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.

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.

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.

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.

Transient Response of BWR Core Flow During Simulated Generator Load Rejection Event

Integral Effects Test (IET) was conducted to investigate the effects of flow redistribution during the generator load rejection event by using the SIRIUS-F facility, which simulates boiling two-phase flow in a BWR core. Owing to the automatic controllers of a recirculation pump inverter and fine-control valves in the facility, the time series of signals of heat flux and mass flux were observed to agree well with those of target rapid flow-decrease events in the previous experimental series. This paper addresses the simulated generator load rejection event, during which the flow and power gradually decrease and the flow takes a turn toward recovery. As a result of the two-parallel channel experiment, mass flux of a hot channel is lower than that of the other during the initial stage. When the void fraction becomes smaller, mass flux of the hot channel is observed to become higher. This phenomenon can be accurately demonstrated with the TRAC-BF1 code as well. The code does, therefore, predict the boiling two-phase flow in a BWR core even at such flow-decrease event. During the event, differential pressure along each channel between the upper and lower plena decreases by several tens of kPa. The relative perturbations of the differential pressure between both channels, however, remain less than 0.4%, which is a significantly small amount. In conclusion, the differential pressures between the upper and lower plena of two-parallel channels are, therefore, identical to each other regardless of the power.

RETRAN Application of Turbine Trip and Load Rejection of Startup Test Analysis for Lungmen ABWR

This study used RETRAN program to analyze the turbine trip and load rejection transients of Taiwan Power Company Lungmen Nuclear Power Plant’s startup test at 100% power and 100% core flow operating condition. This model includes thermal flow control volumes and junctions, control systems, thermal hydraulic models, safety systems, and 1D kinetics model. In Lungmen RETRAN model, four steam lines are simulated as one line. There are four simulated control systems: pressure control system, water level control system, feedwater control system, and speed control system for reactor internal pumps. The turbine trip event, at above 40% power, triggers the fast open of the bypass valves. Upon the turbine trip, the turbine stop valves close. To minimize steam bypassed to the main condenser, recirculation flow is automatically runback and a SCRRI (selected control rod run in) is initiated to reduce the reactor power. The load rejection event causes the fast opening of the bypass valves. Steam bypass will sufficiently control the pressure, because of their 110% bypass capacity. A SCRRI and RIP runback are also initiated to reduce the reactor power. This study also investigated the sensitivity analysis of turbine bypass flow, runback rate of RIPS and SCRRI to observe how they affect fuel surface heat flux, neutron flux and water level, etc. The results show that turbine bypass flow has larger impacts on dome pressure than RIPS runback rate and SCRRI. This study also indicates that test criteria in turbine trip and load rejection transients are met and Lungmen RETRAN model is performing well and applicable for Lungmen startup test predictions and analyses.

Post-Test Investigations of BFBT Critical Power Tests With Trace

The validation of heat transfer models of safety analysis codes such as TRACE is very important due to the strong interaction of the thermal hydraulics parameters with the core neutronics. TRACE is the reference system code of the US NRC for LWR. It is being developed and extensively validated within the international CAMP-program. In this paper, the validation of heat transfer models of TRACE related to the prediction of the critical power is presented. The validation is based on a large number of critical power tests performed in the NUPEC BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility in Japan. These tests were analysed with the TRACE Version 5 RC 2. The comparison of predictions with the experimental data shows good agreement. The developed TRACE model and the comparison of experimental data with code results will be presented and discussed.