Thermal-Hydraulic Studies of the Steam Separation in Horizontal Steam Generator at PGV Test Facility

A new design of nuclear power plant (NPP) with pressurized water reactor “NPP-2006” was developed in Russia. It represents the evolutionary development of the designs of NPPs with VVER-1000 reactors. Horizontal steam generator PGV-1000 MKP with in-line arrangement of the tube bundles will be used in “NPP-2006”. PGV test facility was constructed at the Electrogorsk Research and Engineering Center on NPP Safety (EREC) to investigate the process of the steam separation in steam generator. The description of the PGV test facility and tests, which will be carried out at the facility in 2009, are presented in this paper. The experimental results will be used for verification of the 3D thermal-hydraulic code STEG, which is developed in EREC. STEG pretest calculation results are presented in the paper.

Experimental Investigation of the Influence of Flow Obstacles on Post-Dryout Heat Transfer in an Annulus

This paper describes the experimental setup, instrumentation and procedures which have been developed in the thermal-hydraulic laboratory at the Royal Institute of Technology (KTH), Stockholm, Sweden, to perform new post-dryout heat transfer investigations in an annulus with flow obstacles. Previous investigations performed in the same laboratory indicated that flow obstacles had a considerable influence on the post-CHF heat transfer. The measured heat transfer enhancement was significantly under-predicted by existing models. However, the net effect of obstacles could not be deduced from the measurements, since reference - obstacle-free measurements - had not been performed. In addition, the number of thermocouples that could be installed inside the heated rod was limited to 8. These deficiencies have been removed in the current approach. Firstly, the present design of the test section allows for measurements both with and without flow obstacles. In this way the net effect of the obstacles will be captured. Secondly, a newly developed technique allowed the installation of 40 thermocouples inside of the heated rod. An additional 40 thermocouples have been installed on the external wall of the heated tube. Therefore, a significant improvement of the accuracy of measurements can be expected. The present arrangement of instrumentation is suitable to perform measurements of heat transfer under both steady-state and transient conditions.

Recent Design Technologies of Moisture Separator Reheater

This paper introduces the development of the current model Moisture Separator Reheater (MSR) for nuclear power plant (NPP) turbines, commercially placed in service in the period 1984–1997, focusing on the mist separation performance of the MSR along with drainage from heat exchanger tubes. A method of predicting the mist separation performance was devised first based on the observation of mist separation behaviors under an air-water test, then developed for the application to predict under the steam conditions, followed by the verification in comparison with the actual results of a steam condition test. The instability of tube drainage associated with both sub-cooling and temperature oscillation, which may adversely affect the seal welding of tubes to tube sheet owing to thermal fatigue, was measured on an existing unit both to clarify the behaviors and to develop a method to suppress them. Both methods were applied to current model MSR and the effectiveness of the methods was demonstrated. A new concept MSR for 1,700 MW class APWR units is put in perspective based on the technologies, alongside a multidisciplinary optimum design evaluating the heat exchanger tube bundle.

Heat Transfer to Supercritical Water in Vertical Annular Channel and 3-Rod Bundle

The results of heat transfer to supercritical water flowing upward in a vertical annular channel (1-rod channel) and tight 3-rod bundle consisting of the tubes of 5.2-mm outside diameter and 485-mm heated length are presented. The heat-transfer data were obtained at pressures of 22.5, 24.5, and 27.5 MPa, mass flux within the range from 800 to 3000 kg/m2·s, inlet temperature from 125 to 352°C, outlet temperature up to 372°C and heat flux up to 4.6 MW/m2 (heat flux rate up to 2.5 kJ/kg). Temperature regimes of the annular channel and 3-rod bundle were stable and easily reproducible within the whole range of the mass and heat fluxes, even when a deteriorated heat transfer took place. The data resulted from the study could be applicable for a reference estimation of heat transfer in future designs of fuel bundles.

Effect of Nanofluid on the Film Boiling Behavior at Vapor Film Collapse

A high-temperature stainless-steel sphere was immersed into Al2O3 nanofluid to investigate film boiling heat transfer and collapse of vapor film. Surface temperature is referred to the measured value of thermocouples embedded into and welded onto a surface of the sphere. A direct contact between the immersed sphere and Al2O3 nanofluids is quantified by the acquired electric conductivity. The Al2O3 nanofluid concentration is varied from 0.024 to 1.3 vol%. A film boiling heat transfer rate of Al2O3 nanofluid is almost the same or slightly lower than that of water. A quenching temperature rises slightly with increased the Al2O3 nanofluid concentrations. In both water and Al2O3 nanofluid, the direct contact signals between the sphere and coolant were not detected before vapor film collapse.

General Safety Design Approach and Main Safety Features of ATMEA1

ATMEA, an AREVA NP and MHI joint venture offers ATMEA1, an innovative and reliable mid-sized evolutionary reactor based on its parent companies’ experience and technology. With a net electrical output between 1100 and 1150 MWe, ATMEA1 is aimed at ensuring competitive power generation. Its operational flexibility and features make it particularly attractive for countries with small or medium size grid. The presentation will focus on and emphasize the general safety design approach and main safety features of this Generation III+ Pressurized Water Reactor (PWR). High level safety, operational objectives, licensability and public acceptance have driven the design: Buildings and all safety systems are being designed in compliance with stringent regulations and worldwide accepted codes and standards, making ATMEA1 a licensable product in any country around the world. Public acceptance is a key element in the nuclear world. Supported by safety features like large commercial airplanes crash protection and resistance to severe accident, ATMEA1 is a robust generation III+ reactor.

CIAU Methodology and BEPU Applications

Best-Estimate calculation results from complex thermal-hydraulic system codes (like Relap5, Cathare, Athlet, Trace, etc..) are affected by unavoidable approximations that are unpredictable without the use of computational tools that account for the various sources of uncertainty. Therefore the use of best-estimate codes within the reactor technology, either for design or safety purposes, implies understanding and accepting the limitations and the deficiencies of those codes. Uncertainties may have different origins ranging from the approximation of the models, to the approximation of the numerical solution, and to the lack of precision of the values adopted for boundary and initial conditions. The amount of uncertainty that affects a calculation may strongly depend upon the codes and the modeling techniques (i.e. the code’s users). A consistent and robust uncertainty methodology must be developed taking into consideration all the above aspects. The CIAU (Code with the capability of Internal Assessment of Uncertainty) and the UMAE (Uncertainty Methodology based on Accuracy Evaluation) methods have been developed by University of Pisa (UNIPI) in the framework of a long lasting research activities started since 80’s and involving several researchers. CIAU is extensively discussed in the available technical literature, Refs. [1, 2, 3, 4, 5, 6, 7], and tens of additional relevant papers, that provide comprehensive details about the method, can be found in the bibliography lists of the above references. Therefore, the present paper supplies only ‘spot-information’ about CIAU and focuses mostly on the applications to some cases of industrial interest. In particular the application of CIAU to the OECD BEMUSE (Best Estimate Methods Uncertainty and Sensitivity Evaluation, [8, 9]) project is discussed and a critical comparison respect with other uncertainty methods (in relation to items like: sources of uncertainties, selection of the input parameters and quantification of their uncertainty ranges, ranking process, etc.) is presented.

Transient Flow Field on Containment Floor Predicted by Two-Dimensional Shallow Equation Solver

For the analysis of debris transport on containment floor, a model to predict the flow field should have a fast-running capability and high accuracy. A model is developed to calculate the transient flow field on the containment floor involving a complex geometry in the advanced pressurized water reactor (PWR) such as Advanced Power Reactor (APR)-1400, which does not have a switchover from injection to recirculation following a loss-of-coolant accident (LOCA). Two-dimensional shallow water equation (SWE) is solved using the finite volume method (FVM). Unstructured triangular meshes are used to simulate the complex structures on the containment floor. Harten-Lax-van Leer (HLL) scheme, one of the approximate Riemann solver, is adopted to capture the dry-wet interface and to determine the momentum flux at the interface. An experiment of a sudden dam break having water reservoir and L-shape open channel is simulated and compared with the calculated result, which supports the validity of the present model. The model is also applied to calculation of the flow field of APR-1400. The calculated flow field can be characterized by the propagation of waves generated by surface level difference and by the reflection of waves from solid wall. The transient flow rates entering to the Holdup Volume Tank (HVT) can be predicted within a practical limit of computational resource.

Differences in the Heat Transfer at the Core-Reflector Boundary of the PBMR Due to the Use of Different Numerical Methods

The PBMR (Pebble Bed Modular Reactor) is a High-Temperature Gas-cooled Reactor (HTGR) concept. One of the exercises of the PBMR benchmark of the Organization for Economic Cooperation and Development (OECD) is a steady state two-dimensional (2D) thermal-hydraulics simulation of a simplified PBMR with prescribed heat sources. Two different programs were used to model this exercise. They predicted similar core temperatures but the side reflector temperatures next to the core differed by more than 30 °C (when using a relatively coarse mesh). The underlying methods define temperatures at either vertices (VC) or at mesh cell centres (CC). A study was undertaken using one-dimensional (1D) implementations of the VC and CC methods to model a horizontal slice through the core. This study revealed the root cause of the different predictions. A modified version of the 1D CC method was developed that essentially predicts the same temperatures as the VC method. The extension of the modified method to two dimensions is under investigation. If the difference in predicted temperatures next to the core can be eliminated or reduced, then the focus can shift to other differences between the results of the two programs.

Newly Developed Two-Tip Optical-Fiber Probe for Accurate Measurement of High-Speed Micro-Droplets: Simultaneous Measurement of Their Velocities and Diameters

Optical fiber probing is very useful and reliable for bubbles/droplets measurement particularly in the gas-liquid two-phase flows that have dense dispersed phase and are impossible to be measured via usual visualization techniques. For the practical purpose of small- or medium-size bubbles/droplets measurement, one of the authors successfully developed a Four-Tip Optical-fiber Probe (F-TOP) and reported their excellent performance in industrial uses. Recently, particular demands for measuring properties of micro bubbles/droplets have increased in researches on multi-phase flows. However, no one succeeded in simultaneously measuring diameters and velocities of high-speed micro-droplets (velocity > 50 m/s; 50 μm < diameter < 500 μm). We made a challenge of measuring such tiny droplets via newly developed optical fiber probe equipped with two tips (Two-Tip Optical-fiber Probe: T-TOP). We have succeeded in this difficult measurement with it. Each optical fiber probe composing the T-TOP is made of a silica optical fiber (125 μm in external diameter, 50 μm in core diameter, 37.5 μm in clad thickness). The optical fiber was fine-drawn using a micro pipette puller, and this yielded a sub-μm-scale tip. The interval of the fiber axes and the gap of the tips were arranged depending on the droplets diameter range. In this paper, we demonstrate the performance of the T-TOP. First, we confirm its practicality in industrial use. The strength of the T-TOP is confirmed by exposure test of high-velocity and high-temperature steam flows. Second, we consider the influence of the flow on the measurement of T-TOP; the optical noise due to probe vibration by the high-velocity gas flow around the T-TOP is considered. Next, we confirm its performance using an orifice-type nozzle (300 μm < droplets diameter < 500 μm; droplets velocities < 40 m/s). We confirm the performance of the T-TOP; the results of T-TOP are compared with those of the visualization of the droplets by using an ultra-high-speed video camera. At the same time, we consider the process of droplet contact with the T-TOP via visualization of ultra-high-speed video camera.