This study recaps the experimental effort to characterize a transverse plane flow field in a tightly-packed, 61-pin, wire-wrapped, hexagonal fuel bundle prototypical for a sodium fast reactor. The motive was to produce high spatiotemporal experimental data for computational fluid dynamics turbulence model validation. The matched-index-of-refraction and laser-based optical measurement techniques were utilized on an isothermal experimental flow facility. Measurements were performed on a transverse plane perpendicular to the axial flow. Fluid flow in the three types of subchannels (corner, exterior, and interior) were quantified. All measurements have been performed at a bundle-averaged Reynolds number of approximately 10,400. Results include flow statistics such as ensemble-averaged velocity, root-mean-square fluctuating velocity, and Reynolds stress. Of interest was the flow behavior around the restriction caused by the wire spacer and hexagonal duct wall, where recirculation regions formed. Regions of maximum and minimum momentum transfer coincided with regions of maximum and minimum fluctuations. These regions highlight locations of maximum and minimum cooling of the fuel pins. The experimental data will be used to benchmark computational fluid dynamics simulations of the sodium fast reactor fuel bundle using Reynolds-averaged Navier Stokes and large-eddy simulation turbulence modeling methods.

When restarting Oskarshamn Unit 3 after the re-fuelling outage 2016 a turbine trip occurred at low reactor power. Inspections of the feed-water system in the turbine building revealed a damaged support. A water-hammer caused by an open valve that should have been closed was suspected to be the cause of the damage. Measurement data from the event showed that the pressure in the piping near the damaged support had decreased slowly to be followed by rapid pressure peaks with high amplitudes. The damaged support was reinforced to regain its function and the plant was restarted but an investigation was initiated to analyse the event. A study of measurement data from the plant showed that this was not a single event. Similar events had occurred earlier. The event turned out to be independent of the valve being open or closed. It was also observed that main pumps that were not in operation at the time in the feed-water system had rotated in the reverse direction during the pressure decrease (normally two out of three pumps are in operation). This event could explain a lot of the support damage in the feed-water system that had occurred since the plant was licensed and that was considered to be caused by operational displacements and vibrations. A fruitful and intense collaboration started between Engineering, Operations and Maintenance to find the root cause of the event. This resulted in a number of observations that led to a general understanding of the nature of the event. The event was caused by depressurisation in the feed-water tank caused by a turbine trip at low power. The gradual pressure drop caused the water in the piping lower down to evaporate (flash), causing a mass-flow upwards. When the evaporation stopped the flow changed direction and fell downwards causing condensation and large pressure peaks. The event could be simulated (by RElAP5) with results that qualitatively compared well to measurements. Simulations were then used to show how the event could be mitigated and ultimately avoided. Hopefully future events will be avoided by changing operation procedures. This has been verified when the plant had to shut-down in November 2017 due to a fuel failure. During the shut-down process no flashing event was detected.

Supercritical CO 2 (S-CO 2 ) Brayton cycle has been identified as a promising power conversion method for the next generation of nuclear reactors due to its high efficiency and compactness. The heat exchanger is one of the most important components for S-CO 2 Brayton cycle, and the printed circuit heat exchanger (PCHE) is supposed to be one of the promising candidates for the heat exchangers in S-CO 2 Brayton cycle. It should be noted that the fluid maldistribution would induce heat transfer deterioration, especially for heat exchangers with micro- or mini-scale channels like PCHE. The thermal-physical properties of S-CO 2 change violently during the heat transfer process, which makes the flow inside PCHE more complex. In this paper, the distribution of S-CO 2 flow inside PCHE would be studied by 2-D CFD simulations. For the working fluids with constant properties, the flow nonuniformity increases with the mass flow rate. For the working fluid with S-CO 2 , the thermal-physical properties change significantly with temperature, and there exist a minimum value in the flow nonuniformity-mass flow rate curves (1.64 × 10 5 ≤ Re in ≤ 1.31 × 10 6 ). Insertion of baffles at manifolds could significantly improve the flow distribution uniformity and reduce the pressure drop. And it has been found that insertion of baffles at the collecting manifold has better performance compared with that at the distributing manifold or both.

ATHAS is the subchannel analysis code, independently developed by Nuclear Safety and Operation Laboratory (NUSOL) of Xi'an Jiao Tong University (XJTU), the original version of the ATHAS is only suitable for pressurized water reactor. This paper is aimed at the characteristics of sodium cooled fast reactor core mentioned above, developing the analysis program AYHAS-LMR, which is applicable for LMFBR. In the design of sodium cold fast reactor, flow blockage must be considered to meet the design criteria. In view of this, a new model is proposed in this paper and added in ATHAS-LMR, considering the flow blockage which changes the flow area and the wetted perimeter. This model is based on the original grid model, we use the grid to simulate the presence of block, increase the resistance of the blockage area artificially. Then, ATHAS-LMR is able to simulate the flow blockage in LMFBR. In this paper, simulations of FFM5B blockage test were done for the high flow case with ATHAS-LMR, and we compare the simulation results with COMMIX-1. The result indicates that ATHAS-LMR is suitable for the analysis of the flow blockage.

An ongoing Sandia National Laboratories’ (SNL) research study is evaluating a potential design of an injectable sacrificial material (SM) system that could contain and cool corium ejected from a reactor vessel lower head failure during a potential severe accident involving melting fuel at a commercial light water nuclear reactor (LWR). An injectable system could be installed at any existing LWR, without significant modification to the cavity or to the drywell pedestal region of the plant. The conceptual design under consideration is a passive system. The SM is being optimized to quickly cool the corium mixture while creating gas to form porosity in the solid, such that subsequent water flooding can penetrate the structure and provide additional cooling. The SM would form a barrier and limit corium-concrete interactions. This three-year project takes a joint experimental and computational approach. In this paper, we will first discuss the success of our small-scale experiments conducted on the interactions between the surrogate corium material (SCM) and SM, used to evaluate the injectable concept. A larger experimental study, currently underway, will further validate the injectable concept, with a focus on accurately measuring interactions. This paper details the modeling study and its progress, including modeling the experiments on a surrogate system and extending the model to bench-scale corium flow from validation experiments. The project’s modeling studies will use the SNL engineering code suite SIERRA Mechanics to understand the interaction of injectable SM and molten corium and predict corium spreading. Spreading is modeled using a level set method to track the front in conjunction with a pressure-stabilized finite element method on the fully three-dimensional mass, momentum, and energy conservation equations. Using this diffuse-interface method, the corium spreading front can be tracked and an appropriate pseudo-solidification viscosity models can be implemented to accurately model the corium spreading physics. Finally, an injectable SM delivery system is discussed along with its deployment to the six-common commercial LWR designs currently operating in the United States. At the end of this project, a simplified model based on SIERRA simulations will be developed for implementation into MELCOR, a severe reactor analysis code, developed at SNL for the U.S. Nuclear Regulatory Commission. This will allow us to demonstrate the ability of the injectable SM system to mitigate the ex-vessel corium spreading, provide containment and negate the release of radionuclides.

The GOTHIC code was validated using three experiments carried out in the PANDA facility in the framework of the OECD/NEA project HYMERES. These tests addressed the mixing of an initially stratified atmosphere by means of a vertical jet in the presence of on obstacle (circular plate). This paper reports on the simulations of three experiments, and complementary, quasi-steady state tests without stratification, where the flow structure above the impingement plate could be observed by means of PIV velocity measurements in a region larger than that considered in the transient experiments. Moreover, simulations of similar tests without obstacle conducted during the OECD/SETH-2 are also discussed. The reference, best-estimate model used for the analyses of the three experiments with different flow rates and initial and pressure boundary conditions was built on the base of a multi-step approach. This was based on mesh and modelling sensitivity studies mostly performed for the complementary tests, to assess the capability to represent the flow structure produced by the jet-plate interaction with different meshes around the plate. Generally, the results show that the use of a coarse mesh and the standard k-ε turbulence model permits a reasonable representation of the erosion process, but with a systematic over prediction of the mixing time. The results with the reference model were more accurate for two experiments with two flow rates and same initial conditions and all complementary tests. For the third test with different initial and boundary conditions, however, poor results were obtained with the reference model, which could only be improved by further refining the mesh. These results indicate that a model “qualified” for certain conditions could be inadequate for other cases, and sensitivity studies are necessary for the specific conditions considered in the analyses.

Numerical investigation of buoyancy effect on forced convective heat transfer to supercritical carbon dioxide flowing in a vertical tube was carried out. When the mass flux is low and wall heat flux high, it shows that the buoyancy effect is obvious, which might redistribute the radial and axial velocity, even M shaped distribution in the radial direction. When the zero-velocity-gradient region corresponding to the M shaped velocity distribution appears in the edge of viscous layer, the production and diffusion of eddy will be weakened, resulting in heat transfer deterioration. According to the extended simulations based on experimental data, reducing the wall heat flux, adding the mass flux or raising the inlet temperature can relieve the deterioration of heat transfer caused by buoyancy effect.

To support the activities carried out by the ENEA Brasimone Research Center on the Lead-cooled European Advanced Demonstration Reactor (LEADER) project, University of Pisa has improved the RELAP5/Mod3.3 code introducing the Lead and Lead-Bismuth Eutectic (LBE) as coolant fluids. This new version of the STH code was applied to perform several analyses for the steam generator HERO (Heavy liquid mEtal pRessurized water-cOoled tubes) installed in the CIRCE (CIRColazione Eutettico) pool-type facility that uses LBE as working fluid. HERO consists of a bundle of seven Double-Wall Bayonet Tubes (DWBT) with a leakage monitor system, and it was tested in the CIRCE facility to support the development of the steam generator proposed for ALFRED (Advanced Lead Fast Reactor European Demonstrator). A preliminary analysis was performed comparing the RELAP5/Mod3.3 results against available experimental data. In particular, an isothermal test of the CIRCE-HERO facility allowed to set-up the RELAP5/Mod3.3 nodalization. As a second analysis, a steady state at full power was investigated. This steady-state condition was employed to perform a sensitivity study on the influence of the thermal conductivity of the stainless steel powder characterizing the HERO steam generator. Finally, the analysis of unprotected and protected loss of flow transients, and the analysis of a protected loss of flow with a simultaneous partial Loss of Heat Sink (LOHS) were performed. These last transients were investigated to assess the behavior of the HERO steam generator, especially to verify the possibility to establish a natural circulation capable to safely remove the “decay heat”. The results of the first analysis (isothermal hydraulic characterization) demonstrated the good capabilities of improved RELAP5/Mod3.3 code to reproduce the forced circulation conditions. The results of the second analysis (steady-state at full power) highlighted the strong influence of the stainless steel powder thermal conductivity on the achieved steady-state conditions. The simulations of the last tests provided the time evolution of the LBE mass flow rate and of the temperatures in different regions inside the pool during unprotected and protected loss of flow transients. In the paper, a critical discussion of the obtained results is performed.

In the framework of the ALFRED design (Advanced Lead Fast Reactor European Demonstrator) for DEMO-LFR, a new concept of steam generator (SG) has been proposed consisting in a double wall bayonet tube bundle which improves the plant safety reducing the possibility of water-lead interaction thanks to a double physical separation between them, and allowing an easier control of eventual leakages from the coolant by pressurizing the separation region with inert gas. In order to support the development of this innovative SG configuration, the ENEA has designed and realized the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section, a mock-up (1:1 in length) which represents the ALFRED SG. This test section, implemented in the CIRCE pool facility, aims to investigate on the thermal-hydraulic features of the system, providing a database for STH codes validation. The experimental campaign consists of high pressure tests at about 180 bar carried out in the framework of the HORIZON2020 SESAME project (Simulations and Experiments for the Safety Assessment of MEtal cooled reactors). The secondary loop has been realized for the HERO SG feeding, consisting in an open loop circuit fed by demineralized water. The system is equipped with a volumetric pump and a heater in order to reach the water nominal working conditions of 335°C at the SG inlet and about 180 bar at the outlet. A preliminary test analysis is carried out by RELAP5-3D © thermal-hydraulic system code. A numerical 1-D model of the HERO SG and the secondary loop has been realized in order to test the loop layout and to characterize the main components from a thermal-hydraulic point of view, defining the start-up procedures for the achievement of the working conditions of the water for the high pressure tests. Furthermore, several simulations are carried out to investigate on the secondary system behavior both for steady states and transients.

In the work, we employed a new method to monitor the behavior of the grounded control rod. To this end, a sensor system based on the two-electrode capacitance is applied to measure the rod position in nuclear heating reactor. Additionally, the finite element method was used to analyze the capacitance of the sensors composed of two-electrode. According to the different electric properties of grounded control rod and air, this apparatus determines the control rod position using an interface electronics method and two -electrode capacitance sensor. The capacitance data obtained by simulations were compared with experiment results to verify the measurement accuracy. The subsequent quantitative analyses on the data indicate the reliability and accuracy of the apparatus for monitoring rod position. It is demonstrated that such a system shows very promising applications for measurements of the control rod position of nuclear reactor.

The description of the fluid flow and the coolant velocity field within a Pressurized Water Reactor (PWR), for the safety analysis is always desirable. In this work, a validation of CFD code ANSYS Fluent performed in the frame of experimental investigations to study the velocity field of coolant flow in a single phase at a reactor vessel down-comer. Under steady-state conditions. Several Experiments using Particle Image Velocimetry, have done at small hemispherical vessel, which designed by the College of Nuclear science and Technology at Harbin Engineering University (HEU) to obtain experimental data for velocity field. The measurements of the fluid flow velocities at two different locations as investigation area in the down-comer under the inlet part, for two different inlets velocities have experimentally calculated and comprehensively simulated using FLUENT. A 3D model of hemispherical cylinder, created by commercial package SolidWorks 2016 and star –CCM code has used as tool for geometry meshing and the simulations process performed with ANSYS-FLUENT code. RANS was first calculated with boundary conditions such as the velocity used at the inlet and the outlet. Two different locations were been selected as in the experimental model and the velocity has measured. The results of CFD analyses demonstrated very good agreement with the experimental data. The results that obtained by the experiment and Fluent help to understand the velocity distribution in the down-comer.

A new neutronics and thermal-hydraulics coupled code named ARTAP is developed to analyze the steady-state and transient characteristics of accelerator driven subcritical system (ADS) in this paper. Monte Carlo simulations for spallation neutron source and deterministic calculations for the subcritical core are performed in the steady-state analysis module. The ADS core is divided into a number of nodes both along the reactor axis and the fuel pellet radius for the calculation of spatial temperature distributions. The power iteration method is adopted to solve the coupled neutronics and thermal-hydraulics problems. The transient analysis module consists of space-time neutron kinetics model and thermal-hydraulics dynamic model, which is calculated by using numerical differentiation formulas (NDFs) method. The new code is verified by comparing its predictions for both the steady-state and transient cases of the OECD/NEA ADS benchmark. Results of numerical simulations indicate that ARTAP is reliable and efficient to be applied for the ADS analysis.

With renewed interest in nuclear power to meet the world’s future energy demand, the Generation IV nuclear reactors are the next step in the deployment of nuclear power generation. However, for the potentials of these nuclear reactor designs to be fully realized, its suitability, when coupled with different configurations of closed-cycle gas turbine power conversion systems, have to be explored and performance compared for various possible working fluids over a range of operating pressures and temperatures. The purpose of this paper is to carry out performance analysis at the design and off-design conditions for a Generation IV nuclear-powered reactor in combination with a recuperated closed-cycle gas turbine and comparing the influence of carbon dioxide and nitrogen as working fluid in the cycle. This analysis is demonstrated in GTACYSS; a performance and preliminary design code developed by the authors for closed-cycle gas turbine simulations. The results obtained shows that the choice of working fluid controls the range of cycle operating pressures, temperatures and overall performance of the power plant due to the thermodynamic and heat properties of the fluids. The performance results favored the nitrogen working fluid over CO 2 due to the behavior CO 2 below its critical conditions.

In predicting the void fraction in subcooled flow boiling, accurate evaluation of single bubble behaviors is of considerable importance. In particular, bubble lift-off velocity affects the void fraction significantly since the bubble disappear quickly due to heat transfer with subcooled liquid if the lift-off velocity is high. In this study, the process of bubble lift-off was experimentally investigated to develop mechanistic correlations for the bubble lift-off velocity. In the development of the correlations, it was assumed that the bubble lift-off velocity in the horizontal direction is proportional to the bubble growth rate and that in the vertical direction is determined primarily by the local liquid velocity evaluated at the bubble center position. Then, impact of the bubble lift-off velocity on the void fraction was explored through numerical simulations. In the simulations, the bubble lift-off velocity in the lateral direction was parametrically changed. It was shown that the mean void fraction decreases with an increase in the lateral bubble lift-off velocity since the bubble condensation is enhanced. It was therefore confirmed that accurate evaluation of the bubble lift-off velocity is important for high accuracy prediction of the void fraction.

In the pebble-bed high temperature reactor under construction in China, called HTR-PM, the spherical fuel elements continuously flow downward in the cylindrical core. After the discharge, the burnup of each pebble is checked at the core outlet and, according to the achieved burnup level, the pebble might be disposed or reinserted into the upper section of the core, distributing randomly in the radial direction and defining a variable number of passes necessary to achieve the average maximum burnup of 90 MWd/kgU. Discrete Element Method (DEM) simulations have been carried out to achieve a clear understanding of the movement of 420,000 fuel pebbles in the HTR-PM core. At the same time, neutronic properties have been investigated for a single pebble and for the full core with Serpent 2 Monte Carlo code in order to perform a parametrization of the one-group microscopic cross sections at the core-level. The pebble movement has been coupled with the neutronic behavior of a single pebble in a dedicated burnup script called Moving Pebble Burnup (MPB), developed in Matlab. 3,000 single pebble burnup histories were simulated to obtain sufficient statistics and insight on the burnup process in the HTR-PM. The decrease of the average burnup gained per single pass implies that a miss-handling of recirculated fuel elements is unlikely to lead to exceedance of the maximum allowed burnup of 100 MWd/kgU. Furthermore, the core demonstrates a self-compensation effect of burnup, meaning that it always compensates burnup under- or over-runs in the successive passes. Finally, it is possible to conclude that the fuel cycle of the HTR-PM, as it has been laid out, is well-designed and feasible.

Betavoltaic cells are nuclear batteries ideal for low-power applications for extended periods of time without maintenance or replacement. Betavoltaics function similarly to photovoltaic (solar) cells where instead of using sunlight, beta particles are used to generate electron-hole pairs within a semiconductor p-n junction to generate current. Even though there have been multiple demonstrations, betavoltaic performance has not been extensively studied. To accurately predict betavoltaic performance, which is important for a device in operation without maintenance for elongated periods, all parameters are required to predict potential fluctuations in cell performance, such as doping densities and resistances for semiconductor variation and absorption coefficients for beta-generated current. However, not all parameters are easily measured, especially when the p-n junction is constantly under irradiation and cannot be separated from the source. Critical parameters were characterized experimentally with the betavoltaic cell by performing capacitance-voltage to determine doping densities and performing current-voltage characterization tests to determine resistances on multiple NanoTritium™ cells, while absorption coefficients were determined from MCNP6 simulations. Experiments indicated that series resistance R s was 1 × 10 6 Ω, while shunt resistance R sh was 2 × 10 8 Ω from I-V characterization, while doping density N D was determined to be 1 × 10 17 cm −3 from C-V characterization. Absorption coefficient α was found to vary with semiconductor material and incoming beta energy and used in conjunction with critical parameters from experimentation to accurately model betavoltaic cell performance similar to experimental results. Both implicit equations and explicit estimations were compared to model betavoltaic cell performance.

One of the requirements for licensing a nuclear power plant in the U.S is the capability to survive and recover from a station blackout according to the U.S Nuclear Regulatory Commission (USNRC). Station blackout is the loss of all off-site and onsite power simultaneously. Therefore, experimental test facilities are being constructed and operated to test the performance of the related safety systems in a nuclear power plant. Design and construction of a test facility creates the need to perform scaling analysis to ensure proper representation of key components and phenomena of interest. One of the main outcomes of the scaling analysis is the quantitative estimation of the Similarity Level (SL), which requires derivation of dimensionless scaling parameters and prediction of appropriate input values for the scaling parameters. To study the performance of the Reactor Core Isolation Cooling (RCIC) system, the Nuclear Heat Transfer Systems (NHTS) Laboratory at Texas A&M University has constructed and is operating a RCIC test facility. This paper presents the scaling analysis with reference to a full-size RCIC system and the RCIC system turbine was used as the main component for scaling. The input parameters for dimensionless scaling parameters were obtained through experimental measurements and CFD analysis. The CFD analysis is for the ZS-1 RCIC system turbine model. The STAR-CCM+ CFD code was used in this study to create and run simulations for steady state normal and abnormal operating conditions for the NHTS-developed CAD models. The input for the dimensionless scaling parameters was estimated. Input parameters were collected both experimentally and from CFD simulations and inserted into these equations. As a result, a high degree of similarity was confirmed, with a minimum of 82% between the NHTS and full-size RCIC systems. The 82% represents the amount of transfer properties conserved between the two systems. Consequently, this high similarity level allows the NHTS RCIC system to be used to study the behavior of the full-size RCIC system under Beyond Design Bases Accident (BDBA) conditions. Future work is to study and model other components of the RCIC system such as the suppression chamber to estimate similarity levels and study their effects on behavior of the system under BDBA.

The development of subchannel models for fuel assemblies and reactor cores requires accurate information on flow distribution, wall friction and loss coefficients in order to accurately predict the pressure, temperature and flow distribution on a subchannel level. This paper discusses the use of Computational Fluid Dynamics (CFD) simulations as a practical tool for characterising inlet velocity boundary conditions, an approximation of wall friction factor and spacer grid pressure loss coefficient, which are of fundamental importance to correctly generate a consistent subchannel model of a given assembly system. The geometry of the simplified PWR assembly presented here is based on the NUPEC PWR subchannel and bundle tests. Comparison of the derived friction factors and grid pressure loss coefficient with published and recommended values are reported. Discrepancies are also explained using additional calculations. A comparison of the overall system pressure drop, by comparing_numerical and analytical solutions, and the local axial pressure distribution at subchannel level are presented. To make a one-to-one comparison between CFD and subchannel solutions, volume-averaging is applied to the CFD results according to the chosen subchannel nodalization. The obtained results show a perfect agreement between the two codes. This outcome reflects the correct approach employed to build two consistent numerical models by properly carrying important information from the high-resolution models (CFD) to the low-resolution models (subchannel code). Furthermore, it has been found that the large discrepancies recorded in the CFD prediction of the grid pressure loss coefficients suggested in the benchmark specifications are mainly because the suggested benchmark values do not take into account the presence of the bounding channel that was present in the experimental facility.

Helically coiled tubes are widely used in many industrial applications such as the steam generator in the high-temperature gas-cooled reactor which is recognized as one of the new generation advanced reactors. The thermophysical properties of fluids exhibit drastic and fast changes in the pseudocritical region so that the flow and heat transfer characteristics of supercritical pressure fluids are greatly different from those at the subcritical pressure. The paper presents results of numerical investigation on turbulent heat transfer of supercritical CO 2 in a helically coiled tube with a tube diameter of 9 mm, a coil diameter of 283 mm and a coil pitch of 32 mm under the constant wall heat flux. Both the RNG k-ε model with enhanced wall function and the SST k-ω model were applied in the simulations, and the results showed that the SST k-ω model agreed better with the experimental results in the literature. Effects of buoyancy and flow acceleration were evaluated. Details of developing heat transfer characteristics at three specific cross sections were analyzed. The heat transfer regularity and mechanism presented in this work can be useful for the design and development of more economic and safer design of the supercritical steam generator.

Passive cooling systems driven by natural circulation are common design features of proposals for advanced reactors. The natural circulation systems are inherently more unstable than forced circulation ones due to its nonlinear nature and low driving force. Any disturbance, e.g. flashing or boiling inception, in the driving force will affect the flow which in turn will influence the driving force leading to an oscillatory behavior. Owing to safety concerns, flashing instability particularly for advanced boiling water reactors has been broadly investigated, and many test facilities have been constructed in the past. A number of numerical analyses of experimental test cases are available. Nevertheless, there exists a need to update the method from one-dimensional system codes to high-resolution computational fluid dynamics (CFD). In the present work flashing-induced instability behavior and flow pattern in the riser of the GENEVA facility, which is a downscale of a reactor containment passive cooling system, is investigated using the commercial CFD code ANSYS CFX. A two-fluid model is adopted for the unstable turbulent gas-liquid flow, and the HZDR baseline closure is used to model interphase mass, momentum, heat transfer as well as bubble-induced turbulence. The simulated fluid temperature, pressure and local void fraction at different heights of the riser are compared with the measured ones. The limitation and possibility of the CFD technique for such complex two-phase scenarios are discussed, and suggestions for improving the predictability of simulations are made.