Proceedings of the Canadian Nuclear Society
The Canadian Nuclear Society (CNS) promotes the exchange of information on all aspects of nuclear science and technology and its applications. This includes nuclear power generation, fuel production, uranium mining and refining, management of radioactive wastes and used fuel.
Study of Turbulence and Pressure Recovery in Heat Pipe Vapor Flow Using the Spectral-Element Method
Heat pipes can efficiently and passively remove heat in nuclear microreactors. Nevertheless, the flow dynamics within heat pipes present a significant challenge in designing and optimizing them for nuclear energy applications. This work aims to explore the vapor core of heat pipes through comprehensive two and three-dimensional simulations, with a primary focus on modeling the pressure recovery observed in the condenser section. The goal is to establish improved correlations for one-dimensional heat pipe codes. The simulations are validated against experimental data from a vapor pipe documented in the literature. The k‐τ turbulence model is employed in the two-dimensional simulations through the open-source spectral-element code Nek5000. This model provides insights into pressure recovery within heat pipes with low computational cost. In addition, Large Eddy Simulations (LES) are used to capture turbulent flow features in a three-dimensional vapor pipe model, utilizing the code NekRS. Using LES is crucial for comprehending the impact of laminar-to-turbulent transition on pressure recovery. A simulation framework is created to model the heat pipe's vapor core, laying the foundation for an enhanced understanding of heat pipe behavior. The ultimate goal is to improve and optimize heat pipe designs, provide data to validate lower-fidelity approaches and enhance their performance in nuclear reactors.
Copyright (c) 2024 Canadian Nuclear Society
How and Why Flow Resistances of Cross Flow over Inline Tube Bundles Vary with Reynolds Number
Helical tube bundles are used for steam generators (SGs) and intermediate heat exchangers (IHXs) of High Temperature Gas-cooled Reactors (HTGRs) due to its compactness, high heat transfer coefficient and thermal expansion adaption performance. Accurate prediction of the pressure drop coefficient of cross flow over inline tube bundles is crucial for the design of such heat exchangers. According to the previous measurements from the wind tunnel experiments, the pressure drop coefficient of cross flow over tube bundles with pitch to diameter ratio of 1.875 (S/D=1.88) decreases as Reynolds number (Re) increases, which agrees well with the Zukauskas’ correlation. In current investigation, the cross flow over tube bundles with S/D=1.875, Re=18000 and 36000 are simulated using Large Eddy Simulation (LES). The dynamic Smagorinsky model is used to close the sub-grid stress term. The numerical method is validated using the measured surface pressure coefficient and the pressure drop coefficient. According to the analysis on surface pressure coefficient distributions, the reason for why pressure drop coefficient decreases with Re increasing is attributed to that the surface pressure coefficient at rear side of the tube increases as Re increases. By analyzing the streamwise force balance on the time averaged separation bubble boundary, the pressure coefficient at the separation streamline (wake bubble boundary) increases as Re increases, which leads to the increase of pressure coefficient at rear side of the tube as Re increases. Finally, by comparing flow field of cases with different Re, the wake bubble is found shrink as Re increases, which results in the increasing of pressure coefficient at the separation streamline (wake bubble boundary) as Re increases and ultimately leads to the decreasing of pressure drop coefficient.
Reliability and Safety Assessment of Isolation Condenser System in BWRX-300
The BWRX-300, an innovative Small Modular Reactor (SMR), leverages over 60 years of experience in boiling water reactor technology. As the 10th generation, it evolves from the Economic Simplified Boiling Water Reactor (ESBWR) to reduce capital costs while ensuring high safety. Its versatile design supports various applications, including electricity and hydrogen production, emphasizing efficiency and fewer staffing needs. A hallmark of the BWRX-300 is its utilization of Passive Safety Systems (PSSs) that leverage natural forces for safety measures, eliminating the dependence on active systems and manual interventions. Despite the proven effectiveness of PSSs, their reliability assessment poses challenges due to the lack of operational data and complexity of their functions. This research introduces an Artificial Neural Network (ANN) based methodology for evaluating the reliability of such systems, specifically focusing on the BWRX-300's Isolation Condenser Systems (ICS). By employing ANN alongside traditional Probabilistic Safety Assessment (PSA) techniques, this study aims to refine the reliability and safety assessments of PSS in SMRs, enhancing the safety evaluations and supporting the sustainable and secure advancement of nuclear energy technology. The integration of ANN model demonstrates the potential of artificial intelligence tools in advancing nuclear safety assessments, thereby contributing to the global efforts towards safer and more sustainable nuclear energy solutions.
Transient Reactor Core Behavior by Multi-Physics Approach with Neutron Kinetics, Fuel Performance and Sub-channel Thermal-Hydraulics Coupling
Steam line break accident and reactivity-initiated accident in the nuclear reactor are representative Non-LOCA accidents in which thermal-hydraulics and fuel behavior are strongly coupled each other. Moreover, those accidents might yield a spatially asymmetric core status that cannot be calculated by conventional system-scale reactor safety analysis. Thus, the multi-scale and multi-physics approach based on sub-channel scaled resolution is applied in reactor core region in order to examine a realistic safety margin. In this study, full core pin-wise non-LOCA safety analysis have been performed. An activation of transient behavior is triggered by the system thermal-hydraulics code, MARS-KS, and core behavior is calculated by sub-channel thermal hydraulics code, CUPID-RV and neutron kinetics code, MASTER. Fuel behavior is calculated by the fuel performance code, FRATPRAN. For the realistic core simulation, the initial condition as the fuel burnup information was obtained by pin-wise steady thermal-mechanical code, GIFT. KAERI developed the MARU platform for easy assessment to simulate the MSMP approach in high-performance computing environment. Two accident scenarios for End-Of-Cycle burnup status are simulated. In SLB simulation, the detailed thermal behavior inside the reactor pressure vessel during the transient such asymmetric power distribution and occurrence of the void fraction at upper plenum. Furthermore, the enhancement of safety margin was quantitatively investigated by obtaining a realistic DNBR distribution. Other accident scenario is control rod ejection accident. By the coupled simulation, the safety margin is quantitatively investigated by comparing the stored energy during the transient.
Numerical Analysis of In-Vessel Melt Retention of CANDU Reactors During a Severe Accident
The objective of the IVR strategy in a CANDU reactor is to maintain the integrity of the calandria vessel and prevent ex-vessel phase conditions, including but not limited to, molten core concrete interactions and significant production of non-condensable gases. To achieve this, accurate simulation of corium convective flows in the calandria vessel is important to demonstrate the performance of vessel structures, corium movement patterns and assess potential breaches that may occur. This research work studies the behaviour of three-dimensional corium flows in a scaled model of the CANDU calandria vessel, encompassing complex phenomena related to multiphase flow, solidification, and heat transfer during crust formation. Ansys-Fluent was used as a solver to study the flow behaviour and heat transfer in a 1/5 scale model of the reactor corium pool and predict its progression. Two relevant validation cases were selected and studied to validate the numerical predictions. The results of this study provide valuable insights into the hydrodynamics of corium in the calandria vessel, particularly near the end shields and subshells of CANDU reactors, . This, in turn, contributes to improved IVR safety strategies and the effective management of accidents in CANDU reactors. This study also highlights the importance of ongoing research and development in modeling and simulation techniques for corium behaviour. As future research progresses, understanding of corium hydrodynamics and heat transfer mechanisms will improve, leading to more accurate predictions and assessments of reactor safety analysis. Integrating advanced computational tools and experimental validation offer valuable potential for improving IVR strategies and ensuring their reliability.
Validation of a Model for Estimating the Retention Efficiency in Pool Scrubbing
For the estimation of particle retention in a liquid pool, a model is developed that is suitable to be implemented in the software package AC² developed by the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH. Within AC², the Containment Code System COCOSYS is used for the simulation of ex-vessel phenomena in the frame of postulated severe accident scenarios in light water reactor containments. As a lumped parameter code COCOSYS is characterized by a low simulation time compared to other simulation methods such as CFD. Validation works show an underestimation of the retention efficiency by the pool scrubbing model available in COCOSYS compared to experimental data. For the validation of the newly developed model APVOR experiments conducted in the SAAB test facility, EPRI experiments performed in the Phase I scrubbing facility and experiments from the SAPOS test facility at Shanghai Jiao Tong University are used. The results show a high agreement between the calculated retention efficiencies and the data obtained in the experiments. With a few exceptions, the deviation is less than ±10 %. A high uncertainty can be traced back to the used condensation model, which should be optimized in future activities. Compared to existing correlative approaches, the developed model shows high agreement with experimental data. The input parameters of the new model can be provided by COCOSYS and a calculation of the particle retention is possible without significant user influence.
Opportunities & Challenges for Two-Phase Flow Microchannel Heat Sinks in Nuclear Systems
The microchannel heat sink (MCHS) has garnered significant interest in recent years due to the demand for high heat flux compact energy systems accompanying the development of increasingly compact electronic systems. While the literature on this topic has focused on the use of MCHS for heat removal from electronics, opportunities have been identified for applications to nuclear systems which could benefit from these highly compact systems. Numerous heat transfer enhancement techniques have been proposed with the aim of improving MCHS performance through modified channel geometries and optimization of flow conditions. The characterization of two-phase boiling flow patterns and correlations to predict heat transfer and pressure drop in the associated flow regimes is an active area of ongoing research due to the inconsistent agreement between experimental data and correlations. Flow instabilities in two-phase boiling pose a significant challenge to the design of these systems due to their potential to induce boiling crisis, disturb control systems, or cause mechanical damage. Such instabilities must be well-understood and accurately predicted to ensure the safe design margins for systems relying on MCHS for thermal management; this requirement is especially important in nuclear applications. This review aims to summarize recent efforts to model and predict flow instabilities in the design of microchannel systems and identify key areas of research that need to be addressed to enable the application of MCHS technology to heat removal in nuclear systems.
Modeling of Vapor Mass Formation under Forced Convective Boiling Condition
The development of a precise and reliable prediction method for a departure from nucleate boiling (DNB) is urgently needed to design and develop new-generation reactors that enable us to establish a carbonneutral society. Currently, the DNB has been predicted by the DNB correlation and a subchannel analysis code for the fuel assemblies in pressurized water reactors. It took a huge cost because the DNB correlation had to be obtained by the mock-up test every time the new type of fuel assembly was considered. Therefore, the prediction method of the DNB based on the mechanism has been required for a long time. In this study, we consider that the formation of a large vapor mass on the heating surface is the primary trigger of the DNB. Therefore, we embark on the development of a new model for vapor mass formation under forced convective boiling. We assume that the primary bubbles coalesced with each other, which are generated from the nucleation sites, and the coalesced bubble is formed. The nucleation sites are assumed to be distributed on the basis of the Poisson distribution. The coalesced bubbles grow on the basis of the 1-D bubble growth and motion equation, and the large vapor mass is assumed to be formed by the merging of coalesced bubbles when the diameter of the coalesced bubble satisfies the criteria of the slug formation by Mishima. The model reproduced the dependency of vapor mass formation heat flux on the bulk velocity and subcooling.
Two-Phase Flow Interfacial Structure in CANDU Reactor Fuel Channel and Feeder Pipe Under Postulated Accident Conditions
Experiments were conducted to study the two-phase flow structure relevant to the flow behaviour of the header, feeder, and fuel channel systems in CANDU reactors during postulated accidents. The test loop consisted of a series of instrumented vertical and horizontal pipes with an inner diameter of 32 mm. The test section of the loop simulated a horizontal fuel channel connected with a vertical inlet feeder and a vertical outlet feeder in the primary heat transport system of a CANDU reactor. Wire- mesh sensors were used to measure two-phase flow interfacial parameters such as void fraction distribution, interfacial area concentration, bubble interfacial velocity, and bubble diameter distribution. The experiments were conducted for a series of flow conditions corresponding to various postulated loss-of-coolant accidents, covering various flow regimes for vertical and horizontal flows. The results of the experiments were used to study and model the effect of flow channel orientation (vertical vs. horizontal) on the flow regime and interfacial structure. Furthermore, a drift-flux analysis was carried out based on this new database, and several drift-flux correlations were evaluated. The high-resolution two-phase flow data obtained in this work is appropriate for benchmarking and developing three-dimensional thermal-hydraulic system codes and computational fluid dynamics codes for CANDU-type reactors.
Comparison of URANS and LES Predictions for the Open Phase of the OECD NEA CSNI Fluid Structure Interaction CFD Benchmark
The OECD NEA CSNI WGAMA CFD Task Group ran a benchmark in 2020 and 2021 to assess the predictive capabilities of coupled fluid structure interaction (FSI) CFD analysis methods. This paper presents the predictions made for the open phase of the benchmark using URANS and LES turbulence modelling approaches, and a comparison of the results to the experimental data. The benchmark comprised a channel containing two inline cylinders in cross-flow. The cylinders were fixed at one end, free at the other, and had measured resonant frequencies and damping properties. The URANS modelling used ANSYS Fluent 2-way coupled to ANSYS Mechanical. The LES modelling used Nek5000, 1-way coupled to Diablo. Comparisons with cross-channel velocity profiles are presented, both for the mean flow and its RMS. Comparisons are also made to the frequency spectra for point measurements of fluid velocity and pressure, and for the accelerations of the free end of each cylinder. URANS predicts the average velocity profiles relatively well, and is able to predict the velocity and acceleration spectra at the shedding frequency. However, the frequency content at the 4th harmonic of the shedding frequency is low in the URANS flow fields, and so does not excite accelerations at the resonant frequency of the cylinders. LES makes better predictions of the average profiles, and the velocity spectra agree well at both the shedding frequency and at higher frequencies. The 1-way coupled LES results show good agreement for acceleration spectra.
Numerical Investigation of Surface Tension and Orifice Shape Effects on Bubble Injection Dynamics using Direct Numerical Simulation
Advanced reactor designs, like Molten Salt Reactors (MSR) are outfitted with a fission gas removal mechanism, such as Xe gas. In this mechanism, bubbles of helium gas are injected into the molten fuel salt to remove the Xe gas. It is important to demonstrate a robust methodology capably of predicting the bubble’s behavior and void fraction within the nuclear reactor to evaluate the efficiency of such systems. The present study applies direct numerical simulation (DNS) with interface capturing approach to model the bubble formation and rise dynamics when injected through oval and circular-shaped orifices with an equivalent diameter of 0.5 mm into quiescent fluid. The computational domain was filled with tetrahedral mesh elements varying the size depending on the proximity to the orifices. A finite element method-based flow solver and level set approach is applied for capturing the air/water interfaces. The bubble departure diameter and time, bubble shapes, and deformability factors are evaluated for both the circular and oval-shaped orifices along with different surface tension values as well. The bubble departure diameter and the temporal evolution of the bubble shapes obtained from the numerical simulation are validated against the experimental data. System-level codes as well as multiphase CFD approach rely on closure laws based on average bubble size and other local parameters. The demonstrated and validated DNS approach can help quantify those required parameters based on fluid conditions, properties as well as injector designs.
Quenching Temperatures During Reflood Transients in a Rod Bundle
During reflood transients, drastic changes in the boiling regime occur at the quenching temperature, where liquid can reestablish physical contact with a hot solid surface. A comprehensive understanding of the quenching temperature and the factors affecting its behavior is essential for assessing the cooling of a rod bundle. In this study, constant reflood experiments were conducted at the NRC/PSU Rod Bundle Heat Transfer (RBHT) test facility to explore the impact of pressure and flooding rate on the quenching temperature and the associated boiling characteristics. The reflood experiments explored in this study clearly show significant effects of flooding rate and pressure. A higher rate of flooding always leads to a faster quenching of the rod bundle. For a given flooding rate, however, the quenching temperature (Tq) tends to vary across the bundle, increasing from lower to higher elevations due to skewed power profiles and enhanced local flow effects. Pressure has a significant impact on the quenching temperature (Tq), especially at lower elevations, where the quenching temperature tends to increase as pressure increases. The effect of flooding rate on Tq depends on the elevation and flow rate. At low elevations, the quenching temperature (Tq) tends to increase with the flooding rate regardless of whether the flooding rate is high or low. At high elevations, the Tq behavior varies differently with the flooding rate.
Preliminary Analysis of Flow Distribution in a 100-MWth Molten Salt Fast Reactor Core with Flow-guide Structures using Computational Fluid Dynamics
A 100-MWth liquid-fueled molten salt fast reactor based on the chloride salt is being developed in Korea Atomic Energy Research Institute for maritime applications. The proposed reactor design has a vacant reactor internal because it does not need any moderator or fuel assemblies. In such structure, a proper distribution of the flow entering a reactor core is one of the challenging issues in reactor design. In this study, we suggest the flow-guide structures including the inlet flow distributor, the flow distribution plate, and the outlet flow collector. To examine the feasibility of the suggested concepts, the preliminary analysis was performed using the computational fluid dynamics (CFD). From the analysis, it was confirmed that the flow distribution inside the reactor can be effectively controlled by changing the number and the distribution of the holes on the flow distribution plate. Based on this result, the design approach for the flow distribution plate was suggested. In addition, the inlet flow distributor and the outlet flow collector significantly contributed to the overall pressure loss. To reduce the pressure drop, the effect of the diameters of each design and the effect of the number of channels on each design were investigated. The overall pressure drop decreased by about 65 % with the 50% increase of the diameter and the 100% increase of the number of the holes.
Simulation of the LIVE-J Experiments using MELCOR: Findings and Current Limitations
During the late in-vessel stage of a Severe Accident (SA) the debris is assumed to accumulate as molten pool in the lower plenum of the reactor and form various layers according to the material density. This is the assumption adopted in the In-Vessel Melt Retention strategies for example. The case of Fukushima Daichii however, recent analyses suggest a different accident progression, where at first a debris bed forms in the bottom of the vessel, which later starts melting from the lower melting point material which permeates through the high-melting point debris bed. This configuration was reproduced experimentally at the LIVE-3D facility at KIT, using ceramic beads and nitrate as simulant materials. Heat up due to radioactive decay is simulated using internal wire heaters, and two experiments are performed: the LIVE-J1 experiment, which focused on the melting of the nitrate into the ceramic debris bed, and LIVE-J2, focused on the heat transfer characteristics inside the debris bed. It is still unclear whether SA codes, such as MELCOR, are capable to simulate this scenario and reproduce correctly the heat-transfer on the wall. Hence, in this paper we validate the MELCOR COR model against the LIVE-J2 experiment. Results are then compared with the ones previously obtained from CFD simulations, showing that the MELCOR model predicts higher pool and lower vessel temperatures. Applying an HTC derived from CFD calculations gives an improved forecast of both.
Multi-Physics Modeling of the OECD/NEA MPCMIV Benchmark Transient Scenario
The Organization for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) is conducting the Multi-physics Pellet Cladding Mechanical Interaction Validation (MPCMIV) benchmark aiming at providing guidelines for multi-physics transient validation. In the benchmark, two cold ramp tests performed in boiling water reactor (BWR) conditions are modeled using neutronics, thermal-hydraulic, and fuel performance in single-physics and multi-physics models. Both cold ramp tests consist of a short irradiation (less than one minute) at fixed R2 power where the rodlet is inserted in the core for a certain amount of time before the reactor is shutdown. In this work, the first cold ramp test is investigated using a multi-physics coupling with MOOSE based codes: GRIFFIN for reactor physics, THM for thermal-hydraulics, and BISON for fuel performance. For THM, the 1-D model consists of the entire in-pile loop system, while BISON models only the fuel rodlet in 2D-RZ and GRIFFIN models part of the in-pile at the location of the reactor core using 3D (only the local domain around the rodlet is modelled by the codes). The first step is to perform verification and validation activities consisting of mesh and time-step convergence studies and comparison to measurements for each single physics model. Due to the amplitude of the incoming neutron current on the in-pile loop being unknown, a calibration is performed to reproduce the coolant temperature measurements. Finally, the coupling temporal cladding axial elongation evolution prediction is compared against measurements, showing good agreement.
Friction Factor and Nusselt Number Correlations for Laminar Convection of Liquid in Helical Tubes with Variable Properties
Helical tubes are widely used in various industries with the advantages of compact structure, good thermal expansion and heat transfer performances. Characteristics of convection in helical tubes are more complicated than that in straight tubes due to the centrifugal forces. In the literature, (μw/μb)n is used to consider the property variation effects of liquid on laminar convection in straight tubes. However, investigations of property variation effects on laminar convection in helical tubes are few in the literature. In the current investigation, forced laminar convection with different kinds of liquid in helical tubes are simulated. Property variation effects on friction factor in helical tubes (ψf,hel) are compared with the property variation effects in straight tubes (ψf,str). Based on order magnitude analysis, the property variation effect ratio (ψf,hel/ψf,str)between convection in helical tubes and convection in straight tube is related to the Nusselt number ratio (Nuhel,c/Nustr,c) with constant properties. Then, the correlation form for friction factor is obtained and coefficients are determined by regression analysis of numerical results. Based on the friction factor correlation, the Nusselt number correlation is obtained. The new friction factor and Nusselt number correlations show good prediction of numerical results and can be used for different kinds of fluid.
Experimental Investigation of Phenomena in Heat Transfer to Water at Near Critical Pressures
The knowledge of heat transfer at near-critical pressures is crucial for the safety design of thermal hydraulic systems like supercritical water reactors. Starting from a supercritical state during normal operation, loss of pressure accidents can lead to thermodynamic subcritical conditions, so that boiling crises can occur if the critical heat flux is exceeded. This phenomenon may leads to a significant increase in the wall temperature of heated surfaces. The following failure of the fuel rod material can cause safety-relevant damages. The safety design of novel reactor concepts, working at thermodynamic supercritical conditions, requires therefore meaningful correlations to predict those phenomena and their impact. Current correlations are however often meaningful at reduced pressures pr < 0.7 but uncertain above. The “High Pressure Evaporation Rig – HIPER” at the Chair of Energy Systems at the Technical University of Munich provides experimental heat transfer data on water at near critical pressures. This facility consists of a closed loop with a vertical smooth test section. 300 kW electrical power can be applied to the current 15.8 mm evaporation tube to achieve CHF. This pipe is used to investigate the pressure and mass flux dependence of CHF. A future optimized test section with an internal diameter of 10.0 mm, will be used to study the influence of the hydraulic diameter. In addition, the new test section will provide accurate data to test current fluid-fluid scaling theories, as comparable experiments are currently being conducted internationally using CO2 and R134a instead of water. In summary, this contribution provides insights into currently conducted experimental data and future experimental investigations on heat transfer at high sub- and supercritical pressures.
Thermal Hydraulic Analysis of the AP1000 Passive Containment Cooling System following LBLOCA using ANSYS FLUENT
The AP1000 power plant is a generation III+ design that uses a simplified, innovative, and effective approach to enhance its safety. Among these systems is the Passive Containment Cooling System (PCCS) in which natural recirculation and steam condensation, water film evaporation, conduction, and radiation are the key heat transfer mechanisms. These phenomena are capable of removing decay heat from the containment during the postulated design basis accidents (DBAs) such as large break loss of coolant accident (LBLOCA) and main steam line break (MSLB). As a matter of fact, during LBLOCA blowdown phase and MSLB, hot steam jet is discharged from the break and flows upward due to buoyancy-induced natural circulation, then condensates on the inner wall of the steel containment vessel, which in turn acts as a heat sink to prevent the containment from exceeding design limits for pressure and temperature. Therefore, it is crucial to investigate the containment response following abnormal condition to ensure its safety and integrity. In this paper, a CFD-based model was developed using ANSYS FLUENT to study the containment response following the vapor generated by the jet effect. Moreover, the model simulates the steam jet under LOCA conditions, and corresponding thermal-hydraulic phenomena such as natural circulation and wall film condensation during the blowdown phase. The results were compared with the available data reported in the AP1000 WGOTHIC containment evaluation code.
Numerical Study on Subcooled Boiling and CHF Phenomenon in Eccentric Annular Channels
In the period of the increasing energy crisis, nuclear energy is of particularly importantance. It is critical to understand the core behaviors of fuel assembly and ultimately increase the CHF by accurately predicting flow and heat transfer phenomena inside the rod bundle. In the reactor core, the flow is quite complicated, including single phase flow, subcooled two-phase flow and Critical Heat Flux(CHF). The precise computational analysis of how to enhance the safety of nuclear reactors and thereby reduce costs becomes especially significant. Using computational Fluid Dynamics (CFD), it is possible to extract three-dimensional characteristics such as pressure, velocity, and temperature, providing theoretical guidance for nuclear reactor design and operation. By comparing Bartolomei's experimental data to the validated two-phase boiling CFD models, the CFD codes with RPI boiling models are applied to the internal and external flow phenomena of the pipe. The subcooled boiling two phase flow cases are compared. The void fraction and heated wall temperature distribution along the flow direction are discussed and compared under internal and external flow and a comparative analysis is conducted for CHF. It was found that the cold wall of the annular pipe has a significant effect on boiling and CHF. The study further extends the comparison to eccentric annular pipe configurations, exploring the performances of subcooled boiling and CHF. The results offer valuable insights into the understanding of the flow and heat transfer in the rod bundle of fuel assembly.
Numerical Codes Validation of Liquid Droplet Entrainment during Reflood via NRC/PSU RBHT Tests
Liquid entrainment behavior in two-phase flow systems has been found to play a significant role in mass and heat transfer analysis. However, liquid droplets, generated through various mechanisms, are present in the vapor flow with sizes and velocities typically spanning different orders of magnitude. This leads to very complex two-phase flow behavior and poses challenge in accurate characterization of the mass and heat transfer processes. The objective of the current work is to provide a comprehensive evaluation of the liquid entrainment models incorporated in two different thermal hydraulic codes, TRACE and COBRA-TF, using the benchmark experimental data obtained from the NRC/PSU RBHT reflood tests. Detailed numerical simulation models have been developed in TRACE and COBRA-TF for the RBHT test facility. Based on extensive numerical calculations performed, it is found that both TRACE and COBRA-TF are able to capture the overall two-phase mass flow rate at the outlet of the test section. However, both codes tend to over-estimate the liquid droplet entrained in the post-dryout heat transfer regime while under-predicting the outlet vapor mass flow rate. Such model deficiencies may lead to bias in the peak cladding temperature predictions. Based on the current code validations, new physics-based sophisticated droplet entrainment models can be developed to realistically capture the underling physics involved in the liquid droplet entrainment process to further enhance the code prediction capability.
Investigation of Reflood Quenching Phenomena in LWR Type SMRs
With recent developments and emergence of the light water reactor (LWR) type Small Module Reactor (SMR), there has been an increasing demand for evaluation of different thermal-hydraulic characteristics of these type of reactors. Considering the commonly adapted integral design feature in LWR type SMR designs, the thermal-hydraulic characteristics may be significantly different than conventional LWR designs, especially Pressurized Water Reactor (PWR) type SMRs. Additionally, it is well known that due to the inherently safe design principle, the establishment of Natural Circulation (NC) is one of the most critical safety features for all advanced reactors, including most of the LWR type SMRs. Experimental tests have shown that one of the greatest challenges to the establishment of NC under accident conditions is the downward quenching phenomena near the exit of the fuel core as well as the potential countercurrent flow limiting (CCFL) phenomena that might develop between rising void in the boiling hot channels and the downward traveling subcooled water in the cold surrounding channels (such as low power or guide tube or instrumentation channels). Integral SMRs, in particular, have an increased probability of these phenomena occurring due to the high upper plenum to reactor core length ratio. The large volume of unheated coolant in the upper plenum will be one of the key driving forces of the downward quenching phenomena and is one of the major differences between SMR and traditional PWR designs. The highly non-uniform power distributions, both axially and radially, tend to help promote such localized downward reflood quenching phenomena as well. This paper will provide clarification and verification of the reflooding phenomena with CFD modeling and recent experimental observations, while also exploring the impacts and potential risks related to the reflood quenching phenomena in the integral designed LWR type SMRs. To further investigate these impacts, first the difference between the experimental phenomena, in particular the ones observed during conventional PWR rod bundle CHF tests using uniform heaters, and the actual reactor design characteristics must be explored, particularly with respects to measurement uncertainty, thermal-hydraulic instability, and operation safety. Second, the role of mixing vane spacer grid is assessed with relation to the exit quenching phenomena. Different spacer grid designs are examined with respect to their functions in promoting or diminishing the downward quenching phenomena as well as their roles in the establishment of natural circulation. In addition, the impacts of downward reflood quenching on CHF (Critical Heat Flux) level, the location of CHF, and the potential of leading to flow instability are also examined.
Evaluation of Physics-Informed Machine Learning Models for Liquid Entrainment during Reflood Transient using NRC/PSU RBHT Data
Improved prediction of liquid droplet entrainment for post-dryout reflood heat transfer through the development of physics-based models is of high priority for enhancing the capability of nuclear reactor thermal hydraulic codes. However, due to the very complex and transient two-phase flow behavior during reflood, theoretical modeling of the mass and heat transport processes are extremely difficult. As a result, most liquid entrainment models available today are empirical correlations which are typically found to over-predict the droplet entrainment. The current paper aims at developing a physics-informed machine learning (PIML) model that could deliver more accurate evaluation of the two-phase flow entrainment behavior with improved model reliability and efficiency. Based on the comprehensive experimental data obtained from the NRC/PSU RBHT reflood tests, various pure-ML and PIML models have been developed and assessed. It is found that both pure-ML and PIML models can capture overall entertainment correctly and significantly improve the precision accuracy as compared to the conventional models. Random forest ML architecture typically had better performance than that of artificial neural network (ANN) architecture. In addition, effects of the newly developed PIML models on TRACE reflood transient simulations were also investigated.
Flow Instability Tolerance Simulation Heater for SMR and Other Rod Bundle CHF Tests Under Low Pressure, Low Flow, and High Quality Conditions
In order to properly define the critical heat flux (CHF) limits for any fuel design, electrically heated simulation heater rods (SHR) are generally designed to allow CHF testing to be done with various representative heat flux profiles and bundle assembly configurations. To achieve high heat flux density and accurate CHF measurement representative of the reactor, with constant input of heat flux during the CHF transient event, and without the non-conservative uncertainty resulting from internal heat conduction, direct simulation heater rods (DSHR) with electrical current applied directly to the outer metal cladding for Joule heating are preferred over indirect simulation heater rods (ISHR). DSHR are normally designed to allow for gradual thermal expansion as CHFs under high pressure conditions are often very localized DNB events with relatively small accumulated expansion throughout the CHF transient process. However, under low pressure, low flow (LPLF) conditions, the system is subjected to flow instability induced CHF events that occur rapidly and could happen throughout the entire channel lengths rapidly within very short period of time (often in seconds or less) and often result in heater rod failure. In addition, such rapid flow and pressure oscillations often result in loss of flow measurement and control. Due to such experimental challenges, there are very few rod bundle CHF data obtained under LPLF conditions. With the development of SMRs, a method of obtaining safe and accurate measurements of rod bundle CHF under LPLF conditions has become imperative. An innovative simulation heater, a Flow Instability Tolerant Simulation Heater (FITSH) is presented with supporting test data in this paper utilizing special designed dynamic balancing mechanism to handle the fast transient event and allow safe and accurate measurements of flow instability induced CHF under low pressure (as low as 1.2 bar), low flowrate, and high-quality conditions. Additionally, initial testing and benchmarking performance results of the FITSH are presented, including response time evaluation and LPLF CHF data.
Development of Data-Driven Eddy Viscosity Closure to Support MSR Flow and Heat Transfer Modeling Based on SAM-ML
Safety analysis of advanced non-LWRs often involves predicting sophisticated thermal-fluid (T-F) phenomena that occur in various reactor components with complex geometric setups. An example is the liquid-fueled molten salt reactor (MSR). The System Analysis Module (SAM), a modern nuclear system code, has developed a multi-dimensional (multi-D) flow model to address the modeling of complicated T-F phenomena in advanced reactors. SAM’s multi-D flow model employs a coarse mesh setup to ensure computational efficiency and consistency with its 1-D components. Consequently, this requires constitutive relations in the model for unresolved fine-scale physics, such as turbulence. In this paper, we introduce a novel neural network architecture that serves as a data-driven eddy viscosity closure to complement SAM's multi-dimensional module, supporting the modeling and simulation of MSR. The newly proposed neural network draws on flow features from neighboring flow regions to predict local eddy viscosity. This novel architecture satisfies the physical constraints of the fluid system (i.e., Galilean invariance) and is mesh-agnostic, thus significantly enhancing its flexibility and generality by accommodating irregular computation domains. We have integrated the developed data-driven closure with SAM through the SAM-ML capability, facilitating the modeling of flow and heat transfer in the MSR's primary fuel circuit. The integrated SAM-ML model demonstrated good agreement with high-resolution CFD results. The network utilizes non-dimensional local flow features as inputs, the developed capability showed good performance for various Reynolds number flow conditions within the fully turbulent flow regime.
Challenges regarding the Experimental Validation of Thermal-Hydraulic Modeling Tools for a European SMR
Interest in SMRs has been growing steadily in recent years, promising market competitiveness through technological optimizations such as simplified passive safety features compared to conventional NPPs. Innovation in passive safety involves deeper studies on underpinning physical phenomena, requiring thermal-hydraulic system codes to support reactor design, safety assessment, and licensing. On the other hand, the predictive capabilities of the safety systems codes need to be proved through the verification and validation process. The present study aims to highlight some of the SMR-relevant phenomena for which existing thermal-hydraulic codes require dedicated validation. This study encompasses a literature review and the analysis of accidental transients with a RELAP5 model of a potential European SMR, characterized by an integral layout and the adoption of plate-type compact heat exchangers for both steam generators and safety heat exchangers. In particular, the investigated phenomena are different phenomena affecting the natural circulation based DHRS for long-term cooling, thermal stratification at the pool’s containment and the influence of the increased amount of internal structures that feature the integral SMR designs. In the case of the DHRS behavior and containment pool cooling, an extensive literature review has revealed the need for a deep experimental approach for validating their analytical models. In addition, the RELAP5 model analysis has shown that inner structures may have certain impact on the accident dynamics.
Study on Eutectic Melting Behavior of Control Rod Materials in Severe Accidents of Sodium-Cooled Fast Reactors (1) Project Overview and Progress Until 2022
It is necessary to simulate a eutectic reaction and relocation behavior of boron carbide (B4C) as a control rod material and stainless steel (SS) during a severe accident in a next-generation sodium-cooled fast reactor designed in Japan. On that account, Japan Atomic Energy Agency with academic and industrial partners have been conducting a research project consisting of thermophysical property measurement of the B4CSS eutectic melt, B4C-SS eutectic melting and relocation experiments, eutectic reaction mechanism investigation, and physical model development for a severe accident simulation code and its application to the reactor. This project has accomplished the development of a basic physical model to simulate the B4CSS eutectic reaction in Phase 1 (2016-2019). The objective of Phase 2 (2020-2024) is to upgrade the physical model for the improvement of reactor analysis accuracy. The model improvement is characterized by implementing a variable B4C mass fraction for the B4C-SS eutectic material. This paper describes the project overview including progress until JFY2022. As one of the state-of-the-art results in the experimental studies, this paper also presents a first freezing experiment using the B4C-SS eutectic material.
Mechanistic Dryout Predictions in Uncrept and Crept CANDU Pressure Tubes with MEFISTO-T Subchannel Analysis Code
In support of core thermal-hydraulic code development and applicability demonstration, OECD/NEA has organized a benchmark on CANDU fuel thermal-hydraulic (CTH). Specifications have been released based on full-scale rod bundle heat transfer experiments in a 28-element fuel assembly, covering both CANDU and BWR design operations. The benchmark database includes a single set of axial pressure drop (single-phase and two-phase) and critical power tests (non-blind phase) released for code calibration purposes. Boundary conditions for the remaining part of the database were released without measurement data (blind phase). The tested geometries include CANDU pressure tubes under both uncrept and crept (i.e. ageing) conditions, as well as various bearing pad heights. The crept geometry creates a significant bypass on the upper side of the fuel lattice and associated crossflows which are notoriously challenging to capture by simulation codes. The provided benchmark database was simulated with Westinghouse subchannel analysis code MEFISTO-T using a two-phase three-field approach of annular two-phase flow. Code calibration parameters related to form loss and drop deposition enhancement due to structural components (endplates and spacer pads) were adjusted based on the limited non-blind database. The axial pressure drop was accurately predicted, including under two-phase conditions. Furthermore, the radial position of dryout was correctly predicted by the code for the provided case, however downstream of the measured position. After calibration, all blind cases were simulated. Further assessments of code prediction performances will be performed after release of the benchmark data.
Examination of the Reason Why Hydrogen Explosion did not Occur at Fukushima Daiichi NPP Unit-2
During the Fukushima Daiichi NPP accident, hydrogen explosions occurred in the reactor buildings of Units-1, 3, and 4, but didn’t happen in the reactor building of Unit-2. One potential reason why Unit-2 avoided the explosion can be attributed to the mitigation of hydrogen concentration inside the reactor building, which occurred due to the accidental opening of the BOP on the operating floor. This hypothesis was quantitatively examined by assessing the flammability of the environment inside the reactor building using GOTHIC code. The hydrogen was assumed to enter the reactor building through the reverse flow of the vent gas from Unit-1 and leakage from the drywell of Unit-2. The conditions of the reverse flow and leakage were evaluated according to the predicted accident scenario based on the onsite measurement data. Gas composition on each floor was assessed by simulating its distribution throughout the entire reactor building via staircases and hatches. This information was subsequently employed to evaluate the gas's flammability using H2-Air-Steam ternary diagram. The results showed that gas exchange through the BOP opening effectively reduced the hydrogen concentration in the reactor building and only the 5th floor environment became flammable for a short period. On the other hand, assuming the BOP were closed, the entire reactor building would become flammable for an extended duration. It is suggested that the main reason for the absence of a hydrogen explosion in the reactor building of Unit-2 might be the opening of the BOP, which corroborated the conventional estimate.
Code Analysis on the Loss of Flow Scenario for a Small Modular Reactor
Based on the accumulated pool type sodium cooled fast reactors technology the SALUS reactor is being conceptually designed by KAERI to maximize the utilization of u r anium resources using fast neutrons. The GAMMA code has been developed to predict the physical phenomena expected following the anticipated as well as postulated accidents in a high temperature gas cooled reactor and further updated and extended to non LWR (light water reactor) such as liquid metal reactor and molten salt reactor. The major results from the analysis on both passive decay heat removal system (PDHRS) performance and unprotected loss of flow (ULOF) scenario for the SALUS were summarized as a part of the development and assessment of the GAMMA+ code. First of all, a modeling on the PDHRS of SALUS was performed using the GAMMA+ code to analyze the design values of DHX and AHX heat exchangers. Second, modeling and safety analysis utilizing the GAMMA+ code for the accident scenario of ULOF has been conducted in order to assess the applicability of the code to the SALUS reactor. In analyzing safety for the ULOF scenario, evaluations were conducted for two scenarios of the power loss of two primary heat transfer system (PHTS) pumps and the pseudo trip of one PHTS pump. This verified that the GAMMA+ code could be used to the safety analysis on the sodium cooled fast reactors such as SALUS with the similar accuracy to the MARS LMR code.
Simulation of Pressurized Conduction Cooldown in a Scaled High Temperature Gas Reactor Experiment using GOTHIC
Among the candidates for the next generation of nuclear power plant is the high temperature gas reactor (HTGR). Since the HTGR does not have the same operational history as water cooled reactors do in the United States, investigation into the response of different accident phenomena is still an active area of research. The current work investigates the pressurized conduction cooldown (PCC) accident in an integrally scaled high temperature gas reactor experimental facility. A PCC accident occurs when the primary circulator loses the ability to provide forced convection to the core. During a PCC accident, coolant that is normally forced downward through the core during normal operation loses momentum after the circulator functionality is lost leading to flow stagnation and eventually results in an intracore natural circulation flow path which brings hot coolant into the upper plenum. Critical instrumentation such as the control rod drive shafts are located in the upper plenum. As such, it is important to understand the upper plenum mixing and temperature profile to better support the safety of HTGRs during these accidents. The upper plenum environment was analyzed by simulating a PCC experiment in the Oregon State University High Temperature Test Facility (HTTF) using the GOTHIC multi-physics software package. GOTHIC integrates coarse-grid multi-phase CFD and traditional system level thermal hydraulics modeling capabilities using a multi-volume, domain decomposition approach that provides an efficient 3D analysis tool. This paper summarizes GOTHIC’s capabilities for HTGR analysis and compares the upper plenum temperature comparisons for the PCC at HTTF.
Numerical Study on Thermal-Hydraulic Characteristic and Performance of 3-D Printed Circuit Heat Exchanger Channel having Swirling Flow
The supercritical carbon dioxide (sCO2) Brayton cycle is a promising power conversion system for various industrial applications including generation Ⅳ nuclear reactors owing to its compact size, high efficiency, and simple layout. Printed Circuit Heat Exchanger (PCHE) not only has high compactness and effectiveness but also can be applied to a wide range of temperature and pressure conditions, making it a very suitable heat exchanger for the sCO2 Brayton cycle. In this study, a three-dimensional (3-D) designed PCHE channel was proposed and numerically studied. It was found that there are swirling flows in the proposed 3-D PCHE channels owing to the interaction between two flows in different directions. This swirling flow made the local heat transfer coefficient high along the circumferential direction of channels. A comparative evaluation of the thermal-hydraulic performance of investigated channels was conducted by comparing the volume goodness factor with identical hydraulic diameter. As a result, the proposed 3-D PCHE channels showed a higher or comparable thermal-hydraulic performance compared to conventional 2-D PCHE channels. Therefore, the proposed 3-D PCHE channel could be suggested as a PCHE design option for the sCO2 Brayton cycle. Furthermore, the double-sided straight channel having the swirling flow is expected to have distinctions other than the performance because it is known that the swirling flow can mitigate not only the fouling in heat exchangers but also the heat transfer deterioration of sCO2.
Experimental Investigation on Intermittent Gas-liquid Two-phase Flow Approaching Tube Bundle
In an intermittent gas-liquid two-phase flow, liquid slugs are generated and flowing and those can cause flow-induced vibration when acting on a structure such as tube bundle. For example, there are some locations in a steam generator where the secondary side steam-water two-phase flow is flowing as such an intermittent flow. This study focuses on the intermittent gas-liquid two-phase flow approaching tube bundle. Air-water twophase flow experiment was conducted. In this separate effect test, vertical upward two-phase flow approaches the horizontal square array tube bundle and intermittent liquid slugs are generated in large fluid space upstream of the bundle. In a low void fraction flow with the void fraction approximately from 40% to 70%, those non-homogeneous or intermittent impact forces are considered to play an important role, in evaluating the flow-induced vibration of the tubes. The two-phase flow behaviors flowing into the tube bundle were visualized. It was observed that gas slugs and liquid slugs colliding intermittently to the first row of horizontal tube bundle. Void fraction, interfacial velocity and liquid chord length distribution were measured using bi-optical probes, upstream of the tube bundle and adjacent locations to the first row of the tube bundle.
Modelling and Simulation of Fission Product Release from Melt Pools
During the early phase of a postulated severe reactor accident, predominantly volatile fission products (FP) are released into the cooling circuit. With progression of the accident, the FP move faster in the melting core material than in the crystal structures of the solid fuel, which also allows the release of less volatile FP. In addition, volatile species can be formed by certain low volatile FP in oxidizing or reducing conditions, which may significantly influence their release rates. An adequate simulation of the release of such FP species is therefore necessary to provide reliable initial conditions for source term analyses. For the representation of the FP release from molten core material in the lower plenum, a new in-house model based on the diffusion equation is being developed, which is intended to be prospectively integrated into the system code AC2 model basis developed by the German Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH. The thermodynamic quantities of the multicomponent system are described in an idealized way. To consider the varying volatilities of different FP species in different atmospheric conditions, the Antoine equation is applied. Convective mass transfer is considered by an approach based on the analogy of heat and mass transfer. Verification analyses are being conducted by calculations of experiments of the Late Phase Source Term Phenomena (LPP) project (5th euratom framework program). Preliminary results show that the release behaviour of relevant FP is plausibly represented by the new model.
Analysing the Ablation of Concrete, Heat and Mass Transfer during Molten Corium-Concrete Interaction Phenomena
Molten corium, a mixture of nuclear fuel, cladding, and structural elements, forms after the meltdown of a reactor core in the event of a severe accident at a nuclear power plant. This paper presents the development and implementation of a new numerical mixing model in OpenFOAM for a comprehensive analysis of concrete ablation during the Molten Corium-Concrete Interaction (MCCI) process. The well-researched CCI-2 experiment served as the benchmark for model validation. The model can forecast heat transfer, concrete ablation, phase change, mass transfer, and the mixing of concrete and corium. Additionally, the effect of concrete melting temperature on concrete ablation was investigated. We observed that concrete with a low melting temperature exhibited significantly reduced ablation, as well as controlled temperature, velocity, and density fields. Specifically, at a low concrete melting temperature of 1380 K, 90 percent of the concrete was ablated. However, when the melting temperature of the concrete was increased to 1480 K and 1580 K, the ablation decreased to 80 percent and 60 percent, respectively. Our results also revealed that melted concrete and corium, despite having different densities, form a single phase/mixer that enhances concrete ablation.
Experimental Study on Two-Phase Flow and Pressure Drop Characteristics in Stratified Porous Bed
Understanding the behavior of coolant flow through porous media in nuclear reactor safety is essential for evaluating the coolability of debris beds formed from Fuel-Coolant Interactions (FCI) during a severe accident. This paper presents an experimental study focusing on the pressure drop within a stratified particulate bed containing varying diameters of spherical glass (1.5 and 6mm) in both single/two-phase flow. Two layers of different sizes of spherical particles are independently packed in a cylindrical test section to create an axially and radially stratified bed. The experiment measures pressure drops in each layer as water and air move upwards through the bed. For this study, the new DEBECO-LT facility (Debris Bed Coolability–Low Temperature) was designed and developed to serve as a platform for controlled experimentation. The concept focuses on a two-phase flow, using water and air with varying porous medium sizes. For accurate real-world replication, two homogeneous and two stratified beds were used. The findings show that in a horizontally stratified bed consisting of small particles above coarse particles, the pressure drop in the upper layer is more significant in comparison to a homogenous bed composed only of smaller particles, and the pressure drop measured at the bottom is comparable to that found in a similar homogeneous bed. However, for the vertically stratified bed, the pressure drops on the left and right sides are nearly equal. A vertically stratified structure can cause lateral flow, redistributing the flow rate throughout various sections of the packed bed.
Lessons Learned and Best Practices on BWR Severe Accident Uncertainty and Sensitivity Analysis in the Framework of the IAEA CRP I31033
IAEA launched the CRP I31033 “Advancing the State-of-Practice in Uncertainty and Sensitivity Methodologies for Severe Accident Analysis in Water-Cooled Reactors” in 2019. One of the main outputs of this CRP is a series of TECDOC documents presenting the methodologies for performing uncertainty and sensitivity analysis developed by each of the 22 participant institutions and applied to different water-cooled reactors, i.e. PWR (including SMR), BWR, CANDU, and VVER, as well as to the QUENCH-06 experiment. This paper presents the BWR group´s main outcomes from this CRP and the contents of the TECDOC document in preparation, including best practices and lessons learned by the 4 organizations contributing to the TECDOC: 1) CIEMAT, 2) GAEC, 3) CNSNS, and 4) ININ. The first two institutions addressed severe accident scenarios in a BWR/3 with Mark I primary containment, while the last two used models of a BWR/5 with Mark II containment. Except for CNSNS, the participants have chosen a short-term (unmitigated) Station Blackout scenario for analysis, and the scope of the analysis is limited to only in-vessel phenomena. In the case of CNSNS, depressurization and RCIC injection are considered mitigating measures, and the analysis includes both in-vessel and ex-vessel phenomena. The MELCOR code was used by the first three participants, and the MAAP5 code was used by ININ. Information on figures of merit and the set of uncertain code input variables with their associated probability density functions are also presented in this paper.
Probabilistic Safety Analysis in Risk Informed and Performance Based Safety Management
The integration of Probabilistic Safety Analysis (PSA) in Risk-Informed and Performance-Based Safety Management marks a significant advancement in nuclear safety. PSA is crucial for supplementing deterministic safety designs, ensuring the reliability of vital safety functions. It balances design features from a safety standpoint and identifies areas needing reinforcement. A comprehensive plant-specific PSA, encompassing both Level 1 and Level 2, covers internal initiating events and external hazards such as fires, floods, extreme weather, and seismic events in various operational modes. Plant staff are encouraged to conduct much of the PSA internally for effective decision-making. Regulatory guidelines mandate specific quality standards for PSA, requiring advanced methods including human factor analysis, thermal hydraulic analyses, and detailed uncertainty and sensitivity analyses. Acceptance criteria set by regulatory bodies, like total core damage frequency and conditional containment failure probability, form part of this analysis. The recommended risk-informed and performance-based regulatory enhancements aim to focus on significant activities, set risk-based performance evaluation criteria, develop measurable parameters for system and licensee performance monitoring, and prioritize results in regulatory decision-making. This approach not only aligns the objectives of licensees and regulators towards common safety performance indicators but also addresses stagnation in nuclear technology development and improves safety records by promoting licensee initiatives. It clearly assigns the responsibility for safety to the plant’s owner and operator, with regulatory bodies overseeing safety objectives implementation and performance monitoring. This shift in safety management underscores the vital role of PSA in contemporary nuclear safety practices, ensuring a balanced, thorough, and responsible approach to nuclear safety.
Investigating Post-CHF Heat Transfer Behavior During Dryout and Rewet Cycles
Due to the better thermo-mechanical and oxidation properties of accident tolerant fuel (ATF) materials, a shift from critical heat flux (CHF)-based design safety criteria to a “time and temperature” criterion is being discussed for light water reactors. To make it possible, however, large uncertainties in the post-CHF heat transfer regime need to be addressed. This work investigates the uncertainty of the transition boiling models adapted in three commonly used nuclear fuel performance and thermal-hydraulic codes: BISON, TRACE, and RELAP. This is achieved by comparing the calculation results against the instability tests performed at the Karlstein Thermal Hydraulic Test Loop (KATHY) facility. The data were collected using a 10×10 rod assembly prototypical of a boiling water reactor (BWR) under natural circulation at operational pressure and power. The heat transfer behavior in the transition regime is divided into two phases: the “dry-out” and the “rewet” phases. The results show that the studied models do not fully capture the differences between these phases. The predicted wall superheat does not agree with the experimental values, while heat transfer coefficients show even greater discrepancy. Furthermore, all models are unable to predict the “failure to rewet” behavior at the end of the heating cycle. Therefore, this study concludes that further improvement of the nuclear safety codes is essential to differentiate and capture the transition boiling heat transfer, which significantly impact the transient clad temperature over time after CHF is achieved.
Mass and Energy Release Analysis of Postulated Main Steam Line Break Accident for APR1400 Using SPACE-ME Methodology
The mass and energy (M/E) release from postulated main steam line break (MSLB) accidents is essential for reactor containment functional design. Korea Electric Power Corporation Engineering & Construction Company, Inc. (KEPCO E&C) is currently developing a novel M/E release analysis methodology, known as SPACE-ME methodology, utilizing the SPACE 3.3 and CAP 3.1 codes. In this study, the M/E release analysis of postulated MSLB accidents in Advanced Power Reactor 1400 MWe (APR1400) was performed using SPACE-ME methodology. The M/E release rates were calculated with various combinations of initial core power conditions of 102%, 75%, 50%, 20%, and 0 % of full power (%FP), break sizes of area fractions ranging from 0.1 to 1.0 for a total area of 0.8542 m2. In addition, single failure, such as main steam isolation valve failure (MSIVF) or one train of containment spray system failure (CSSF) is assumed for each postulated MSLB accident. Subsequently, the containment response analysis to containment pressure and temperature (P/T) were conducted with the M/E release results by the stand-alone CAP 3.1 code. The highest containment peak pressure was obtained as 388,389 Pa (56.33 psia or 41.63 psig) at the MSLB accident with MSIVF, 102 %FP, and AF 0.3. The highest containment peak temperature was found to be 449.27 K (349.02 °F) in the MSLB accident with MSIVF, 102 %FP, and AF 0.4. The margins of containment peak P/T are determined as approximately 24.6% and 1.4%, respectively, compared to environmental qualification limits for design basis accidents. In addition, the containment peak pressure is lower than that of calculated by KIMERA methodology, while the containment peak temperature is higher. Consequently, this paper presents the SPACE-ME methodology is valid to analyze the M/E release of the postulated MSLB accidents.
Formation Behavior of Gaseous Iodine from Sodium Iodide under SFR Severe Accidental Condition
Many experimental and analytical studies on the source term such as fission products released from nuclear fuels to the external environment have been carried out since the severe accident in Fukushima Daiichi Nuclear Power plant. Currently, source term research is being also focused in the field of advanced nuclear power plant such as sodium-cooled fast reactor (SFR). In SFR, sodium (Na) chemistry should be considered for source term evaluation because Na is used as coolant in terms of excellent heat transport capability. Iodine as one of important fission products is reactive with Na. Therefore, sodium iodide (NaI) as reaction product is formed in the reactor vessel under SFR severe accidental condition. If considering the transport of NaI from the reactor vessel to the containment vessel with air atmosphere, almost NaI is likely to behave as aerosol. However, it was reported that gaseous iodine as non-aerosol was formed from some NaI depending on the atmospheric conditions. In this study, thermal behavior of NaI in different atmospheric conditions was investigated using thermal analysis technique to understand the formation behavior of gaseous iodine. Based on the results, it was revealed that the thermal behavior of NaI in inert atmosphere is phase change and approximately 10 mass% of I2 could be formed from NaI in 20 vol% oxygen atmosphere. The chemical equilibrium simulation was also performed to understand the mechanism of I2 formation. The estimated value of I2 formation in 20 vol% oxygen atmosphere was comparable to the result of thermal analysis.
IAEA CRP I31033: Severe Accident Uncertainty and Sensitivity Analyses for Reactors of PWR and SMR Types and Relevant Insights
In 2019, the International Atomic Energy Agency (IAEA) launched the five-year Cooperative Research Project (CRP) I31033 to advance the understanding and characterization of sources of uncertainty and investigate their effects on the key figures-of-merit (FOMs) of the severe accident code predictions in watercooled reactors (WCRs). Twenty-two institutions from 18 Member States participated in the CRP and, as a result, five TECDOCs are being developed: each for WCRs of four types [PWR (inclusive of watercooled SMR design), BWR, CANDU, and VVER] and one for the QUENCH-06 experiment. Respective TECDOCs address a specific exercise and outline relevant research results pointing to best practices on the uncertainty and sensitivity (U&S) analyses of the currently available severe accident codes. For the uncertainty analysis exercise, the CRP participants defined their own analysis scope and framework, including target plant, severe accident code, accident scenario, and relevant FOMs. General conclusions were made based on the analysis of the results in view of the best-practice application of the U&S methods. Among them, this paper summarizes the main results of the U&S exercise performed for reactors of PWR and SMR types within the context of the IAEA CRP and relevant insights.
Void Redistribution Measurement Downstream of Part-length Rods in Rod Bundle with High-Energy X-Ray CT
The fuel rod bundle of boiling water reactors (BWRs) is implemented with part-length rods (PLRs) to improve the pressure drop and optimize the power profile. The complex rod bundle structure can affect the void fraction distribution. Prediction of void fraction distribution is essential for achieving high burnup of fuel rods and optimizing the thermal margins of the rod bundle. This paper focuses on detailed void behavior downstream of PLRs under BWR operating pressure conditions. The test section is a 5×5 rod bundle with five PLRs arranged in the center, side, and corner of the rod bundle. The heated lengths of the full-length rod and PLR are 3708 mm and 1854 mm, respectively, and the power profile is uniform in both axial and radial directions. A boiling flow experiment is conducted at a pressure of 7.2 MPa with inlet mass flux, inlet water temperature, and rod bundle power as test parameters. The cross-sectional void fraction distributions are measured with a minimum axial interval of 25 mm using a linear accelerated-driven high-energy X-ray CT system with a working distance of 4 m in the axial direction and a pixel resolution of 80 microns in CT reconstructed images. The experimental results show the void redistribution behavior downstream of the PLRs, i.e., the void fraction reaches a localized maximum just downstream of the PLR end and then decreases due to the sudden expansion of the flow channel. Further downstream, it recovers to the comparable void fraction of the surrounding subchannels.
Assessment of Two-Phase Flow Predictability of 3D Components in MARS-KS and TRACE in Rod Bundle Geometry
This study aims at assessing the 3D components in MARS-KS and TRACE, in order to confirm the appropriate capability of both system codes under the two-phase flow conditions in the rod bundle geometry. For the assessment, two-representative separate effect tests, namely GE 3X3 and PSBT bundle, were employed, and code-to-code comparison has been made against the experimental results, statistically. From the assessment, it was revealed that both codes did not predict the phasic distribution of bundle, appropriately. It was observed that TRACE significantly overpredicted vapor distribution of the experiments compared to MARS-KS. The overcalculation of TRACE was attributed from the restricted crossflow calculations compared to MARS-KS, as it imposed large interfacial drag applying drift flux model for the crossflows. In case of MARS-KS, it also featured inappropriate crossflow behaviors when compared to the experimental data. This result was obvious since both codes neglected the presence of secondary flows due to substantial flow disturbances by rod bundle. Employing the subchannel mixing model into MARS-KS, it was observed that more accurate phasic distribution could be simulated by the code, consequently, improving the overall code predictability. These findings clearly indicated that the code predictions were significantly affected by the accurate crossflow predictions in bundle-scale level. Hence, the improvements to the crossflow models of both codes were necessary for bundle-scale analyses.
Development of an Advanced Freeze Valve System for Molten Salt Reactor
Molten Salt Reactors (MSR) offer distinct advantages, including low system pressure, small-sized core, and minimal nuclear proliferation risk. Yet, MSRs are facing challenges such as high system temperature and the corrosiveness of the coolant material. These harsh conditions render conventional mechanical valves inadequate for MSR flow control. To overcome this problem, Oak Ridge National Laboratory (ORNL) has designed a specialized freeze valve system to address coolant transport within their MSRE (Molten-Salt Reactor Experiment) projects and proven in operation for over 20,000 hours. Despite its success, issues like slow response time for opening and closing, limited experiences for small diameter pipes only, and leakage detected at the end of the project due to improper field reform calls for further enhancements. Leveraging the lessons learned from the MSRE project, our current initiative is to explore the development of an advanced freeze valve system suitable for a new 100 MWth-class MSR. This system is envisioned to support carbon-free marine power systems, including ship propulsion and offshore floating power plants. This work is preparatory; we are going to present study of the MSRE freeze valve in deep, basic concepts and the preliminary proof of concept efforts to mitigate previous limitations, aiming to deliver a prototype that excels in response time, leak prevention, flow rate, and ultimately aiming to enhance MSR safety and efficiency.
Modeling of Heterogeneous Nucleation from Surface Nuclei and Application to the Evaluation of the Void Ratio in a Molten Salt Reactor
In molten salt reactors (MSRs), the nuclear fuel is dissolved in the liquid salt coolant. This design allows prolonged exposure of the fuel to the neutron flux, facilitating the transmutation of minor actinides. Thus, MSR research is being conducted with the goal of reducing the radioactive toxicity of nuclear waste. In this design, the gaseous fission products pose a challenge because they are poorly soluble and tend to precipitate as bubbles. These bubbles affect neutronic reactivity and hence reactor stability. In particular, the heterogeneous nucleation of gas bubbles on the vessel walls is studied. When liquid salt is poured into the vessel, small amounts of filler gas, called nuclei, can be trapped in the cracks in the walls. The dissolved gaseous fission products can diffuse to such nuclei, leading to their growth and eventually to the detachment of bubbles. These detached bubbles migrate in the salt where they can remain for tens of seconds, modifying the void ratio and thus the reactivity in the core. Our aim is to evaluate the effect of this heterogeneous nucleation on the void ratio. A theory of the nucleation process is developed, including mass transfer in a turbulent flow, the kinetics of gas exchange with the bubbles, and the forces triggering detachment, in order to determine the frequency of bubble production. The transport of the detached bubbles in the salt and their behavior in the reactor are modeled to finally give an estimate of the void ratio due to these bubbles.
MELCOR Model of iPWR and Application to Emergency Planning Zone Study
This study undertakes MELCOR modelling and analysis of a generic integral pressurized water reactor (iPWR) system with the goal of informing a broader study on emergency planning zone (EPZ) sizing. The iPWRs are one of the small modular reactor technologies that have high technical readiness and are potentially being considered for near term deployment. To understand possible iPWR accident scenarios/behaviors important to support emergency preparedness, this study uses the severe accident analysis code MELCOR. Station blackout (SBO) was selected as the base case along with other hypothetical accidental scenarios such as loss of steam generator feedwater, failure of the decay heat removal system, breaks at operating pool, and breaks in the containment vessel. The analyses investigate accident progression in the reactor vessel, containment, and operating pool/reactor building, as well as radionuclide releases for the assessed cases to provide insights on EPZs. The simulated cases in this study provide a spectrum of mechanistic source terms that can fit into the broader probabilistic framework of accident frequency that could inform decisions on EPZ sizing.
Evaluation and Demonstration of Predictive Capability Maturity Quantification (PCMQ) Method for Assessing Software Validation and Applicability
The use of any modeling and simulation (M&S) tool, such as GOTHIC, MAAP, RETRAN or VIPRE, requires the tool be qualified for use for a given application. Generally, this process requires the tool to adequately address the important phenomena with a sufficient level of accuracy; however, these determinations tend to be largely qualitative and rely heavily on expert knowledge and engineering judgement. That approach may be sufficient for well-established applications, but not first-of-a-kind (FOAK) applications, such as advanced reactor design, where expertise and operational experience may not exist. Predictive Capability Maturity Quantification (PCMQ) type methods have the potential to provide a quantitative metric for evaluating adequacy considering the applicability, scaling, relevance, and coverage of the available evidence as well as the accuracy of the M&S tool. PCMQ can potentially also identify which experiments/benchmarks are most applicable for a given application, whether the available benchmarking is sufficient, and what improvements or additional benchmarks would provide the most value. The current work presents a feasibility study with a proof-of-concept to evaluate the process, identify potential constraints or limitations, and assess the credibility of applying PCMQ for an industry tool like GOTHIC∗. Two use cases are presented – a simpler Separate Effects Test (SET) and a more complex Integrated Effect Tests (IET) – that GOTHIC has already been benchmarked against. This approach provides a controlled environment for evaluating PCMQ since the applicability of GOTHIC for those scenarios is well-established and can be confirmed with comparisons to the experimental data.
Experimentation and Modeling of Subcooled Flow Boiling from the Perspective of Large Bubble Formation
The authors’ previous studies are summarized to discuss that mechanistic or semi-mechanistic prediction from the perspective of large bubble formation (LBF) is possible for the occurrences of the onset of significant void (OSV) and the departure from nucleate boiling (DNB) in subcooled flow boiling. From the analogy with the flow regime transition from bubbly flow to slug flow in gas-liquid two-phase flows, it may be assumed that the global bubble coalescence leading to large bubble formation takes place in subcooled flow boiling when the local void fraction exceeds the critical value about 0.3. In addition, the commercial CFD codes are a satisfactory tool to calculate the local void fraction accurately. From these facts, it is discussed that the CFD codes can be used to carry out mechanistic and conservative prediction of the DNB heat flux even under the complex situations such as the flow channel of complex geometry with mixing vanes, inclined channel, and rolling and/or pitching motion.
Analyses of an Integral Pressurized Water Reactor During Postulated Accident Conditions Using the System Code Package AC²
In the frame of this work, the model of a generic integral Pressurized Water Reactor (iPWR) with 160 MWth is analyzed regarding the possibility of severe fuel damage during a postulated accident. This is relevant due to a growing interest on building iPWR and consequently intensified research on this reactor type. The simulation model is developed using the system code package AC² developed by GRS gGmbH focusing on the thermohydraulic and core degradation modules. The accident tolerance during the postulated failure of selected components is analyzed. Overall, the code is able to calculate plausible results for the generic model of the iPWR. Under design conditions, the passive core cooling mechanisms are efficient enough to prevent further degradation. Postulating the failure of additional selected components results in endangered core integrity. The heat up and steam oxidation at the top part of the core starts after the core level lowered, as expected. In case of core melting and relocation, the late-phase module AIDA is assessed with regard to possible melt retention and the influence of the necessary heat transfer coefficient in the lower plenum for a successful external cooling. Depending on the progression of the postulated accident, one important factor might be the fuel cladding, which is usually a Zr-based alloy. This cladding material possibly contributes to a temperature escalation due to the exothermic steam reaction. As a possibility, the accident tolerance could be further increased using alternative materials. Therefore, the application of slower oxidation kinetics of near-term candidates of Advanced Technology / Accident Tolerant Fuel (ATF) is investigated.
Conceptual Design of Component Test Facility (CTF) for Investigation of FACE Phenomenon in Molten Lead
Lead Fast Reactors (LFR’s) are being identified as Gen IV advanced reactor technologies. Despite its significant advantages for its high boiling point, excellent cooling properties and chemical inertness, there are research challenges associated with development of the technology. It includes its opacity, high density, high temperatures and the potential for corrosion/erosion when it contacts with structural steels/components. Hence it is necessary to test the materials/components in molten lead at reactor conditions to investigate the LFR phenomena and to validate the modeling and simulation tools. This paper focusses on the design of a separate effect facility for testing pump designs/impellers in molten lead at different flow rates at reactor operating conditions. The conceptual design of the facility to operate at different pressure heads has been discussed with numerical analysis. The Component Test Facility [CTF] is a pump facility to test the components like pump impeller and valves for Flow Accelerated Corrosion and Erosion (FACE) at nominal operating conditions in LFR’s. It is a pool type pump system which accommodates integrated flow obstruction and the drain system to achieve the required pressure drop. The baseline concept includes the flow obstruction system using Porous medium to achieve pressure drop and the drain system with Venturi to measure mass flow rates. In this work, computational investigations have been performed using COMSOL Multiphysics to estimate the flow properties and pressure head in the pump with an axisymmetric model for different flow resistances. Simulations are carried out using Turbulence modeling with Free and Porous media flow module. The results were used to calculate the system curve for the pump facility and estimate the flow velocity for different mass flow rates. Also, flexible accommodation of different pump curves to test has been discussed.
Development of Thermalhydraulic Modelling and Simulation Capabilities in CNL in Support of Advanced Reactors
Within Canada, several advanced reactor concepts are undergoing vendor design review or site license application with the Canadian Nuclear Safety Commission (CNSC). Canadian Nuclear Laboratories (CNL) is a major hub of R&D that supports the CNSC, SMR stakeholders, industry, and academia. CNL is developing a unique multiphysics modelling code suite to support lab-scale experimentation and assess reactor safety under normal conditions and during accident transient events. This paper showcases ongoing efforts at CNL towards the development of standalone and coupled thermalhydraulics modelling capabilities to simulate postulated scenarios of relevance to gas-cooled advanced reactors. The main focus of this study is to showcase the participation of CNL in the application of standalone CFD for modelling the mixing in the lower plenum of the Oregon State University High Temperature Test Facility, within the HTGR T/H benchmark organized by OECD/NEA. The CFD results for the lower plenum mixing (problem #1 of the benchmark) using fixed boundary condition (Exercise #1), and the progress made in the coupling of Siemens STAR-CCM+ and the CNL system thermalhydraulics code ARIANT are discussed. Based on the significant progress achieved to date, CNL is well positioned to contribute to the next-phase Exercise #2 within the OECD/NEA HTGR T/H international benchmark on the modelling of lower plenum mixing.
Thermal-Hydraulic Behavior in Containment as Influenced by Steam Injection Location: Insights from the OECD-ATLAS3 C5.2 Separate Effect Tests
This study presents the C5.2 test series, a part of the OECD/NEA ATLAS phase 3 project, conducted at the CUBE (Containment Utility for Best-Estimate Evaluation) facility. The focus was on analyzing thermal-hydraulic behaviors during constant-rate steam injection, emphasizing heat transfer between fluid and structure. The research included two steam injection scenarios: RUN1, via the reactor coolant pump discharge position, and RUN2, through the main steam line position. Extensive measurements of containment pressures, fluid and wall temperatures, and relative humidity were undertaken to provide comprehensive insights on the thermal-hydraulic phenomena. The highlight of the present study was the observation of thermal stratification variations according to steam injection location and the changing role of the compartments as passive heat sinks, influenced by the formation of distinct thermal boundary layer. The test results demonstrate how the role of compartments as passive heat sinks varies with thermal stratification, influenced by the location of the break within a containment. These findings offer a valuable database for validating containment safety analysis codes. Additionally, the study reveals that the lumped model code, commonly used in containment safety analysis, may produce non-conservative calculations depending on the patterns of thermal stratification.
Preliminary Assessment of Cooling Performance in Small Modular Reactor Metal Containment Vessel according to the Filling Gas Types.
As an innovative feature of small modular reactors (SMRs), SMRs have employed a metal containment vessel (MCV) surrounding a reactor pressure vessel (RPV). According to the double-vessel structure, a gap is formed between MCV and RPV. The atmospheric condition of the gap is maintained either by filling it with inert gases or by creating a vacuum to prevent heat loss in normal operation and to enhance the reactor safety in a transient state. In particular, the type of inert gases in the gap affects the condensation performance within the SMRs. However, as research on analyzing condensation performance under mixed environments of inert gas and steam in the gap during accident situations is insufficient, further studies are needed. Thus, this study focused on evaluating the condensation performance within the MCV under different gap-filling gases, including argon, nitrogen, and vacuum. The experiment was conducted under transient conditions simulating scenarios such as automatic depressurization valve (ADV) operation. Experimental setup and conditions were established based on preliminary studies utilizing MELCOR and computational fluid dynamics (CFD) codes. Our findings indicate that the vacuum gap exhibited superior cooling performance compared to other non-condensable gases (NCGs). While our study provides valuable insights, it is essential to acknowledge its limitations, including experiments and a defined test matrix based on calculations. Overall, this research will contribute to understanding heat transfer mechanisms and optimizing cooling performance in SMRs, advancing their development across various applications.
Transient Simulation of a Steam Rankine Cycle using CATHARE-3 System Thermal-hydraulic Code: Application to Superphenix Fast Reactor
In this paper, a simplified model of the Superphénix (SPX) steam Rankine cycle is drawn and implemented in the CATHARE-3 transient two-phase thermal-hydraulic system code. First, a comparison of several levels of simplification of the Superphénix Rankine cycle is made using the CYCLOP steady-state pre-conceptual design tool. Then, the most simplified thermodynamic cycle that reproduces accurately the physical parameters at key points of the cycle at the nominal state is modeled using CATHARE-3. The model is compared with the nominal state of the Rankine cycle simulated by CYCLOP. It is observed that physical parameters at nominal state are consistent with CYCLOP results (within 10% discrepancy for 15/17 parameters analyzed). The model is then used to simulate a transient of reactor normal operation: a power decrease from 100 to 40% of Nominal Power (NP) by acting on the secondary sodium flow rate in the Steam Generator (SG). The model shows an expected physical behavior since the whole system adapts itself well to a change in the power exchanged in the SG, physical parameters in the cycle such as pressure and temperature stabilizes at lower values which is what occurs in an actual Rankine cycle at partial load. In the next future, it will be necessary to use transient experimental data of the SPX Rankine cycle in order to evaluate more quantitatively the model under dynamic conditions.
Comparison Between EDF MAAP5.04 and ASTECv3 Codes on Hypothetical Severe Accidents in an Integral PWR
This paper presents a comparison between EDF MAAP 5.04 and ASTECv3 codes in simulating postulated Severe Accident (SA) scenarios in a generic integral PWR characterized by a submerged containment and about 60 MWe. The code crosswalk was performed by EDF and Tractebel within the Horizon Euratom project SASPAM-SA (Safety Analysis of SMR with PAssive Mitigation strategies – Severe Accidemt). MAAP is a deterministic code licensed by EPRI whilst ASTEC is an integral code developed by IRSN. Both codes were validated through intense and extensive validation activities and recognized with MELCOR as reference tools for SA deterministic studies in nuclear reactors. The aim of this work is to investigate capabilities and limitations of SA codes, as well as identify causes leading to relevant deviations in results. This is because the phenomena that occur in SMRs may be different from those that occur in commercial large reactors and may not be captured correctly by models currently implemented. Comparisons performed between EDF MAAP5.04 (SMR code version developed by EPRI and improved by EDF) and ASTEC V3.1 include accident progression from initial events to long-term in-vessel retention of the corium. Particular attention was also paid to the solution adopted to reproduce the strong vessel-containment interaction typical of SMRs. The main differences observed in the results were related to core degradation evolution and corium slumping mechanism in the lower plenum. Discrepancies in core degradation kinetics depend partly on how debris formation is modeled and partly on the different amounts of power exchanged between vessel and containment predicted by the codes.
Robust Abnormality Diagnosis Model Considering Nuclear Power Plants Data Trends
The development of artificial intelligence applications in the real world has been driven by enhancements in computer performance, the accessibility of big data, and advancements in deep learning algorithms. For instance, artificial intelligence in the autonomous driving holds the potential to enhance driving convenience. However, accidents can occur when artificial intelligence encounters differences between training and testing environments such as handling complex road or extreme weather conditions. Even in the nuclear field, applying abnormality diagnosis models using artificial intelligence to real nuclear power plants may lead performance degradation due to data differences between the simulator and real nuclear power plant variables. Therefore, we aim to develop an abnormality diagnosis model that trains on simulator data and tests on real plant data using neural networks and fuzzy logic. Conventional artificial intelligence models excel at responding to microscopic data changes, but they are vulnerable to macroscopic changes in test cases that deviate from training situations. To overcome these limitations, we use fuzzy logic that considers data trends in plant variables. Simulation plant variables are transformed into input values representing the degree of change, and input values are passed into fuzzy membership functions like decrease, maintenance, and increase. Consequently, the plant variables are converted into different types of variables that represent how much has changed compared to the past by fuzzification. Fuzzy values enhance robustness of an abnormality diagnosis model on real power plant data with data differences, even when trained solely on clear simulation data.
A Comprehensive Study of Aerosol Transport and Deposition Behavior in Severe Reactor Accidents: Experimental and Modeling Approaches
Radioactive aerosols may escape into the environment as the outcome of core meltdowns caused by severe accidents. The transport and deposition of these aerosols during such events can have significant and wide-ranging consequences. This study intends to advance the core damage assessment technology, with a particular focus on Level 2 probabilistic safety assessment, by providing a source term migration evaluation method using narrow circular channels. To achieve this, an experimental setup has been established that represents the leak paths via containment vessel penetrations and reactor buildings, as well as the environmental and flow conditions during severe accidents. Experiments were carried out to investigate the effects of flow Reynolds number, aerosol particle size, and leakage path diameter and length on the decontamination factors (DFs). In addition to experimental endeavors, a theoretical model has been developed to investigate the deposition and transport of accidental source terms, considering the re-entrainment effect. Both the particle concentration and the deposited particle numbers on the surface are considered position and time dependent in the model, so it can predict the transient DFs along with the average DFs that are obtained from the experiment. The model results are compared and validated with the experimentally obtained DFs. These experimental and theoretical results reveal that DF increases with flow path length and Reynolds number while decreasing with flow path diameter and particle mass median diameter at higher Reynolds number ranges.
A Three-Dimensional Multiphysics CFD Analysis of Cobalt Adjuster Units in CANDU Reactors – In Core Operation
Adjuster rods are essential components for Cobalt-60 production through core irradiation in CANDU reactors. As operational demands increase, especially at elevated power levels and extended durations, a rigorous evaluation of their performance under heightened thermal conditions becomes imperative. This study presents a comprehensive thermal hydraulic analysis focusing on cobalt adjuster assemblies within the reactor core. Leveraging advanced 3D multi-physics computational fluid dynamics (CFD) modeling, adjusters were analyzed under normal operating conditions and during a total loss of moderator inventory (LOMI) accident. The developed model integrates a conjugate heat transfer analysis, facilitating the comprehensive consideration of heat transfer mechanisms within both fluid and solid domains. Furthermore, the CFD modeling employs a multi-dimensional mesh to accommodate the diverse length scales of various components, ranging from the adjusters' bundles to the calandria vessel. The model’s predictions were found to accurately capture the flow characteristics, including the transition between momentum-driven and buoyancy-driven flows for different operational scenarios. Moreover, the temperature distribution in different components was predicted, and the critical hotspots were identified. The outcome of this analysis provides an insight into the flow pattern and heat transfer affecting the adjuster units and can be utilized as an operational, design optimization, or regulatory guideline.
A Computational Investigation on Fuel Dispersal Phenomena during Large Break Loss of Coolant Accident in LWRs
In the event of cladding rupture, which could occur in light water reactor fuel assemblies during a loss-ofcoolant accident (LOCA), fuel particles, along with fission gases, can be expelled into the reactor core from the fractured fuel rod. This expulsion of fragmented fuel particles, referred as fuel dispersal, is the subject of investigation to evaluate the safety implications of increasing fuel burnup in light water reactors, with a specific focus on fuel fragmentation, relocation, and dispersal. Particle trajectories and the resulting mass distribution of the settled particles within the reactor pressure vessel can pose a long-term cooling challenge for the reactor core. The fuel dispersal phenomenon is significantly influenced by the ejection characteristics of the fuel fragments, as well as the size and shape of the cladding rupture and fuel rod depressurization history during LOCA transients. In this study, the transport of fuel particles within a scaled 5 × 5 lattice of a pressurized water reactor rod bundle geometry is modeled through a two-fluid Eulerian framework that treats the gas and solid phases as interpenetrating continua. The required boundary conditions are evaluated from the fuel performance code BISON in a postulated large-break LOCA scenario. The modeling framework considers solid fuel particles as granular matter, interacting with the gaseous dry steam phase and fission gases through the governing interfacial momentum and energy exchange between the gas and solid phases. The simulation results provide the volume fraction of the solids settled on the bottom surface of the fuel bundle, quantifying the deposition within the bundle geometry.
The Influence of Fuel Bundle Misalignment and Pressure Tube Aging on Mass Flow Rate Distribution in CANDU Fuel Channels
This work presents a numerical investigation of a CANDU fuel channel using Computational fluid dynamics (CFD) methodology. The developed CFD model is employed to examine the impact of bundles misalignment and pressure tube aging on mass flow rate distribution in CANDU fuel channels. Practically, in the station, the bundles in the channel are introduced without any specific alignment, resulting in endless potential misalignment configurations. This work focuses on three distinct misalignment configurations, with two of them representing the most challenging scenarios (A & B) in terms of flow obstruction and creating a convoluted twisted flow path, while the third scenario (C) allows for a continuous flow path in-between the fuel rods. All configurations were examined at the designed pressure tube diameter with no aging and compared to the perfectly aligned fuel bundle case. To study the aging effect, the same model was used while incorporating axial variation in pressure tube diameter. The results indicated that fuel bundles misalignment can lead to a reduction in mass flow rate ranging from -6% to -16%, depending on the misalignment configuration. Conversely, the pressure tube aging resulted in a 9% increase in the mass flow rate. This increase was primarily attributed to a nearly doubled clearance between the fuel bundles outer diameter and the inner diameter of the pressure tube. The outcome of this analysis provides an insight into parameters that might influence the flow rate inside fuel channels.
Analysis of IBLOCA Scenarios in ATLAS: Simulation Approaches with the TRACE Code
In 2005, the USNRC proposed the Intermediate Break causing Loss of Coolant Accidents (IBLOCA) as a design basis event for assessing the effectiveness of emergency core cooling systems. Since then, the number of experiments in integral test facilities related to these scenarios, which were very limited up to that date, has increased. Korea Atomic Energy Research Institute (KAERI) has included IBLOCA scenarios in the ATLAS facility experimental programs. Specifically, two tests were performed under the frame of OECD-ATLAS2 project in 2017. Test B3.1 represents a pressurizer surge line IBLOCA, and Test B3.2 is a direct vessel injection (DVI) line IBLOCA. The size of these breaks corresponds to 11.4 % and 8 % of the cold leg flow area, respectively. Both tests were performed to evaluate the effectiveness of available injection systems, including the safety injection pumps (SIP) and the safety injection tanks (SIT). Parallel to experimental studies, simulations can contribute to understanding thermal-hydraulic phenomenology when reproducing experimental and hypothetical scenarios. In this framework, this work aims to analyze both scenarios with the TRACE thermal-hydraulic code. First, the experiments are simulated with TRACE to verify its capability to reproduce the sequence of major events and relevant phenomena. During the transient simulations, thermal-hydraulic parameters such as the system pressures, temperature, collapsed water levels, and natural circulation flows in the system were studied. Subsequently, sensitivity analyses on the break size spectrum of an IBLOCA and its location are conducted to evaluate the effects and identify “cliff-edge effects”.
Testing of Reflood Heat Transfer Under Very-Low Flooding Rate Conditions at the NRC-PSU RBHT Facility
While numerous reflood tests have been conducted in various test facilities in the past, very few experiments have examined very-low inlet flowrates, which are more likely in a Small Modular Reactor (SMR) long term cooling scenario. In this study, various sets of data were collected in the Rod Bundle Heat Transfer (RBHT) test facility located at the Pennsylvania State University focusing on very-low-flooding-rate (0.25<Vin<2.3 cm/sec) test conditions subjected to different system pressures and degrees of inlet subcooling. These benchmark data sets will further expand the ability of thermal-hydraulics codes to evaluate nuclear reactor safety, and additionally confirm the facility is capable of further testing in this very-low reflood rate condition. The very-low-flooding-rate test conditions were carefully chosen to identify the effects of inlet subcooling, system pressure, bundle power, and reflooding rate. The test series included a demonstration of test repeatability with a change in droplet laser camera location to observe the effects of bundle position on droplet entrainment characteristics. Droplet entrainment data will be presented in a future paper. It was observed that the very-low reflood condition extends overall bundle quench and decreases liquid carryover and droplet entrainment substantially when compared to other reflood rate conditions. Decreased inlet subcooling extends the bundle quench time and increases liquid carryover. It is anticipated that the data will expand the capability of thermal-hydraulics codes down at these very-low reflood conditions. Further testing will be possible now that this facility has been benchmarked at these very-low reflood conditions.
Evaluation of Particle Size Distribution Models in STAR-CCM+ and OpenFOAM for Dispersed Two-Phase Flows
The Eulerian-Eulerian framework used in Multiphase Computational Fluid Dynamics (M-CFD) is becoming an increasingly valuable tool for the design and analysis of advanced Light Water Reactors and many other applications. To produce valuable results, the Eulerian framework requires a proper model for the total interfacial area available for mass, momentum, and energy transfer between the two phases. Particle Size Distribution (PSD) models are important to incorporate within this framework to properly characterize the distribution of the dispersed phase in two-phase flows. Within the field of M-CFD, various codes exist such as the commercial software STAR-CCM+ and the open-source software OpenFOAM. Likewise, various PSD models are available within each software thanks to the ability to introduce new models, specifically in open-source codes. The method of moments is a popular PSD modeling approach, with each variation taking different approaches and making different assumptions as to the shape of the distributions, leading to a lack of consensus on their applicability. To address this issue, a set of baseline closures was chosen to evaluate the performance of the various PSD models when applied to the TOPFLOW experimental database. Breakup and coalescence models of different complexities were also tested to assess their applicability and accuracy to different flow regimes. Satisfactory agreement was obtained with certain breakup and coalescence models with the S-Gamma PSD model in STAR-CCM+ while most models within OpenFOAM obtained good agreement when using the LogMoM PSD model.
Fukushima Daiichi Nuclear Power Plant Accident Analysis Considering the Thermal Stratification and Containment Leakage
During the Fukushima Daiichi Nuclear Power Plant accident in 2011, it was reported that several units of the containment vessel had failed, and a large quantity of radionuclides had been released into the environment. However, the detailed accident progression of such a containment failure, which includes core melt, reactor vessel failure and following containment vessel behavior, still has large uncertainties. Especially for Unit 2 and Unit 3, they had succeeded in the initial core cooling, but at last lost their cooling system and fell into the severe accident to release fission products into the environment. Nowadays, much information has been obtained by the internal inspection into the containment of Fukushima Daiichi Nuclear Power Plant. To clarify the uncertainties in the accident scenario, considering the information and several insights already obtained by previous research, the latest accident scenarios in Unit 2 and Unit 3 of the Fukushima Daiichi Nuclear Power Plant accident are suggested and tested by the severe accident analysis code, MAAP* in this study. It is shown that Unit 2 and Unit 3 both accident scenarios would have resulted in the thermal stratification in the suppression pool which encouraged the containment pressure response in the early phase of the accident. In addition, containment vessel leakage would have occurred, affecting the containment depressurization.
Novel Approach for Modeling Effective Delayed Neutron Fraction and Transient Reactivity Feedback Effects for Circulating Fuel Reactors
The point reactor kinetics (PRK) model in GOTHIC∗ has been modified to consider the motion of delayed neutron precursors (DNPs) and Decay Heat Precursors (DHPs) when calculating reactor power. In a solid fuel reactor, delayed neutron precursors (DNPs) born during fission are typically treated as a time-lagged source of neutrons. Meanwhile, in fluid fueled circulating fuel reactors (CFRs), DNPs and DHPs are transported and decay outside the traditional “active core” region. This results in a reduction of the delayed neutron fraction, β, relative to the stationary fuel case. However, the reduction in β does not act equally on all DNP groups. It depends on the half-life relative to the loop transit time for the CFR as this determines where a DNP decays relative to the core. Due to the dependence on loop transit time, the reactivity feedback effect is not constant, but rather a transient phenomenon that must be considered when modeling startup, shutdown and flow anomaly scenarios. GOTHIC includes a detailed tracer transport equation that includes the effects of advection, molecular and turbulent diffusion, and radioactive decay, including both progeny formation and decay heat release within the fluid. This allows for time-dependent concentration for each DNP group to be calculated throughout the system. Integrating the tracer transport and PRK models in GOTHIC provides a novel approach to modeling the transient reactivity feedback effect due to the motion of DNPs in CFRs. This paper summarizes the new capability with validation against the Molten Salt Reactor Experiment (MSRE) data.
Approach for Rod Burst Ratio Evaluation during Loss-Of-Coolant Accident with DRACCAR software
In the frame of the “Reduction of Radiological Consequences of design basis and extension accidents” (R2CA) European Union’s project, the French Institute for Radiation and Nuclear Safety (IRSN) has developed an approach to evaluate more realistically the rod burst ratio (RBR) during loss-of-coolant accidents (LOCA) and the associated radiological releases. The approach is based on the IRSN DRACCAR and ASTEC software. The DRACCAR model consists in a reactor coolant system thermalhydraulic model coupled to a core model using 3D thermal hydraulics and describing each fuel assembly by several 2D (r,z) lumped rods. This model can capture distinctive fuel assembly responses by coupling thermal hydraulics to thermomechanic. Radiological consequences can then be evaluated by a chain between DRACCAR and ASTEC software. Some highlights on DRACCAR core model are provided by discussing results obtained on a PWR intermediate break loss-of-coolant accident scenario. Several burst criteria are tested on this demonstrative case through a sensitivity study. In this work, the influence of burst criterion selection on the rod burst ratio prediction was highlighted and the rod burst prediction varied strongly with the difference between predicted cladding temperatures and burst temperature criterion. Therefore, the methodology should be improved by managing uncertainties quantification associated to LOCA phenomena as in the Best-Estimate Plus Uncertainty core coolability analysis. Finally, some prospects for the DRACCAR software are identified to bring some advances of the presented approach.
Coupled Thermal-Hydraulics and Neutronics Simulation using Sub-Channel CFD and SERPENT
The novel coarse-mesh CFD tool Subchannel-CFD (SubChCFD) is coupled with the Monte Carlo neutron transport code SERPENT using a segregated multi-physics coupling algorithm. The multi-physics model is used to simulate a soluble-boron-free small modular reactor assembly from the design proposed by Alzaben et al. The coupled solution is found to converge within 4 iterations, with successive keff values of within 20 pcm. The final coupled SubChCFD solution shows good agreement with a full-scale CFD model from the CFD software STAR-CCM+, with the exception of a small gap region between two oversized control rod guide tube and burnable poison rods, which induces a periodic flow instability that SubChCFD is unable to model.
Severe Accident Analysis of High Pressure Sequences for APR1400 using CINEMA Computer Code
As an integrated computer code development in Korea, CINEMA (Code for INtegrated severe accidEnt Management Analysis) has been developed for a severe accident analysis from an initiation event to a containment failure. The basic goal of this code development is to design a severe accident analysis code package by exploiting the existing domestic DBA (Design Basis Analysis) code system for the severe accident analysis. The CINEMA computer code is composed of CSPACE (Compass, Safety and Performance Analysis CodE for nuclear power plants), SACAP (Severe Accident Containment Analysis Package), and SIRIUS (SImulation of Radioactive nuclide Interaction Under Severe accident), which are capable of core melt progression with thermal hydraulic analysis of the RCS (Reactor Coolant System), severe accident analysis of the containment, and fission product behavior analysis, respectively. For the purpose of CINEMA verification for a real Korean power plant application, severe accident of high-pressure sequences of TLOFW (Total Loss of Feed Water) and SBO (Station Blackout) has been analyzed in this study. This best estimate analysis has been performed to estimate the efficiency of the CINEMA computer code and the predictive qualities of its models from an initiating event to a reactor vessel failure during severe accidents.
Development of a Source Terms Migration Model for CDA Bubble
The probability of a core disruptive accident (CDA) is considered to be extremely low for sodium-cooled fast reactors (SFRs) because of their high inherent safety design. However, from a defense-in-depth perspective, it is still worthwhile to study CDA sequences for further assurance of safety and reliability of SFRs. During a CDA, a large CDA bubble rapidly expands inside the sodium pool, rises from the core, and cover the gas region, providing a potential migration path for radioactive materials. Source terms released provisionally to the cover gas region on the order of a few hundred milliseconds are called instantaneous source terms. In this study, we propose an instantaneous source term migration model that provides a simplified evaluation of the amount of source terms absorbed by coolant sodium during the ascent of the CDA bubble. In this model, a particle motion model within the CDA bubble based on the basic momentum equation is used to calculate the amount of source terms escaped from the bubble boundary. In addition, a model analogous to the aerosol scavenging by precipitation is used to assess the amount of source terms absorbed by entrained sodium droplets that occur during rapid expansion of the CDA bubble. The model is further validated by a past source term migration experiment in which a high-pressure, large bubble expanded and rose in a sodium pool. Good agreement with the measured retention factor of a source term demonstrates the reliability of the developed model.
Mechanistic Modeling of Departure from Nucleate Boiling in Subcooled Flow Boiling at Near-Critical Pressure Conditions
The Departure from Nucleate Boiling (DNB) leads to a drastic deterioration in heat transfer and thus limits the cooling capability through boiling processes in technical applications. Supercritical power plants, such as SCWRs, inherently operate without a two-phase flow and therefore without the risk of boiling crises during standard operating conditions. Yet, transient scenarios such as startups, shutdowns, or accidents can cause a transition into high subcritical conditions, potentially leading to the occurrence of DNB. Since the immediate and significant increase in wall temperature can lead to destruction of the reactor material, an accurate DNB prediction is highly relevant for a safe design and operation. In the pressure range up to reduced pressures of 0.7, the prediction of boiling crises has been investigated intensively. At higher pressures, however, only few models describe the Critical Heat Flux (CHF) accurately. In this study, a mechanistic model that is based on the modeling of subcooled nucleate boiling is implemented numerically by using established heat partitioning approaches. Further focus is put on the implementation of different sub models and boiling closures (e.g., nucleation site density, bubble departure diameter, etc.). The modifications are investigated and discussed regarding their performance in the near critical pressure region with reduced pressures pr > 0.7. The analysis compares the modeling results with experimental CHF data from literature within a broad parameter and pressure range.
Comparison Analysis on Spreading Distance of Spill Molten Salt with Molten Core Material using Analytical Model
As a part of small reactor developments for the GEN-IV (Generation IV) reactors, a MSR (Molten Salt Reactor) is selected for reactor technology and safety aspects worldwide. In a severe safety of the MSR, the most important thing is a loss of the molten fuel salt with the coolant in the primary system. For this reason, the spreading and cooling process of the spill molten fuel salt is very important for MSR safety. The corium spreading and cooling in PWR (Pressurized Water Reactor) was widely studied for severe accident mitigation. The developed methodology for corium spreading may be used for spreading of the molten salt in MSR. This is focused on preliminary comparison analysis on spreading distance and time of the spill molten salt in MSR with corium in PWR using a simple analytical model with or without crust formation. In general, the spreading of the molten salt is not big difference with the corium spreading without crust formation, because there is no big difference in material property of viscosity. The spreading is restricted by inertia due to a low viscosity and a high released rate of the spill molten salt or the corium, that is, gravity-inertia spreading regime. The spreading distance and time of the molten salt with crust formation in MSR are longer than those of the corium spreading in PWR, because of radiation heat loss to the environment.
Keynote: CEA Fukushima Investigations for Ex-Vessel Debris Retrieval and Decommissioning
Fuel debris retrieval is one of the important challenges in view of Fukushima Daiichi decommissioning. Indeed, hundreds of tons of fuel debris will have to be cut and collected in the 3 units that have been subject to core meltdown in March 2011. A large R&D effort has been supported by the Japanese stakeholders in support of this retrieval. In this context, CEA has used its severe accident and decommissioning research expertise to launch research activities in support of fuel debris retrieval. A first series of works deal with acquiring knowledge on these fuel debris: experiments have been carried out to simulate Molten Core Concrete Interaction or to fabricate fuel debris simulants or depleted uranium-containing prototypes. These samples have been thoroughly analyzed and their properties have been determined. A second line of research deals with fuel debris cutting. CEA has successfully applied and improved its laser cutting technology to fuel debris. Mechanical cutting has also been studied. One of the important safety issues linked to debris cutting is linked to the generated radioactive aerosols and particles. Dedicated research programs have been carried out to characterize these releases and study mitigation techniques.
Advancements in High-Temperature Measurement Techniques to Aid Development and Enhance Monitoring of Advanced Reactors
This manuscript includes research results focused on high-temperature measurements crucial for characterizing the behavior of advanced reactor components, such as heat pipes, as well as providing online monitoring of advanced reactors that operate at temperatures exceeding 800 °C. The study employs Type II Fiber Bragg Grating (FBG) sensors, systematically testing them with various fiber coating and jacket materials. The results obtained from FBG measurements are compared against conventional type K thermocouple measurements. Temperature transients at various rates, as well as cycled measurements were carried out to shed light on the repeatability, accuracy, and potential hysteretic behavior at high temperature conditions found in advanced reactors. This paper underscores the promising outcomes of this research, positioning both FBG and distributed Rayleigh scattering techniques as invaluable tools. These techniques can not only aid in the development of advanced reactor components, such as sodium heat pipes, but also emerge as integral elements in the instrumentation and control strategy for monitoring the operation of advanced reactors. Of particular significance is the application of these techniques to microreactors, which have potential as enabling technologies for remote reactor operation. Enabling remote operation would set microreactors apart from conventional reactor designs, opening new avenues for low-cost and efficient reactor management. The results show that type-II FBGs are a promising technique for temperature measurements up to 800 °C, here tested with two different sensor packaging designs, showing no appreciable spectral distortions. When the temperature was cycled between 700 °C and 800 °C, however, a hysteretic behavior arose.
Plaid-BC Algorithm for Direct Numerical Simulation of Boiling Phenomena
Transition boiling and boiling crisis scenarios are highly important to analyze across many industries including nuclear power, electronics cooling, and chemical processing. However, the scarcity of data describing these processes leads to a poor understanding of the underlying physical mechanisms which trigger them and high variability in system level analyses. Hence there exists a significant need for new studies of these phenomena. To contribute to these efforts, we select the massively parallel multiphase CFD flow solver PHASTA to generate high fidelity simulation data of near critical heat flux scenarios. However, some enhancements must be made to the existing code to properly capture the complex physics of these flows. First among these changes is the capability to locally transport mass across the phasic interface. The current approach computes the average heat flux entering each bubble and uniformly added the generated vapor volume. As the boiling transitions to higher void fraction, the current algorithm does not reflect the local nature of boiling process and would fail to model conditions closer to critical heat flux. The PLAID (Parallel Lattice Algorithm for Interphase Dynamics) approach fulfills this need by overlaying a coarse three-dimensional Cartesian grid over the computational mesh and computing heat flux toward interfaces independently within each large Cartesian cell. The algorithm is presented herein with details on heat flux and mass transfer computations. Some verification cases are presented as well, including the case of a stationary bubble growing in superheated liquid compared to the previous, bubble-centric boiling algorithm.
Comprehensive Verification and Validation of 2D and 3D RANS Turbulence Models in the Mixing Region of Three Parallel Channels Across Diverse Reynolds Numbers
This research assesses the feasibility and effectiveness of utilizing Reynolds-Averaged Navier-Stokes (RANS) models to precisely ascertain the velocity profiles within the mixing region formed by the convergence of three parallel channels. In this study, the geometry of these channels is thoroughly analyzed across three distinct cases of inlet velocities, encompassing laminar, turbulent, and transitional flows, to provide a comprehensive understanding of the dynamics involved. The outcomes of this analysis will also contribute to bolstering the initial design phase for the replacement reactor at the National Institute of Standards and Technology (NIST), specifically the NIST Neutron Source (NNS). OpenFOAM v10 serves as the foundational platform for analyzing three distinct RANS turbulence models. Employing the PIMPLE algorithm, this study integrates and examinses the K-Epsilon Relizable, K-Epsilon, and KOmega SST turbulence models in its comprehensive analysis. Parameters such as velocity, pressure, vorticity, turbulent intensity are analyzed in comparison to experimental data recorded. Further, a Grid Convergence Study is performed to obtain numerical uncertainties for all models. The K-Omega SST model demonstrated superior performance compared to the other three models tested, delivering results the most closely align with the experimental data within the scope of RANS modeling. Nonetheless, significant discrepancies in transient validation highlight the limitations of RANS models in accurately simulating highly turbulent crossflow systems.
Measurements of Boiling Two-Phase Flow Dynamics in a Heated Tube Using Neutron Transmission Imaging
In BWRs, the boiling two-phase flow structure determines the heat transfer characteristics in the core. It is also crucial for evaluating cooling characteristics during severe accidents. Therefore, the boiling two-phase flow has been studied by many researchers for reactor core design and safety assessment. In this study, neutron imaging is applied to visualize the boiling two-phase flow dynamics in a heated tube, and the flow structure is evaluated by the void fraction distributions estimated from the neutron transmission images. The test section is a stainless-steel circular pipe, and it is installed vertically in the neutron beam. A high-speed neutron imaging system with a high-speed camera and an image intensifier is applied to observe fast two-phase flow phenomena. The axial flow development in the heated tube could be visualized, and the spatio-temporal characteristics were evaluated.
Experiments on Passive Heat Removal of Fully and Partially Immersed Containments by Laminar and Turbulent Convection at Very High Rayleigh Numbers
Passive heat removal systems are essential components in modern small nuclear reactor (SMR) designs, aimed at enhancing overall plant safety during incidents. Technical details vary depending on the design and positioning of the containment in relation to the ultimate heat sink. In some configurations, steam condensation occurs at the containment wall, which is fully or partially immersed in water pools during a hypothesized loss of coolant accident. The heat transfer from the outer containment wall to the surrounding water depends on the Rayleigh number, assuming either single-phase laminar or turbulent natural convection. Experimental data for validating numerical models and correlation approaches for free convection are available up to Rayleigh numbers of Ra ~ 1012. However, for vertically aligned cylindrical containments with great lengths, Rayleigh numbers may reach up to Ra ~ 1016. Therefore, the application of established correlations for Ra > 1012 in safety analysis codes requires validation through experimental data. To achieve Rayleigh numbers in the order of Ra = 1015, steam was injected into a stainless-steel containment simulator (9 m high, 0.4 m wide, 0.128 m thick) submerged in water within the PAD vessel of the THAI+ facility. Temperature measurements were taken in both the fluids and structures, and velocity profiles in the water were measured using PIV. The experimental results are presented, providing insight into the performance and behavior of passive heat removal systems in scenarios involving vertically aligned cylindrical containments of significant height.
CFD Simulation of High Temperature Corium Interaction with Coolant in Pre-flooded Cavity
Many efforts have been concentrated on the study of the severe accident to guarantee the integrity of containment function when the reactor vessel is penetrated by high temperature corium. The present study focuses on modeling of ex-vessel fuel-coolant interactions such as jet breakup and debris bed formation by using a commercial CFD code, STAR-CCM+. The corium is assumed to have a very high temperature near 3000 K and injected directly into the pre-flooded cavity. As a preliminary study, the major objectives of the study are to evaluate the capability of the CFD code for the simulation of ex-vessel phenomena in extreme high-temperature and to establish the required CFD models. The heat transfer and phase change of coolant are described by the intrinsic boiling curve. The formation and movement of steam are modeled by adopting the volume-of-fluid (VOF) model and adaptive mesh control. The corium breakup mechanisms are incorporated by using detached eddy simulation (DES) turbulence model. The solidification process is considered by changing the viscosity of the corium with temperature change. The initial conditions and properties of corium are determined from the experimental condition of TROI test, which had been performed by the KAERI. The simulation results show that a large volume of steam is generated along the corium jet, which forms a steam pocket. It is observed that the jet breakup mechanisms are continuously exerted by the highly turbulent motion of the steam. This study is expected to contribute to expanding the insights on ex-vessel phenomena occurring in real plant situation.
Performance Study of High-temperature Latent Heat Storage for Integration with Advanced Nuclear Reactors
Integrating thermal energy storage (TES) with advanced nuclear reactors enhances the flexible use of nuclear energy, facilitating its expansion beyond conventional electricity generation. High-temperature TES, capable of storing heat above 400 ℃, emerges as a vital carbon-free energy solution for decarbonizing industrial sector, especially when combined with low-emission energy sources like advanced nuclear reactors. Idaho National Laboratory (INL) has recently developed a novel design for high-temperature latent heat storage system, called Heat pipe Integrated Thermal Battery (HITB), and efforts are underway to experimentally demonstrate the concept. HITB employs liquid-metal heat pipes to establish thermal linkage between TES and nuclear systems without direct fluid exchange, minimizing the potential risk of the integrated systems such as cross-contamination. HITB is designed to achieve high charging and discharging efficiency as well as high energy storage density by employing metal alloys as Phase Change Material (PCM). Initial proof-of-concept experiments are being undertaken using an aluminum alloy (Al59%-Mg35%-Zn6%). This paper discusses current progress of the HITB project at INL, seeking to develop high-temperature TES for versatile integration with advanced nuclear reactors, and shares insights from the performance evaluations via numerical modeling and analysis.
The Impact of Thermal-Hydraulic Information on Nuclear Fission Product Behavior Simulation Using the SIRIUS Module in the CINEMA Code
The Code for Integrated Severe Accident Evaluation and Management (CINEMA), a comprehensive national code for severe accident analysis, was developed and improved in Korea to simulate the presumed progression of the accident of light water reactor systems. The CINEMA code consists of four key modules to simulate different accident events, including CSPACE (Core meltdown progress simulation coupling with Safety and Performance Analysis CodE) for in-vessel phenomena, SACAP (Severe Accident Containment Analysis Package), SIRIUS (SImulation of Radioactive nuclide interaction under severe accident), and MASTER, which facilitates coordination between individual modules. The study focuses on identifying suitable options for generating thermal-hydraulic data files necessary for future uncertainty analysis of SIRIUS systems using independent calculation. The simulations were conducted for a duration of 72 hours, focusing on a large break loss of coolant accident (LBLOCA) scenario in the Optimized Pressure Reactor 1000 (OPR1000). Analysis was conducted by altering the input conditions of thermal-hydraulic data file, specifically the STEP and TIME conditions, to investigate the influence of the SIRIUS module on nuclear fission product behavior simulation. The results show that an increase in the size of the STEP or TIME option in suspended aerosols resulted in an increase in the maximum amount of suspended aerosol, and the maximum value increased exponentially with increasing option size. Upon analyzing the differences in results stemming from the selection of the STEP and TIME options, it was observed that initially, both options yielded nearly similar and stable computational outcomes.
Comparative Analysis of Particle Retention Models in Filtered Containment Venting Systems
In the European AMHYCO project which aims at an enhanced H2/CO combustion risk management, a high -pressure accident sequence, such as a station blackout with release of high amounts of fission and activation products, is investigated. By comprehensive plant simulations using the German lumped parameter system code AC2 it is shown that large amounts of steam and non-condensable gases can lead to a continuous pressurization of the containment beyond the design pressure in case of an extensive failure of safety sytems. To decrease the containment pressure with simultaneous retention of particles from the atmosphere the gas stream must be filtered before being released to the environment. Therefore, the majority of the European nuclear powerplants are equipped with a filtered containment venting system (FCVS). The most commonly used FCVS combines a venturi scrubber in a liquid pool with a metal fiber filter. For the particle retention in a venturi scrubber inside an FCVS, different numerical models are available. These models can be broadly categorized into empirical models, which are based on experimental data, and mechanistic models, which also rely on theoretical studies. By using selected models to analyze different experiments, it is shown that they cannot reproduce the results sufficiently. In this context, based on the currently available experimental database, a comparative analysis of several models with different experiments is performed. The intention of the work is to evaluate the numerical models with regard to their applicability and quality of results, including selected experiments, and to show the limits imposed by influencing external factors.
Experimental Investigation of Heat Pipes as a Passive Cooling System for a Spent Fuel Pool: The Effects of Air Velocity and Temperature
A passive cooling system, utilizing heat pipes and an air cooling tower, has been designed, which can serve as one of accident mitigation systems in a nuclear power plant, even in scenarios where electricity is unavailable, such as Station Black Out (SBO) accident. Particularly designed for application in a spent fuel storage pool, the system plays a crucial role in maintaining the cooling of spent fuels, effectively managing the continuous release of decay heat. The heat pipe features a unique design with condenser tubes acting as manifolds to increase the effective heat transfer area. To assess the feasibility and thermal performance of the heat pipe under accident conditions, an experimental facility was established. This facility included a water pool with a cartridge heater (heat source), a heat pipe, and an air circulation chamber (heat sink). Pressure transmitter and thermocouples were strategically placed at the evaporator and condenser sections of the heat pipe. The heat transfer performance of the heat pipe was evaluated based on the measurement results, and an analysis of the thermal characteristics inside the heat pipe was conducted as a function of air velocity and air temperature. The findings indicated a proportional relationship between air velocity and heat transfer performance, with an inverse relationship observed between air temperature and the performance. Furthermore, the temperature oscillations in both the inner fluid and on the outer surface were observed especially at the evaporator section, attributed to the differences in thermal-hydraulic characteristics between pool boiling and falling film boiling regions.
Improving Accuracy of a Machine Learning Approach for Forecasting Severe Accidents
As the progression of a severe accident is highly nonlinear and complex, there is a need to develop an accurate and accelerated prediction method that can support the decision-making of the operators. With this background, a machine learning-based severe accident prediction tool was designed. The main objective of this study is to find a time series forecasting strategy that enables an accurate prediction of thermal-hydraulic (TH) variables in a loss-of-component-cooling-water (LOCCW) accident. The datasets used for training the surrogate model were produced by the Modular Accident Analysis Program (MAAP). The surrogate model receives a set of information containing: ten TH variables that can be observed in the main control room, whether the safety components have failed or not, and whether the mitigation actions are implemented or not. Depending on whether the model predicts one TH variable or all ten TH variables, the model can be classified as a Multi-Input-Single-Output (MISO) or Multi-Input-Multi-Output (MIMO) model. Three types of deep neural network architectures have been tested to construct the surrogate model. It was found that the prediction accuracy of the ten TH variables was enhanced in most of the test scenarios by adopting the MISO strategy instead of the MIMO strategy. Not only were the peak values of the time series predicted more accurately, but the divergence issue at the latter part of the scenario was also resolved.
Numerical Modelling of Two-Phase Critical Flow in a Steam Generator Guillotine Tube Rupture Scenario
Critical flow is a phenomenon that occurs when the mass flow from an opening reaches a choked state, and the downstream pressure does not affect the flow rate. This phenomenon is well understood in fluids that do not undergo a phase change during rapid depressurization. It is crucial to understand critical flow for the safe operation of nuclear power plants (NPP) and validated safety analysis modelling. Precise measurement and modelling of two-phase critical flow (TPCF) at the break location are essential for designing reliable emergency core cooling systems (ECCS) and containment safety systems. It is also necessary for advanced reactor designs that rely on passive ECCS. In this work, simulations of TPCF are carried out using the Apros system code and the ANSYS FLUENT CFD code. This paper compares numerical results gathered pre-commission of the CRAFTY facility used for TPCF experiments at Lappeenranta-Lahti University of Technology LUT University. This facility studies critical flow discharges during a primary-to-secondary leak scenario caused by a guillotine break in a steam generator tube. The Apros simulations are based on a 6-equation model. The CFD simulations utilize the Eulerian-Eulerian two-fluid approach. Two different discharge tube inner diameters, D=13 mm and 17 mm, are simulated. Several L/D ratios are studied. The results reveal the critical pressure ratio at which two-phase critical flow conditions occur in tubes with varying L/D ratios. The study offers valuable insights into the impact of L/D ratios on critical mass flow, void fractions, pressures, and flow velocities.
Evaluation of Interface Capturing Schemes of VOF Method through Application to Bubble Flows with Single Orifice
Mechanistic multi-phase fluid computations with the VOF method are becoming possible for the scale of engineering equipment in recent years. However, the performance of the interface capturing schemes (ICSs) on which these computations are founded are generally not well evaluated for the realistic multi-phase flow involved in engineering problems in user’s interest. This study, therefore, aims to evaluate the performance of the ICSs with growing popularity, the THINC and the THINC/WLIC, with the evaluation axes of the volume of bubbles which is of the great importance in the reactor physics and the nuclear thermalhydraulics. A single bubble rising and an experiment of bubble flow with single orifice is analyzed with the above two ICSs, followed by the qualitative and quantitative comparison. While the qualitative bubble flow pattern and the total bubble volume in the results with THINC schemes agree with that of the experiment, the simulation result with the THINC/WLIC scheme shows a significant numerical dissipation of the bubble volumes, an unphysically tiny bubbles and an unintended continuous gas injection. The cause of these numerical issues is identified as the slight numerical diffusion of VOF and its subsequent dispersion by the shear stress, which has not been clearly recognized by the preceding VOF advection tests.
Integral Effect Test on Intermediate Break Loss-of-coolant Accident with Coupling RCS and Containment in ATLAS-CUBE Facility
During an accident such as a loss-of-coolant accident (LOCA) or a steam line break (SLB), the released coolant from a reactor coolant system (RCS) increases a pressure and temperature (P/T) inside a containment, which is a crucial parameter to threaten an integrity of the containment wall and the equipment qualification (EQ). The ATLAS-CUBE integral effect test facility has been constructed to experimentally investigate the thermal hydraulic interaction between the RCS and the containment. This study focused on the C1.2 test in the OECD/NEA ATLAS phase 3 international joint project simulating an intermediate-break loss-of-coolant accident (IBLOCA) accident using the ATLAS-CUBE facility. The test result showed that the RCS was effectively cooled down during the transient by an activation of the safety injection system, which supplied the safety injection water to contribute the recovery of the coolant inventory in the reactor core. The mass and energy (M/E) supply from the RCS increased the pressure and the steam-gas mixture temperature of the containment, while the condensation of the steam on the passive heat sink wall contributed to cooling down the steam-gas mixture in the containment. The increasing rate of the containment pressure became smaller as the break flow changed to the two-phase flow with the subcooled liquid flow. The present integral effect test data can be used to evaluate the capability of thermal hydraulic safety analysis codes for the RCS and the containment, with considering the multi-dimensional or the multi-compartment models.
Integrated Simulation of Rod Ejection Accidents in Boron-Free Pressurized Water Reactors Using S5K Neutronics and VIPRE-01 Sub-channel Solvers
Transient neutronics and core-thermal subchannel simulations of rod ejection accidents were performed for the Rolls-Royce Small Modular Reactor equilibrium core design. Transient neutronics simulations were performed using SIMULATE5-K (S5K). Pin-power reconstruction capabilities in S5K allowed for predictions of the prompt Radial Averaged Peak Fuel Enthalpy (RAPFE) rise and total RAPFE encountered in rod ejection accidents. Rod ejection events were analysed across various core initial power levels, cycle statepoints, and ejected rod worths. This was done to determine relationships between fuel failure criteria and ejected rod worth, to aid future design choices about appropriate control rod worth limits. Power shapes from the S5K simulations were then utilized in VIPRE-01 (subchannel analysis code) to obtain predictions of the Minimum Departure from Nucleate Boiling Ratio (MDNBR) in a wide range of postulated rod ejection accidents. A S5K-VIPRE interface tool was used to generate extremely detailed VIPRE-01 decks, which directly incorporated S5K predicted pin-power shapes into the models. This interface tool was used to investigate the impact of VIPRE model fidelity on the MDNBR predictions. Two VIPRE model geometries were investigated: one where the assembly containing the ejected rod was modelled in full subchannel fidelity and another where the eight assemblies surrounding the ejected rod were also modelled in full subchannel fidelity (including full S5K supplied pin-wise power profiles). It was found that the most limiting fuel failure criterion was the MDNBR and that there were significant differences between the predicted MDNBRs in the two VIPRE model geometries. Particularly, it was found to be likely that the most limiting assembly in terms of MDNBR performance was likely to not be the assembly from which the control rod was ejected from.
MIDI, a Facility Dedicated to Study the Thermal Hydraulic Characteristics of Spent Fuel Pools
In the framework of the post-Fukushima actions, a research program called DENOPI has been conducted by the French Institute for Radiation Protection and Nuclear Safety (IRSN) to study the behavior of spent fuel pools during loss-of-cooling accidents. This project, supported by the French Government and carried out in collaboration with partners, is composed of experiments and modeling works. As part of this project, the test facility named MIDI was implemented in the IRSN laboratories at Cadarache (France). MIDI facility is an experimental tool dealing with the thermal hydraulic phenomena that may occur at an integral scale in a spent fuel pool undergoing a loss-of-cooling accident before the stored fuels get uncovered. This paper first describes MIDI, a small-scale mock-up of a spent fuel pool, focusing on the design strategy and a presentation of its instrumentation. Then, it presents an overview of the main results. Within the DENOPI project, observations and obtained data highlight the competition during the accidental transient of three vaporization modes: the evaporation at free surface, the nucleate boiling onto the fuel assemblies and the gravity-driven flashing of superheated water above the heating zone. The operationality of the MIDI facility has been demonstrated. As a first step, additional works are planned as part of the new OECD POLCA (Pool during Loss-of-Cooling Accident) project due to start in 2024. In a general way, MIDI experimental results constitute an useful database for computational validation and for supporting the safety demonstration of large pools in natural convection configurations.
CSNI Code Validation Matrix (CCVM) Current Status and Development Needs
For Anticipated Operational Occurrences (AOOs), Design Basis Accident (DBA), Beyond Design Basis Accident (BDBA) and Severe Accidents (SAs) as postulated accidents, computer code predictions are relied on nuclear reactor safety analysis. The computer codes are required to go through a rigorous Verification and Validation (V&V) and Uncertainty Quantification (UQ) process. The codes must be validated against data from scaled-down experimental facilities, both Separate-Effect Test Facilities (SETFs) and Integral-Effect Test Facilities (IETFs), to confirm their predictive capability with qualitative and quantitative accuracy. The experiments therefore play a key role to provide the needed knowledge and data. For decades, with internationally recognized experts, OECD/NEA/CSNI/PWG-2 and Working Group on Analysis and Management of Accidents (WGAMA) have taken a key role to identify: 1) Relevant phenomena important for reactor safety analysis; 2) Phenomena occurrence versus experimental facilities in NEA Member Countries; and 3) suitability of the experimental facilities, as they pertain to the phenomena of interest, for code assessment. The information has been summarized in several Committee for the Safety of Nuclear Installations (CSNI) Code Validation Matrices (CCVMs) with Phenomena Identification Ranking Tables (PIRTs) that list SETFs, and IETFs relevant to each of different reactor types. The CCVM aims to gather the best set of openly available data for code validation possibly with uncertainty assessment of the code constitutive models and calculated results. The accessibility to the CCVM suggested data is essential for the development of reliable safety assessed codes, to support knowledge management and transfer to young experts and newcomers. This paper summarizes the CSNI’s CCVM development activities related to Thermal-Hydraulic (TH) and SA reactor safety analysis codes. The paper underlines the current needs for various new designs, including Small Modular Reactor (SMRs), through the PIRT process and highlights experimental data needs for the key accident phenomena.
Thermal-Hydraulic Study of the Converter Facility in the FRM II Research Reactor using Serpent 2 and Ansys CFX within the Framework of the LEU Conversion Program
The Forschungs-Neutronenquelle Heinz Meier-Leibnitz (FRM II) actively engages in international efforts toward the decrease of Highly Enriched Uranium (HEU) usage in civil cycles. This goal can be achieved by changing the current fuel system to one with high-density, low-enrichment uranium (LEU) fuel. This work is part of the reassessment of the secondary sources located in the moderator tank and are, therefore, likely to be affected by a change in the fuel material. Therefore, investigating the current secondary sources in terms of neutronics and thermal-hydraulics is of utmost importance as it serves as a validation basis for a future assessment. The converter facility is one of the secondary sources located in the moderator tank of the FRM II reactor. Its purpose is the production of fission neutrons from the thermal neutrons that migrate from the reactor core toward the converter plates. These fission neutrons are used for different applications, such as tumor irradiation for palliative patients, radiation hardness testing for electronics, etc. The converter plates exhibit an asymmetric power density distribution due to their tilted geometry and proximity to the fuel element. In this work, the neutronics and thermal-hydraulic aspects of the converter facility and the influence of the asymmetric heating profile on the cooling of the plates are evaluated using Serpent 2 and Ansys CFX. The thermal-hydraulic assessment is the first realistic calculation that uses the asymmetric heating profile. The coolant temperature in the middle of the cooling channel between the two plates reaches about 50°C. The power density asymmetry causes one side of the facility to be warmer than the other, with a difference of approximately 10°C in maximum temperatures.
Development of a Condensation Model for Realistic Analysis of Heat Removal Behavior by Passive Heat Sinks within a Containment
The Colburn-Hougen condensation model, proposed in 1934 based on gas kinetics, assumes that the gas mixture around the condensation surface is in a static state. Therefore, it does not reflect the turbulent effects that occur when a strong flow field is formed around the condensation surface. In this study, we developed a new condensation model by introducing the turbulence diffusion coefficient to account for turbulent effects by convective flow of steam mixed with a non-condensable gas. In the original Colburn-Hougen model, turbulent effects were merely considered in the Sherwood number correlation used for calculating the mass transfer coefficient. The developed condensation model adopted the Cebeci-Smith model to calculate a turbulent diffusion coefficient, which is added to the molecular diffusivity. For application to the basic form of the Colburn-Hougen model, we derived a modified Sherwood number correlation by calculating the mass transfer coefficient from the condensation rate measured in the flat plate condensation experiments. The developed model has been validated against the COPAIN, CONAN, and SETCOM experiments, in which local heat fluxes were measured under forced convection conditions. In most cases, the results of the developed model closely predicted the experimental heat fluxes, and the average standard deviation of the relative error was 24.1%. Furthermore, the developed model better predicted the local heat flux along the condensation surface compared to the original model.
Comparative Study of Cold Ramp Tests Simulations with the Thermo-Mechanical Code TESPA-ROD and the Subchannel Code CTF
The OECD/NEA published the Multi-physics Pellet Cladding Mechanical Interaction Validation (MPCMIV) Benchmark (here Rev. 3 of the specifications). This benchmark is based on experimental data from the R2 research reactor in which nearly 150 ramp tests were specifically designed for PCI/PCMI problems. While the benchmark is aimed mainly at the validation of multi-physics phenomena, the so-called “Tier 4” exercises focus on the validation of traditional fuel performance codes, using boundary conditions from measurements, namely the coolant pressure, mass-flow and inlet temperature, and the axial linear heat rate. The CTF code is an advanced version of the well-known subchannel code COBRA-TF, co-developed at the North Carolina State University (NCSU) and Oak Ridge National Laboratory (ORNL). In the last decade the fuel model in the thermal-hydraulics subchannel code COBRA-TF has been further developed and improved. However, it remains a simplified model compared the ones found in dedicated codes. The TESPA-ROD code analyses the thermo-mechanical behavior of the fuel rod for RIA-transients, LOCA-transients, PCMI-transients, and interim storage. The fuel rod is represented in a 1½ dimensional spatial resolution. The first and second cold ramp tests of the Benchmark are simulated with both codes and their results compared. The parameters of interest are the fuel rodlet gap width, the centerline temperature, and the radial temperature distribution. The comparison allows to identify differences in the models and correlation used in the rod models of CTF and TESPA-ROD. This work is the first step toward a tri-code multi-physics coupling between neutronics, thermal-hydraulics and thermo-mechanics. The next step will be the offline coupling between CTF and TESPA-ROD.
Numerical Investigation on Gas and Powder Flow Characteristics of Pebble Bed of Fusion Blanket
The breeding blanket, which is core component for fusion device, plays an important role in tritium breeding and energy conversion. Ceramic particles are usually used as tritium breeders (e.g., Li4SiO4, Li2TiO3 particle) and neutron multipliers (e.g., Be/Be12 particle) inside the blanket, and pebble bed structure is utilized to hold ceramic particles. The helium pressure drop inside the ceramic pebble bed is closely related to the design of the tritium purge structure of the blanket, the flow characteristics of helium and powder directly affect the tritium breeding and extraction efficiency. Hence, there is a great need to promote research in the flow behavior of purge gas and dynamic behavior of the crushed particles. Based on the two-way coupled Computational Fluid Dynamics and Discrete Element Method, the velocity distribution, pressure distribution and inlet and outlet pressure drop of helium in pebble bed structure under different inlet flow rates were obtained. The velocity and quantity distribution of lithium powder at different moments and inlet flow rates were numerically investigated. The results indicated that helium flow velocity distribution along the z-axis fluctuates in the opposite direction to the axial voidage distribution, the helium dimensionless flow velocity keeps nearly the same in the whole pebble bed. The magnitude and distribution trend of the dimensionless helium velocity is independent of the magnitude of v0. The average absolute velocity of powder increased with the increase of the inlet gas flow rate, while the average dimensionless velocity decreased with the increase of the helium inlet flow rate. The results in the paper can provide technical recommendations for fusion ceramic blanket design.
Computational Fluid Dynamics Investigation of the Impact of 6% Crept Pressure Tubes on Flow Behavior, Fuel Temperatures, and Pressure Tube Wall Temperatures of a Single CANDU 37M Fuel Bundle
A Computational Fluid Dynamics (CFD) simulation with heat transfer was carried out in this work on a single CANDU 37M fuel bundle sitting in a pressure tube that has experienced 6% diametral creep. The investigation focused on the predicted coolant flow behaviour, fuel temperature, and pressure tube wall temperature to explore the effects of the crept pressure tube on these parameters. The findings reveal that around 29% of the coolant flow bypasses the bundle within the 6% crept pressure tube, resulting in a temperature difference of roughly 20 K between the fuel element positioned at the bottom compared to those at the top. Moreover, the temperature profile of the crept pressure tube wall was observed to be uneven, with the bottom exhibiting approximately 30 K higher temperature than the upper crept pressure tube wall.
Potential Applications and Recommendations of Dynamic Event Tree Analysis using DICE(Dynamic Integrated Consequence Evaluation)
Over the past six years, Kyung Hee University has completed the development of DICE (Dynamic Integrated Consequence Evaluation), a dynamic event sequence analysis tool. For the analysis of design basis accidents, the tool incorporates the MARS-KS, a regulatory verification code in Korea, while for severe accident analysis, it integrates the MELCOR from the U.S. NRC as its physical module. The tool comprises a diagnosis module for determining control tasks, a reliability module for assessing their functionalities (normal/abnormal), and a scheduler has been configured to exchange inputs and outputs between these modules at a specified interval. Each module is structured as dynamic link libraries to ensure flexible extension with the scheduler. By following to the specified data structure protocol defined by DICE, the integration of external modules becomes relatively straightforward. This paper explores leveraging these structural features to propose potential applications for risk assessment, where both DSA (Deterministic Safety Analysis) and PSA (Probabilistic Safety Assessment) are conducted within a unified computational process. The paper suggests enabling approaches for addressing challenges that are difficult under conventional PSA, such as complex multi-unit assessment, cyber security, autonomous operation development, and so on. Based on these benefits, authors discuss considerations for standardizing the dynamic event sequence analysis.
Study on Eutectic Melting Behavior of Control Rod Materials in Severe Accidents of Sodium-Cooled Fast Reactors (3) Material Analysis of Boron Carbide Immersed in Molten Stainless-Steel
Interest in eutectic reactions between boron carbide, used as control rod material and stainless steel used such as cladding tubes, is growing to improve the accuracy of severe accident analysis codes. Immersion experiments of boron carbide pellets into molten stainless steel were performed between 1,773 and 1,973 K. The eutectic melting behavior of the pellet was investigated by observing the cross-section of the pellet using an optical microscope, a scanning-type electron microscope. The elemental distribution and crystal structure of the reaction products were analyzed using an energy-dispersive X-ray spectrometer, X-ray diffraction, and electron–beam diffraction by a transmission electron microscope to clarify the eutectic reaction behavior. Based on the thickness reduction of the pellet cross-section, the reaction rate constants between boron carbide and stainless steel were evaluated under various contact temperatures and times to develop an evaluation formula of the reaction rate constants, which would be incorporated into severe accident analysis codes.
Source Term Analysis in the Fuel Degradation Phase of PHEBUS FPT-1 Experiment using MELCOR 2.2
A severe accident in a nuclear power plant is accompanied by complex phenomena within the containment building due to the release of various fission products and combustible gas. An accurate prediction of the source term is crucial for preventing the release of radioactive materials outside the containment building and for implementing appropriate mitigation measures under severe accident conditions. In this study, we analyzed the parameters related to the core degradation and the release fractions of fission products by benchmarking the PHEBUS FPT-1 experiment using the MELCOR 2.2 code. The simulation results were compared with experimental data of the temperatures of fuel and cladding, hydrogen production rate, and the release fractions of fission products to the containment vessel for source term assessment. The fission products discussed in this study include Xe, Te, Cs, and I classes. The predicted release fractions of xenon, tellurium, and cesium were similar to those measured in the experiments, while the release fraction of iodine was underestimated by about 20%. This indicates that further efforts in modeling with the MELCOR code are required to reproduce iodine behavior observed in the FPT-1 test. Uncertainty and sensitivity analyses are also deemed necessary for future work.
Preparing the MACCS Code for Advanced Reactor Applications
The MACCS code was created by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission and has been used for emergency planning, level 3 probabilistic risk assessments, consequence analyses and other scientific and regulatory research for over half a century. Specializing in modeling the transport of nuclear material into the environment, MACCS accounts for atmospheric transport and dispersion, wet and dry deposition, probabilistic treatment of meteorology, exposure pathways, varying protective actions for the emergency, intermediate and long-term phases, dosimetry, health effects (including but not limited to population dose, acute radiation injury and increased cancer risk), and economic impacts. Routine updates and recent enhancements to the MACCS code, such as the inclusion of a higher fidelity atmospheric transport and dispersion model, the addition of a new economic impact model, and the application of nearfield modeling, have continuously increased the codes capabilities in consequence analysis. Additionally, investigations of MACCS capabilities for advanced reactor applications have shown that MACCS can provide realistic and informative risk assessments for the new generation of reactor designs. Even so, areas of improvement as well as gaps have been identified that if resolved can increase the usefulness of MACCS in any application regarding a release of nuclear material into the environment.
Towards CFD Modelling of the Lower Plenum Mixing Test of the OSU High Temperature Test Facility
Prismatic High Temperature Gas-cooled Reactors (HTGRs) are a concept approaching deployment, often as Small Modular Reactors (SMRs). Deploying these reactors requires modeling and simulation tools that have been validated for these systems but most System Thermal Hydraulic (STH) modeling and simulation tools were originally developed and validated for Light Water Reactors (LWRs). To provide a set of validation and verification problems, an OECD/NEA benchmark is underway that is based on the High Temperature Test Facility (HTTF) at Oregon State University (OSU). This is a 1:4 scaled integral effects experiment on the General Atomics’ (GA) MHTGR design. The OECD/NEA benchmark consists of three separate problems to be analyzed. One of these is the Lower Plenum (LP) mixing exercise. This problem can be tackled in two cases: a code-to-code comparison study with fixed boundary conditions and a code-to-experiment comparison study with best estimate boundary conditions. In the framework of the European collaborative project GEMINI4.0, NRG is currently participating in the code-to-code study where the boundary conditions try to mimic the full power conditions of the experiment. The commercial code STAR-CCM+ is employed for the calculations. Preliminary tests showed that the problem cannot be solved with a steady state RANS solver. This will be briefly demonstrated in the paper. After that, the focus of the current work is with Unsteady-RANS (URANS). The benchmark requires participants who run transient simulations to show grid independence as well as temporal convergence. The present article aims to showcase results for the coarse mesh with various time step sizes to show this temporal convergence.
Coupled CFD-STH Modelling of the ATHENA Lead-Cooled Pool Type Facility
The Advanced Thermal-Hydraulic Experiment for Nuclear Applications (ATHENA) is a multipurpose pool-type facility located in Romania, designed for testing large to full-scale components of the Advanced Lead Fast Reactor European Demonstrator (ALFRED). This facility is representative in height of the ALFRED reactor and incorporates a barrel structure to drive the lead flow path within the pool. The internal structure's purpose is to prevent thermal stratification within the main vessel, which is a safety-relevant component in the reactor. Computational tools are required to analyse the thermal-hydraulics of pool facilities and reactors. Computational Fluid Dynamics (CFD) codes have the capabilities to simulate local phenomena (e. g., thermal stratification, fluid mixing and local distributions) by directly solving the three-dimensional Navier-Stokes equations. On the other hand, System Thermal Hydraulic (STH) codes solve a set of zero- and one-dimensional equations, even though at a considerably lower computational cost. Hence, a multiscale model of ATHENA has been developed in order to improve the capabilities to simulate the phenomena occurring within the facility at the different spatial scales, exploiting the strengths of both code families. In this paper, an in-house developed coupled tool that combines ANSYS CFX and RELAP5 Mod 3.3 has been applied to analyse the behaviour during a transient event involving the trip of the Main Pump with the consequent reduction of lead flow. In the work, the first step toward the identification of the strengths and weaknesses of the multiscale approach is made, and possible further model developments to validate the tool for safety related transients are proposed.
Experimental Investigation on Helium Drag Effect in Terms of Natural Circulation Enhancement for Passive Molten Salt Fast Reactor (PMFR)
A helium bubbling system employed in molten salt reactors for the elimination of insoluble fission products improves a natural circulation performance of working fluid owing to an intensified drag force of helium bubbles. In this regard, to evaluate the extent of natural circulation enhancement, adiabatic two-phase natural circulation experiments were performed. In this experiment, the variation of major thermal-hydraulic parameters such as a velocity of the working fluid and volume fractions of gas-phases was observed according to the modification of sensitivity variables including the width of riser channel, the type and quantity of gas, and the viscosity of working fluid. A complicated bubbles’ behavior was captured through a high-speed visualization within the wide range of gas injection rate from 1 to 15 lpm. Based on the experimental findings, the width of the riser channel showed a major effect on the natural circulation performance. Moreover, helium resulted in a superior enhancement in the circulation compared to air because of a larger density difference with the working fluid. As the viscosity of the working fluid increases, fluid velocity decreased non-linearly. Experimental outcomes showed that the helium bubbling system can elevate the natural circulation performance of working fluid. This study will contribute to investigating the helium bubbling effect with respect to the circulation performance of working fluid in a passive molten salt fast reactor (PMFR).
Influence of ATF Coatings on the Critical Heat Flux in a Vertical Rod Bundle during Flow Boiling of Water at Low Pressure
This paper issued the experimental testing of ATF cladding materials as rod bundles in loss of coolant accidents of pressurized water reactors from the onset to the departure of nucleate boiling under low pressure conditions. A CrN-coated rod bundle is compared with an untreated Zircaloy-4 rod bundle as a promising candidate from ATF material research in terms of corrosion/ hydrogen generation rates. To clarify the influence of ATF-grade coatings, the heat flux required to reach the boiling crisis is used as a benchmark. With halved and doubled reference conditions (20 K subcooled mass flux of 200 kgs-1m-2 at 1400 mbar), the influence of the CrN coating on the steady state temperature and maximum heat transfer is shown in consecutive heating steps until the boiling crisis is reached. The experimental setup consists of a directly electrically heated and magnetically compensated 5-rod bundle within a vertical 43.7 x 43.7 mm flow channel with controlled boundary conditions and a temperature excursion-triggered rapid shutdown of the transformer, which enables multiple non-destructive CHF experiments with identical cladding tubes. The outer and inner diameter of each rod is 9.5 and 8.36 mm with a layer of 8 μm CrN and a heated length of 300 mm. The sight glasses in the flow channel with high-speed cameras and transformer-synchronized ring buffer enable the observation of the heated surface at the time of the CHF. The investigated material influence of heated surfaces can be used for the further development of mechanistic CHF prediction and CFD models.
Current Status of Cr-coated Claddings in Severe Accident Codes and Evaluation within the Frame of the QUENCH-ATF-1 Benchmark of OECD Project QUENCH-ATF
Zirconium-based claddings have been widely used for decades due to their good mechanical properties, corrosion resistance, and low neutron absorption under operational conditions. However, at elevated temperatures, these claddings exhibit negative characteristics, such as exothermic oxidation reactions in steam or air environments. Research and development of conventional cladding and furthermore Accident Tolerant Fuel materials has been motivated by the need to establish an additional safety element in nuclear power plants. These new materials should exhibit significantly slower oxidation kinetics compared to the Zr-based alloys. Among others, the coating of claddings with is being intensively investigated and tested for increased accident tolerance. Its oxidation rate is several orders of magnitude smaller than that of typical zirconium-based cladding, assuming the coating is in the protective state. Afterwards, the kinetics approaches Zr-alloy behavior. In addition to experiments, code modelling capabilities for DBA and DEC analyses must be improved for the analysis of systems that include such materials. Suitable oxidation models must be implemented to predict the ATF behavior under DBA and DEC conditions. Additionally, different material properties like density and thermal conductivity or heat capacity have to be considered. The ability to model Cr-coating varies among severe accident codes, including AC²/ATHLET-CD, ASTEC, MAAP, SAMPSON, and MELCOR. This will be described in detail in the paper. To validate the performance of integral codes that model Cr-coated cladding behavior, the benchmark on the experiment QUENCH-ATF-1 within OECD/NEA QUENCH-ATF project is considered. The boundary condition of the sequence is an extended DBA scenario, with a PCT close to the eutectic temperature of the Cr/Zr binary system. In the paper the results concerning thermal evolution and hydrogen generation will be discussed and the model basis for Cr-coated cladding will be evaluated.
Numerical Investigations of Ball Valve Performance in Variable Valve Diameters, Openings and Flow Regimes
A United States Nuclear Regulatory Commission’s (USNRC) study found that 65% of Nuclear Power Plant (NPP) system failures are due to human errors [1]. The Strategic Management Analysis Requirements and Technology (SMART) valve initiative by the Department of Energy (DOE) is a part of the Light Water Reactor Sustainability (LWRS) program to improve the energy economics of the existing light water reactors (LWR) [2]. SMART valves can improve plant safety by reducing human intervention, balancing thermal dispatch, and reducing operation costs. An NPP has complex hydraulic networks containing various valves. Regarded as the backbone of an NPP, hydraulic networks are responsible for core-cooling, heat transfer, thermal extraction, and dispatch while maintaining their safety. This study presents three different-sized full-port ball valves’ performance in both water and steam flow conditions. Using Star-CCM+, different flow regimes at various valve positions are analyzed, and performance parameters are compared to strategize the automated SMART valve network control. It is observed from the performance parameters that the ball valves are suitable to control the flow from 45° to 90° with loss coefficient values below 100 and follow a distinct pattern across all the sizes. The rate of change for flow coefficients per degree gives us an overview of the precision of control needed for flow regulation, predominantly in wider openings, where a slight actuation of the valve causes significant changes in flow parameters.
Intelligent Control System Design for Implementation of Control Strategies for Supercritical CO2 Brayton Cycle-Cooled Reactor System
The Supercritical CO2 (S-CO2) Brayton cycle-cooled nuclear reactor system offers advantages such as high efficiency and a compact configuration, both essential to enhance the competitiveness and sustainability of nuclear energy. Robust control strategies are required to maintain its performance and stability during operational and accidental conditions. For the implementation of advanced control methods, an integrated simulation platform SCTRAN/CO2-SIMULINK is adopted. Thermal hydraulics and turbomachinery calculations are carried out using system code SCTRAN/CO2, while control actions are taken using SIMULINK and data between them is exchanged in run-time. Backpropagation and least-squares regression method is used for Adaptive Neuro-Fuzzy Inference System (ANFIS) to tune the parameters of FIS controllers. These controllers are used to control the coupled dynamics of the turbine shaft speed and compressor mass flow rate of the cycle. Furthermore, the compressor inlet temperature controller is designed to keep the system in a supercritical state, and the turbine inlet temperature controller is developed to regulate the reactivity in the reactor core to keep the turbine inlet temperature constant. The use of an intelligent control system with the integrated simulation platform has demonstrated promising results for load following operation of the cycle.
Optical Fiber Temperature Sensor Performance Test in a Metallic Capillary Tube at High Temperatures
Accurately measuring temperature at elevated temperature ranges is crucial for many industrial purposes. With the deployment of several advanced nuclear reactors on the horizon, the ability to instrument in these harsher environments at high temperatures and get accurate readings is needed. One potential solution is optical fiber temperature sensors. These sensors are capable of measuring temperatures with great spatial and temporal resolution up to around 600 °C. For use in instrumenting next generation reactors the performance up to 1000 °C is of interest. This work presents tests on fiber optic temperature sensors operating on the Rayleigh backscatter principle with cross validation from thermocouples. The results of each measurement technique were compared at high temperature to determine the accuracy of each instrument and to determine if optical fiber temperature sensors can be used beyond their typical operating range, targeting advanced reactor conditions.
COAL Reflooding experiments during a Loss Of Coolant Accident : effect of the water flow rate and the residual power with ballooned rods
During a loss of coolant accident (LOCA) in a pressurized water reactor, the drying of the fuel assemblies leads to an increase in the fuel temperature and a deformation of the fuel rod cladding. The COAL experiments focused on the coolability issue of a partially deformed fuel assembly during water injection with the safety systems using a 7x7 bundle of electrically heated rods. The relocation of the fragmented fuel in the balloons is taken into account by a local increase in power by a factor of 1.5, and the effect of the flow area restriction is provided with various flow blockage {intact geometry up to moderate and long ballooning (100 and 300 mm) with different blockage ratios (80 and 90%)}. These experiments, in the frame of the PERFROI project, were launched by the “Institut de Radioprotection et de Sureté Nucléaire” (IRSN) with the support of the “Agence Nationale pour la Recherche” (ANR), Electricité de France (EDF) and the US-Nuclear Regulatory Commission. This paper presents the thermal hydraulics parameters and the main results of some experiments carried out in a facility of the STERN Laboratories. We studied the effect of the inlet water flow rate which is the consequence of the amount of water entering the reactor core after the break of the primary circuit and the effect of the residual power as a function of the moment of availability of the safety pumps after the reactor scram. These results are used to improve and validate the heat exchange models of thermal hydraulics codes dealing with the complex reflooding processes in such a configuration.
Computational Fluid Dynamics Analysis of Phase One of the OECD‐NEA / NRC Benchmark on Liquid Metal Fast Reactor Thermal Hydraulics
Computational Fluid Dynamics (CFD) analyses of a 61-pin Liquid Metal Fast Reactor (LMFR) fuel bundle are performed in Fluent and compared to the experimental data available in the Organization for Economic Cooperation (OECD) Nuclear Energy Agency (NEA) / Nuclear Regulatory Commision (NRC) Benchmark on LMFR Thermal-Hydrualics. Porous media CFD models are evaluated and compared to benchmark and subchannel code pressure drop results. The results of one bundle length (476.0 mm) are presented for all modeling strategies. The uncertainties due to mesh sizing, turbulence model, and porous media parameters are estimated. Porous media parameters were calculated based on the experimental results, leading to good agreement with CFD results. This study shows that porous media CFD modeling provides good results and would benefit from further development.
Establishing a LOCA Fuel Safety Evaluation Methodology for Assessment of Advanced Technology Fuel Performance Using FRAPTRAN-TE-1.5
To aid its Member States in the development and performance assessment of Accident Tolerant and Advanced Technology Fuels (ATFs), the International Atomic Energy Agency (IAEA) initiated a Coordinated Research Project (CRP) on Testing and Simulation for ATFs (ATF-TS) in 2021. Tractebel has participated in this IAEA CRP, with the goal of developing a best estimate plus uncertainty (BEPU) hot rod LOCA fuel safety evaluation methodology for assessing the performance of ATF with Cr-coated cladding under Loss-of-Coolant Accident (LOCA) conditions. This methodology was based on the hot rod thermal hydraulic boundary conditions computed by the system thermal hydraulic code TRACE for a large-break LOCA of a typical Pressurized Water Reactor (PWR), and a modified fuel rod code FRAPTRAN-TE-1.5. The original FRAPTRAN-1.5 code was improved by incorporating a high-temperature creep model and a burst model, crucial for simulating Cr-coated cladding ballooning and burst during LOCA. The updated FRAPTRAN-TE-1.5 code was utilized to simulate selected LOCA burst tests with Cr-coated Optimized ZIRLO cladding, conducted within the ATF-TS CRP, to calibrate the creep model parameters and quantify the associated uncertainties. The updated FRAPTRAN-TE-1.5 code was subsequently used to predict the expected behaviors of the integral single rod LOCA tests performed at the Halden reactor (IFA-650.9 and IFA-650.10) with uncoated and Cr-coated cladding, for the purpose of validating the models. Lastly, the updated FRAPTRAN-TE-1.5 code was employed to develop a BEPU hot rod FSEM based on the hot rod thermal hydraulic boundary conditions for the reference fuel rod and to predict the expected behavior of the ATF with Cr-coated cladding for the PWR large-break LOCA scenario, in comparison with the behavior of the reference fuel rod with un-coated cladding. This paper presents the developed hot rod LOCA fuel safety evaluation methodology and its application to assess the performance of the ATF with micro-cell fuel pellet and Cr-coated cladding.
Computational Fluid Dynamics Analysis of Phase One of the OECD‐NEA Benchmark on CANDU Thermal-Hydraulics
High-resolution Computational Fluid Dynamics (CFD) analyses of a 28-element CANadian Deuterium-Uranium (CANDU) fuel bundle were performed in OpenFOAM and compared to the experimental data available in the Organization for Economic Cooperation (OECD) Nuclear Energy Agency (NEA) Benchmark on CANDU Thermal-Hydraulics. Porous media models of the full experimental facility (including 12 fuel bundles) are investigated and compared to benchmark pressure drop results. The uncertainties due to mesh sizing and turbulence models are investigated alongside the porous media approximation. This study is expected to show that more development is needed if porous media modeling is to accurately flow behavior in CANDU fuel bundles.
Coupled Computational Fluid Dynamics (CFD) and Finite Element (FE) Thermal Stress Analysis of Thermal Fatigue Phenomena within a T-junction
The assessment of thermal fatigue phenomena in T-junctions, within the primary and secondary thermal hydraulic circuits of nuclear power plants (NPPs), remains a significant computational challenge. To accurately model the complex physics of fluid thermal mixing, conjugate heat transfer (CHT) and thermal fatigue, it is necessary to use high-fidelity, one-way coupled, finite volume (FV) computational fluid dynamics (CFD) and finite element (FE) thermal stress analysis, modelling and simulation (M&S) methods. In this paper, we compare different M&S approaches to performing one-way coupled CFD, CHT and stress analyses of the thermal stress within the pipe wall of a T-junction for a temperature transient test case with fixed mass flows. These high-fidelity multiphysics models are a necessary first-step in developing mechanistic approaches for analysing thermal fatigue due to cyclical thermal stresses induced by the transient thermo-fluid behaviour within T-junctions. In the first approach, a one-way coupled, CFD, CHT, and thermal stress analysis is performed using only the commercial off-the-shelf (COTS) CFD code STARCCM+. We compare the results with a second approach that uses a one-way coupled sequential STARCCM+ CFD simulation and Abaqus FE thermal stress simulations via the SIMULIA Co-Simulation Engine. The comparison between these two approaches shows small differences for the predictions of von Mises stresses, normal stresses, and displacements. Values predicted by STAR-CCM+ around the pipes’ junction, where strong thermal mixing occurs within the fluid, are in very good agreement with the values predicted by Abaqus and within few percentage points of one another. These differences increase at locations corresponding to the system constraints such as the flange locations.
Study on Eutectic Melting Behavior of Control Rod Materials in Severe Accidents of Sodium-Cooled Fast Reactors (2) Modeling of Multi-Phase Eutectic Reaction Behavior
The eutectic reaction between boron carbide (B4C) and stainless steel (SS) and the relocation of reaction products in a damaged core are important phenomena to be considered in the safety analysis of severe accident (SA) in sodium-cooled fast reactors (SFRs). In this study, a multiphase eutectic reaction model was developed to simulate eutectic reactions and relocation of reaction products in the damaged core and incorporated into the fast reactor safety analysis codes SIMMER-III and SIMMER-IV. The eutectic reaction model was coupled with the multi-fluid model, the fluid dynamic model of the SIMMER codes, because the reaction products cause thermal and hydraulic interactions with multiphase core materials in the damaged core. In addition, the equation of state and thermophysical property models of the SIMMER codes were extended to allow for the B4C concentration dependence of the physical properties of the eutectic reaction products. The developed eutectic reaction model was validated for its capability by analyzing a B4C-SS eutectic reaction experiment in which a B4C pellet was immersed in a molten SS pool. In the validation analysis, quantitative comparisons were made on the measured thinning of a B4C pellet due to the eutectic reaction and the distribution of a boron component dispersed in the molten SS pool along with the melt of the reaction product. Validation results indicate that the SIMMER-III code incorporating the developed eutectic reaction model has the potential capability to evaluate the eutectic reaction and the relocation behavior of reaction products during SA of SFR.
Preliminary Study of Microreactor Emergency Planning Zone
This study examines the dispersion patterns of a generic nuclear microreactor accident. The investigation involved a comprehensive analysis of χ/Q values across various receptor distances and directional orientations over time intervals ranging from immediate post-accident periods to 30 days. The computer code ARCON 2.0 was used to perform all simulations for the current study, with a meteorological data of approximately 1 year. Key findings indicate a consistent trend of elevated χ/Q values at a receptor distance of 15 meters, emphasizing the significance of proximity to the accident epicenter in assessing potential environmental impacts. Moreover, a discernible decrease in χ/Q values was observed within the near-field dispersion zone (15 to 50 meters), followed by stabilization in dispersion patterns beyond this range. Comparative analysis revealed notable differences in χ/Q values across different post-accident time intervals, suggesting a convergence towards equilibrium in dispersion trends over time. Future efforts intend to integrate these findings with accident source term data from TRIGA reactors to evaluate Emergency Planning Zone (EPZ) size determination.
Uncertainty Quantification (UQ) for the Analysis of Thermal Stresses within a T-junction Pipe Wall Using a Structured Gaussian Process (GP) Method
Thermal fatigue of components within the primary and secondary circuits of nuclear power plants (NPPs) is important to understand from a structural integrity perspective. T-junction components within NPPs may be subject to near-wall temperature changes that can induce thermal stresses within pipe walls, eventually leading to fatigue. This paper focusses upon the multiphysics aspects of thermal fatigue phenomena and aims to perform efficient uncertainty quantification (UQ) of one-way coupled, finite volume computational fluid dynamics (CFD) and finite element thermal stress simulations. The aim is to understand the uncertainties associated with the thermal stresses, within the pipe wall of T-junctions, due to uncertainties in transient inlet flow conditions. To minimise the required number of computationally demanding coupled CFD and thermal stress analyses, the structured Gaussian process (GP) based surrogate modelling approach is adopted in this work. The surrogate models of wall temperatures and thermal stresses are trained using the structured GP algorithm with data generated by transient coupled CFD and thermal stress simulations. The GP is then used to propagate the uncertainties in the inlet flow conditions to the temporal evolution of temperatures and stresses at probes located at the pipe inner wall. The GP predictions show that, due to the uncertainties in the transient inlet flow conditions, the maximum temperature and stress can vary between ~210/245°C and ~ −1260/−1660 MPa, respectively, for the temperature transients considered in this study. The proposed GP-based UQ methodology allows for efficient UQ and can be applied to coupled CFD and thermal stress analyses of T-junctions.
A Preliminary Severe Accident Analysis of a Generic Small Boiling Water Reactor
Canadian Nuclear Laboratories (CNL) has been conducting research to support the deployment of the various types of small modular reactors (SMRs) in Canada. Among the different SMR technologies, Canada is poised to start deploying several small, natural-circulation-cooled boiling water reactors (BWRs). These types of small-BWRs utilize two major systems that are natural-circulation-driven: the isolation condenser system (ICS) for decay heat removal, and the passive containment cooling system (PCCS). While the use of such passive safety systems is likely to reduce the overall likelihood and severity of accidents, it is still critical to understand how these passive systems could be impaired, and where in accident space that is possible. These then help to inform what mitigating equipment, accident management guidelines, and emergency preparedness plans would need to be in place to address the lower frequency extreme events. These efforts include a thorough understanding of the severe accident behavior of the small BWR and a good estimation of the associated radionuclide source terms. This paper summarizes CNL’s initial efforts on the severe accident analysis of a generic small BWR. The analysis was performed using the MELCOR severe accident analysis code with a focus on two major types of events, 1) a station blackout accident (SBO) with impaired ICS, and 2) a small-break loss-of-coolant accident (LOCA). The effectiveness of ICS and PCCS were assessed with particular interest in the impact of released hydrogen on the performance of these systems.
Interface Resolved Simulation of Two-Phase Flow within a Steam Separator Geometry
Steam separators have been an integral component in the current operating Boiling Water Reactor (BWR) fleet and will continue to play a role in new BWR reactor designs. Mechanistic models used in the design process may utilize the insights gained from high resolution experiments which can increase safety/operational margins. However, direct observation of the complex two-phase flow within steam separators at operational conditions is both difficult and expensive to achieve. An alternative approach to real-world experimentation exists in the form of numerical experiments using Direct Numerical Simulation (DNS). In this work, a model of a steam separator is simulated using the PHASTA two-phase incompressible finite element code. PHASTA can directly resolve the motion of droplets/bubbles using an interface capturing approach via the Level Set method. Specifically, the effect of the pickoff-ring is examined through metrics like the outlet flow quality and steam carryunder. Time-averaged statistics are also acquired for future use in system model improvements. By using massively parallel HPC systems, high Reynolds number flows can be fully resolved, although it is still well below prototypical BWR conditions.
A Summary of the Uncertainty and Sensitivity Analysis of a Generic CANDU 6 Plant by Means of Severe Accident Codes in the Framework of the IAEA CRP I31033
IAEA launched in 2019 the Cooperative Research Project (CRP) I31033 entitled “Advancing the State-of-Practice in Uncertainty and Sensitivity Methodologies for Severe Accident Analysis in Water-Cooled Reactors”, with the main goal to advance the understanding and characterization of sources of uncertainty and investigate their effects on the key figure-of-merits (FOMs) of severe accident code predictions in water cooled reactors (WCRs). The 5-year CRP included participants from 22 organizations representing 18 Member States and generated 5 International Atomic Energy Agency (IAEA) technical documents on uncertainty and sensitivity (U&S) analyses of severe accidents scenarios in different water-cooled reactors (PWR, BWR, CANDU, VVER, and SMR), as well as the QUENCH-06 experiment. With respect to the CANDU (CANada Deuterium Uranium) pressurized heavy water reactor analyses, three organizations in three Member States (Canada - CNL, South Korea - KAERI, and Romania - UPB) participated in the CRP. Each partner used their own severe accident and uncertainty tools to evaluate relevant input parameters and their impact on common figures of merit (e.g., hydrogen generation, event timing and fission product releases). All three CANDU participants studied a station blackout in a generic CANDU 6 station, with similar analysis assumptions, and used similar plant normal operating conditions. This paper presents a summary of results of the U&S analyses performed by the CANDU group within the IAEA CRP and discusses all relevant insights and lessons learned in view of the application of the U&S methodologies for severe accident analyses in CANDU.
Advanced Thermal Monitoring in Scaled Reactors using Deep Learning-Enhanced Drone IR Imaging
For enhance surveillance capability within nuclear power plant and assist expert in hazardous region, this study presents a combined technology of drone-equipped infrared (IR) imaging and segregation deep learning for remote and smart monitoring of nuclear power plant. The test facility, designed to replicate the thermal-hydraulic phenomena of the OPR1000/APR1400 at lower pressures and temperatures, was monitored using drones equipped with dual vision and IR cameras. This setup allowed for the collection of detailed imagery under various test scenarios, including normal operation, transient operation such as all reactor coolant pumps trip and station blackout conditions. The captured images were analyzed using advanced deep learning algorithms, including object detection and instance segmentation models. These models were trained to identify normal states and abnormal states within the reactor model, providing enhanced insights into the dynamic thermal-hydraulic behavior of the system. The application of color-coded bounding boxes and segmentations in the imagery further facilitated the identification of subtle anomalies and component faults, which are challenging to detect through traditional monitoring methods. Our approach demonstrates an advancement in the use of AI and drone technology for real-time monitoring and analysis in safety-critical nuclear reactor environments. The combination of these technologies in a scaled IET setting offers a more comprehensive and automated method for thermal-hydraulic analysis, contributing to improved safety and efficiency in nuclear reactor operations.
On the Influence of Model Uncertainty in Thermal Stratification Simulations of Pool-Type Sodium-Cooled Fast Reactors using Computational Fluid Dynamics
Thermal stratification in the upper plenum of pool-type sodium-cooled fast reactors (SFR) can occur during transient and accident conditions, potentially impacting the natural circulation performance and adversely affecting heat removal performance. Having reliable simulation tools and a good understanding of simulation uncertainty is crucial for designing a robust reactor. In this research, we demonstrate an approach to quantify the effect of model uncertainty on thermal stratification simulations using computational fluid dynamics (CFD). A 3D Reynolds Averaged Navier-Stoke (RANS) model is established based on the SUPERCAVNA facility, in which the flow scenario is an approximate representation of the complex flow in the upper plenum of SFR. To assess the influence of model uncertainty, the turbulent viscosity is treated as a Gaussian random field, in which the hyper-parameters are inferred in phenomena-based separate effects tests and then applied in a forward context to the application of interest. The uncertain mixed-convection phenomenon is investigated by quantifying the variation of temperature distribution; the uncertainty mixing flow rate at different elevation levels is also accessed. The preliminary result shows that the size and shape of the flow recirculation zone are significantly affected by a small perturbation of the eddy viscosity field. As a result, large variations in temperature profile and mixing flow rates are observed.
Numerical Study of Two-Stage Safety Injection Tank in a Small Modular Reactor
Numerical analyses were performed for a two-stage safety injection tank (TSIT), which is able to passively inject coolant with a single tank in two stages in case of a loss of coolant accident (LOCA) in a small modular reactor. At the early stages of the accident, coolant of the TSIT is injected at a high flow rate into the reactor. When the TSIT water level is reduced lower than the predetermined level, the TSIT injection flow rate is instantaneously reduced to a lower flow rate in a passive manner. Subsequently, coolant of the TSIT is injected at a low flow rate for a long period of time until the TSIT is depleted. The effects of the break size, installation height of the tank, installation height of the injection hole, and pressure-loss coefficient of injection line on the TSIT injection performance were examined numerically. With pertinent design parameters, the TSIT could satisfy its design requirements in an integral reactor, by providing two-stage coolant injection for at least 72 hours after a LOCA.
Data-Driven Prediction of Preheater Steady-State Conditions Based on Support Vector Regression
The determination of steady-state operating conditions is a fundamental and complex task in the computation of heat exchangers. Traditional approaches often involve intricate models based on conservation laws—mass, energy, and momentum. These methods, while thorough, tend to be intricate, labor-intensive, and heavily dependent on empirical data, often making them less efficient and adaptable in diverse applications. Contrasting sharply with these conventional techniques, machine learning and data-driven methods stand out for their simplicity, user-friendliness, and swift response times. Despite their successful implementation in various domains, their integration into heat exchanger computations remains surprisingly limited. To bridge this gap, this study introduces the application of Support Vector Regression (SVR), a supervised machine learning algorithm. This research revolves around training a model with datasets generated from heat exchanger design software, aiming to accurately predict the steady-state conditions of preheaters and evaluate the robustness of the model. The results reveal that SVR is remarkably effective in forecasting the steady-state operations of preheaters. The model demonstrates an impressive accuracy level, with the maximum deviation between the predicted outcomes and actual computational results not exceeding 2%. Moreover, the model exhibits good robustness. It maintains reliability and accuracy even in the face of input fluctuations. Such robustness not only validates the effectiveness but also highlights its potential for practical deployment in real-world settings.
Nuclear Power Plant Condenser Improvement
Electrical output of a reactor unit is determined by the amount of heat produced by the reactor, efficiency of the turbine cycle and by the condenser cooling water temperature. The amount of heat produced by the reactor is largely determined by the reactor operating licence and the turbine cycle efficiency is determined by the turbine design. Bruce Power operates one of the largest nuclear power plants in the world with eight operating reactor units, which provide ~30% of Ontario power with low carbon footprint. The condenser cooling water source for the reactors is Lake Huron with temperature variation from approximately 0°C in the winter to 25°C in the summer. The electrical output of each reactor unit at the currently licenced reactor power is reduced by as much as 50-60 MWe in the summer, compared to the winter months. The reduction will be even higher when the licenced reactor power is increased after power uprates are implemented and the average Lake Huron water temperature is increased due to climate change. This paper describes conceptual engineering investigation into improvement of the plant condenser performance. The investigation utilizes the approach based on the extensive analysis of station data, the Heat Exchange Institute Standards for Steam Surface Condensers and model simulations using the Performance Evaluation of Plant System Efficiencies (PEPSE) software code. Several condenser modifications are evaluated based on the cost-benefit analysis and extensive operating experience of other nuclear plants.
Preliminary Study on the Integration of Organic Rankine Cycle with the Low-Temperature Heating Reactor
Nuclear heating reactor operates at a low temperature (usually below 200 ℃), which can be used for heating or combined with power generation, desalination and other devices. The Organic Rankine Cycle (ORC) is often used for the power generation of low temperature heat sources such as industrial waste heat and geothermal energy. However, there has been little research focused on nuclear energy. Consequently, this paper introduces a novel system that integrates a nuclear heating reactor with the ORC. The ORC system selects n-heptane as the working medium, which is a dry working fluid with relatively high critical temperature and excellent thermodynamic properties. In this paper, ASPEN PLUS software is used to model the ORC in detail, and the thermal performance of the recuperative Rankine cycle is studied. In this study, we focus on optimizing the performance of ORC system by adjusting two key parameters: the evaporation pressure and the condensing pressure. The objectives are to maximize the net output work, thermal efficiency, and exergy efficiency. Additionally, we analyze the impact of several critical parameters on the system's overall performance.
System-Level Thermal-Hydraulics Analysis of the Proposed NIST Neutron Source Design
The National Institute of Standards and Technology (NIST) is considering replacing its existing reactor, known as NBSR, with a cutting-edge test reactor. This new test reactor, the NIST Neutron Source (NNS), is currently in the conceptualization phase. The current design of the NNS consists of a 20 MW compact reactor that uses light-water cooling, features a pool-type configuration, and incorporates a heavy water reflector tank. The core consists of a 3x3 rectangular lattice of assemblies fueled with low-enriched U-10Mo curved plates with aluminum cladding, where each assembly contains 21 fuel plates. This paper provides system-level thermal-hydraulics analysis results for the proposed NNS under various normal operation and accident conditions. The system-level model of the NNS is developed using thermal-hydraulics code TRACE V 5.0 Patch 5. The model contains the core components, coolant piping, the reactor pool, and the primary coolant pumps to provide the thermal and hydraulic behavior of the coolant in the system and throughout the core. The goal is to establish safety margins for the safe operation of the NNS and assess the evolution of pre-determined accident conditions. Analyses include hottest channel temperature distributions and thermal safety margins such as Critical Heat Flux Ratio and the Onset of Flow Instability Ratio, both during state-state and transient conditions, such as Loss of Flow Accident, Loss of Coolant Accident, and Reactivity Insertion Accidents. The natural circulation cooling capability of the NNS is also assessed in this paper.
Prediction of Single Channel Critical Heat Flux by Machine Learning Approaches
Prediction of critical heat flux is crucial for the design and safety of nucleate boiling systems. Recent advancements in predicting critical heat flux center around machine learning methods, with a notable challenge being the size of the training dataset gathered from past experiments. Recently, NEA launched an international benchmarking endeavor for ML/AI based CHF modeling and prediction using the critical heat flux database provided by the US Nuclear Regulatory Commission (NRC) based on the 2006 LUT database. In this study, a large dataset covering mainly single channel CHF data with test settings and conditions similar to the NRC dataset is utilized to develop a machine learning framework for predicting critical heat flux. This machine learning framework is then applied to predict single tube CHF under various conditions, which is then compared with experimental results. The objectives include feature selection and importance analysis, model optimization, and achieving high-performance metrics. Firstly, we conducted data pre-processing including cleaning, scaling, and feature selection to ensure the high-quality datasets for training the model. Next, we developed a deep neural network (DNN) and optimized it through fine-tuning hyperparameters to train the data effectively. The overall R-squared value is outstanding for the unseen dataset, with a reasonable RMSE value. Finally, we employed the SHAP (SHapley Additive exPlanations) approach to attribute credit for the model's predictions to individual input features. The analysis suggests the potential removal of less important features to simplify the model and reduce computational complexity.
Numerical and Experimental Study of Radiative Heat Transfer with Radiation Shielding for Application of Metal Containment Vessel Design in Small Modular Reactor Development
Small modular reactors (SMRs) are gaining attention for their safety, low construction costs, and modularity. Many SMRs adopt metal containment vessels (MCVs) for enhanced safety, allowing residual heat removal through the outer wall during accidents. However, the high thermal conductivity of MCVs leads to increased heat loss during normal operation. To address this, creating a vacuum between the Reactor pressure vessel (RPV) and MCV can minimize convective heat transfer but amplifies radiative heat transfer. Preliminary experiments have shown that radiative heat loss accounts for about 60% of total heat loss. Thermal radiation shielding experiments were conducted to reduce radiative heat loss. Two parameters, shielding diameter and material, were chosen for the experiment to show the effect of geometry and emissivity. Experiments were conducted with aluminum (Al) shielding and SS304 shielding to determine the heat loss due to emissivity. Experiments were conducted by changing the shielding diameter to determine the heat loss due to the shielding position between the RPV and MCV. In the main experiment, Al shielding demonstrated a 10% reduction in heat loss compared to SS304, and a smaller shielding diameter resulted in a 3% decrease. The RPV temperature was maintained at 320 ℃, and calculations at steady-state were used to evaluate MCV heat loss to the surroundings. These findings suggest that radiative heat loss should be considered in SMRs with MCV.
Investigation of Thermal Limitations in Heat Pipes Using Three Different Working Fluids in a Heat Pipe Reactor
Heat pipe reactors are being considered for electricity generation in remote locations because they are small, transportable, cost-effective as well as suitable for mining sites (for mining activities), individual industrial heat and power scenarios, distributed hydrogen generation and integrated energy solutions. However, one of the most important considerations when using heat pipes to cool a reactor is the thermal limits of the heat pipes. If heat is not extracted from the heat pipes, this can lead to a risk of accidents. It is therefore important to create a safe design that has accounted for the thermal limits of heat pipes. In addition, removal of heat from fuel is one of the fundamental safety requirements in the CNSC’s regulatory framework. In this study, the limits of heat pipes in a heat pipe reactor are investigated. Three different working fluids used in heat pipes were selected: potassium, sodium and lithium. The results for these three working fluids are compared. Additionally, the paper discusses the safety analysis challenges associated with heat pipe limitations.
CFD Analysis of Hydrogen within a Filter-Vented Canister Containing Fukushima Fuel Debris
This paper describes experimentally verified CFD analysis that examines safety of a possible design for passive filter vented interim storage of fuel debris from the Fukushima Daiichi Nuclear Power Station. Retrieved fuel debris will be firstly placed into a small cylindrical “unit can.” Four unit cans will be stacked inside a canister whose lid contains two high efficiency filters in series. Loaded canisters will be temporarily stored in a storage facility at the power station site. Since the fuel debris contains water, hydrogen is generated in the canister due to radiolysis and it must be discharged through filter to avoid forming a flammable mixture within the canister. In this study, behavior of the hydrogen generated in the canister was estimated by CFD analysis for understanding the hydrogen discharge performance from the canister based on the conceptual design. The filter model and the canister lid model etc. were developed for the CFD analysis and the validity of these models were confirmed based on experimental results obtained by Fauske & Associates. From the CFD analysis, the hydrogen concentration distribution in the canister based on the conceptual design was not indicated locally stagnant or flammable. Furthermore, we estimated the hydrogen discharge performance from the canister when the unit-cans are stacked in an unorderly manner in the canister, and it was found that the hydrogen discharge performance was not affect by the unorderly stacked unit cans.
Fuel Fragment Jamming Inside Fuel Rods Following a LOCA
The US nuclear industry is designing longer fuel cycles requiring fuel enriched beyond the current limit of 5 wt% and resulting in burnups beyond the current rod average limit of 62 GWd/MTU. As burnup increases, fuel fragments leading to potential issues caused by fine fragmentation, relocation, and dispersal (FFRD). Following a LOCA, burst of high burnup fuel rods could disperse fine fragments into the fuel assembly coolant channels, which can challenge core cooling. The amount of fuel dispersal will be limited by fuel fragment interactions upstream of the burst opening that can lock or form a bridge across the opening. A review of available literature on particle jamming is presented to demonstrate that dispersal of high burnup fuel fragments upon burst following a LOCA will be limited by this phenomenon.
Morphology Change of CRUD Formed on Fuel Cladding by Deposition Time in Pressurized Water Reactor Environment
In this study, to investigate morphology and thermal property change of fuel CRUD, simulated CRUD deposition experiment was conducted for three different period (24 hours, 72 hours and 168 hours) in PWR primary water condition. The specimens were examined by scanning electron microscopy with focused ion beam instrument and energy dispersive X-ray spectroscopy. To investigate the effect of CRUD morphology on its thermal properties, thermal conductivity of CRUD was calculated from difference of cladding wall temperature which measured before and after the CRUD deposition experiment. In 24 hours exposed specimen, CRUD layer was usually composed of equiaxed nickel ferrite particle and few needle like nickel oxide particles. Whereas in 72 hours and 168 hours exposed specimen, composition of needle-like nickel oxide particles was increased. It can be reasonably explained by particle formation energy of nickel ferrite and nickel oxide particle. Because nickel ferrite formation is spontaneous reaction in PWR primary water, it forms prior to nickel oxide, which is not preferred reaction in the environment. Since nickel oxide formation energy decrease with temperature increase, when thickness of CRUD layer increased enough so that temperature of CRUD layer increased, nickel oxide particles are formed.
Research on Resistance and Flow Characteristics of 7-Rod Bundles with Simple Support Grid at Reynolds Numbers of 1000-10000
The grid positioning triangle fuel assembly is a potential fuel design form in the fourth generation reactor, especially in lead/lead bismuth reactors. Although different from the critical heat flux strengthening brought about by the complex form of PWRs grids, it is still necessary to study the resistance and flow characteristics of liquid metal in lead bismuth fuel assemblies with simple support grids, in addition to their positioning role. Due to the opacity of liquid metals, this study adopted the particle image velocimetry (PIV) and refractive index matching (RIM) method, utilizing the similar flow characteristics of water as working fluids at the same Reynolds number to conduct experiments and measured resistance characteristics. The influence of simple support grids on velocities, as well as the resistance characteristics of simple support grids, were studied in the range of Reynolds numbers from 1000 to 10000. The results show that, the SST k-ω turbulence model in CFD showed consistency with PIV experimental measurements over a wide range of Reynolds numbers, making this turbulence model more applicable. And the simple support grid resistance characteristic has been studied, a new grid CV expression has been obtained. The research results can provide support for the calibration or model improvement of the fourth generation fast reactor subchannel code and CFD tool.
An Integrated Fiber Optic and Acoustic Leak Detection Technologies for Nuclear Power Plants Water Supply and Reactor Cooling Systems
Leak Detection Systems (LDS) play a critical role in ensuring the safety and operational integrity of nuclear reactors. The detection of small leaks is of paramount importance in this context, as even minor breaches might lead to the release of radioactive materials, posing significant environmental and health risks. Furthermore, small leaks often precede larger failures, making early detection crucial for preventing catastrophic incidents. The development of leak patterns from small, undetected leaks underscores the need for advanced detection technologies that can identify and localize these leaks with high precision. The current technologies employed for leak detection in nuclear reactors face several critical limitations, including their extremely slow response times, inability to detect small or early-stage leaks, and lack of precise localization capabilities. These shortcomings pose significant risks, as undetected or delayed leak identification can lead to severe hazards and operational inefficiencies. This paper presents a solution to these issues that integrates advanced acoustic leak detection technique using fiber optic sensors (FOS) and fiber optic leak detection algorithms such as Raman, Rayleigh, Brillouin, Interferometric sensing, and Fiber Bragg Grating (FBG) based methods. With the highly sensitive fiber optic sensor, the acoustic leak detection systems are capable of detecting acoustic pressure wave, generated by even the smallest leaks and propagates inside the pipeline. Additionally, distributed fiber optic sensing with various algorithms offer high-resolution, real-time monitoring of temperature, strain, and vibration along the length of the water supply line or reactor cooling systems (RCS) to detect leak based on any external phenomena change (including temperature, strain, and vibration) resulting from any leak event, Together, these technologies provide a comprehensive and effective solution for early leak detection and precise localization, significantly enhancing the safety and reliability of nuclear reactors. By integrating these technologies, nuclear reactors can achieve a comprehensive leak detection strategy to safeguard the water supply line and reactor cooling systems with early and precise leak detection that enhances safety, reduces the risk of radioactive contamination, and ensures the long-term reliability of the reactor infrastructure. The integration of acoustic leak detection and distributed fiber optic leak detection technologies provides a multi-layered defense against leaks, enabling the timely identification of potential issues before they escalate into serious problems. This system will be expanded to detect leaks in the reactor and steam generators in future investigations.
Subchannel Simulation of Rod-Bundle Two-Phase Flow Experiments with Flica5
Flica5 is the latest CEA’s thermal-hydraulic software system for computing subchannel two-phase flow in Light Water Reactors. It embeds modified correlations from its predecessor Flica4 as well as new numerical architecture and solvers. This paper describes a simulation with Flica5 of BFBT and PSBT test series, which are reference OECD/NRC rod-bundle experimental databases. The void fractions measured during the BFBT tests are compared with the ones predicted with Flica5. A statistical method is applied in order to identify the best-fitted and most versatile closure laws. Comparisons between the experiments and the simulations show that 95 % of the discrepancies are below 4%. Both the steady-state single sub-channel and the steady-state bundle benchmarks from the PSBT tests are simulated with a certain set of closure laws in Flica5. In both cases, predicted void fractions are in good agreement with the experimental ones. Flica5 tends to underestimate the void fraction in the sub-channel cases while overestimating it in the bundle cases.
Experimental Investigation on Aerosol Scavenging by Electrically Charged Spray with Preinjected Water Mist for Fukushima Daiichi Decommissioning
Retrieving the resolidified debris in both aerial and submerged conditions is crucial in the context of Fukushima Daiichi (1F) decommissioning. This involves segmenting the debris into manageable pieces through assorted techniques, such as laser cutting. During these operations, radioactive aerosols, mostly submicron-sized, may generated, necessitating a containment spray system for removal. The challenge is the aerosols within the Greenfield gap (0.1-1 μm) are hard to remove with conventional sprays. Hence, a new method using water mist to enlarge aerosol sizes before removal with water sprays was proposed. Further, spray droplet changing is a useful technique to improve the aerosol scavenging efficiency through the additional electrostatic interactions for facilitating the aerosol capture. In this study, experiments are conducted in our UTARTS facility to improve the aerosol removal efficiency by developing an electrostatic-charged spray system with water mist. A ring brass electrode near the spray nozzle's outlet, charged by a high-voltage power supply, induces charges on the droplets. Prior to spray activation, water mist is injected to agglomerate aerosol particles for their size enlargement. Results indicate accelerated aerosol removal with electrically charged sprays, especially as voltage increases. However, the benefit plateaus at higher voltages (e.g., higher than 15 kV) due to droplet charge saturation. Pre-injected water mist also enhances removal efficiency under current electro-spraying conditions. The findings obtained from the current study will be beneficial to the optimization of spray system design for effective aerosol scavenging during the 1F decommissioning.
Loss-of-Heat-Sink Transient Simulation with RELAP5/Mod3.3 Code for the ATHENA Facility
ATHENA (Advanced Thermal-Hydraulic Experiment for Nuclear Applications) is a large multipurpose pool-type lead-cooled facility, which is under construction at the Mioveni site, in Romania. It has been identified by the FALCON (Fostering ALfred CONstruction) Consortium to conduct integral tests, to characterize large to full-scale ALFRED components, and to investigate the main thermal-hydraulic phenomena involved in a pool-type system. ATHENA is representative of ALFRED in terms of the difference in height of the thermal barycenters of the heat source and heat sink, i.e., 3.3 m, in order to reproduce the buoyancy forces in the system. As foreseen also for ALFRED, thermal stratification inside the main vessel is limited, even in natural circulation, by the presence of an internal structure, named barrel, that drives the fluid towards the main vessel, avoiding the stagnation of the fluid adjacent to the external wall. Firstly, a steady-state thermal-hydraulic characterization of the facility is presented, along with the description of the numerical model realized with RELAP5/Mod3.3. Then, the work focuses on the transient analysis considering as a reference scenario a Loss-of-Heat-Sink (LOHS) accidental transient, during which the Core Simulator (CS) and the Main Circulation Pump (MCP) are supposed to continue to work. The response of the system following the isolation of the Main Heat Exchanger (MHX) from the secondary loop is analyzed, and the effect of the delay in the scram of the reactor (reducing the CS power to a value representative of the decay heat) on the pin cladding and system maximum temperature is evaluated. The pre-test simulation presented in this paper have shown that the facility is able to achieve a safe shutdown condition thanks mainly to the high thermal inertia of the lead pool.
Three-Dimensional Lagrangian-Eulerian Modeling and Analysis of Fuel-Coolant Interaction
Fuel-Coolant Interaction (FCI) is a phenomenon in which hot molten fuel interacts with surrounding coolant during a nuclear reactor severe accident and the potential explosive outcome of such an event could threaten the reactor containment integrity. In this study, a multi-field, three-dimensional computational model has been developed to assess the plant safety by an energetic FCI. The molten fuel is modeled in Lagrangian frame and the liquid and vapor of coolant fluid are modeled in Eulerian space. The Lagrangian and Eulerian fields are coupled semi-implicitly. The breakup of molten fuel jet is modeled based on the physical mechanisms of Rayleigh-Taylor instability at jet front and Kelvin-Helmholtz instability at jet lateral surface. The interfacial heat and momentum transfer between phases are modeled based on flow regimes. The Lagrangian particles of fuel is treated as a parcel and its size is allowed to expand as the fuel breaks up. Verification of the model has been carried out for experimental data of prototypical molten fuel: KROTOS and TROI. The validation calculations showed that the present model is able to adequately simulate an energetic fuel-coolant interaction. Three-dimensional calculations of ex-vessel steam explosions in a PWR reactor cavity showed that the wall impulse in case of off-center entry of melt jet by 1 m is 50% higher than that of melt entry at the center of cavity.
ARIANT and RELAP5-3D Simulation of OECD-NEA Benchmark Against the Oregon State University High Temperature Test Facility Data
Canadian Nuclear Laboratories (CNL) is developing the system thermalhydraulics code ARIANT (AlgoRIthm for Analysis of Network Thermalhydraulics) to extend the code’s capabilities for advanced reactor and small modular reactor analysis. As part of the ongoing modelling capabilities development, CNL is participating in the OECD-NEA prismatic High Temperature Gas Reactor thermalhydraulics benchmark based on measurements from the Oregon State University (OSU) High Temperature Test Facility (HTTF). The objective of this study is to present preliminary results of code to code benchmarking efforts undertaken at CNL using two system thermalhydraulic codes (ARIANT and RELAP5-3D). ARIANT and RELAP5-3D simulations were performed for the pressurized and depressurized conduction cooldown (PCC, DCC) tests and the code results were compared against each other. For both codes, CNL developed a ring model that simulates the HTTF. The ARIANT and RELAP5-3D predicted results are in reasonable agreement with each other. Based on the analyses undertaken, it can be inferred that ARIANT and RELAP5-3D can model the phenomena that are expected to occur in postulated accident scenarios in prismatic-block gas-cooled reactors.
Experiments and CFD for Wire-Wrapped Fuel Assembly Blockages
Liquid metal cooled fast reactors are under development world-wide with the aim to contribute to an increased sustainable nuclear future. These reactors often employ wire-wrapped fuel assemblies in the core. For the safety assessment of such reactors, a reasonable prediction of the flow and temperature field in the fuel assemblies is required, both for operational as well as for accidental conditions. This is usually achieved using a combination of experimental work and various types of numerical simulation. This paper will describe the collaborative work performed by European partners with respect to the effect of blockages in wire-wrapped fuel assemblies. To begin with, results for an unblocked situation will be presented for experiments and simulations, showing satisfactory comparison. After that, initial work performed on solid blockages will be described. It will be shown that first simulations of such blockages showed large discrepancies with experimental results. As a result, a number of sensitivity analyses revealed the major contributors to these discrepancies, which has led to the design of a porous blockage experiment and simulation campaign, which at the same time is more representative of the postulated reactor conditions. Finally, the first results of the ongoing porous blockage campaign will be shown.
Physics-Informed Neural Networks for Two-Phase Flow Simulations: An Integrated Approach with Advanced Interface Tracking Methods
Physics-informed neural networks (PINNs) are emerging as one of the promising artificial intelligence approaches for solving complex thermal-hydraulic simulations. A critical challenge in these simulations is an accurate representation of the gas-liquid boundary with different interface tracking methods. While numerous studies in conventional computational fluid dynamics (CFD) have addressed this issue, there remains a notable absence of research within the context of PINNsbased two phase flow simulations. Therefore, this study aims to develop a robust and generic PINNs for two-phase flow by incorporating governing equations and three advanced interface tracking methods—specifically, the Volume of Fluid, Level Set, and Phase-Field methods—into the adaptive residual differential (ARD) PINNs framework that has been previously proposed and validated. To further enhance the performance of the PINNs in simulating two-phase flow, the phase field constraints and the time divide-and-conquer strategies are employed for restricting neural network training within the scope of physical laws. The ARD PINNs then is optimized by minimizing both the residual and loss terms of partial differential equation. By incorporating the three different interface tracking methods, it efficiently handles high-order derivative terms and captures the phase interface. The case of single rising bubble in two-phase flow is simulated to validate the robustness and accuracy of the ARD PINNs. The accuracy of the simulation is compared with the velocity, pressure, and phase field obtained from the CFD simulation. The results indicate that the ARD PINNs coupled with these interface tracking methods offers a satisfactory performance in simulating rising bubble phenomenon.
Cooperative Nuclear Safety Research Activities at the Nuclear Energy Agency in Response to the Fukushima-Daiichi Accident
Since the 2011 Tōhoku earthquake and tsunami, which severely damaged the Fukushima Daiichi Nuclear Power Station (FDNPS), the Nuclear Energy Agency (NEA) has been leading international efforts in nuclear safety research. These activities, coordinated through the Working Group of Accident Management and Analysis (WGAMA) of the Committee on the Safety of Nuclear Installations (CSNI), aim to facilitate the safe decommissioning of FDNPS and enhance safety measures for nuclear power plants (NPPs) globally. Numerous joint projects, including the Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF), Preparatory Study on Analysis of Fuel Debris (PreADES), and others, focus on analyzing the accident and preparing for decommissioning. The most recent initiative, the Fukushima Daiichi Nuclear Power Station Accident Information Collection and Evaluation (FACE) project, launched in July 2022, targets key technical challenges identified in recent investigations. However, significant challenges remain in managing the long-term impacts of the FDNPS accident. Continued international collaboration, particularly through ongoing projects like FACE, is crucial to inform decommissioning efforts and enhance accident management for existing and future reactors. Looking ahead, the NEA plans to organize an international workshop in September 2024 to explore future research directions in accident management, leveraging insights from the Fukushima Daiichi accident. This workshop aims to guide future research efforts and improve safety measures for both operating and future nuclear reactors. Overall, international collaboration and ongoing research endeavors are essential for addressing the complex challenges posed by the Fukushima Daiichi accident and enhancing global nuclear safety standards.
Hydraulic experiments on a 1/5 Scale Model of APR1000 reactor
The hydraulic characteristics at the reactor core are important information for licensing and design processes. The APR1000 reactor contains 177 fuel assemblies in the core region, and the core inlet flow of each fuel assembly must be presented with high accuracy through experimental methods. Core flow rates are significantly influenced by flow conditions and the hydraulic resistances of the internal structures. In this study, the Advanced power reactor COre Flow & Pressure (ACOP) test facility was constructed at a reduced scale of 1/5 to simulate the entire internal flow of the APR1000 reactor. The design concept and applied scaling method are briefly introduced, and preliminary experimental results are presented for core inlet flow and outlet pressure distributions under both balanced and unbalanced flow conditions.
Validation of Evaluation Model for Thermal Absorber System in High Temperature Gas-Cooled Reactor Nuclear Hydrogen Production System
Japan Atomic Energy Agency (JAEA) is planning a nuclear hydrogen production demonstration test by connecting a hydrogen production to the HTTR (High Temperature engineering Test Reactor), a 30MWt High Temperature Gas-cooled Reactor (HTGR) test reactor located at JAEA’s Oarai Research Institute. The objective of the test is to establish coupling technologies between HTGR and a hydrogen production plant. One of key technologies to be developed is a thermal absorber consists of a steam generator and an air cooler that can mitigate rapid fluctuations in helium temperatures during loss of heat load in the hydrogen production plant. The present study aims to develop an analytical model for a thermal absorber that can be used for safety analyses of the HTTR for coupling a hydrogen production plant. A set of closed loop model is newly introduced to a RELAP5-based system analysis code to enable evaluation of steam-water natural circulation behavior between the steam generator heated by helium and the air cooler cooled by ambient air. Experimental data obtained from a mock-up test facility of thermal absorber is compared with calculated results of the developed model. As a results of the comparative study, the code reproduced similar tendency in the transient thermal hydraulic behavior in the thermal absorber during the loss of chemical reaction test. The code also conservatively evaluated the variation in the helium gas temperature at the outlet of the steam generator from the initial condition compared to that observed in the test results. In conclusion, the system analysis code was validated for the application to the thermal absorber using the results of loss of chemical reaction test in the mock-up test facility.
Application of the GIF Safety Design Criteria and Safety Design Guidelines on Natural Circulation Capability to Next Generation Sodium-Cooled Fast Reactor in Japan
This paper describes the application of safety design criteria (SDC) and safety design guidelines (SDG) developed in the Generation-IV International Forum to the natural circulation of sodium to sodium-cooled fast reactors (SFRs) recently designed in Japan. It was consequently confirmed that the SFR designs conformed to the SDC and SDG without any specific gaps. Against design-basis accidents, the decay heat removal system (DHRS) is designed to have a combination of one train of a direct reactor auxiliary cooling system (DRACS) and two trains of a primary reactor auxiliary cooling system (PRACS) in a loop-type SFR, and a combination of one train of a DRACS and four trains of an intermediate reactor auxiliary cooling system (IRACS) in a pool-type SFR. Each DHRS design has a fully passive system with natural circulation capability by adopting reduction of system pressure loss and securing the difference in height between components. The natural circulation capability was confirmed by scaled experiments, one-dimensional plant dynamics calculations, and three-dimensional computational fluid dynamics simulations.
Investigating the Influence of PCM Properties, Containment Geometry, and Initial Conditions on the Transient Behavior of a PCM-Based Heat Removal System for Passive Molten Salt Fast Reactor
This study presents a comprehensive numerical analysis of melting phenomena and natural convection heat transfer in a heat removal system utilizing phase change material (PCM) for application in Passive Molten Salt Fast Reactor (PMFR). The integration of PCM-based systems in PMFR is crucial for enhancing safety and efficiency by mitigating temperature fluctuations and managing heat dissipation during transient conditions. The investigation focuses on the dynamic interplay between the phase change process of the PCM and the natural convection heat transfer mechanisms within the reactor system. A lumped-parameter model is developed to simulate and analyze the behavior of the PCM as it undergoes melting in response to varying heat loads. The study explores the influence of key parameters such as PCM properties, containment geometry on the overall performance of the heat removal system. The results highlight the effectiveness of the proposed PCM-based heat removal system in maintaining temperature stability and optimizing heat dissipation in PMFR. This research advances our understanding of PCM applications in passive safety systems for molten salt reactors. The insights gained contribute to designing and optimizing passive safety features in advanced reactor systems, potentially advancing the field of advanced nuclear engineering.
Detailed Assessment of Combustion isk and PAR Efficiency in the Late Phase of a Severe Accident within the European AMHYCO Project
The European AMHYCO project (Euratom 20192020, GA No 945057) aims at enhancing the understanding of H2/CO combustion risk within the containment of a nuclear power plant in the late phase of a severe accident. The goal is to incorporate this knowledge into severe accident management guidelines and recommendations for long-term operation upgrades. Based on a critical review of existing methodologies and practices related to gas combustion risk, as well as the identification of accident sequences where the containment integrity may get challenged, experimental investigations were conducted to close knowledge gaps related to combustion characteristics and the operation of passive autocatalytic recombiner under late phase conditions. To prepare the basis for the further assessment and refinement of existing SAMGs, systematic detailed analyses of the most challenging scenarios and possible mitigative measures were conducted for three generic European PWR containment designs, namely KWU, Westinghouse, and VVER. For each reactor type, one Loss of Coolant Accident and one Station Blackout scenario were selected for detailed analyses. Both scenarios cover a range of in-containment atmospheric conditions from potentially flammable at medium pressure to a steam-inertized atmosphere at high pressure, including the late phase with an active filtered containment venting system. This paper outlines the employed methodology using a consecutive analysis chain consisting of three levels with increasing detail (system codes, 3D GOTHIC™ and CFD) to assess containment pressurization, efficiency and/or options of individual mitigation measures with respect to H2/CO combustion risk and equipment and instrumentation survivability. As a common basis, the system code nodalization schemes and 3D models are developed from detailed CAD geometries. The paper summarizes the status of the work with a focus on the comparative assessment of the impact and effectiveness of mitigative measures (PARs, sprays, FCVS) on the combustion risk in the late phase. Concluding, challenges and lessons-learned are summarized.
Evaluation of Various Correlations for Direct-Contact Heat Transfer in Dispersed Flow Film Boiling Using DRACCAR Code
Following the core uncovery after a hypothetical loss of primary coolant accident (LOCA), a flow of steam with dispersed droplets forms downstream of the quench front. This flow is crucial for cooling nuclear fuel rods that are not in direct contact with water, representing a typical Dispersed Flow Film Boiling (DFFB) scenario. Depending on accident severity, fuel rod deformation may partially block subchannels, diverting steam towards less deformed areas and potentially impacting cooling. As a result, characterizing heat trans-fer under such conditions is essential to enhance the nuclear reactor safety during accidents like LOCA. The COLIBRI experimental loop was designed to assess the heat flux released by steam/droplet flow in intact and deformed fuel rod subchannels. Experimental results obtained with this facility helped evaluate various correlations for direct-contact heat transfer calculations in DFFB within the DRACCAR code. In this article, the DRACCAR code is used to compare different correlations for heat transfer due to droplet impacts onto the wall calculation. This comparison was made on four COLIBRI experiments, involving two different droplet mass flow rates (5 kg/h and 11.5 kg/h) and two levels of subchannel deformation (intact and 90% blocked).
On Flow Distribution and Heat Transfer in a Graphite Moderated Molten Salt Reactor
Graphite moderated nuclear reactor cores experience irradiation induced dimensional change as a function of graphite temperature and fast neutron fluence. In this paper, the impact of changes in channel geometry of a liquid fueled Molten Salt Reactor (MSR) is simulated with Monte-Carlo method based neutronics. Thermal-hydraulic phenomena are subsequently investigated with system code thermal hydraulics and Computational Fluid Dynamics (CFD). Specifically, gaps between graphite blocks introduced by irradiation-induced shrinkage are investigated, demonstrating the corresponding neutronic penalty on reactivity, and risk of semi-stagnant flow for a specific core lattice design. A comparison of CFD and 1D thermal hydraulics demonstrates the effect of flow entrainment between channels and gaps, and buoyancy assisted mixed convection in regions of laminar flow. Results indicate that the fuel temperature in gaps is approximately 25% higher than fuel in the adjacent flow channel for a representative 150 MWth design.
Interpretation from Sample Analysis Results Obtained in Fukushima Daiichi NPS to Accident Progressions -First Trial-
Particles attached on surface of samples collected from the unit 2 of Fukushima Daiichi Nuclear Power Station were analyzed with SEM and TEM in detail. From their composition, microstructure and crystal structure, the formation process of these particles were summarized. From these results, the authors extracted the conditions inside the reactor necessary for particle formation and tried to interpret the inside reactor status during the accident. The results suggested that Si was supplied into core region before fuel melt and that the hot space was formed in the core region in the early stages of the core meltdown. Regarding the status at the time when the bottom side of the reactor pressure vessel was damaged, it was suggested that the in-reactor condition were such that the particles were rapidly cooled from above 1700°C or higher under a steam-rich atmosphere.
Validation of NEPTUNE_CFD for Single and Two-Phase Natural Convection
Computational Fluid Dynamics (CFD) is widely used in nuclear engineering for safety related studies, providing a mechanistic solution approach in support of the system and sub-channel numerical scales applied in the safety demonstration. Co-developed by four of the biggest players of the French nuclear industry, the NEPTUNE CFD code is specialized in nuclear thermal-hydraulics applications allowing to simulate single and two-phase flows on complex geometries. The increasing protagonism of passive systems in new generation reactors demands specific studies regarding its operation and safety. Passive systems are engineered to rely on natural physical phenomena, such as buoyancy or natural convection and to operate without the need of active mechanical components or external power sources. This paper focuses on the validation of NEPTUNE CFD over experimental data of Qureshi and Gebhart [1] - vertical flat plane immersed in water - and semi-empiric correlations from the literature to simulate single-phase natural convection to validate the use of NEPTUNE CFD for simulating passive systems over a wide range of Rayleigh numbers from laminar to turbulent regimes.
Mechanism Study on Condensation Heat Transfer on Shell Side of Spiral Wound Heat Exchanger : Visualization of Film Behavior
Because of compact design and easy to modulization, the Spiral Wound Heat Exchanger (SWHX) is used as a steam generator and passive residual heat removal system (PRHRS) in light water Small Modular Reactors (SMRs) and nuclear ships. When used as a PRHRS, the shell side is mainly connected with the primary system. In an accident situation, two phase steam flows down through the shell-side and its cooled down by condensation phenomena. To design proper PRHRS, it is important to evaluate and predict the condensation performance. In this study, to analysis the condensation heat transfer mechanism on shell side of SWHX, heat transfer evaluation and visualization experiment are conducted. In order to reflect real SWHX to test-section, 6 coiled tubes are installed in one layer, and it are cut in 60° to mimic spacing bars which support coiled tubes to make stable structure. The experiment is conducted with pressurized steam. The steam is condensed at little super-heated condition and the condition covers velocity of 4 ~ 18 m/s and pressure of 3 bar. Coolant is injected to each coiled tube with same conditions. At the heat transfer experiment, heat transfer coefficients of each tube are measured. In addition, film behavior is observed by high-speed camera (PHANTOM) at frond and side of test-section. Visualization experiment is conducted with Nd:YAG laser and light source. From the experimental results, condensation heat transfer trend of each tube is discussed with film behavior.
GAMMA+ Simulation of Turbulent Heat Transfer Over a Backward-Facing Step
GAMMA+ (General Analyzer for Multi-component and Multi-dimensional Transient Application) code is a safety analysis code for nuclear system design and accident analysis. The GAMMA+ code has been developed for the simulation of both one-dimensional and multi-dimensional flows, the latter one being particularly important for components in advanced reactors such as molten salt liquid fuel reactors. In our previous work over the years, we have implemented the standard k-ε model into the GAMMA+ code for multi-dimensional turbulent flow. We have validated a series of cases under turbulent conditions such as flow in a pipe, flow in a plate, flow over a backward-facing step, and flow in a cavity. Most of those validations were performed with adiabatic conditions, and the heat transfer through the fluid was solved for the natural convection in cavities. However, the conjugated heat transfer with a wall has not yet been validated. In this work, we simulate the flow over a backward-facing step under turbulent conditions to validate the capabilities of the GAMMA+ code. A constant heat flux condition with the wall conduction is chosen for the validation. We compare the results of the GAMMA+ code with a commercial CFD code (CFX) and experimental data. By comparing the results of the GAMMA+ code with the CFX, we verify the standard k-ε model in the GAMMA+ code performs well. The validation results show that the GAMMA+ code predicts the overall trends of the experiment data well in terms of velocity distribution, turbulent intensity distribution, and fluid temperature distribution.
Experimental Study of the Effect of the Heating Profile on the Characteristics of the Geysering Instability
Two-phase flow instabilities can be observed in diverse two-phase flow systems such as nuclear power, cryogenic fuel pipelines for aerospace engines, thermosyphon in solar energy collectors and during the transport of two-phase flows in pipelines. Under low flow rate and heat flux conditions a particular type of instability can occur named as geysering instability. The geysering instability is a compound relaxation instability driven by condensation and the changes in the hydrostatics pressure. In this work, we investigate the effect of non-uniform heat distribution in the dynamics of geysering or condensation-induced instability. By fixing the total power in the heated section, the control of the heating profile can influence the vapor generation. We study the characteristic of the oscillation wave shape, amplitude and period, and reversal flow magnitude. The test section consists of 2m horizontal heated section followed by a 5 m vertical inverted u-tube section. The heated section is divided into 5 independently controlled regions which allow to control the heating profile. The effect of the heat profile, peak value, and peak value location are discussed. Increasing the heat flux in the last segment of the test section, while maintaining the total power fixed, leads to the suppression of the instability. Contrary, increasing the heat flux in the first segment of the test section leads to a more pronounce oscillation.
Experimental Investigation of Erosion and Corrosion of Coated and Uncoated Steels Under High Velocity Lead Flow
The choice of molten lead as a chemically inert and low-pressure coolant for a nuclear reactor can provide enhanced safety features. However, development is still necessary in areas such as materials performance and corrosion control. In this paper, we investigate the erosion/corrosion of coated and uncoated 316L and 316H stainless steels under high velocity lead flow using the newly commissioned BULLET (Bangor University Lead Loop Erosion/corrosion Testing) facility. In this loop, lead flow at a rate of 3.3 ms-1 was established across the sample surface for a period of 500 hours, at 693K and an oxygen content of 10-8 wt%. Post-test materials characterisation of the samples using Scanning Electron Microscopy with Energy Dispersive X-ray Spectroscopy (SEM-EDS) is presented. In particular, the effect of the coating on the erosion/corrosion of the candidate materials is investigated.
Effect of Heater Sizes on Critical Heat Flux and Enhancement using Discretized Heating Elements with Constant Temperature Wall Condition
Considering the limitations of previous studies on heater size effects on CHF and the actual geometry of conventional chips, the rectangular-type heaters are fabricated and experimentally investigated. To exclude the effect of pressure, sub-cooling, and heater orientation, the experiments are conducted as saturated boiling under atmospheric pressure conditions using upward facing heaters. Seven different surfaces by changing the geometries of micro-pillar structures are fabricated and tested to further enhance CHF. In each surface, four different heater configurations are fabricated by changing both heating width (the shorter one of two side lengths) and length (the longer one). Among the experimental cases, maximum CHF of 3043 kW/m2 (Surface: D40G20H20+SiO2, 4X15 mm2) and minimum CHF of 696 kW/m2 (Smooth SiO2, 16X40 mm2) are experimentally confirmed. Among 7 groups, group 2 showed a minimum CHF difference of 232 kW/m2, and group 7 showed a maximum CHF difference of 1316 kW/m2 for give heater configurations. For surfaces with poor wettability, width shows dominant effect on CHF whereas samples with good wettability showed similar dependency of both heating length and width. Besides, we find different behavior of post-CHF condition for samples with small effective heater size. They maintain nucleate boiling phenomena for certain period without rupture and even go back to temperature before CHF in some case. Based on this, we thought the heating elements that can maintain constant temperature could enhance CHF further. We fabricated discretized heating elements (DHE) to induce constant temperature wall condition. Using DHE, we confirmed CHF enhancement of 16%.
Large-Scale Ex-Vessel Corium Spreading Calculation Using a Particle Method for Estimating Fuel Debris Distribution in Fukushima Daiichi Unit 1: Effect of Cooling Efficiency on Spreading
Estimating fuel debris distribution is one of the indispensable approaches for retrieving fuel debris spread within the Primary Containment Vessel (PCV) in Fukushima Daiichi Nuclear Power Plant Unit 1 (1F1). Recent findings from an internal investigation conducted by Tokyo Electric Power Company Holdings (TEPCO) suggest the possibility that molten corium has spread over a vast region of the PCV due to significant Ex-Vessel Corium Spreading. This study utilized a large-scale Ex-Vessel corium spreading simulation, employing a method that couples Moving Particle Hydrodynamics (MPH) with Large Eddy Simulation (LES). Since reliable simulation requires known boundary conditions, some information was referred from Unit 3. Furthermore, sensitivity analyses were performed concerning the dynamic viscosity of corium ranging from 1.0×100 to 1.0×10-3 (Pa.s), the heat transfer coefficient of vapor surrounding the pedestal, and the magnitude of decay heat, respectively. Calculation results indicated that the magnitude of cooling efficiency, consisting of vapor cooling and heat loss by radiation heat transfer, plays a significant role in determining the termination of spreading due to the formation of a solid crust at the free surface. In contrast, the encapsulated corium below the free surface maintained its liquid phase due to decay heat even after spreading termination. Calculation results were compared to TEPCO findings regarding thickness distribution and thermal neutron flux distribution, providing rational possibilities for fuel debris distribution in 1F1.
Understanding MELCOR CVH/FL Governing Equations for Future AI Applications
Severe accidents (SA) in nuclear systems have been studied with reliable system codes such as MELCOR and MAAP. These codes enable the investigation of physical phenomena during SAs by numerically solving thermal-hydraulic governing constitutive equations in a configured geometry with detailed design information. Despite the utility, these system codes have challenges. Configuring the input geometry for packages such as control volume hydrodynamic (CVH), flow path (FL), and cavity (CAV) takes significant amount of time. In order to address the challenge, machine learning (ML) techniques, specifically physics-informed neural networks (PINNs), have been regarded as a possible methodology. PINNs enables physics interpretation by adopting physics equation as a loss function and achieves meshless calculation by incorporating auto-differentiation along with loss function. Due to distinguished characteristics, PINNs is expected to bring simplification to the input for system code, along with the capability of multi-physics analysis for including multiple thermal-hydraulic fields of database. With regards to PINN study, comprehending the governing equation with the logic behind is crucial. In advance to applying PINN to MELCOR, this study aims to assess the feasibility of developing a system code which computes the momentum and the mass of water inside the tank by obtaining the governing equation from MELCOR CVH/FL. Through Python, the forementioned MELCOR governing equation of such flow behavior was assessed by embodying semi-implicit method of calculation.. The accuracy of the developed model was verified by comparison with the MELCOR code in a case study of gravity-driven water tank flow. In conclusion, further application of the governing equation is expected in the future.
Multi-Scale Coupled Simulation of E-SCAPE at Isothermal Conditions
Liquid metal-cooled reactors are a Generation IV reactor type that has the advantage of high power density owing to the high thermal conductivity of metals. The thermo-hydraulic phenomena that occur in the reactor pool, such as flow patterns, thermal diffusion, and stratification in different scenarios (steady operation, accidents, non-critical transients, etc.) have been and still are a topic of great interest in various research projects. One such reactor is MYRRHA, a flexible fast-spectrum research reactor cooled by lead-bismuth eutectic (LBE) in a pool configuration under design in Belgium. In support of its development, the European Scaled Pool Experiment (E-SCAPE), a 1/6th scale model of the same, was developed within the collaborative European framework of the THINS and MYRTE projects. This LBE mock-up, constructed and operational at SCK CEN in Belgium, serves to provide relevant data that gives insight into the thermo-hydraulic phenomena and provides the possibility to carry out numerical code validation and support the overall design of MYRRHA. In the framework of the collaborative European project PASCAL, NRG aims to perform multi-scale simulations of asymmetric accident scenarios (i.e. Single Loop Heat Exchanger Failure and Single Pump Failure) of E-SCAPE by coupling the in-house System Thermal Hydraulic (STH) code SPECTRA to the commercial Computational Fluid Dynamics (CFD) code STAR-CCM+ via the in-house coupling tool myMUSCLE: MultiphYsics MUltiscale Simulation CoupLing Environment. In a previous article, the construction of the standalone models of SPECTRA and STAR-CCM+ was reported. The simulation results were compared to a steady state isothermal E-SCAPE experiment at a mass flow rate of 93.2 kg/s and a temperature of 180 °C, showing good agreement of both models with each other and the experiment. In this article, the next step is taken by performing coupled calculations of the same isothermal case and comparing to the standalone and experimental results. The calculations reveal stable solutions that are well in agreement with the standalone results.
Computational Fluid Dynamics Analysis of the Core of the Motel Facility
This paper presents the computational fluid dynamics analysis of the core of the Modular Test Facility (MOTEL), an integral test facility recently commissioned in the nuclear safety research laboratory of LUT University. The MOTEL is a model of an SMR that resembles the design of the NuScale type SMR. The current implementation of MOTEL represents a prototypical integral pressurized water small modular reactor (SMR) operating with natural circulation. The facility can operate at a pressure of up to 40 bar and a temperature of 250 °C. It has a total height of 7.4 m and a diameter of 711 mm, with a maximum heating power of 1000 kW. The MOTEL core comprises 132 electrically heated heater rods, 145 dummy rods, and 16 measurement probes. The diameter of the core is relatively large. Only one-fourth of the MOTEL's core geometry was simulated for this study using the ANSYS FLUENT CFD tool. Various core power distributions were analyzed. Flow turbulence was solved by employing Shear Stress Transport (SST) k-ω turbulence models. Results show that CFD can help evaluate phenomena that are difficult to reproduce or costly to analyze experimentally at a lower cost and with reasonable accuracy, making it a key advantage in the nuclear industry.
Flow-Induced Vibrations in Nuclear Steam Generators: Simulation of the AMOVI Experiment with Different FSI Approaches
The interaction between cooling fluid and solid structures (rods, tubes) in nuclear power plants may lead to flow-induced vibrations (FIV), causing material fatigue, fretting wear, and eventually loss of component integrity. This can trigger further safety issues as well as substantial standstill costs due to longer or unplanned outages. Within the European GO-VIKING (Gathering expertise On Vibration ImpaKt In Nuclear power Generation) project, experimental and numerical efforts are performed to improve the understanding of FIV in reactor cores and steam generators, as well as to develop and validate advanced numerical approaches for their prediction. In the French CEA’s AMOVI experiment, FIV in different tube configurations, exposed to a cross-flow, are investigated. GRS performed first analyses on AMOVI using three FSI methods: the full-order model ANSYS CFX – Mechanical, the reduced-order model ANSYS CFX-MOR, and the ANSYS CFX Rigid Body approach. In addition, results generated with OpenFOAM using a model similar to the Rigid Body approach are described. The tube vibration parameters (displacements in time and frequency domains), calculated with all these methods, are discussed and then compared with experimental data. Further, conclusions are made and steps to improve the presented results are identified.
Application of Risk Informed Safety Margin Characterization to the Analysis of a Pressurized Water Reactor
The sequence Core Damage Frequency (CDF) of a pressurized water reactor has been quantified in three initiating events that are Medium Break Loss of Coolant Accident (MBLOCA), Small Break Loss of Coolant Accident (SBLOCA) and Steam Generator Tube Rupture (SGTR), following a realistic methodology called risk informed safety margin characteristic (RISMC). The surrogate plant analyzed in the study is a typical pressurized water reactor. The plant adopted two Westinghouse Three-Loop Pressurized Water Reactors with rated thermal power of 2,830 MWt. The phenomenon identification and ranking table (PIRT) is applied for uncertainty analysis. The mitigation actions as described in plant specific Probabilistic Safety Assessment (PSA) include cooldown and depressurization, emergency cooldown and depressurization, high head safety injection (HHSI), high head safety recirculation (HHSR), low head safety recirculation (LHSR) and Refueling Water Storage Tank (RWST) replenishment. These mitigation actions are analyzed by thermal hydraulic system analysis code RELAP5-3D to determine the successfulness of the actions. The uncertainties of input parameters of the plant conditions are included, and the time of mitigation action executed is treated as one of the input uncertainties. The results of realistic methodology show a decrease in CDF for all three analyzed events in comparison with conventional methodology. The differences between three initiating events are also discussed.
Initial Results of Scaled MHTGR Lower Plenum CFD Analysis
The High Temperature Test Facility (HTTF) at Oregon State University (OSU) is a 1:64 volume, 1:1 temperature scaled experimental facility used to model the General Atomics Modular High Temperature Gas Reactor (GA-MHTGR) design during steady state and accident transient scenarios. One item of particular concern during steady state operation for this design is the potential existence of “hot-spots” in the lower plenum which could, over long-term operation, pose a risk to the integrity of the lower plenum support posts. Using Star-CCM+ for CFD modeling, this paper will investigate the velocity and temperature profiles within the lower plenum of the HTTF facility using fixed boundary conditions based on full power steady state test conditions. The objective of this modeling is to identify potential areas of hot-spot formation. This is part of the ongoing Thermal hydraulic code validation benchmark for high temperature gas-cooled reactors using HTTF data, a Nuclear Energy Agency (NEA) Verification and Validation (V&V) Benchmark project using high quality experimental data from the HTTF experiments at OSU to quantify accuracy and uncertainty in thermal hydraulics calculations using CFD and systems codes; later references to “The Benchmark” in this paper shall refer to this project.
Thermal-Hydraulic Simulations of Turbulent Flows using the Immersed Boundary Method for Innovative Nuclear Reactor Safety Devices
In the framework of safety-system developments for new generations of nuclear power plants, we model turbulent flows inside industrial passive-safety devices, based on the fluidic diode principle, for design optimization. Considering a statistical processing or an expert advice, this geometrical optimization would require a specific meshing of each evaluated shape. Facing a large number of calculations and to avoid the tedious re-meshing task, we use the fictitious domain approach allowing us to dissociate the computational domain from the geometrical one. Using in-house industrial software, we are not able to modify the equations governing the physical problem in the code. Hence, we choose the Penalized Direct Forcing (PDF) method. It consists of a penalized forcing term, written in a direct forcing formulation, added to the governing equations to take account immersed boundaries. Considering industrial turbulent flow problems, calculated by the RANS approach and unresolved boundary layers, we need an appropriate treatment of immersed walls. An immersed wall model, based on the power law, is chosen because of its ability to model the velocity profile within the entire boundary layer, to deal with non-uniform distributions of distances to the immersed wall and to be explicit. Consequently, we develop a finite element version of the power law approach and its variants in our PDF method, validate this approach on analytical test cases and compute industrial cases. In particular, we highlight the need of turbulent viscosity control near the immersed wall and the impact of the power law variants on the mechanical constraint estimation on the immersed wall.
Analysis of the Metal Containment Vessel Wall Thickness Effect on the Coolability of the Flooding Safety System using the MELCOR Code
To remove decay heat during an accident, flooding safety system (FSS), a passive safety system (PSS), was designed for small modular reactors (SMRs) employing a metal containment vessel (MCV). The FSS provides emergency coolant and submerges the reactor modules (RMs) installed in the reactor module cavity during an early phase of the accident. The sensible heat from primary coolant during the blowdown phase and decay heat from the core during the boiloff phase are removed by the heat capacity of the MCV wall before the sufficient flooding. Thus, the thickness of the MCV wall was expected to affect accident sequences. Thus, in this study, the effects of the wall thickness of the MCV on the decay heat removal and accident sequences were investigated through numerical analysis using MELCOR code. A natural circulation SMR whose thermal power is 330 MWt and a loss of coolant accident were selected as the reference reactor and the postulated accident. The effect of the variables such as wall thickness and delay time for cavity flooding were evaluated. The thickness and the delay time ranged from 5 cm to 10 cm and from 0 minute to 75 minutes, respectively. As the results, the thicker MCV wall absorbed more heat during time before flooding, but the MCV pressure after flooding was higher due to the higher MCV inner surface temperature. Moreover, the thicker MCV expedited the achievement of recirculation due to the smaller pressure difference between the reactor pressure vessel (RPV) and the MCV.
Critical Review of the Film Flow Model by Okawa et al.: Analysis and Improvement Potentials
The film flow model by Okawa et al. is critically investigated. The basis of the study is a database of dryout experiments from the literature, which contains 4460 data points from 21 different sources, covering all influence parameters over a wide parameter range and comprising water as well as non-water fluids. Even though the model is well established and works generally well, three assumptions may be improved. Firstly, the film velocity should be considered in the interfacial shear force for the derivation of the film thickness. Secondly, a linear dependence between the shear entrainment rate and the film thickness is assumed, which cannot be supported by the physically-based Kelvin-Helmholtz approach. And thirdly, an equilibrium between the shear entrainment rate and the deposition rate is assumed for the calculation of the initial entrained fraction at the onset of annular flow. Using an alternative approach for the calculation of the initial entrained fraction, it can be shown that this assumption is not justified. Employing the proposed approaches leads to a more precise description of the conditions at the onset of annular flow, which is the starting point of the integration of the film mass balance in the film flow model, and therefore offers the potential for more accurate predictions of dryout.
Experimental Study on Pool Boiling of Aqueous Surfactant Solutions on Micro-Structured Surfaces for Boiling Enhancement
Surfactants have been widely used in various industrial fields. Simply adding a small amount of surfactant can greatly enhance the boiling heat transfer of water, which can improve the performance of heat exchanger and has great application prospects in industry, but its mechanism is not fully understood. Many possible factors have been proposed, but their relative importance remains to be examined. In this paper, the experimental study found that when the surfactant concentration was low, the change in boiling performance was mainly due to the non-coalescence of growing bubbles on the wall rather than the change of physical properties of the solution. The underlying reason is that surfactants alter the mobility of the vapor-water interface. At high heat flux, the non-coalescence bubbles form wet foam, making it impossible for bulk water to pass through to wet the heating surface, thereby reducing CHF.
Experimental Investigation of Flow Boiling Regimes in a Horizontal Heated Pipe
Many industrial applications rely on heat transfer processes when a fluid is boiled. Research on boiling and flow boiling has been of paramount interest since the 40s but despite this fact, the physical mechanisms controlling the heat transfer process remain under scrutiny. In the case of flow boiling inside pipes, three regimes have been distinguished, namely convective flow boiling, nucleate flow boiling and a liquid deficiency boiling regime. At low heat fluxes and low working pressure, convective flow boiling is dominant, and the heat transfer coefficient depends on the mass flux and the thermodynamic quality. At high heat fluxes and high pressures, nucleate boiling is dominant, and bubbles produced at the wall are attributed the control of the heat transfer. In addition, experimental studies have shown that increasing the heat flux during flow boiling can led to a limiting value of the heat transfer coefficient before the boiling crisis. These three regimes have been the focus of diverse experimental studies. In this work, we focus on experimentally isolating working conditions where only one dominant mechanism play a dominant role. This work provides a new experimental framework for studying each heat transfer regime individually.
Development of a Flow Boiling Heat Transfer Correlation for a Helically Coiled Tube
Helically coiled tubes have been widely used in heat exchangers across various industries. Particularly in the nuclear sector, many small modular reactor (SMR) designs adopt helical coil heat steam generators. Heat transfer in a helically-coiled tube has a unique feature due to the centrifugal force generated by the fluid motion. This in known to enhance the boiling heat transfer inside the tube. However, accurately predicting the heat transfer performance of helically coiled tubes remains a significant challenge. This is primarily attributed to the complex boiling mechanisms influenced by centrifugal forces. Existing research on boiling heat transfer correlations has largely been based on experimental data from vertically oriented straight tubes. In this study, we investigated the effect of centrifugal force on boiling heat transfer in a helically coiled tube; at first we devised a new dimensionless number defined by centrifugal force over gravity and, then, demonstrated its effect on boiling heat transfer. Using the findings, we incorporated the dimensionless centrifugal force number into a new boiling heat transfer correlation, of which the basic structure was adopted from the Kandlikar correlation for saturated flow boiling. We showed the enhanced accuracy of the new correlation compared to existing ones and its applicability to boiling heat transfer within a straight tube as well.
IAEA Benchmark CRP - I31038: Open Phase Liquid-Metal CFD Results for Fuel Pin Simulator
In the IAEA Coordinated Research Project “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop” (CRP - I31038) participants benchmark their simulation capabilities versus the experimental data obtained in the NACIE-Up facility (NAtural CIrculation Experiment-UPgrade) at ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development). The fuel pin simulator consists of 19 wire-wrapped pins. The liquid metal coolant first passes through an unheated cold section and then through an electrically heated section. In the heated section either all pins or only a fraction of the pins are heated to simulate power gradients in a nuclear reactor. During the experiments both the axial and the lateral temperatures are measured at suitable locations. Twelve results from different organizations contributed to the open phase benchmarking, where specifically the heating of all 19 pins, 9 pins and the 7 inner pins was studied. This will contribute to the validation and verification of the computational fluid dynamic tools for reactor design. We have tried to deliver results that require minimal numerical resources, well-resolve all relevant geometric features and match the accuracy of the provided experimental data. In the second benchmark meeting at IAEA the results of the contributors were compared. It was confirmed that our modelling strategy leads to credible and accurate results at low numerical cost. In the paper we present details on the modelling approach and compare the numerical results to experimental data and preliminary CFD data resulting from all participants in the work package. Our simulation is ranked among the most reliable results.
Experimental Investigation of Two-Phase Flow Instability in HTR-10 Steam Generator
For ultra-high temperature operation of the 10 MW High-Temperature Gas-cooled test Reactor (HTR-10), the operation parameter of its steam generator will deviate from the original design. Two-phase flow instability may occur. To investigate this phenomenon, experiments were carried out using two full-scale helical tubes. A segmented electrical heating approach was employed for these two tubes. The stability boundary was searched by adjusting inlet resistance coefficient or heating power. This study examined the stability of the HTR-10 OTSG at low power levels and explored the influences of operational parameters. Out-of-phase pulsation type density wave oscillation was observed during the stability experiments. The stability of HTR-10 OTSG was verified at 10% power level under non-uniform heating (close to convection heating). The systems with high inlet resistance coefficients became stable as increasing power level. Consequently, HTR-10 OTSG can operate stably during ultra-high temperature operation conditions. For the influence of heating power distribution, the stability under uniform heating was more conservative compared to non-uniform heating. The influences of inlet resistance coefficient, heating power, system pressure and inlet temperature on the stability were investigated under uniform heating. HTR-10 OTSG became stable when increasing inlet resistance coefficient or decreasing heating power. Within system pressure range of 1.5 MPa to 3.5 MPa, increasing system pressure made the systems more stable. Flow instability may occur easier with an increase in inlet temperature in the range of 50℃ to 210℃.
The Conceptual Design of Integrated Emergency Operating Procedures with Hierarchical Structure
The nuclear power plant is a system that demands a high level of safety. In such safety-critical systems, human error consistently emerges as a key issue, particularly in emergency operating situations where operator mistakes can lead to severe accidents. Commercial nuclear power plants are equipped with various procedures to support operators. In emergency situations, operators use emergency operating procedures to diagnose and mitigate accidents by manipulating components and systems. Traditionally, paper-based procedures were predominantly used, but with advancements in instrumentation and control technology, digital main control room facilities have been introduced, leading to the adoption of computerized procedures. Despite technological progress in instrumentation and control, the structure of procedures still retains many characteristics of traditional paper-based procedures. As instrumentation and control technology evolves, human-system interfaces have improved, but the procedure structure continues to rely heavily on traditional paper-based procedure structure. Operators compare information from procedures with plant data to perform their tasks, making operator training and experience crucial factors. If the system are to advance further, monitoring the plant status and selectively providing information, it is anticipated that more optimized actions could be offered. In this study, we aim to develop a system by analyzing and reconstructing existing procedures to provide an integrated form of procedural structure. The integrated procedure system offers hierarchical information on main objectives, safety function states, relevant systems, and operation constraints, allowing operators to collect information at a glance. This paper proposes a conceptual design for a next-generation procedural system that supports operators, using hierarchal structure and procedure integration.
Development of Reactor Vessel Thermal–Hydraulic Analysis Method in Natural Circulation Conditions - Investigation of Interwrapper Gap Model -
A dipped-type direct heat exchanger (D-DHX) installed in the hot pool of the reactor vessel (RV) has been investigated as one of the decay heat removal systems under natural circulation to enhance a safety of pool-type sodium-cooled fast reactors. The D-DHX provides the cold sodium to the hot pool. The cold coolant from the D-DHX flows into assemblies and an interwrapper gap (IWG) between them in the core, and removes the heat from assemblies. Thus, the behavior of cold sodium is a large scale in the whole RV. It is required to predict a complex three-dimensional thermal–hydraulic behavior in the RV by detailed modeling. For the design study, a computational fluid dynamics code needs to have an acceptable accuracy and low computational cost performance. In this study, we investigate the coarse mesh modeling of IWG using correlations for the purpose of the development of a practical model which can reduce the computational cost maintaining the prediction accuracy. An influence of combinations of the coarse mesh and the correlation for pressure loss in the IWG on the thermal–hydraulics in the RV and the core temperature distribution is revealed through the numerical analysis of a sodium experimental apparatus PLANDTL-1. As a result, the coarse mesh modeling of IWG with the pressure loss correlation evaluated the thermal–hydraulics in the RV. Temperature distribution in the core generally correlated between numerical and experimental results. To improve the evaluation accuracy of the temperature distribution in the simulated core, the further study about numerical simulation with detail mesh or alternative pressure loss correlation in the IWG region of the top of simulated core where the cold sodium enters will be needed.
Source Term Assessment and Preparation of Countermeasures in Fukushima Daiichi NPS Unit-1 in Case of Further Degradation of Support Function of Pedestal Caused by Earthquakes
Partial loss of concrete in the inner lower part of the pedestal of Fukushima Daiichi NPS Unit-1 was confirmed during the primary containment vessel (PCV) internal investigation made from March 2022 to April 2023. In previous studies of the pedestal base seismic robustness, it was concluded that the possibility of pedestal collapse as the result of earthquakes was low, and further degradation of support function to the superstructure would be limited. It is necessary to address such an event, however, and source term assessment was carried out and sufficient countermeasures were prepared in order to reasonably mitigate potential dose exposure. The source term was assessed with reference to the five-component linear equation in DOE HDBK-3010-94. Amounts of radioactive materials were assumed based on results from severe accident analysis codes (MAAP/ SAMPSON), muon tomography, and an actual contaminant sample collected during previous PCV internal investigations; and based on findings obtained during construction work, done using an abrasive water jet (AWJ), for the route on which an investigation machine was to approach the PCV. Airborne release of the radioactive aerosols was considered based on information from construction work using the AWJ and vibration experiments. The source term assessment concluded no significant dose exposure would occur. Although low risk of dose exposure was confirmed, uncertainties of the assessment would exist. Thus, countermeasures were discussed. Nitrogen gas is normally injected into PCV in order to have an inert atmosphere in it. If radioactive aerosols are generated in the PCV, the radioactive aerosols might be released from the PCV via the nitrogen gas. As the result of comparing other risks, it was decided that the injection of nitrogen gas should be temporally stopped when large earthquakes occur.
Real-Time Modeling of Gas-Liquid Two-Phase Flow through Machine Learning and Optimization Based on Experimental Data
This work aims to enhance the gas-liquid two-phase flow model to improve the accuracy of predicting fission product removal via pool scrubbing, in case of a severe accident at a nuclear power plant. The complexity and computational cost associated with calculating detailed flow parameters under pool scrubbing conditions pose challenges. To address this, development of a real-time simulator was considered, to account for the need for fast calculations to facilitate model refinements. Gas-liquid two-phase flow calculations were performed based on Navier-Stokes equations in particle simulations. The resulting database of particle simulations were used to train a regression forest model. Sensitivity analysis identified key parameters in the trained model, and they were optimized through multi-parametric Bayesian optimization by comparing statistical flow parameters with the experimental data. With such data assimilation, the regression model was refined to output more realistic simulation results. This study highlights the potential of the use of machine learning in the areas of realistic real-time analysis, offering a promising method for improving the prediction capabilities of gas-liquid two-phase flow.
Comparison of System Thermal Hydraulic and CFD Analyses of NACIE-UP Asymmetric Test
Liquid Metal Fast Reactors (LMFRs) are expected to play an important role in the future of nuclear energy as they offer higher fuel efficiency, high power density and excellent safety characteristics due to the optimal heat transfer performance of the metallic coolants. The NACIE-UP loop facility, installed at ENEA Brasimone R.C. in Italy, is specifically designed to perform experimental campaigns for natural circulation testing; in particular to investigate the shift from forced to natural circulation in a test section with lead-bismuth eutectic (LBE). In 2017, ENEA conducted transient tests which provided an excellent opportunity to validate both integral values and local temperature distribution, allowing code validation by comparisons of reactor system codes and CFD simulations with experimental data. In the framework of a benchmark organised by an IAEA Coordinated Research Project, NRG will perform multi-scale simulations of asymmetric transient tests by coupling the NRG’s system code SPECTRA to a commercial CFD code. To this end, this paper presents the comparison results obtained by SPECTRA and CFD for two steady-state conditions, ADP10 and ADP06. Both codes demonstrate reasonably good agreement in predicting the wall and bulk temperatures of the fuel pin simulator (FPS). The CFD model can predict the detailed temperature variations across three measurement planes in the FPS, especially evident in the non-uniform power distribution test, ADP06. The 3D effects introduced by the wires can be addressed accurately in the CFD regarding the pressure drop in the FPS. Comparative analyses between numerical and experimental results demonstrated a good agreement, providing positive feedback on the feasibility of coupled simulation to enable a smooth and stable coupling.
Long Straight Closed Two-Phase Thermosiphons - Experiments of IKE and Validation of ATHLET Models
In recent years, there has been a global focus on passive residual heat removal using thermosiphons or heat pipes for spent fuel storage pool and wet storage cooling. In the event of a station blackout with loss of active cooling modes, a passive heat removal system based on two-phase closed thermosiphons (TPCT) can effectively maintain the cooling of spent fuel pools due to their high heat transport capacity and simple design. GRS develops and validates computer codes like AC2 to simulate all relevant nuclear power plant phenom-ena during normal operation, incidents, accidents, and severe accidents. In 2016, the University of Stuttgart and GRS initiated an R&D cooperation to enhance the thermal-hydraulic code system ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), enabling it to model thermosiphon and heat pipe operations. After addressing key model gaps, such as properties of water/steam in near-vacuum conditions and intro-ducing new heat transfer correlations for boiling and condensation under such conditions, in 2020, the re-search focus expanded to study complex thermosiphon configurations. Experiments in four different TPCT test facilities with one 3/10-meter-long straight TPCTs, with one/nine straight/bended TPCTs were and are performed. For two selected experiments (3-meter-long straight TPCT and nine 10-meter-long straight TPCTs) post-test calculations with improved German ATHLET program (as part of coupled code system AC2 of the main modules ATHLET, ATHLET-CD and COCOSYS of GRS) are presented and the results are compared to experimental measurements. The correlations taken from literature fit the general operation behaviour quite well but doesn’t fit for 3-meter-long to 10-meter-long TPCTs in all details. Therefore further model improvements based on semi-empirical correlations of the new experiments will be done.
Validation of the System Code CATHARE3 on Critical Flow Experiments in the Framework of the OECD-NEA Atrium Project
When applying the Best Estimate Plus Uncertainty (BEPU) methodology for the safety analyses of nuclear reactors, one of the major issues is to quantify the uncertainties associated to the physical models in thermal-hydraulic codes. A good practice guideline for Inverse Uncertainty Quantification (IUQ) was therefore developed during the OECD-NEA SAPIUM (Development of a Systematic APproach for Input Uncertainty quantification of the physical Models in thermal-hydraulic codes) project in 2020. A first application of this guideline is now carried out within the OECD-NEA ATRIUM (Application Tests for Realization of Inverse Uncertainty quantification and validation Methodologies in thermal-hydraulics) project, which was launched in 2022. The goal is to perform practical IUQ benchmark exercises to evaluate the applicability of the SAPIUM best-practices and suggest possible improvements. In this article, we describe part of the work performed at CEA on the first benchmark exercise on critical flow. In particular, we focus on the validation of the system code CATHARE3 against the available experimental data and the associated sensitivity analyses performed to better understand the simulation results and prepare the IUQ process. The 324 chocked flow experiments come from three different facilities: Sozzi-Sutherland, Super Moby-Dick and Marviken-CFT. The simulations are in very good agreement with the experimental data (maximum discrepancy of 23.3% on the critical flowrate). Based on the sensitivity analyses, two main influential parameters are identified: the wall-to-liquid friction and the flashing models in CATHARE3. The flashing is dominant for relatively short nozzles (L/D ≤ 18). For longer nozzles, the wall-to-liquid friction becomes more and more influential.
Advection-Diffusion of Scalars with the Incompressible Schrödinger Flow
The description of fluid motion is typically carried out either considering continuum matter or collision models (Lattice Boltzmann methods), with the former being the most common: in this framework, differential balance equations are derived from conservation principles and are solved to obtain density, velocity, pressure and temperature fields. If viscous forces are neglected, the fluid behaves as inviscid, and it obeys the Euler equations: their direct numerical simulation is challenging because of their mathematical structure, which imposes tight constraints on the mesh size and the temporal discretization. Thus, powerful computers are needed to deal with the long computational times of direct numerical simulations of fluids, and still, computational times are often unsuitable for multi-query scenarios and parameter sensitivity studies. A third approach to dealing with inviscid fluids was born in the 1920s when Madelung proposed its interpretation of quantum mechanics based on an analogy between the Schrödinger equation and the compressible Euler equations obtained by adopting a suitable coordinate transformation. This framework has been recently analysed for simulating incompressible inviscid flows, providing a novel methodology for solving the Euler equations, known as Incompressible Schrödinger Flow. Its different mathematical formulation allows a less tight numerical discretization; investigating this approach can give a new perspective on the description of fluids. This work focused on the application of this method in the advection-diffusion of scalars (e.g., temperature or concentration) in inviscid incompressible flows, adopting as velocity that computed by the Incompressible Schrödinger Flow; furthermore, it will be investigating the possibility of inserting into the “quantum” counterpart equation a potential term related to gravity and buoyancy forces.
Assessment of the Rankine Vortex Model for Modeling the Gas Entrainment in a SFR Reactor
Since decades, the CEA has been involved in the development of 4th generation reactors cooled by sodium. The dedicated experimental platform PLATEAU was erected to study the different issues and built a results database for the code validation. As experiments with sodium are complex, the tests are led in mock-ups using water as a simulant fluid. The MICAS mock-up, model of the upper plenum, aims at studying the gas entrainment at the free surface. Gas is prohibited in the reactor core due to neutron effects, which may induce reactivity variations. Since vortices are difficult to handle by numerical codes in complex geometries such as reactor ones, criteria have been developed based on the Rankine vortex model to forecast the air entrainment. However, in some case, numerical results do not correlate the experimental ones and those discrepancies are may be due to the vortex model. This article aims at comparing the vortices characteristics calculated using the Rankine vortex model with experimental results in order to improve the gas entrainment criteria. The parameters of the Rankine model are determined using the velocity fields measured by the Particle Image Velocity (PIV) method and the vortices diameter is calculated using the raw PIV images.
Study on Eutectic Melting Behavior of Control Rod Materials in Severe Accidents of Sodium-Cooled Fast Reactors (4) Analyzing Eutectic Melting and Relocation Dynamics in B4C-Stainless Steel Using the Moving Particle Semi-Implicit (MPS) Method
In severe nuclear accidents, eutectic reactions cause stainless steel (SS) cladding and boron carbide (B4C) to melt early, leading to control rod failure and eutectic melt relocation. To simulate this, a modified moving particle semi-implicit (MPS) method is used for 2D and 3D simulations of a real control rod B4C-SS specimen, considering mass diffusion and eutectic reaction criteria based on Fe-B phase diagram. The eutectic reaction model implemented in the MPS code was validated against our prior experimental data on B4C-SS. The objective of the study is to numerically measure the boron concentration within the solidified eutectic melt. The results show that once eutectic reaction occurs, the boron diffuses into the SS wall. Melting initiates at the B4C and SS interface, with melt flow following SS cladding penetration. As temperature increases, boron concentration within the melt proportionally rises due to enhanced unidirectional diffusion into SS cladding. However, within the SS layers at a specific temperature, boron concentration decreases away from the interface due to concentration gradients. The current model accurately replicated the eutectic melt's geometry, aligning closely with experimental observations. It also reflected a nearly 13% rise in boron's atomic composition within the eutectic melt, similar to the approximately 15% increase noted in our experiments. The updated MPS method indicates a computational capability of the eutectic reaction model used to effectively analyze control rod eutectic reactions, simulating severe accidents, and their subsequent relocation to understand the effect of B4C ingress into the core.
New Gap Conductance Model for the ATHLET System Code
The ATHLET system code, which is part of the AC2 software package, is developed by the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) to analyze the behavior of nuclear reactors, ranging from normal operating conditions and design basis accidents to design extension conditions including severe accidents with core melt. Recent developments by GRS are aimed at enhancing the predictive capabilities of the existing fuel rod model of ATHLET. The existent simplified gap conductance model of the system code had been observed to overestimate nuclear rod temperatures, particularly under conditions of high burnup, when an unsuitable initial guess for the gap size is supplied. To address this concern, a more sophisticated gap conductance model has been considered and implemented. This new model dynamically evaluates gap sizes, incorporating aspects such as clad elastic mechanical deformations, as well as burnup-dependent fuel deformations, including swelling, densification, and radial fuel relocation after cracking. This enhanced model has been successfully integrated into the recent code version ATHLET 3.4. The present research gives an overview of the new models considered and implemented in ATHLET. Additionally, it summarizes the verification and validation activities that underpin the code's improved predictiveness for the fuel temperature distribution within nuclear rods. Finally, we discuss further developments planned for the near future for fuel behavior models within ATHLET.
Steam Condensation in Mini-Channels: A Benchmark of Available Correlations
This study takes place in the framework of the development of innovative heat exchangers. We consider a heat exchanger with vertical mini-channels (hydraulic diameter <3mm) operating in forced circulation, with pure steam at the inlet of the primary side. In such conditions, gravity and surface tension drive the condensation mechanism, but the latter effect is generally neglected in the modeling of the heat transfer coefficient. Indeed, condensation at mini-channel scale exhibits specific flow regimes that require dedicated modeling. Considerable effort has been done in the last two decades in order to model friction coefficients and heat transfer coefficients. Nevertheless, all the experimental studies took place with refrigerant fluids, whose properties significantly differ from those of steam. Therefore, transposition of previous work is delicate. A 1-dimensional model is used to investigate the effect of the closure laws on condensation and the results are compared with the results of the system-scale code CATHARE. In the latter model, surface tension effects are not taken into account. The results show that the decrease in hydraulic diameter favors quicker condensation, despite the increase in wall friction. Operating conditions are also explored. The results show a major influence of the heat transfer closure law on the results, e.g. the condensation length. This work highlights the lack of unicity in the available literature regarding steam condensation in innovative heat exchangers. Dedicated experimental work on steam is needed in the future.
3D Modeling of Passive Safety Condensers with CATHARE3 for a High-Power Pressurized Water Reactor
In the wake of the Fukushima nuclear accident, passive safety systems have gained in appeal to improve a nuclear power plant’s resistance to accidents without requiring external power supplies. SAfety COndensers (SACO) are a promising example of these new systems. These are secondary heat exchangers, immersed in a tertiary pool, attached to the steam piping of Steam Generators (SG). In the event of a failure in the normal water supply to the SG, they take the place of emergency pumps, which require a power supply. Their purpose is to extract residual power from the core by condensation of the secondary steam produced at the SG, and return it in liquid form. Then, they preserve the SG liquid inventory, preventing from drying out. For several years now, CEA, in collaboration with EDF, has been involved in modeling SACO of various designs (straight or “C”-shaped vertical tubes, in a calandria or a small or large pool ...) for new reactor concepts. In particular, the 3D modeling of immersed exchangers in a pool, using the CATHARE3 code, challenges the conventional 0D/1D modeling and shows the interest of 3D spatial discretization to better take into account the lateral feed of exchangers. As a result, several 3D SACO models, based on different experimental designs and adapted to the reactor power under consideration, are proposed. There are compared with each other on a typical secondary depressurization transient. Also, the CATHARE3 3D modeling is discussed on a typical station black-out transient for a given type of SACO design.
Experimental Investigation of Critical Heat Flux and Post-Critical Heat Flux in a Carbon Dioxide Cooled Tube at High Subcritical Pressures.
Supercritical fluids such as water or CO2 are an attractive choice in various engineering applications (e.g. in energy conversion systems or nuclear power systems such as SCWR or SMART reactor concept). The exact knowledge of the heat transfer in a trans-critical pressure is important for the design of those systems. In subcritical pressure especially knowing how to avoid the reduction of heat removal when exceeding the critical heat flux (CHF). A series of experiments are carried out with the SCARLETT (Supercritical CARbon dioxide Loop at IKE StuTTgart) test facility at the Institute of Nuclear Technology and Energy Systems of the University of Stuttgart, focusing on investigations of CHF and post-CHF heat transfer with CO2 as a working fluid. Four test sections have been used – each with a different internal diameter. In the test section with a 4 mm internal diameter and a heated length of 2000 mm and the test section with 8 mm internal diameter and heated length of 1640 mm, the wall temperature was monitored every 40 mm along the tube with Pt100 sensors. The other two test sections with 10 mm and 6 mm internal diameters and heated lengths of 2000 mm were equipped with glass fiber in a stainless-steel capillary for semi-continuous wall temperature measurement and thermocouples for the reference temperature measurement. All test sections were instrumented accordingly for pressure, differential pressure, mass flow, inlet, and outlet temperature measurements. The paper shows the experimental results in vertical upward flow with reduced pressures of Π>0.7, mass fluxes ranging from 250 to 3000 kg/m2s, and heat fluxes from 40 to 200 kW/m2.
A Novel OpenFOAM Library for Magneto-Hydrodynamics Studies in the Nuclear Fusion Field
The intricate interplay between externally imposed magnetic fields and the dynamic behaviour of conducting fluids constitutes a challenging magneto-hydrodynamic (MHD) multi-physics problem of significant importance within the nuclear field. Studying such phenomena is crucial for comprehending and optimising processes inherent to this field, encompassing scenarios like magnetic confinement fusion reactors. The current open-source OpenFOAM solver, albeit validated, has limited modelling capabilities as it is suited only for incompressible, isothermal, and laminar flows; thus, there is interest in developing a complete MHD library that can tackle fusion-relevant conditions, involving, for example, the MHD flow in liquid metal or molten salts blankets and the confinement of the hot plasma. This work presents the development and verification of magnetoHDFoam, a new OpenFOAM computational library designed for simulating the complex fluid dynamics of MHD flows, with a particular emphasis on practical applications within the domain of the nuclear field. Compared to the existing OpenFOAM library, the presented solver has been designed for transient simulations of compressible fluids, both for laminar and turbulent regime with (eventually) temperature and density variations. For verification, the new library is tested against well-established benchmark tests and analytical cases (Hartmann flow, Shercliff flow, Hunt flow and thermal Shercliff flow) representing MHD phenomena within fusion reactors, focusing on conductive fluids under externally imposed magnetic fields. The library exhibits good agreement with the expected results, confirming its capability to reproduce the effects of magnetic fields on fluid flow regimes. As such, it represents a useful computational tool for the in-depth study of MHD phenomena in the context of nuclear fusion research, offering significant insights into the behaviour of conducting fluids in the presence of magnetic fields.
Simulation of Transient Melting of CANDU Debris Bed under Severe Core Damage Accident
In a Loss-of-Coolant Accident (LOCA), the fuel channels inside of a CANada Deuterium Uranium (CANDU) reactor would collapse to the bottom of the calandria vessel and continue to heat and melt due to decay heat generation. The transient melting process of the terminal debris would impact the melt pool thermohydraulic behaviours and consequently affect the exiting heat flux, which is a key parameter for reactor safety assessments under severe accidents. Here, the transient heating and melting of a collapsed CANDU solid debris bed after moderator dry-out is simulated using computational fluid dynamics incorporating an enthalpy-porosity melting model. A time-varying decay heat source is considered, and the transient evolution of the debris bed temperature profile, flow field, and exiting heat are presented. Results from the current study can assist future modelling efforts on melt migration and debris bed deformation.
A Novel Approach for Parametric Dynamic Mode Decomposition: Application to the DYNASTY Experimental Facility
Learning models from data has become quite popular recently due to the significant developments in artificial intelligence and the availability of large amounts of data. Nevertheless, this topic has already been addressed in the past by methodologies belonging to the Reduced Order Modelling framework: in particular, one of the most famous equation-free techniques is the Dynamic Mode Decomposition (DMD), an algorithm designed to learn the best linear model underlying some physical phenomena starting from the knowledge of time series datasets. Hence, this method can find the best state matrix of a dynamical system to advance in time even beyond the span of the original dataset. In its standard formulation, DMD cannot deal with parametric time series, and a different linear model must be derived for each realization of the parameter: this work proposes a novel approach to tackle this problem by treating the linear DMD operators as snapshot data, from which a reduced representation can be obtained. Then, it is only sufficient to learn the map between the parameters and the modal coefficients of the reduced expansion of the DMD operators. In this perspective, this approach stands as a generalization of linearization methods to learn a broader class of transient phenomena within the same framework. This novel methodology has been applied to a RELAP5 model of the DYNASTY experimental facility built at Politecnico di Milano, which studies the natural circulation established by internally heated fluids for Generation IV applications, especially in the case of Circulating Fuel reactors.
Designs of Printed Circuit Steam Generator for the Modular Integrated Gas-cooled High Temperature Reactor
Advanced reactors, such as the Horizontal Compact High Temperature Gas Reactor (HC-HTGR), require compact steam generators for near-term deployment as an alternative to the traditional shell-and-tube steam generators to achieve a relatively high level of compactness for the reactors. The recently proposed Modular Integrated Gas-cooled High Temperature Reactor (MIGHTR) that is under development by one of the U.S. Department of Energy’s Advanced Reactor Demonstration Program projects is an HC-HTGR design, aiming to increase its reactor building power density by four to five times. Therefore, printed circuit steam generator (PCSG) has gained interest as a compact steam generator potentially applicable to the MIGHTR. Previous work on printed circuit heat changers is primarily focused on single-phase flow applications, and thus two-phase experimental data involving flow boiling in such heat exchangers are hardly available. The present work presents a prototypic PCSG design by employing zigzag-shaped flow channels to minimize its volume and pressure drop. A genetic algorithm (NSGAII) was utilized to perform PCSG volume optimization. Based on the optimization results, two lab-scale PCSGs were designed for testing under close-to-prototypic conditions with a reduced thermal duty of 24 kW each to investigate their thermal performance and density wave oscillations on the cold side due to phase change. A test facility that couples an existing high-temperature helium loop with a newly designed water/steam loop is currently under construction to test these PCSGs with an anticipated hot-side helium inlet temperature of 687 °C at 3.0 MPa and a cold-side feedwater inlet temperature of 193 °C at 7.0 MPa.
Experimental Validation of Hydrogen within a Filter-Vented Canister Containing Fukushima Fuel Debris
This paper describes experimental evaluation of hydrogen removal from possible designs for passive filter vented storage of fuel debris from the Fukushima Daiichi Nuclear Power Station. Retrieved fuel debris will be first placed into a small cylindrical “unit can.” Four unit cans will be stacked inside a canister whose lid contains two high efficiency filters in series. Since the fuel debris contains water, hydrogen is generated in the canister by radiolysis and it must be discharged through filters to avoid forming a flammable mixture within the canister. A first group of experiments was designed to investigate hydrogen removal through a series of filters to create a performance curve for the hydrogen mole fraction difference across the filter as a function of its flow rate through the filter. These data were used to provide the value for the hydrogen flow resistance across the filter and to confirm models that include mass transfer to and from filter surfaces and between the interior and exterior of the canister lid. A second group of experiments was designed to investigate the distribution of hydrogen within a full size canister using lid designs tested in the first experiment group. These data confirmed that hydrogen will be essentially well-mixed in a canister, and mixing would be even better with decay power present. The conclusion is that possible designs for passive filter vented interim storage are safe with respect to flammable gas behavior.
Optimal Design and Preliminary Transient Analysis of Coupling System of Micro-Nuclear-Reactor with Energy Storage System
Micro nuclear reactors, particularly those utilizing heat-pipe-cooling technology (HPRs), have emerged as a compelling solution for reducing carbon emissions, attributed to their compactness, versatility, and economic advantages. These reactors are distinguished by their resilience against single-point failures, their ability to passively remove heat from the core, and their modular nature, making them ideally suited for diverse applications, including remote grid-connected communities and mining operations. However, the deployment of HPRs in such settings introduces operational challenges, notably the difficulty in adapting to the fluctuating energy demands due to the reactors' inherent thermal inertia and the cooling system's response characteristics. This study introduces a novel conceptual framework for a hybrid energy system that integrates HPRs with battery-based energy storage, aiming to enhance the system's adaptability to variable energy demands. The proposed system incorporates a comprehensive model encompassing the nuclear reactor core, heat pipe cooling apparatus, Brayton cycle energy conversion mechanism, and a supplementary battery storage unit. Through a detailed evaluation, the study assesses the hybrid system's capacity to respond to dynamic energy requirements effectively. Preliminary results from this investigation reveal that the integration of HPRs with battery storage significantly augments the system's load-following capabilities, mitigating the challenges associated with sudden power fluctuations inherent in standalone HPR systems.
Development of Fission Gas Release Modeling in CTF for High Burnup Fuel Transient Analysis
As part of the high energy fuel modeling and simulation, existing computational tools must be adapted to maintain applicability and accuracy at high burnups and increased enrichments. For high burnups in pressurized water reactors (PWR), there are ongoing studies on steady state fission gas release (FGR) and its effect on the fuel rod steady state operation. As burnup increases, fission gases trapped in the fuel matrix migrate to the fuel rod edge, increasing fuel brittleness, and if released present a possibility of rod overpressurization and clad ballooning. These fission gases also present a non-negligible effect on heat transfer through the fuel rod gap region. Increases in gas inventory in the gap following release from steady state and transient operation changes the timing and likelihood of cladding liftoff during transient conditions. This change in rod gap conductance significantly changes the timing of cladding quench following DNB in transient conditions. An improvement to the CTF reactor core thermal hydraulic analysis code on its capability of modeling FGR for high burnup fuel is proposed. The FRAPGR FGR model is added as a new code feature in steady state modeling, moving CTF towards applicability at high burnup above a fuel rod average burnup of 62 GWD/MTU for PWRs. The improvement in calculation of rod internal pressure and gap heat transfer is quantified and the updated code is verified against available benchmark cases. Current work is presented, and future work is proposed for additional improvements to the CTF code for high burnup fuel transient analysis.
The Preliminary Analysis of a Typical Fluoride-Salt-Cooled High-Temperature Reactor in the Loss of Flow Accident with the Influence of Rolling Motion
The solid-fuel-type MSR, Fluoride-salt-cooled High-temperature Reactor presents potential advantages in the use of passive safety systems that ensure the safe response of the reactor during anticipated operational occurrences, design basis events, and beyond design basis events. The FHR applied in ocean conditions faces a strong influence on the thermal-hydraulics. Using the system analysis code, ASYST with the implementation of thermal-hydraulic models in the consideration of the ocean motions effects, the safety evaluation is performed on the typical FHR design, MK1 PB-FHR, with the proposed concept designs of Direct Reactor Auxiliary Cooling System (DRACS). The influence of rolling motion condition is explored on the safety characteristics in the protected Loss of Forced circulation Accident (LOFA).In the analyzed rolling motion condition, mass flow rates in the core pebble bed, DHX, and DRACS loop fluctuate, signifying flow direction reversals. Additionally, DRACS' passive heat removal capability declines under rolling motion. While both DRACS loops operate, core outlet and fuel maximum temperatures remain stable. However, if one DRACS loop fails due to freezing, notable differences in thermal-hydraulic parameters arise between static and rolling motion conditions.
BWRX-300’s Nominal Pressure and Temperature Testing (2): Steam-Water Two-Phase Flow Evaluations in Large Diameter Chimney Based on Void Fraction Measurement using Double-Plane Wire-Mesh Sensor
The BWRX-300 is a water-cooled, natural circulation, small modular reactor developed by GE Hitachi Nuclear Energy in cooperation with Hitachi-GE Nuclear Energy. BWRX-300 is equipped with passive safety systems and leverages the design and licensing of USNRC-certified ESBWR. Like the ESBWR, the BWRX-300 uses a tall chimney above the core to provide natural circulation head to drive sufficient core flow and ensure fuel is operated within its thermal limits. The TRACG system code has been qualified for calculating thermal limits in BWRs including the natural circulation ESBWR. To support extending the TRACG qualification to the large-diameter simplified chimney of BWRX-300, we have obtained additional data and shown the effectiveness of the measurement data in the previous study. This study aims to validate the void fraction estimation with the TRACG for the large-diameter chimney based on the measurement data using a Double-plane Wire-mesh Sensor (DWS) in a down-scaled chimney of around 0.5 m diameter. In Hitachi Utility Steam Test LEading facility (HUSTLE), void fraction distributions and velocities of the gas-liquid interface were measured by the DWS at nominal BWRX-300 operating conditions of 7.17 MPa and 289 ℃. The flow regimes in the chimney were discussed based on the void fraction distributions and representative scale of the dispersed phase measured by the DWS. Furthermore, the average void fraction in the chimney was estimated by the TRACG and showed good agreement with the HUSTLE test results. Consequently, we confirmed the validity of the TRACG estimation for the large-diameter chimney.
Steam Generator Model Design Parameter Sensitivity Study Using Machine-learning Tools
This study focuses on design parameter sensitivity studies pertaining to a steam generator (SG) model, using both Python and machine-learning tools. The SG model is a mathematical representation (including fluid flow and heat transfer equations/models/correlations) of a steam-generating unit in a pressurized water reactor (PWR)-type small modular reactor (SMR) system. Design studies involve changing the model’s input design parameters (e.g., temperature, pressure, mass flow rate) to observe the resulting effects on the output of the system (e.g., heat transfer coefficient [HTC], Nusselt number, heat transfer performance). Sensitivity studies analyze the degree to which system output and/or desired parameters (e.g., HTC or heat transfer performance) are sensitive to changes in input parameters. By using machine-learning tools such as the Risk Analysis Virtual Environment (RAVEN) developed at Idaho National Laboratory (INL), detailed design parametric sensitivity studies and model optimization were performed. Six input parameters—namely, the pressure, temperature, and mass flow rate for the inlet of the primary-side (hot fluid) and secondary-side (cold fluid) of the SG—were randomly perturbed via RAVEN’s Monte Carlo Sampler module, using uniform distributions (±1% relative changes). The analysis results give valuable insights into SG system performance and optimization, and provide justification for researching optimized sensor placement to effectively monitor and obtain experimental data.
BWRX-300’s Nominal Pressure and Temperature Testing (1): Analysis of Test Results from Various Large Diameter Chimney Tests at Elevated Pressure and Temperature
The BWRX-300 is a water-cooled natural circulation small modular reactor developed by GE Hitachi Nuclear Energy in cooperation with Hitachi-GE Nuclear Energy. BWRX-300 is equipped with passive safety systems and leverages the design and licensing of USNRC-certified ESBWR. Like the ESBWR, the BWRX-300 uses a tall chimney above the core to provide natural circulation head to drive sufficient core flow and ensure fuel is operated within its thermal limits. The TRACG system code has been qualified for calculating thermal limits in BWRs including the natural circulation ESBWR. To support extending the TRACG qualification to the large-diameter simplified chimney of BWRX-300, test data have been obtained in Hitachi Utility Steam Test LEading facility (HUSTLE) using two down-scaled chimneys of 0.5 m and 0.35 m in diameter, at nominal BWR operating conditions of 7.17 MPa and 289 ℃ and flow rates scaled from BWRX-300 chimney operating conditions. Some previously published data for 0.5 m chimney were compared with additional test results for a smaller 0.35 m diameter chimney and with other similar tests. The test data were compared with some theoretical models. The study analyzed how the test data scale with various parameters including the chimney diameter and the results provide additional insight into the behavior of two-phase flow in large diameter vertical channels, especially in the high void-fraction range.
First Steps in Developing a Sampling-based Approach to Incorporate Uncertainties in CFD Modeling of Involute Fuel Element Research Reactors
To support the global non-proliferation efforts, the two European high-performance involute fuel element research reactors Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRM II) operated by Technical University of Munich (TUM) in Garching, Germany, and the R´eacteur `a Haut Flux (RHF) located at the Institut Laue-Langevin (ILL) in Grenoble, France, are working towards the conversion of their fuel elements from currently highly enriched to low enriched uranium. The design of such new fuel elements strongly benefits from Steady-State Thermal-Hydraulic (SSTH) safety parameter calculations, which take into account various sources of uncertainties like manufacturing tolerances, measurement inaccuracies and modeling assumptions. While there are well-established methods to determine the nominal SSTH safety parameters at FRM II and RHF, namely by using the 1D thermal-hydraulic code PLTEMP/ANL (PLTEMP), developed at Argonne National Laboratory, and the commercial Computational Fluid Dynamics (CFD) software Ansys CFX (CFX), the investigation of uncertainties requires the implementation of new analysis techniques. For that purpose, the Python tool TAILLEFER is currently being developed at TUM and ILL. Based on Monte-Carlo sampling, TAILLEFER allows for model-independent Sensitivity Analysis (SA) and Uncertainty Propagation (UP) studies of scalar-valued response functions with respect to parameter spaces of arbitrary shape. With the application of TAILLEFER to models implemented in Python or PLTEMP already being shown in the past, the work presented here aims at demonstrating how TAILLEFER can be used with CFX. Using a simplified CFD model of a FRM II cooling channel, the method for code-coupling is explained and verified, and a detailed interpretation of the SA and UP study results is provided.
Modeling the Impact of Surface Characteristics on Wall Boiling Performance in Subcooled Flows
In the nuclear industry, boiling is used for removing heat from the core of both pressurized and boiling water reactors, due to the much greater efficiency in heat transfer compared to single-phase forced convection. While it is known that heat transfer effects can be emphasized through surface engineering and impacted by fouling, the ability to generally model both these effects are still lacking. Improved understanding of the fundamental effects that surface morphology and wettability have on the boiling process is needed, where empirical correlation-based approaches, as well as a first generation of partitioning approaches do not incorporate the necessary mechanisms The MIT boiling formulation has attempted to expand the reach of partitioning models by including a more consistent representation of all mechanisms that impact heat transfer at the wall. The approach aims at representing the boiling phenomena directly from the measured wall characteristics represented from the density of active nucleation sites and quantified static bubble contact angles. Leveraging novel experimental data produced at MIT for varying surface characteristics, we evaluate the ability of the MIT boiling approach to represent the boiling performance directly from surface characterization, and we further identify current limitations and recommendations for future developments. The final goal of the work is to allow predicting the evolution of thermal performance due to fouling during the reactor’s operational lifetime and allowing for optimization of surface design to improve heat transfer capabilities.
An Approach to Establishing Operator Trust Through an Explainable Diagnostic Model
If an abnormal state caused by a problem with a specific component within a nuclear power plant is not addressed, it can result in a reactor shutdown, leading to significant economic losses. Operators identify such issues and implement appropriate abnormal operating procedures to prevent accidents. Recently, there has been research on artificial intelligence models to assist operators in diagnosing these abnormal states more efficiently. However, it is essential to provide additional information for operators, who are the final decision-makers, enabling them to make independent judgments even in cases where the model provides incorrect prediction information. This study utilizes explainable artificial intelligence to offer operators the trustworthiness information of the diagnosis results. When the neural network model diagnoses the abnormal state, the parameter relevance calculated about the model diagnosing each abnormal state. The trustworthiness of the diagnosis result is further evaluated by employing another classification model using this parameter relevance. This dual-step process establishes the overall trust of the model for operators. By emphasizing transparency and providing operators with detailed insights into the causes behind diagnosis results, our study aims not only to increase operators' trust in the model but also to contribute to the operation of a stable nuclear power plant.
Wall-Modelled Large Eddy Simulation (LES) for the Separated Effect Flow Accelerated Corrosion and Erosion (SEFACE) Facility
SEFACE is a purpose-built rotating disk facility to investigate material damage resulting from turbulent high temperature heavy liquid metal (HT-HLM) flow. In this study, we conducted a large eddy simulation (LES) to analyze the turbulence dynamics near the rotating disc, which operates in a Reynolds number range of 1 to 10 millions. We employed the algebraic approach proposed by (Shur et al, 2008) from the wall-modeled Large Eddy Simulation (WMLES) capability in ANSYS FLUENT 21.0R2 to simulate the test chamber at the SEFACE facility. To validate our time-averaged boundary layer flow solution, we compared it with the established boundary layer profile in the viscous sublayer and logarithmic region. Furthermore, we explore the flow structure within the asymmetric test chamber, taking into account the influence of turbulence jets on the flow pattern. Although recirculation and turbulent jets had a minor impact on the mean flow properties near the rotating disk, we observed persistent fluctuations at higher y+ values. Note that these fluctuations may also be influenced by the substantial Reynolds number in our analysis. Furthermore, we identified the presence of adverse pressure gradient (APG) boundary layer characteristics in rotating disk flows, emphasizing the need for consideration when applying an appropriate algebraic wall model. However, deviations were negligible at lower y+ values, suggesting near-wall behavior similar to that of the boudnary layer flow without APG. This implies that a non-APG wall model may still be suitable for engineering evaluations. Lastly, we discovered interconnected wall shear stress streaks in scenarios with smaller compartments, possibly due to the gap effect. These flow statistics related to wall shear stress will prove valuable in the interpretation of experimental results in the future.