Subcooled boiling has been investigated by using the RPI wall boiling models in the last two decades. High accuracy of such models has been achieved by improving the submodels for interphase actions or tuning the model parameters. However, the reliabilities of the models are still suspicious due to the limited model validation and the experimental data based model calibration. The applicability of calibrated model parameters in the new experiment data can not be assured. The effects of model parameters need to be calibrated were treated as the uncertainties of these parameters. The critical parameters that dominate the prediction of subcooled boiling were selected by using the hierarchy analysis. After that, the input samples for uncertainty analysis were obtained by an efficient Monte-Carlo sampling technology — Latin Hypercube Sampling based on the hypothetic normal distribution. Then, the samples were transferred into the FLUENT code for CFD calculations. Results from CFD code were extracted for statistical analysis. Besides, the uncertainties from boundary conditions were also analyzed to quantify the effects of experimental uncertainties. The dependency of the predicted subcooling parameters and the input parameters can be obtained. A PIRT table can be drawn from the generated correlation coefficients between inputs and outputs to quantify the importance of model parameters on subcooled boiling.

In a BWR, the cooler feedwater is injected in the downcomer of the reactor vessel where it mixes with the recirculation water that is separated from the generated steam at the exit of the core. In order to avoid high temperature fluctuations on the reactor vessel and the structures present in the downcomer, this mixing shall be performed in a controlled manner. In order to improve the design of the feedwater spargers in Olkiluoto Unit 1 and 2 in Finland, the thermal mixing of the feedwater with the reactor water in the vicinity of the spargers has been studied with transient Computational Fluid Dynamics (CFD) calculations. Unsteady Reynolds-Averaged Navier-Stokes (URANS) calculations have been performed to simulate about 40 s of the water mixing. The time history of the variations of the water temperature on the structures in the downcomer has been analyzed. The calculations have been run for several designs of the replacement spargers. The designs have been assessed using transient statistics of the temperature fluctuations. Minimizing the maximum amplitude of the temperature variations, the so-called peak-to-peak value, has been a design objective.

This Paper presents an improved estimation of reactor core baffle temperature distribution, during operation, at the nominal power level to address swelling problems of the reactor internals. Swelling is the main limiting factor in the reactor core internals long term operation of VVER-1000 nuclear units. The material irradiation-induced swelling and creep models are very sensitive to temperature distribution in metal, thus a more detailed analysis of the core baffle metal thermohydraulic cooling characteristics is required. A framework for CFD analysis of VVER-1000 reactor baffle cooling is presented. Firstly, an analytical model was developed to obtain boundary conditions and simplify CFD analysis, i.e. the real geometry of the cooling channels was replaced by equivalent elements, the core was presented as porous body with the appropriate characteristics. Secondly, the CFD analysis was performed using 60–degree symmetry, which included: core, baffle and core barrel, it is limited by the height of the baffle. Core is simplified as a homogeneous body with considering of spatial volumetric energy release. Core baffle is presented as monolithic body with considering of gamma-ray heat generation. Model includes a simplified geometry of connecting studs, considering cooling flow of the coolant through the nuts grooves. Calculated convection coefficient and temperature are in good agreement with analytical model, and give a more convenient result comparing to RELAP5/mod3.2. Obtained results were used to estimate baffle swelling process. Due to the less conservative results in temperature distribution swelling and creep deformations significantly decreased.

The Transient Reactor Test Facility (TREAT) is a high enriched, graphite moderated, air cooled reactor built for experimental transient fuel testing. Recently, the reactor was returned to operation after having been shut down since 1994. Transients at TREAT are controlled largely by control/transient rod movement and temperature feedback that is attributed to the core’s graphite-fuel matrix. To date, TREAT simulations use the standard ENDF/B-VII.1 graphite thermal neutron scattering cross sections that assume an ideal crystalline form for the core’s graphite. Historically, it has been reported that the use of these cross sections may result in a −2000 pcm discrepancy when attempting to predict TREAT criticality [1]. In this work, a multi-physics simulation of a TREAT transient is performed using the standard ENDF/B-VII.1 graphite thermal scattering cross section libraries and compared with results using graphite libraries that assume a porous graphite structure and a corresponding density consistent with TREAT graphite. The transient simulation methodology couples a full-core transient Monte Carlo calculation in the Serpent code with feedback calculated from temperature estimates derived using the computational fluid dynamics code OpenFOAM. Steady state simulations show that use of the “porous” graphite libraries allows predicting TREAT criticality to within a few hundred pcm. In the current transient simulations, the reactor’s time dependent power behavior is successfully reproduced. With this model, observables such as maximum fuel temperatures and temperature-dependent flux spectra are calculated, using both the traditional ENDF/B-VII.1 and the “porous” graphite thermal scattering libraries.

Gas entrainment (GE) from cover gas, which is an inert gas to cover Sodium coolant in a reactor vessel, is one of key issue for Sodium-cooled fast reactors (SFRs) design to prevent unexpected effects to core reactivity. In this study, evaluation method has been investigated for surface dimple depth growth of unstable drifting vortex dimples on the liquid surface in the reactor vessel. By using a computational fluid dynamics (CFD) code, analyses have been conducted to estimate the drifting vortexes on water experiments in the circulating water tunnel. The unstable drifting flow vortexes on the water surface were generated as wake vortexes behind a plate obstacle. Downward flow velocity was induced by bottom slit flow pass along the flow channel. In CFD analysis, flat water surface assumption was used with free-slip model for the development of a conventional design method. The gas core lengths from the surface vortexes were evaluated based on existing non-dimensional numbers method by using the stream-viewer code provided from Japan Atomic Energy Agency (JAEA). In this paper, the results of comparison between experiments and analyses were discussed.

The experiments on upward flow of supercritical water in a vertical annuli and 3-rod tight bundle simulator made of 485-mm heated-length tubes of 5.2-mm OD and 4.5-mm ID with four helical ribs of 0.6-mm height, 1-mm width, and axial 400-mm pitch are presented. Heat transfer and pressure drop under various operating conditions (inlet pressure and temperature, flow mass rate and heat flux) were investigated. Longitudinal wall temperature profiles made it possible to determine the place and flow thermal state of heat transfer deterioration (HTD) onset. Analysis of the obtained data (about 200 regimes) proved their good enough agreement with the correlations previously derived by the authors both for the heat flux rate (q/G) b of HTD beginning and for pressure drop in round tubes and annular channels. These correlations were updated to correct the results of their prediction. Computational fluid dynamics and its counterpart computational heat transfer were used for modeling the above-mentioned thermohydraulic processes studied in the first part of the work by finding the most adequate flow turbulence model and optimized domain meshing. The accepted model was benchmarked by some data on heat transfer and pressure drop in tubes and annular channels cooled by SCW.

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.

Helical steam generators are proposed for use in a number of advanced nuclear reactor designs. The cross-flow around the helical tubes is a complex flow-field, and accurate knowledge of this flow is necessary for estimating pressure drop, heat transfer, and risk of flow-induced vibration. However, legacy data for helical tube cross-flow are scarce, and building new large-scale experiments that investigate relevant phenomena can be costly. Thus large uncertainties must currently be taken into account in the design of these systems. Numerical modeling with CFD can provide improved insight into the flow phenomena to reduce this uncertainty, but choosing a methodology can prove difficult. LES methods provide high-fidelity data, but require immense computational time to perform even an investigatory calculation for a moderate-sized sector, let alone for many design iterations. URANS methods offer significantly lower computational time, but it can be difficult to confidently justify the accuracy of a particular model without validation, particularly given the highly three-dimensional and complex flow-field present here. To better establish a basis for URANS turbulence modeling, an LES simulation was performed using Nek5000, a massively-parallel spectral element code developed at Argonne National Laboratory, for the geometry of a legacy helical tube bundle experiment. Data from this high-fidelity LES simulation were compared with URANS simulations using a number of turbulence models with the commercial code STAR-CCM+. Turbulent kinetic energy in the flow channels as well as bundle pressure drop were compared. The values of these key parameters were found to vary significantly between different turbulence models, with some models predicting pressure drops and kinetic energies well below those seen in LES. Some models were identified that showed good potential for predicting helical tube bundle flow phenomena. Further work, at a wider range of flow velocities, will be useful to further solidify the range of applicability of these models.

The paper focuses on the extension of the NEK-2P Wall Heat Transfer model, which was initially developed for the analysis of Critical Heat Flux (CHF) under Dryout (DO) conditions to the simulation of CHF under Departure from Nucleate Boiling (DNB) conditions. The paper presents results of recent NEK-2P analyses of several CHF experiments including both DO and DNB conditions. The CHF experiments analyzed have measured the axial distribution of wall temperatures in two-phase boiling flow in a vertical channel with a heated wall. The axial distribution of the calculated wall temperatures is compared with the corresponding experimental data. Reasonably good agreement with measured data is obtained in predicting the CHF location and post CHF wall temperature magnitudes illustrating the ability of the NEK-2P code and Extended Boiling Framework (EBF) models to simulate the CHF phenomena for a wide range of thermal-hydraulic conditions.

In order to study the dynamic characteristics and fluid structure interactions of tubular structures under the action of fluid in reactor, such as fuel rod bundles and heat transfer tube bundle of steam generator, the dynamic equations and the acoustic wave equations of structures are discretized by finite element method. The acoustic wave equations are simplified from continuity equation and momentum equation of fluid field. Based on the fluid structure coupling method, the dynamic characteristics of the tube under the internal flow, external flow and combined action of internal and external flow are calculated respectively. The influence of flow field domain, element type and grid number on the dynamic characteristics of the tube is also analyzed. Secondly, based on the computational fluid dynamics and computational structural dynamics, the interaction between the two physical fields of fluid and structure is considered simultaneously. The finite volume method is used to discretize the fluid control equations and the turbulent flow is investigated using the large eddy simulation method (LES). The Newmark algorithm is used to solve the structural dynamic equations. Combined with the dynamic mesh control technique, a numerical model for flow induced vibration of three-dimensional flexible tube is established. Finally, the flow induced vibration of a three dimensional flexible single tube and a square arrangement tube bundle is calculated using the numerical model. By comparing with the existing research results, it is found that the numerical simulation results are in good agreement with the experimental results. Thus, the correctness of the model is verified. It is also shown that the numerical model established in this paper can be used to simulate the dynamic characteristics and flow induced vibration of tubular structures.

With the development of the computational capability, Computational Fluid Dynamics (CFD) is used more and more for design and safety issues related to reactor thermal hydraulics. A series of CFD codes have been developed and utilized to analyze the reactor thermal hydraulics behavior. GASFLOW-MPI is a well-validated parallel scalable CFD software and has been widely utilized to predict the complex physical phenomena in nuclear power plant containment. Furthermore, the simulation results are well accepted by the nuclear authorities in several countries all over the world. Turbulent mixing is a basic but important issue for thermal hydraulics safety analysis. In order to capture more unsteady turbulent information under the acceptable computational cost, the LES and RANS hybrid turbulent model Detached Eddy Simulation (DES) model has been developed in the CFD code GASFLOW-MPI. In this paper, two typical turbulent cases are simulated and compared with the experimental data to validate the DES model. The first case is the turbulent backward-facing step flow which is a well-known turbulent benchmark. The second case is a hydrogen turbulent jet which is a typical turbulent flow for hydrogen combustion risk evaluation in nuclear containment. The numerical results show that the averaged flow profiles agree well with the experimental data.

The radial porosity generally have a higher value at the container wall than that in the core part. Consequently, the fluid flows are mal-distributed in packed bed with significant bypass flow at the wall, which lowers the convective heat transfer performance inside the packed bed. To overcome this drawback of packed bed, we developed an effective way to construct the radial layered composite packed bed, which can easily realize placing small particle at the near wall region and large spheres in the core region. Therefore, smaller pores forms close to the container wall and larger channels presented in the core part. This could result in a much homogenous radial porosity distribution, which is benefit to restrain the bypass flow near the wall. In present paper, the packing procedure is simulated by the discrete element method (DEM). Radial layered composite packed bed and traditional packed bed with uniform spheres are compared on radial porosity distribution. By altering the size of spheres ratios in the near wall region different radial layered packings are also generated and compared. Then, the geometries of these packed bed are imported into the computational fluid dynamics (CFD) simulation. The fluid flow inside the packed bed is investigated. It finds that the radial layered packed bed has a lower pressure drop with the ordered packing structure. And a much homogenous fluid flow distribution is obtained than the traditional, which is benefit for the heat removal inside the packed bed. This would be useful for the optimum design of packed bed in industry applications.

The finite element method based on unstructured mesh has good geometry adaptability, it has been used to solve reactor physics problems, manual description of geometric modeling and meshing makes the current finite element code very complicated, it greatly restricts the application of this method in the numerical calculation of reactor physics. Using the CAD pre-processing software ICEM-CFD, three dimensional geometry is divided into tetrahedral or hexahedral meshes, two dimensional geometry is divided into triangular or quadrilateral meshes, the main code of neutron calculation for nuclear noise analysis based on finite element method is developed. The steady state parameters are calculated and tested through benchmark problem, the test results show that the code has the corresponding computing capabilities. Finally, the neutron noise spectrum is calculated for the 3D PWR benchmark problem published by IAEA, and the noise distribution under given frequency is given.

The FRED code is an in-house tool developed at the Paul Scherrer Institut for the so-called 1.5-D nuclear fuel performance analysis. In order to extend its field of application, this code has been re-implemented as a class of the OpenFOAM numerical library. A first objective of this re-implementation is to provide this tool with the parallel scalability necessary for full-core analyses. In addition, the use of OpenFOAM as base library allows for a straightforward interface with the standard Open-FOAM CFD solvers, as well as with the several OpenFOAM-based applications developed by the nuclear engineering community. In this paper, the newly developed FRED-based Open-FOAM class has been integrated in the GeN-Foam multi-physics code mainly developed at the École polytechnique fédérale de Lausanne and at the Paul Scherrer Institut. The paper presents the details of both the re-implementation of the FRED code and of its integration in GeN-Foam. The performances and parallel scalability of the tool are preliminary investigated and an example of application is provided by performing a full-core multi-physics analysis of the European Sodium Fast Reactor.

Control Rod Hydraulic Drive System (CRHDS) is a new type of built-in control rod drive technology, and the Integrated Valve (IV) is the key control component of it. The pulse water flowing into the control rod hydraulic mechanism (CRHDM) is controlled by the IV to drive the hydraulic cylinders to move in a predefined sequence to make the control rod perform step-up, step-down and scram functions. Flow resistance of the IV flow channels is the key design parameter of IV which influences the step motion of the hydraulic cylinder and thus affects the performance of the CRHDS. Experiments on the flow resistance of IV flow channels at different working temperatures were conducted to obtain differential pressures of IV under various temperature and flow rate operating conditions. Based on the experimental conditions and results, three dimensional flow field analysis of the IV flow channels was carried out to get the flow field distribution and hydraulic parameters of the IV flow channels. Flow resistance of the IV flow channels at different working temperatures were obtained using the calculation results and agree well with the experimental results. It verified the correctness of the CFD model. On the basis of the numerical simulation results, the velocity and pressure distribution schemes in the IV flow channels under different working temperature conditions were compared and analyzed. The research results show that the flow resistance of the IV in-rod flow channels remains largely unchanged at different working temperatures, the peak flow velocity appears at the entrance of the valve core section which is also the main flow resistance loss area. The theoretical model was then applied to analyze the influence of the design parameters which include the valve core size, the angle between flow channels, etc., on the total flow resistance of IV at high temperatures. And the analysis results show that, the angle between flow channels has little influence on the flow resistance coefficient. The increase of valve core radius can significantly reduce the total flow resistance of IV flow channels. Numerical simulation on one out-rod flow channel is also carried out, which shows that the flow resistance in out-rod flow channel is much lower than the corresponding in-rod flow channels. The research results can give guidance for the design and optimization of the IV.

The paper describes methods used for evaluation of pressurized thermal shock (PTS) at UJV Rez and selected results. The UJV Rez participated on the one hand in a number of international programs focused on development of PTS methodology, benchmark testing etc. On the other hand UJV has worked in several projects evaluating PTS for nuclear power plants in the Czech Republic and abroad. The introductory part of the paper also briefly mentions a development of PTS methodology in the world, its current status and trends. The core of the paper describes thermal hydraulic part of the PTS evaluation process, which usually contains system thermal hydraulic (TH) analysis and mixing calculation. Already at the system TH calculation level, a simple 2D model of reactor downcomer (DC) is used in UJV Rez analyses — as this approach enables the prediction of 2D temperature and velocity fields in DC and improves the prediction of natural circulation flow or its stagnation in individual loops. For the subsequent mixing calculation the Computational Fluid Dynamics (CFD) computer codes are used nowadays. Comparison of 2D temperature fields from 2D system TH analysis and CFD calculation shows surprisingly good agreement for most cases. The paper shows such comparison for several representative cases and takes a more detailed look into one of them.

In the concrete cask, the canister is sealed with lids by welding, and has high sealing performance. But considering long-term storage, there is a concern about loss of the sealing performance due to stress corrosion cracking (SCC). In the concrete cask, unlike the metal cask, it is not mandatory to constantly monitor helium pressure between the lids. However, it is useful from the viewpoint of improving safety during the long-term storage to install a helium leak detector in the canister inside the concrete cask. Currently, we are developing the leak detector utilizing the phenomenon that the surface temperature of the canister changes when helium leaks out of the canister. As part of developing the leak detector of the canister, leak tests were performed using a small canister model as a pressurized vessel and a 1/4.5 scale cask model of the actual cask including the canister. This leak detector utilized the phenomenon that canister bottom temperature (TB) increases and canister lid temperature (TT) decreases when the internal pressure of the canister decreases. In computational fluid dynamics (CFD) calculation, focused on this phenomenon, the influence of the internal pressure and physical properties of internal gas in the canister were examined by calculating conditions of three kinds of pressure and two types of gas (air and helium). The main purpose of the CFD calculation was to confirm the results of the experiment, and we grasped the phenomenon occurring in the canister and elucidated its mechanism. For the CFD calculation, a commercial CFD software, STAR-CCM+ ® (ver.12.06.010) by Siemens PLM Software Company, was used. A CAD file used for the calculation simulated also the shape inside the canister (e.g. basket, fuel rods). A polyhedral mesh was used for a calculation mesh. In the small canister model, a mesh of its ambient air was not generated, and heat transfer between the canister surface and the ambient air was calculated from a heat transfer correlation equation. On the other hand, in the 1 / 4.5 scale cask model, the mesh of its ambient air was generated, so that the heat transfer on the surface of the canister was calculated according to the actual heat transfer phenomenon. The internal gas and the ambient air of the canister were ideal gas, and buoyancy due to density change was taken into consideration. A realizable k-epsilon model was used for a turbulence model, and a DO model was used for a radiation model.

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.

The main steam stop valve (MSSV) is installed in the main steam line in thermal and nuclear power plants. The MSSV is a safety valve that instantaneously shuts off the steam flowing into the steam turbine in an emergency. However, as high-speed steam flow goes through the MSSV during even the rated operation, acoustic sound or noise is generated in the MSSV. Moreover, there is a possibility that flow-induced acoustic resonance occurs in the MSSV. Flow-induced acoustic resonance must be suppressed to decrease the sound noise. Reducing the pressure loss of the MSSV is also an important issue that cannot be neglected with respect to the plant thermal efficiency. Therefore, we have developed the MSSV which can suppress the flow-induced acoustic resonance and its pressure loss. To develop this MSSV, we conducted scale air tests and computational fluid dynamics (CFD) analyses that are described in this paper. Mach and Strouhal number of the test conditions were the same as those of an actual plant. Reynolds number was sufficiently large to obtain the developed turbulent flow. An unsteady compressible CFD analysis was also conducted using large eddy simulation as a turbulence model. We developed new tilted triangular tabs and installed them in the MSSV to suppress the intense vortex generation and pressure loss. As a result, the sound noise due to the flow-induced acoustic resonance was completely attenuated and pressure loss was reduced compared to the case using the current tilted tabs. CFD results also showed that the tilted triangular tabs could suppress the generation of intense vortexes and the flow-induced acoustic resonance.

The direct and indirect hydro-dynamic loads in the suppression pool of the BWR fleet were developed well before 2000. These loads were based on scaled experiments and numerical solutions using one dimensional models. The analysis was cumbersome and conservatisms were added in multiple steps. As these loads spread throughout the containment, as global vibrations, they are generally a part of all structural verification inside the containment. This fact has made it hard and unpractical to challenge and revise these loads, as a change could lead to significant re-work. A consequence of this is that loads in the suppression pools have seldom been revisited, regardless if they cause local or global vibrations. This is problematic when new equipment is needed. The design of this equipment suffers from the very conservative loads. Experience has shown that these loads can be challenged and refined using updated techniques, leading to significantly lower loads. This was realized during the modernization of Swedish plant Oskarshamn Unit 2 with Mark II containment, where loads following pool swell proved to be particularly challenging and it was decided to investigate possibilities to reduce the conservative loads. Due to the large scale of the condensation pool coupled with the transient and small scaled condensation of steam at the drywell vent pipe nozzle full CFD resolution is not feasible. Instead lumped models in GOTHIC was used to increase resolution from the 1D approach of normal containment analysis, to a resolution that can account for the features of the condensation pool. This showed that the pool swell was less uniform than initially thought, leading to fewer objects affected as well as lower loads on objects that suffered from these loads. A full CFD analysis was then used to resolve phenomena working on even shorter time scales leading to a complete rework of all local loads. The loads addressed using updated codes and modelling techniques was pool swell impact (PSI), pool swell drag (PSD) and local drag loads due to pressure relief valve opening, LV/SRV. The current work shows that using updated modelling techniques and aligning results with previous analysis and documentation, it is possible to reduce loads for some events in the suppression pool without violating safety for the power plant. The results from the GOTHIC model shows that the cylindrical shape of the pool will create an uneven velocity distribution radially at pool swell resulting in much smaller loads at the outer boundary. CFD analysis of the LV/SRV event shows that the loads are reduced in comparison with previous methodology and this is mainly because of shadow effects. The calculations also verifies assumptions used in the previous methodology.