Experiments at the mixing test facility ROCOM for benchmarking of CFD codes

For the validation of computational fluid dynamics (CFD) codes, experimental data on fluid flow parameters with high resolution in time and space are needed.Rossendorf Coolant Mixing Model (ROCOM) is a test facility for the investigation of coolant mixing in the primary circuit of pressurized water reactors. This facility reproduces the primary circuit of a German KONVOI-type reactor. All important details of the reactor pressure vessel are modelled at a linear scale of 1:5. The facility is characterized by flexible possibilities of operation in a wide variety of flow regimes and boundary conditions. The flow path of the coolant from the cold legs through the downcomer until the inlet into the core is equipped with high-resolution detectors, in particular, wire mesh sensors in the downcomer of the vessel with a mesh of 64 × 32 measurement positions and in the core inlet plane with one measurement position for the entry into each fuel assembly, to enable high-level CFD code validation. Two different types of experiments at the ROCOM test facility have been proposed for this purpose. The first proposal concerns the transport of a slug of hot, under-borated condensate, which has formed in the cold leg after a small break LOCA, towards the reactor core under natural circulation. The propagation of the emergency core cooling water in the test facility under natural circulation or even stagnant flow conditions should be investigated in the second type of experiment. The measured data can contribute significantly to the validation of CFD codes for complex mixing processes with high relevance for nuclear safety.     

Validation of the CFD code fluent by post-test calculation of a density-driven ROCOM experiment

During the last years, boron dilution events with the potential of reactivity transients were an important issue of German PWR safety analyses. A coolant with a low-boron concentration could be collected in localized areas of the reactor coolant system, e.g., by separation of a borated reactor coolant into highly concentrated and diluted fractions (inherent dilution) which can occur during reflux-condenser heat transfer after a small break loss of coolant accident with a limited availability of the emergency core cooling systems.During the course of follower core assessments, TÜV NORD SysTec appraises safety analyses of boron dilution events presented by the utilities. These analyses are based on the simulation of boron dilution and transport processes in conjunction with a number of dedicated experiments. The analyses demonstrate that boron dilution events cannot lead to recriticality of the core. Hence, the boron concentration at the core inlet has to be determined.TÜV NORD SysTec applies the CFD code FLUENT for the investigation of boron dilution events in pressurized water reactors. To affirm the FLUENT abilities for the simulation of boron dilution events, a validation against the ROCOM experiment T6655_21 with a density-driven coolant mixing was performed. This validation proves that FLUENT is able to appropriately simulate the effects of boron transport and dilution such as streaks of coolant with lower density in the downcomer. Deficits were identified in the simulation of fluid layering in the cold leg, which fortunately have a rather small influence on the predicted core inlet concentration. Therefore, the boron concentration in the reactor core can be determined with sufficient accuracy to solve the safety issue, regardless of the core becoming critical or not.     

Experiments on slug mixing under natural circulation conditions at the ROCOM test facility using high-resolution measurement techniques and numerical modeling

ROCOM is a four-loop test facility used for the investigation of coolant mixing in the primary circuit of pressurized water reactors. Recently, a new sensor was developed for an improved visualisation and quantification of the coolant mixing in the downcomer. This new sensor array spans a dense measuring grid and covers nearly the whole downcomer. In the presented work, special emphasis was given to the comparison of the data of this sensor with the results of calculations using the Computational Fluid Dynamics (CFD) code ANSYS CFX. A coolant mixing experiment during natural circulation conditions has been conducted. The underlying scenario of this experiment is based on a boron dilution scenario following a SBLOCA event. The corresponding CFD code solution has been obtained using the Best Practice Guidelines. All main effects observed in the measurement are described by the calculation. The detailed comparison reveals that the calculation underestimates the coolant mixing inside the reactor pressure vessel.The measurement data, boundary conditions of the experiment and facility geometry can be made available to other CFD code users for benchmarking.     

Modeling of containment atmosphere mixing and stratification experiment using a CFD approach

An experiment on containment atmosphere mixing and stratification, which was originally performed in the TOSQAN facility in Saclay (France), was simulated with the Computational Fluid Dynamics code CFX4.4. The TOSQAN facility consists of a large cylindrical vessel in which gases are injected. In the considered experiment, steam, air and helium were injected during different phases of the experiment, with steam condensing on some parts of the vessel walls. During certain phases, steady states were obtained when the steam condensation rate became equal to the steam injection rate, with all boundary conditions remaining constant. In the present work, three such intermediate steady states were simulated independently. The essential purpose was to reproduce the non-homogeneous structure of the vessel atmosphere, given that condensation is simulated in such a way to obtain the proper condensation rate. A two-dimensional axisymmetric model of the TOSQAN vessel for the CFX4.4 code was developed. The flow in the simulation domain was modeled as single-phase. Steam condensation on vessel walls was modeled as a sink of mass and energy. Calculated profiles of temperature, steam concentration, and velocity components are compared to experimental results and discussed. The comparison suggests that atmosphere mixing and stratification in an NPP containment at accident conditions could be successfully simulated using the proposed CFD approach.     

Thermal hydraulic performance analysis of the printed circuit heat exchanger using a helium test facility and CFD simulations

The thermal-hydraulic performance of the PCHE was investigated using the KAIST helium test loop. Experiments were performed in the helium laminar region with 350 < Re < 1200. The hot/cold side inlet conditions were 25–550 °C/25–100 °C over the operating pressure of 1.5–1.9 MPa, respectively. Mass flow rates were controlled in the range of 40–100 kg/h. Pressure drop and temperature difference were measured at the inlet and outlet of the hot and cold sides. A global Fanning factor correlation and a global Nusselt number correlation were proposed using information only at the inlet and outlet of the hot and cold sides. A three-dimensional (3-D) numerical simulation was performed using FLUENT, a commercial computational fluid dynamics (CFD) code, to compare simulation results to the KAIST helium test data and to obtain the local Nusselt number in the PCHE. CFD predictions showed good agreement with experimental data. A local pitch-averaged Nusselt number correlation was proposed using local temperature, pressure, surface heat fluxes, and properties provided by CFD simulations. The system analysis code, GAMMA, was also utilized to identify which correlation was more applicable for system analysis. It turns out that the proposed local pitch-averaged Nusselt number correlation from CFD simulations is more appropriate than the global Nusselt number correlation developed from experimental data.     

Qualification of the CFD code Trio_U for full scale reactor applications

The article presents a procedure to qualify the Trio_U code for the prediction of the boron concentration at the core inlet of a French 900 MWe pressurized water reactor under accidental conditions (inherent dilution problem).¹ The objective of this procedure is to ensure that the validation calculations are performed with the same modelling hypotheses as the full scale reactor analysis, for which usually no experimental data are available. A density driven ROCOM experiment as well as an UPTF Tram-C3 experiment have been used for the qualification of the Trio_U code. Both experiments present similar thermal hydraulic conditions as the reactor case. The predicted boron concentration at the core inlet of the reactor shows that the potential return to criticality might not be excluded in the case of a small break LOCA. Further neutronic calculations are necessary to confirm this result.     

Actual flow calibration of a feedwater flowmeter using a high Reynolds number facility at NMIJ

The results of calibration tests of the feedwater flowrate of ultrasonic flowmeters used in a nuclear power plant for variety of upstream conditions obtained using the new high Reynolds number calibration facility at NMIJ are described. In this examination, the measurements are performed for five pattern pipe layouts with one or two elbows. The flow conditioners installed upstream of the flowmeter are the tube bundle type and the Mitsubishi, which are normally used in nuclear power plants. The calibration result for each flowmeter are largely different for each flow conditioner and each upstream pipe layout, except in some special cases. Moreover, the trend of the correction factor with Reynolds number is not uniform for each case. Furthermore, some differences were observed for individual flowmeters. It is recommended that the feedwater flowmeter, especially when used to perform measurement uncertainty recapture, is calibrated based on the actual pipe layout and the Reynolds number corresponding to the actual nuclear power plant conditions.     

Detailed CFX-5 study of the coolant mixing within the reactor pressure vessel of a VVER-1000 reactor during a non-symmetrical heat-up test

As part of the reactor dynamics activities of FZK/IRS, the qualification of a detailed 3D CFD model of a reactor pressure vessel is a key step in safety evaluations for improving predictive capabilities and acceptability of commercial CFD tools in reactor physics. The VVER-1000 Coolant Transient Benchmark, initiated by OECD, represents an excellent opportunity for validation. In this work a CFD model for the complete VVER-1000 reactor pressure vessel is presented. Due to computational limits simplifications of the core and of some other geometrical details are introduced. The simulated scenario is the heat-up of one coolant loop in case of the isolation of a steam generator while the reactor is operating at a low power level. Two transient runs with a first and second order approximation of the spatial discretization are performed. Unexpectedly, the first order method reveals better agreement with measured data.     

Technical developments at the KIT gyrotron test facility

Parasitic beam tunnel oscillations have been discovered on some of the series production gyrotrons for W7-X and also on the coaxial pre-prototype gyrotron for ITER. Solutions to remedy these problems have resulted in a modified beam tunnel design, technologically close to the existing beam tunnel. The new design has successfully been tested on both the coaxial and also the f-step-tunable gyrotrons and has subsequently been implemented on one of the W7-X series-production-tubes presently undergoing factory acceptance tests in Karlsruhe.The ECRH test loads at KIT are operated under normal atmospheric conditions. Several loads have eventually failed in 1 MW long pulse experiments and KIT has therefore started to design its own loads. The first KIT-load is based on a fixed conical mirror and an aluminum cylinder coated with a lossy material for increased absorption. The new load has so far successfully been used during the acceptance tests of two 1-MW CW gyrotrons. Nevertheless a new load based on pure (uncoated) stainless steel absorbers is being developed as a backup solution for the ongoing high priority gyrotron testing.A superconducting magnet capable of rapid field changes between 4.15 and 5.67 T for frequency step-tunable gyrotrons has been procured, has demonstrated a (static) field of 7.2 T and its capability of rapid field-changes.     

Study on the flow distribution in prismatic VHTR core with a multi-block experiment and CFD analysis

VHTR is a gas-cooled, graphite-moderated reactor that was developed for the production of hydrogen and process heat. In the prismatic VHTR core with hexagonal graphite blocks, the bypass gap, which is an interstitial gap between the hexagonal blocks, inevitably forms due to the installation tolerance and irradiation shrinkage of the core blocks. Core bypass flow is the ineffective coolant flow that does not pass through the coolant holes within the fuel block. The core-bypass flow distribution varies according to the fuel cycle because the graphite blocks are deformed by irradiation and heating. Since the bypass flow reduces the core thermal margin, it should be evaluated precisely to guarantee reactor safety. In this regard, experimental and computational studies were carried out to evaluate the core-bypass flow distribution. A multi-block air test facility was designed and constructed in Seoul National University to measure the flow and pressure distributions according to the block combination and cross-flow gap size. The effects of the cross-flow phenomenon on the flow and pressure distribution in the core were investigated. The experimental data were used to validate a CFD model for the analysis of bypass flow characteristics in detail.     

Fluid mixing and flow distribution in a primary circuit of a nuclear pressurized water reactor—Validation of CFD codes

The EU project FLOMIX-R was aimed at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity.This report will focus on the computational fluid dynamics (CFD) code validation. Best practice guidelines (BPG) were applied in all CFD work when choosing computational grid, time step, turbulence models, modelling of internal geometry, boundary conditions, numerical schemes and convergence criteria. The strategy of code validation based on the BPG and a matrix of CFD code validation calculations have been elaborated. CFD calculations have been accomplished for selected experiments with two different CFD codes (CFX, FLUENT). The matrix of benchmark cases contains slug mixing tests simulating the start-up of the first main circulation pump which have been performed with three 1:5 scaled facilities: the Rossendorf coolant mixing model ROCOM, the Vattenfall test facility and a metal mock-up of a VVER-1000 type reactor. Before studying mixing in transients, ROCOM test cases with steady-state flow conditions were considered. Considering buoyancy driven mixing, experimental results on mixing of fluids with density differences obtained at ROCOM and the FORTUM PTS test facility were compared with calculations. Methods for a quantitative comparison between the calculated and measured mixing scalar distributions have been elaborated and applied. Based on the “best practice CFD solutions”, conclusions on the applicability of CFD for turbulent mixing problems in PWR were drawn and recommendations on CFD modelling were given. The results of the CFD calculations are mostly in-between the uncertainty bands of the experiments. Although no fully grid-independent numerical solutions could be obtained, it can be concluded about the suitability of applying CFD methods in engineering applications for turbulent mixing in nuclear reactors.     

Pre- and post-test calculations to natural circulation experiments at the integral test facility ISB-VVER using the thermalhydraulic code ATHLET

In 1995 at the integral test facility ISB-VVER in Elektrogorsk near Moscow natural circulation experiments were performed, which were scientifically accompanied by the Forschungszentrum Rossendorf. These experiments were the first of this kind at a test facility, which models VVER-1000 thermalhydraulics. Using the code ATHLET which is being developed by ‘Gesellschaft für Anlagen und Reaktorsicherheit’, pre- and post-test calculations were done to determine the thermalhydraulic events to be expected and to define and tune the boundary conditions of the test. The conditions found for natural circulation instabilities and cold leg loop seal clearing could be confirmed by the tests. Besides the thermalhydraulic standard measuring system, the facility was equipped with needle shaped conductivity probes for measuring the local void fractions.     

An integral effect test of a complete loss of reactor coolant system flow rate for the SMART design using the VISTA-ITL facility and its simulation with the MARS-KS code

An integral effect test was successfully performed to provide data to assess the capability of the system analysis code to simulate a complete loss of reactor coolant system (RCS) flow rate (CLOF) scenario for the SMART (System-integrated Modular Advanced ReacTor) design. The steady-state conditions were achieved to satisfy initial test conditions presented in the test requirement, its boundary conditions were accurately simulated, and the CLOF scenario in the SMART design was reproduced properly using the VISTA-ITL facility. The natural circulation flow rate in the RCS was about 12.0% of the rated RCS flow rate and the flow rate in the passive residual heat removal system (PRHRS) loop was about 10.6% of its rated value in the early stage of the PRHRS operation. In this paper, the major experimental results of the CLOF test are discussed. The test results were analyzed using the best-estimate system analysis code, MARS-KS, to assess its capability to simulate a CLOF scenario for the SMART design.     

The first DC performance test and analysis of CC conductor short sample at ASIPP conductor test facility

The first DC performance experiments of ITER correction coil (CC) conductor short sample have been carried out in the conductor test facility of Institute of Plasma Physics, CAS (ASIPP) in January this year. Those experiments aim to investigate the DC performance of ITER CC conductor. The tested conductor short sample is bended as a half circle with the diameter of 270 mm to meet the background magnetic field shape. The half circle part of sample is longer than the final twist pitch. The current sharing temperature (T_cs) in the 3.86 T external magnetic field (B_ex), ≤12 kA could be measured including the critical current (I_c) run. There is no obvious impact of 1000 cycles on DC performance. Those measured T_cs results are in agreement with the expected results from strand scaling.     

Analysis of a loss of residual heat removal system during mid-loop conditions at PKL facility using RELAP5/Mod3.3

When a nuclear power plant is in shutdown conditions for refuelling, the reactor coolant system water level is reduced. This situation is known as mid-loop operation, and the residual heat removal (RHR) system is used in this situation to remove the decay power heat generated in the reactor core.In mid-loop conditions, some accidental situations may occur with not a negligible contribution to the plant risk, and all involve the loss of the RHR system. Thus, to better understand the thermal–hydraulic processes following the loss of the RHR during shutdown, transients of this kind have been simulated using best-estimate codes in different integral test facilities. This paper focuses on the simulation, using the best estimate code RELAP5/Mod3.3, of the experiment E3.1 conducted at the PKL facility. This experiment consists of the loss of the RHR system when the plant is in mid-loop conditions for refuelling and with the primary circuit closed. In particular, in this experiment the physical phenomena to investigate are the mechanisms of heat removal in presence of nitrogen and the deboration in critical parts of the primary system.     

Analysis of the hot gas flow in the outlet plenum of the very high temperature reactor using coupled RELAP5-3D system code and a CFD code

The very high temperature reactor (VHTR) system behavior should be predicted during normal operating conditions and postulated accident conditions. The plant accident scenario and the passive safety behavior should be accurately predicted. Uncertainties in passive safety behavior could have large effects on the resulting system characteristics. Due to these performance issues in the VHTR, there is a need for development, testing and validation of design tools to demonstrate the feasibility of the design concepts and guide the improvement of the plant components. One of the identified design issues for the gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream components. To provide flow details, the analysis presented in this paper was performed by coupling a VHTR model generated in a thermal hydraulic systems code to a computational fluid dynamics (CFD) outlet plenum model. The outlet conditions obtained from the systems code VHTR model provide the inlet boundary conditions to the CFD outlet plenum model. By coupling the two codes in this manner, the important three-dimensional flow effects in the outlet plenum are well modeled while avoiding modeling the entire reactor with a computationally expensive CFD code. The values of pressure, mass flow rate and temperature across the coupled boundary showed differences of less than 5% in every location except for one channel. The coupling auxiliary program used in this analysis can be applied to many different cases requiring detailed three-dimensional modeling in a small portion of the domain.     

Analysis of the Peach Bottom flow stability test number 3 using the coupled RELAP5/PARCS code

Nowadays, the coupled codes technique, which consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into system codes, is extensively used for carrying out best estimate (BE) simulation of complex transient in nuclear power plants (NPP). This technique is particularly suitable for transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. Such complex interactions are encountered under normal and abnormal operating conditions of a boiling water reactors (BWR). In such reactors Oscillations may take place owing to the dynamic behavior of the liquid-steam mixture used for removing the thermal power. Therefore, it is necessary to be able to detect in a reliable way these oscillations. The purpose of this work is to characterize one aspect of these unstable behaviors using the coupled codes technique. The evaluation is performed against Peach Bottom-2 low-flow stability tests number 3 using the coupled RELAP5/PARCS code. In this transient dynamically complex neutron kinetics coupling with thermal-hydraulics events take place in response to a core pressure perturbation. The calculated coupled code results are herein assessed and compared against the available experimental data.