G delayed neutron-emission groups, L temperature lumps (regions) and a linear reactivity feedback for each temperature. G-L reactor models are applicable for modeling four out of the six Generation IV reactor types (Gas Fast Reactor, Lead Fast Reactor, Sodium Fast Reactor and Very High Temperature Reactor).New analytical stability and bifurcation results of general G-L reactor models and of the G1-L2 reactor model are derived. These analytical results add a different perspective to the previous studies of Generation IV reactors, most of which are based only on numerical analysis. In particular, the codimension-1 bifurcations of general G-L reactor models are characterized, and a classification of the linear stability region in the G1-L2 model is derived.Then, stability and bifurcation analysis of a G1-L2 model of a Generation IV HTR-type microreactor is performed, using both analytical methods and the numerical continuation method. The nominal design point of this microreactor is shown to be globally asymptotically stable, while in other regions of the parameter space, a bistability of the operating steady state and a limit cycle, or of two limit cycles, is demonstrated.

Feasibility study of a resistive-type sodium aerosol detector with ZnO nanowires for sodium-cooled fast reactors

In sodium systems, leakage is one of the safety concerns; it can cause chemical reactions, which may result in fires. There are contact and non-contact types of leak detectors, and the conventional method of non-contact type detection is by gas sampling. Because of the complexity of this method, there has always been a need for a simple gas sensor, and the resistive-type nanostructure ZnO sensor is a promising option with various advantages. In this study, a ZnO sensor was fabricated, and the concept was tested as a leak detector using a dedicated experiment facility. The experiment results showed distinctive changes in resistance with the presence of sodium aerosol under various conditions. Replacing the conventional gas sampling with the ZnO sensors is expected to enable identification of the leakage location if used as a point-wise instrumentation and to greatly reduce the total cost, making the system simple, light, and effective. For further study, more tests will be performed to evaluate the sensitivity of key parameters under various conditions.

A safety enhanced sodium-cooled MOX fueled fast reactor core concept

ABSTRACT We have developed a 750MWe sodium-cooled MOX fueled fast reactor core with enhanced safety against typical ATWS. Design goals were set as follows: 1) negative void reactivity for the loss of flow (ULOF); and 2) burnup reactivity less than 1 $ for the transient over power (UTOP). To achieve the goals the following were considered: 1) adoption of the axially heterogeneous core with a sodium plenum and a gas expansion module to reduce void reactivity; and 2) addition of minor actinides (MAs) to the internal blanket to reduce burnup reactivity. Configurations of the core were optimized as follows. First, void reactivity was reduced by a parameter survey for the inner core height, B-10 content of the upper shield, and standby position of the backup control rods. Second, burnup reactivity was reduced by a parameter survey for the MA content in the internal blanket, etc. We designed the safety enhanced fast reactor core that did not undergo sodium boiling if the ULOF occurred and we maintained fuel integrity for the UTOP by the transient analysis. By installing a self-actuated shutdown system in the backup control rods, the improved operational reliability of this shutdown system and the improved safety can be expected.

Nuclear data uncertainty influence on the breeding ratio in sodium-cooled fast reactor systems

Nuclear data are a main uncertainty source in neutron transport simulations making their consideration in reactor safety necessary. This arises anew with the state-of-the-art reactors known as Generation IV. Some of the reactors suggests providing the reactivity margin below the effective delayed neutron fraction excluding prompt criticality accidents, and the breeding ratio is the key factor in this. Consequently, assessing a degree of the breeding ratio accuracy is of interest. Therefore, in this work, the breeding ratio uncertainties are analyzed by performing a sensitivity and uncertainty analysis of the MET1000 and MOX3600 models with respect to nuclear data using SCALE. As a result, the breeding ratio uncertainties are obtained approximately equal to 2% as the main contributors are 239Pun,γ, 238Un,γ, and 238Un,n'. The uncertainty sources between the models are compared, and 16O preponderantly increases the total uncertainty not directly by its uncertainty but by its impact on the spectrum.

Simulation of decay heat removal by active means following emergency shutdown in SFRs

Decay heat removal (DHR) is an important safety function in sodium-cooled fast reactors (SFRs). Dedicated systems are provided in the plant for performing DHR functionalities. Transients following reactor shutdown need to be studied to understand the dynamics of DHR systems, design the systems, and establish reactor safety. DHR systems can operate by active as well as passive means, thanks to the phenomenon of natural circulation. In the present work, a DHR system riding on the Balance of Plant (BoP) of a typical medium-sized SFR has been modeled and incorporated into the in-house developed plant dynamics code DYANA-P. DYANA-P already has models of primary and secondary circuits of SFR. With this code enhancement, the simulation of DHR in an integrated manner became possible. In particular, thermal models of the Moisture Separator Tank (MST), Decay Heat Removal Condenser (DHRC), and hydraulic model of the DHRC flow circuit were developed. Two-phase one-dimensional mass, momentum, and energy conservation equations with homogeneous flow assumptions were used to create the models. Simulation of SCRAM in a typical SFR has been carried out to demonstrate the applications of the modified code. Two different DHR deployment schemes were studied and compared – 1. Using DHR systems designed for high pressure of ∼170 bar, 2. Using DHR systems designed for low pressure of ∼15 bar. In the first scheme, reactor cooling was achieved at the desired rate of ∼20 K/h. In the second scheme, the reactor cooling rate will be higher at ∼60 K/h.