CFD analysis of thermal-hydraulic behavior of supercritical water in sub-channels

Investigations on the thermal-hydraulic behavior in the supercritical water-cooled reactor (SCWR) fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding and ability to predict the heat transfer behavior of supercritical fluids. In this paper, computational fluid dynamics (CFD) analysis is carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical SCWR fuel assembly using commercial CFD code CFX-5.6. Three types of sub-channels, e.g. regular sub-channel, wall sub-channel and corner sub-channel, are analyzed. Effects of various parameters, such as boundary conditions and pitch-to-diameter ratios, on the mixing phenomenon in sub-channels and heat transfer are investigated. The turbulent mixing in tight lattice (P/D = 1.1) is lower than that in wide lattice (P/D > 1.1), whereas, the effect of pitch-to-diameter ratio on the turbulent mixing is slight at P/D > 1.1. The amplitude of turbulent mixing in wall sub-channel is slightly higher than that in regular sub-channel and is close to that in corner sub-channel. The mixing coefficient in the sub-channel at P/D ≥ 1.2 is in the range from 0.022 to 0.028. The results also show unusual behavior of turbulent mixing in the vicinity of the pseudo-critical point, and further investigation is needed. The mass mixing due to cross flow in wall sub-channel is much stronger than that in regular sub-channel at a same pitch-to-diameter ratio. The mass mixing in wall and regular sub-channels, especially at small pitch-to-diameter ratio, brings an unfavorable feedback to the heat transfer and strengthens the non-uniformity of the circumferential distribution of heat transfer. The strong mass mixing in corner sub-channel should be paid attention.     

Core thermal–hydraulic design

The core thermal–hydraulic design for the HTTR is carried out to evaluate the maximum fuel temperature at normal operation and anticipated operation occurrences. To evaluate coolant flow distribution and maximum fuel temperature, we use the experimental results such as heat transfer coefficient, pressure loss coefficient obtained by mock-up test facilities. Furthermore, we evaluated hot spot factors of fuel temperatures conservatively.As the results of the core thermal–hydraulic design, an effective coolant flow through the core of 88% of the total flow, is achieved at minimum. The maximum fuel temperature appears during the high-temperature test operation, and reaches 1492 °C for the maximum through the burn-up cycle, which satisfies the design limit of 1495 °C at normal operation. It is also confirmed that the maximum fuel temperature at any anticipated operation occurrences does not exceed the fuel design limit of 1600 °C in the safety analysis.On the other hand, result of re-evaluation of analysis condition and hot spot factors based on operation data of the HTTR, the maximum fuel temperature for 160 effective full power operation days is estimated to be 1463 °C. It is confirmed that the core thermal–hydraulic design gives conservative results.     

Quantitative assessment of thermal–hydraulic codes used for heavy water reactor calculations

The RD-14M large LOCA test, characterized by a reliable set of experimental data, was selected for an international standard problem exercise (SPE) entitled “Intercomparison and validation of computer codes for thermal–hydraulics safety analyses”. The activity was performed within the frame of International Atomic Energy Agency's (IAEAs) Technical Working Group on Advanced Technologies for Heavy Water Reactors (TWG-HWR). In this study, the recently improved fast Fourier transform based method (FFTBM) was used for accuracy quantification of RD-14M large LOCA test B9401 calculations of six participants using four different thermal–hydraulic codes. In addition, developing the capability to calculate the accuracy as a function of time-continues-valued accuracy, did further improvement of FFTBM. Namely, in the past only single valued accuracy parameters for selected time windows and time intervals were calculated. The objective of the study was to demonstrate that the new FFTBM is a powerful tool for quantitative assessment of thermal–hydraulic codes. For demonstration, the test from the facility simulating heavy water reactor was used. The blind accuracy analysis was completed based on solely experimental and calculated data. However, short discussions were held with the representative from Italy (co-author, here) regarding phenomenological windows, variables and void fraction weights selection. In general, the open accuracy analysis confirmed the results obtained in blind accuracy analysis. The main conclusions from accuracy analysis agree with the conclusions from the SPE intercomparison report, which was written independently. Finally, the results suggest that the accuracy of the best calculations of the RD-14M test is comparable with the best calculations of light water reactor experiments.     

Treatment of the thermal–hydraulic uncertainties in the pressurized thermal shock analysis

The pressurized thermal shock (PTS) analysis is a quantitative analysis to calculate the vessel failure probability of the embrittled reactor pressure vessel. The PTS analysis consists of three major parts, such as the probabilistic safety analysis (PSA), the thermal–hydraulic analysis (T/H), and the probabilistic fracture mechanics (PFM) analysis. Because each analysis involves many parameters and assumptions associated with the uncertainties, it is important to identify and incorporate them into the analysis. Though the PSA and PFM analysis can be easily treated statistically, the thermal–hydraulic analysis results are very difficult to be treated statistically. Instead, sensitivity analyses of the thermal–hydraulic inputs were performed to understand the significance of the variation in the thermal–hydraulic inputs to the PFM analysis. In this study, the existing PFM code was modified to incorporate the uncertainties in the thermal–hydraulic inputs for the PFM analysis. The effects of the uncertainties in the thermal–hydraulic inputs for the vessel failure probabilities were evaluated using the modified code. The results showed the effects of uncertainties in the thermal–hydraulic inputs on the vessel failure probabilities are not significant for the ranges of the transient types. Even for the larger uncertainties, the effects on the vessel failure probabilities are small. Also, the effects of the thermal–hydraulic uncertainties vary depending on the transient characteristics such that the effects are greatest for the pressure dominant transient. Within the transient, the relative increases in the failure probabilities are greatest for the circumferentially oriented semi-elliptical flaws. It was found that the results of the sensitivity analysis using one standard deviation are conservative enough to bound the analysis results considering the uncertainties in the thermal–hydraulic inputs.     

Local thermal–hydraulic behaviour in tight 7-rod bundles

Advanced water-cooled reactor concepts with tight lattices have been proposed worldwide to improve the fuel utilization and the economic competitiveness. In the present work, experimental investigations were performed on thermal–hydraulic behaviour in tight hexagonal 7-rod bundles under both single-phase and two-phase conditions. Freon-12 was used as working fluid due to its convenient operating parameters. Tests were carried out under both single-phase and two-phase flow conditions. Rod surface temperatures are measured at a fixed axial elevation and in various circumferential positions. Test data with different radial power distributions are analyzed. Measured surface temperatures of unheated rods are used for the assessment of and comparison with numerical codes.In addition, numerical simulation using sub-channel analysis code MATRA and the computational fluid dynamics (CFD) code ANSYS-10 is carried out to understand the experimental data and to assess the validity of these codes in the prediction of flow and heat transfer behaviour in tight rod bundle geometries. Numerical results are compared with experimental data. A good agreement between the measured temperatures on the unheated rod surface and the CFD calculation is obtained. Both sub-channel analysis and CFD calculation indicates that the turbulent mixing in the tight rod bundle is significantly stronger than that computed with a well established correlation.     

CFD analyses of liquid metal flow in sub-channels for Gen IV reactors

Some of the limitations of Reynolds Averaged Navier Stokes (RANS) based Computational Fluid Dynamics CFD codes in computing the flow and temperature field in a rod-bundle are well known. An in-house validation campaign has indicated that the Baseline Reynolds Stress Model (BSL-RSM) with automatic wall treatment is preferred for RANS analyses of a rod-bundle using the CFX code.As a first step in the present paper, the employed CFX code has been assessed with the analyses of a liquid sodium flow in a rod-bundle as in the TEGENA (TEmperatur- und GEschwindigkeitsverteilungen in Stabbündel mit turbulenter NAtriumströmung) experiment. For this RANS analysis, the full cross-section is modelled to avoid numerical issues associated with symmetric boundary conditions.The influence of pitch-to-diameter ratio (p/d) and rod arrangements on thermal-hydraulics is analyzed by applying the assessed modeling approach. For this purpose, rod-bundles with different p/d are arranged in a square and triangular lattice. The computational sub-channels make use of periodic boundary conditions. RANS computed axial velocity normalized with the friction velocity shows the presence of a logarithmic outer region for both arrangements. Similar behavior was reported based on a Large Eddy Simulation (LES) approach. The analyses reveal that the intensity of secondary flow increases with decreasing p/d for both arrangements. RANS analyzed normal Reynolds stresses normalized with centerline velocity in the smallest gap of rod-bundle reveal their anisotropy. Furthermore, the analyses show that the Nusselt numbers increase with p/d for described flow conditions and for both arrangements. Following observations of flow oscillations in a tight lattice rod-bundle as in Hooper's experiment, as a final step, unsteady RANS simulations for hydraulic analyses using a rod-bundle with small p/d are presented with two commercial CFD codes, namely, CFX and STAR-CCM+. In particular, the analysis of Hooper's hydraulics experiment with a tight lattice rod-bundle having a p/d of 1.1 demonstrates the existence of flow oscillations or instabilities as inferred in the experiment.     

Effects of wire-spacer shape in LMR on thermal–hydraulic performance

In this work, analyses of three-dimensional flow and convective heat transfer in wire-wrapped fuel assemblies with different shaped wire-spacers have been carried out using Reynolds-averaged Navier–Stokes equations with shear stress transport turbulence model. Three cross-sectional shapes of wire-spacer, circle, hexagon and rhombus have been tested. All the assemblies have been analyzed for single pitch of wire-spacer with periodic boundary conditions applied at inlet and outlet of the calculation domain. It is found that the assemblies exhibit the directional periodicity in radial gradients between the adjacent subchannels due to presence of wire-spacer. The overall pressure drop is highest in case of rhombus shaped wire-spacer assembly followed by hexagonal shaped. Although circular shaped wire-spacer gives lowest peak temperature as well as lowest overall temperature difference in the assembly, the rhombus shaped wire-spacer assembly gives highest Nusselt number on fuel rod surface.     

Transient analysis of Ghana Research Reactor-1 using PARET/ANL thermal–hydraulic code

PARET/ANL (Version 7.3 of 2007) thermal–hydraulic code was used to perform transient analysis of the Ghana Research Reactor-1. The reactivities inserted were 2.1 mk, 4 mk and 6.71 mk. The results obtained are similar to experiment and theoretical studies performed to demonstrate that the reactor is safe to operate. The PARET/ANL (Version 7.3 of 2007) could not simulate the reactivities above 5 mk insertions which were successfully performed in earlier theoretical and experimental studies. This may be attributed to different fluid flow and heat transfer regimes within the flow channels of the reactor that were considered by the codes.     

Transient thermal–hydraulic simulations of direct cycle gas cooled reactors

This work concerns the design and safety analysis of gas cooled reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbo-machine trip, 10 in. cold duct break, 10 in. break in cold duct combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation. The turbo-machine contribution is discussed and can offer a help or an alternative to “active” heat extraction systems.     

Analysis of potential waterhammer at the Ignalina NPP using thermal–hydraulic and structural analysis codes

The RBMK (Russian acronym for ‘channeled large power reactor’)-1500 reactors at the Ignalina nuclear power plant (NPP) have a series of check valves in the main circulation circuit (MCC) that serve the coolant distribution in the fuel channels. In the case of a hypothetical guillotine break of pipelines upstream of the group distribution headers (GDH), the check valves and adjusted piping integrity is a key issue for the reactor safety during the rapid closure of check valve. An analysis of the waterhammer effect (i.e. the pressure pulse generated by the valves slamming closed) is needed. The thermal–hydraulic and structural analysis of waterhammer effects following the guillotine break of pipelines at the Ignalina NPP with RBMK-1500 reactors was conducted by employing the RELAP5 and PipePlus codes. Results of the analysis demonstrated that the maximum values of the pressure pulses generated by the check valve closure following the hypothetical accidents remain far below the value of pressure of the hydraulic tests, which are performed at the NPP and the risk of failure of the check valves or associated pipelines is low. Sensitivity analysis of pressure pulse dependencies on calculation time step and check valve closure time was performed. Results of RELAP5 calculations are benchmarked against waterhammer transient data obtained by employing structural mechanics code BOS fluids.     

CAST3M/ARCTURUS: A coupled heat transfer CFD code for thermal–hydraulic analyzes of gas cooled reactors

The safety of gas cooled reactors (High Temperature Reactors (HTR), Very High Temperature Reactors (VHTR) or Gas Cooled Fast Reactors (GFR)) must be ensured by systems (active or passive) which maintain loads on component (fuel) and structures (vessel, containment) within acceptable limits under accidental conditions. To achieve this objective, thermal–hydraulics computer codes are necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Some key safety questions are related to the evaluation of decay heat removal and containment pressure and thermal loads. This requires accurate simulations of conduction, convection, thermal radiation transfers and energy storage. Coupling with neutronics is also an important modeling aspect for the determination of representative parameters such as neutronics coefficient (Doppler coefficient, Moderator coeffcient, …), critical position of control rods, reactivity insertion aspects, …. For GFR, the high power density of the core and its necessary reduced dimension cannot rely only on passive systems for decay heat removal. Therefore, forced convection using active safety systems (gas blowers, heat exchangers, …) are highly recommended. Nevertheless, in case of station black-out, the safety demonstration of the concept should be guaranteed by natural circulation heat removal. This could be performed by keeping a relatively high back-up pressure for pure helium convection and also by heavy gas injection. So, it is also necessary to model mixing of different gases, the on-set of natural convection and the pressure and thermal loads onto the proximate or guard containment. In this paper, we report on the developments of the CAST3M/ARCTURUS thermal–hydraulics (Lumped Parameter and CFD) code developed at CEA, including its coupling to the neutronics code CRONOS2 and the system code CATHARE. Elementary validation cases are detailed, as well as application of the code to benchmark problems such as the HTR-10 thermal–hydraulic exercise. Examples of containment thermal–hydraulics calculations for fast reactor design (GFR) are also detailed.     

Development of a thermal–hydraulic system code for simulators based on RELAP5 code

A thermal–hydraulic system code for simulators, RELAPSIM, was developed at NSSE based on RELAP5. The development procedure consists of three major parts. Firstly, time control function was added into the code to meet real-time calculation needs. Secondly, controlled dynamic data communication was improved, so that thermal–hydraulic parameters can be easily modified for further applications. Finally, functions controlling the computation procedure were embedded to achieve a full capability to simulate multiple operations, such as start-up, shutting down or freeze. This paper describes the main features of the new code. The results of code assessment and code application are presented and discussed.     

Flow measurement techniques in heavy liquid metals

This paper summarizes several flow measurement systems qualified in the operation of different lead-bismuth loops in the KArslruhe Liquid Metal LAboratory (KALLA) during the last 5 years. There are several experimental techniques which were well proven in air and water and thus could be transferred similarly to liquid metals: these techniques are split into measuring local quantities as temperature, pressure e.g. by means of pressure taps or velocities using Pitot and Prandtl tubes or the Ultrasound Doppler velocimetry (UDV) for local flow velocities, as well as an integral quantity like the flow rate. Since the knowledge of the flow rate acts in terms of the operational safety of nuclear liquid metal systems as one of the most crucial parameters, this aspect is discussed widely herein. Unfortunately, as liquid metals are opaque, an optical access is not possible. Instead, one can take advantage of the high electric conductivity of liquid metals to measure integral and local quantities, like electromagnetic flow meters and miniaturised permanent magnetic probes for local velocity measurements. In this context especially the electromagnetic frequency flow meter (EMFM) is discussed as a prospective and reliable option to measure the flow rate without demanding extensive precognitions with respect to the fluid-wall interface behaviour.This article describes some of the techniques used in KALLA for different liquid metals, explains the measurement principle and shows some of the typical results obtained using these techniques. Also the measurement accuracy as well as the temporal and spatial resolution of each device is discussed and typical error sources to be expected are illuminated. Moreover, some hints for a correct placement of the individual sensor in the liquid metal environment are given.     

Source–receptor relationships for heavy metals in the European atmosphere

Atmospheric concentrations of heavy metals (HMs), in particular As, Cd, Cu, Pb and Zn, were studied in an effort to contribute to the understanding of European source-receptor relationships. A comparison was made between the ambient concentrations measured at 11 background aerosol monitoring stations (in Denmark, the Czech Republic, Finland, Norway and Sweden) and the corresponding HM concentrations estimated by the Heavy Metals Eulerian Transport (HMET) meteorological dispersion model. The collected samples were analysed with Particle Induced X-ray Emission analysis (PIXE) except the Finnish samples which were analysed with Inductively Coupled Plasma – Mass Spectrometry (ICP-MS). The available data covers the period 1985–1994. The comparison showed that the European emissions of As, Cd and Pb seem to be fairly well estimated. On the other hand, the European Zn emissions are underestimated by a factor of 3 or more, while the Cu emissions appear to be slightly overestimated. The HMET dispersion model also made it possible to select occasions for which the sampling sites had a substantial contribution of HM from the highly polluted “Black Triangle” region (on the borders between the Czech Republic, Poland and Germany). The time evolution of the sources of HMs within this source region could be studied by applying various statistical receptor models on the extensive data set from two Danish stations, Keldsnor and Tange, covering the period 1985–1994. Four source types were clearly discerned throughout the 10 year time period. These sources were: soil dust; sea spray; general combustion and oil combustion. The strong time-dependence observed for the contribution from the Black Triangle region emphasizes the importance of keeping the emission inventories continuously updated if HMs deposition calculations and HMs emissions reduction protocols are to be based on dispersion modelling approaches.     

Uncertainty and sensitivity analyses of the Kozloduy pump trip test using coupled thermal–hydraulic 3D kinetics code

The modeling of complex transients in nuclear power plants (NPP) remains a challenging topic for best estimate three-dimensional coupled code computational tools. This technique is, nowadays, extensively used since it allows decreasing conservatism in the calculation models and performs more realistic simulation and more precise consideration of multidimensional effects under complex transients in NPPs. Therefore, large international activities are in progress aiming to assess the capabilities of coupled codes and the new frontiers for the nuclear technology that could be opened by this technique. In the current paper, a contribution to the assessment and validation of coupled code technique through the Kozloduy VVER100 pump trip test is performed. For this purpose, the coupled RELAP5/3.3-PARCS/2.6 code is used. The code results were assessed against experimental data. Deviations between code predictions and measurements are mainly due to the used models for evaluating and modeling of the Doppler feedback effect. Further investigations through the use of two “antagonist” uncertainty GRS and the CIAU methods, were considered in order to evaluate and quantify the origin of the observed discrepancies. It was revealed on one hand that relative error quantification discrepancies exist between the two approaches, and further enhancements for both methods are needed.     

Thermal–hydraulic feasibility analysis on uprating the HTR–PM

In order to meet energy demand in China, the high temperature gas-cooled reactor–pebble-bed module (HTR–PM) is being developed. It adopts a two-zone core, in which graphite balls are loaded in the central zone and the outer part is fuel ball zone, and couple with a steam cycle. Outer diameter of the reactor core is 4.0 m and height of the core is 9.43 m. The helium inlet and outlet temperature are 250 and 750 °C, respectively. The reactor thermal power is 380 MW. Preliminary studies show that the HTR–PM is feasible technologically and economically. In order to increase the reactor thermal power of the HTR–PM, some efforts have been made. These include increasing the height of reactor core, optimizing the thickness of fuel zone and better selection of the scheme of central graphite zone, etc. Basic design concepts and thermal–hydraulic parameters of the HTR–PM are given. Measures to increase the thermal power are introduced. Thermal–hydraulic analysis results are presented. The results show that, from the viewpoint of thermal–hydraulics, it is possible to increase the reactor power.     

液态金属在堆芯子通道内的湍流换热

采用Speziale-Sarkar-Gatski(SSG)雷诺应力模型对液态金属在堆芯子通道内的流动、传热过程进行计算流体动力学(Computational Fluid Dynamics,CFD)模拟,研究雷诺数(Re)、分子普朗特数(Pr)、格拉晓夫数(Gr)、节径比(P/D)等无量纲参数对湍流换热的影响。比较无量纲对流换热系数(Nu)可以看出,CFD预测值与实验值及经验关系式符合得较好。对各种不同无量纲参数下的计算结果进行分析发现:在P/D和Re数相同条件下,三角形子通道的壁面温度分布比方形更均匀,换热情况更好;提高Re数,增大P/D,选用Pr数大的冷却剂,可有效改善温度和换热的周向分布不均情况;在Re数大于10 000的条件下,浮力对液态金属换热的影响可忽略不计。     

CFD analysis of thermal-hydraulic behavior in SCWR typical flow channels

Investigations on thermal-hydraulic behavior in SCWR fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding and ability to predict the heat transfer behavior of supercritical water. In this paper, CFD analysis is carried out to study the flow and heat transfer behavior of supercritical water in sub-channels of both square and triangular rod bundles. Effect of various parameters, e.g. thermal boundary conditions and pitch-to-diameter ratio on the thermal-hydraulic behavior is investigated. Two boundary conditions, i.e., constant heat flux at the outer surface of cladding and constant heat density in the fuel pin are applied. The results show that the structure of the secondary flow mainly depends on the rod bundle configuration as well as the pitch-to-diameter ratio, whereas, the amplitude of the secondary flow is affected by the thermal boundary conditions, as well. The secondary flow is much stronger in a square lattice than that in a triangular lattice. The turbulence behavior is similar in both square and triangular lattices. The dependence of the amplitude of the turbulent velocity fluctuation across the gap on Reynolds number becomes prominent in both lattices as the pitch-to-diameter ratio increases. The effect of thermal boundary conditions on turbulent velocity fluctuation is negligibly small. For both lattices with small pitch-to-diameter ratios (P/D < 1.3), the mixing coefficient is about 0.022. Both secondary flow and turbulent mixing show unusual behavior in the vicinity of the pseudo-critical point. Further investigation is needed. A strong circumferential non-uniformity of wall temperature and heat transfer is observed in tight lattices at constant heat flux boundary conditions, especially in square lattices. In the case with constant heat density of fuel pin, the circumferential conductive heat transfer significantly reduces the non-uniformity of circumferential distribution of wall temperature and heat transfer, which is favorable for the design of SCWR fuel assemblies.     

Research results on the corrosion effects of liquid heavy metals Pb, Bi and Pb–Bi on structural materials with and without corrosion inhibitors

The following article is a summary of research results of structural material corrosion in liquid heavy metals in the Czech Republic. The effect of alloying elements on the corrosion resistance of various types of steel in lead (Pb), bismuth (Bi) and lead–bismuth (Pb–Bi) was examined. Furthermore, the effect of temperature, temperature gradient, liquid–alloy composition and the flow velocity on a particular material corrosion resistance with and without the use of a corrosion inhibitor Ti and Zr was described. Types of steel affected by inhibition and their maximum operating temperatures are listed.     

Thermal–hydraulic analysis for a sodium-heated steam generator using a multi-shell method

A multi-shell analysis method has been applied to predict the Thermal–hydraulics in a steam generator of a liquid metal reactor. The method is intended to improve the calculation accuracy for temperature profiles at a wide range of heat exchange rates in a large sized steam generator for the future plant. The calculation accuracy has been examined using experimental 50 MWth steam generator test data that were measured during 1970's–1980's. The calculation results of temperature profiles were compared with experimental data at 20, 30, 50, 75, 90, and 100% of nominal operating condition. Responses for the stepwise flow rate variations were also evaluated. In conclusion, the calculation accuracy for the temperature profile in the steam generator was improved by using the multi-shell analysis method for a wide range of heat exchange rates.