Takuji Nagayoshi

Spacer effect model for subchannel analysis : Turbulence intensity enhancement due to spacer

Mean velocity and velocity fluctuation in a test channel that consisted of five subchannels with and without ferrule-type spacer were measured using air as a working fluid, to clear turbulence intensity enhancement due to spacer. Measurements were performed at Reynolds number of 0.5-1.2x10 5 , which simulated vapor flow velocity of annular-dispersed flow in BWR condition. It was confirmed that magnitudes of velocity fluctuations in radial direction were proportional to Reynolds number and square root of friction factor downstream from a spacer. New spacer effect model to describe turbulence intensity enhancement due to the spacers was developed. In the model, dependence of the velocity fluctuation on ferrule thickness was correlated by blockage ratio. It was found that the present spacer model is applicable to prediction of turbulence intensity enhancement due to spacer.

Three-dimensional time-averaged void fraction distribution measurement technique for BWR thermal hydraulic conditions using an X-ray CT system

We have developed a void fraction distribution measurement technique using the three-dimensional (3D) time-averaged X-ray computed tomography (CT) system to understand two-phase flow behavior inside a fuel bundle for boiling water reactor (BWR) thermal hydraulic conditions of 7.2 MPa and 288 °C. As a first step, we measured the 3D void fraction distribution in a vertical square (5 × 5) rod array that simulated a BWR fuel bundle in the air–water test. A comparison of the volume-averaged void fractions evaluated by the developed X-ray CT system with those evaluated by a differential pressure transducer showed satisfactory agreement within a difference of 0.03. Thus, we confirmed that the developed system could be used to get 3D imaging of the vertical square rod array used in the test under the BWR operating pressure condition. In the next step, we did a verification test using the vertical pipe (11.3 mm ID) for BWR thermal hydraulic conditions. A comparison of the cross-sectional-averaged void fractions evaluated by the X-ray CT system with those evaluated by the drift-flux model showed good agreement within a difference of 0.05. We confirmed that the evaluated void fraction distribution forms in the horizontal cross section changed with the quality in response to the flow regime transition.

Simulation of Multi-dimensional Heterogeneous and Intermittent Two-Phase Flow by Using an Extended Two-Fluid Model

A new gas-liquid two-phase flow simulation method has been developed based on the extended two-fluid model, which has capabilities of both the two-fluid and the interface-tracking models. The VOF (Volume of Fluid) technique has been introduced for suitable interface calculations. Interfaces of free surface and large bubbles are calculated directly by solving transport of a steep void fraction gradient corresponding to interface, while averaged behavior of microscopic dispersed bubbles and droplets are calculated in the two-fluid model scheme. It is expected that the present method can treat effects of significant kinetic interaction between the phases directly without empirical correlations. The calculated propagation of wet front in a dam break problem is close to experimental data. The predicted flow patterns of complex gas-liquid two-phase flow in a flat tube are quite similar to observations with a video camera. The present simulation will be a useful tool for predictions of integral behavior of thermal-hydraulic phenomena in large-scale nuclear power plants.

Development of a transient boiling transition analysis method based on a film flow model

A new single-channel, transient boiling transition (BT) prediction method based on a film flow model has been developed for a core thermal-hydraulic code. This method could predict onset and location of dryout and rewetting under transient conditions mechanically based on the dryout criterion and with consideration of the spacer effect. The developed method was applied to analysis of steady-state and transient BT experiments using BWR fuel bundle mockups for verification. Comparisons between calculated results and experimental data showed that the developed method tended to predict occurrence of rewetting earlier, however, onset time of BT and maximum rod surface temperature were well predicted within 0.6 s and 20°C, respectively. Moreover, it was confirmed that consideration of the spacer effect on liquid film flow rate on the rod surface was required to predict dryout phenomena accurately under transient conditions.

Preliminary test of an ultrasonic liquid film sensor for high-temperature steam–water two-phase flow experiments

A prototype liquid film sensor for high-temperature steam–water experiments has been developed. The sensor shape simulates a boiling water reactor (BWR) fuel rod. The pulse-echo method can be utilized to measure the thickness of the liquid film covering the sensor surface. A piezoelectric element is soldered onto the inside of the sensor casing which consists of two curved casing pieces. After the piezoelectric element is attached, the two casing pieces are laser welded together. It is confirmed that the temperature rise at the time of the laser welding does not influence soldering of the piezoelectric element. The pressure proof test shows that the sensor can be used at a high-pressure condition of 7 MPa. Simple air–water experiments are done at atmospheric pressure to confirm the liquid film thickness can be measured with the sensor. The fluctuation of the liquid film thickness is satisfactorily captured with the sensor. The minimum and maximum thicknesses are 0.084 and 0.180 mm, respectively. The amplitude of the waveform at 286 °C is predicted by the calculation based on the acoustic impedance. It is expected that the sensor is able to measure the liquid film thickness even at BWR operating conditions.

Safe and Simplified Boiling Water Reactor (SSBWR)

A safe and simplified BWR (SSBWR) has been developed as an innovative future reactor to provide a super-long life core of 20 years and to realize a passive core safety system with infinite grace period. Operability and maintainability can be largely improved by using the super-long life core, cutting the number of active components, and using a one-batch core with no exchange of fuel assemblies, which can also significantly reduce the possibility of nuclear proliferation. Np-237 of MAs (Minor Actinides) can be effectively transmuted using the very hard neutron spectrum of SSBWR and high level radioactive wastes can be reduced.Copyright

Application results of a prototype ultrasonic liquid film sensor to a 7 MPa steam–water two-phase flow experiment

A prototype ultrasonic liquid film sensor was applied to a high-temperature steam–water two-phase flow experiment. The liquid film sensor was vertically installed in a loop which was connected to HUSTLE, a multi-purpose steam source test facility. The hydraulic diameter of the measurement section was 9.4 mm. The output waveforms of the sensor were acquired with a digital oscilloscope. The fluid temperature and system pressure were kept at 288 °C and 7.2 MPa, respectively, during the experiment. The pulse-echo method was used to calculate the liquid film thickness. The cross-correlation calculation was utilized to determine the time difference between the pulse reflected at the sensor surface and the pulse reflected at the liquid film surface. The time-averaged liquid film thicknesses were less than 0.055 mm in the annular flow condition. The increase of the time-averaged thickness was small with the change of the gas momentum flux. The film thicknesses measured with the sensor were compared with the past experimental results; the former were smaller than one-fourth of the thickness estimated as the mean film thickness. The comparison results suggested that the continuous liquid sublayer thickness was measured with the liquid film sensor.

Measurement of Three-Dimensional Time-Averaged Void Fraction Distribution in Rod Bundle in Air-Water System by X-Ray CT Technique

We have been developing a void fraction distribution measurement technique using the three-dimensional (3D) time-averaged X-ray CT (computed tomography) system to understand two-phase flow behavior inside a fuel assembly for BWR (boiling water reactor) thermal hydraulic conditions of 7.2 MPa and 288 °C. Unlike CT images of a normal standstill object, we can obtain 3D CT images that are reconstructed from time-averaged X-ray projection data of the intermittent two-phase flow. We measured the 3D void fraction distribution in a vertical square (5 × 5) rod array that simulated a BWR fuel assembly in the air-water test. From the 3D time-averaged CT images, we confirmed that the void fraction at the center part of the channel box was higher than that near the channel box wall, and the local void fraction at the central region of a subchannel was higher than that at the gap region of the subchannel. A comparison of the volume-averaged void fractions evaluated by the developed X-ray CT system with those evaluated by a differential pressure transducer in a void fraction range from 0.05 to 0.40 showed satisfactory agreement within a difference of 0.03.Copyright

Numerical Simulation of Fluid Mixing Phenomena in Boiling Water Reactor Core Using Advanced Interface-Tracking Method

Thermal-hydraulic design of the current BWR is performed by correlations with empirical results of actual-size tests. Then, when the reactor of new design is developed, an actual size test is required to confirm or modify the correlations. Development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for BWRs using innovative two-phase flow simulation technology. In this study, detailed two-phase flow simulation code using advanced interface tracking method : TPFIT is developed. In this paper, the TPFIT code was applied to simulation of two-phase flow in modeled 2 subchannels of BWRs rod bundle, and the existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data.

Numerical Evaluation of Fluid Mixing Phenomena in Boiling Water Reactor Using Advanced Interface-Tracking Method

Thermal-hydraulic design of the current boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test that simulates its design is required to confirm or modify the correlations. Development of a method that enables the thermal-hydraulic design of nuclear rectors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, detailed two-phase flow simulation code using advanced interface tracking method: TPFIT is developed to get the detailed information of the two-phase flow. In this paper, firstly, we tried to verify the TPFIT code comparing with the existing 2-channel air-water mixing experimental results. Secondary, the TPFIT code was applied to simulation of steamwater two-phase flow in modeled two subchannels of current BWRs rod bundle. The fluid mixing was observed at a gap between the subchannels. The existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data. From the data, pressure difference between fluid channels is responsible for the fluid mixing, and effects of the time averaged and fluctuating pressure difference must be incorporated in the two-phase flow correlation for fluid mixing.Copyright