Akira Ohnuki

Experimental study on transition of flow pattern and phase distribution in upward air–water two-phase flow along a large vertical pipe

Abstract In order to investigate the dependency of gas–liquid two-phase flow on pipe scale, the transition characteristics of flow pattern and phase distribution were studied experimentally in upward air–water two-phase flow along a large vertical pipe (inner diameter D: 0.2 m, the ratio of pipe length to diameter L/D: 61.5). The experiments were conducted under the flow rate: 0.03 m/s ≤ superficial air velocity (at top of test section) ≤ 4.7 m/s, 0.06 m/s ≤ superficial water velocity JL ≤ 1.06 m/s. Flow pattern was observed and measurements were performed on axial differential pressure, phase distribution, bubble size and bubble and water velocities. The scale effect was discussed with small-scale data (D: 0.025–0.038 m). The flow conditions at which coalescence starts are almost the same as those found in small-scale pipes, but no large bubbles are observed in the region L/D 20. The churn flow is dominant in the large vertical pipe under the conditions where small-scale pipes have slug flow. The transition of phase distribution corresponds to the change of flow pattern. Large coalescent bubbles affect the phase distribution as similar to small-scale pipes but the following remarks are concluded as the scale effect: (1) under a low JL where small-scale pipes have a wall-peak phase distribution, a core-peak phase distribution is established, where some large eddies including bubble clusters fill up the pipe, (2) the large coalescent bubbles are developed along the test section via the churn bubbly flow where the phase distribution is a core peak one, whereas Taylor bubbles in small-scale pipes are generated at the vicinity of gas–liquid mixing region or are developed from the bubbly flow with a wall-peak phase distribution, (3) the wall-peak in the large vertical pipe is lower even under the same bubble size. The lower peak is considered to be related to the lower radial velocity gradient of water and the larger turbulent dispersion force.

Scale effects on countercurrent gas-liquid flow in a horizontal tube connected to an inclined riser

The scale effects of a flow path under countercurrent flow limitation (CCFL) (air/water or steam/saturated water) in a horizontal tube connected to an inclined riser have been studied. The studied geometry simulates that of a PWR hot leg. An analytical model with a two-fluid model was developed based on flow observation results in small scale experiments and then assessed with various scale experiments under various pressures to evaluate the scale effects. The assessments with small scale experiments revealed that the region controlling the flow limitation was shifted from the horizontal tube near the bend to the inclined riser part as the length of inclined riser increased. The degree of the shift became weaker for larger scale experiments under the assumption that the ratio of interfacial friction factor to wall-to-gas friction factor was maintained to be the same as that for small scale experiments. The degree of the shift was not affected by the change of pressure (0.3 MPa → 1.5 MPa).

Model Development for Bubble Turbulent Diffusion and Bubble Diameter in Large Vertical Pipes

Multi-dimensional analyses have been expected recently with expanding computation resources for gas-liquid two- phase flow analyses of advanced nuclear systems such as passive safety systems and natural-circulation-type reactors. However, the applicability of previous constitutive equations for multi-dimensional analyses has not been fully investigated especially for the effects of flow path scale because the equations have been assessed for small-scale experiments. In this study, we analyzed the scale effects by the multi-dimensional two-fluid model code using data in 38 mm and 200 mm diameter pipes. We clarified a key-parameter to model the scale effects and developed models for the effects on phase distribution. The scale effects can be classified by the relative relationship between bubble diameter db and turbulent length scale lT . Bubble-induced turbulence is increased under that db is smaller than lT and bubble coalescence is predominated rather than breakup under that lT is about three times larger than db and under higher void fraction. Based on these findings, we established new models for bubble turbulent diffusion and bubble diameter. The applicability was promising through assessments against the 38 mm and 200 mm pipes under different flow rates and against databases for 60 mm, 100 mm and 480 mm pipes.

Development of Analytical Procedures of Two-Phase Flow in Tight-Lattice Fuel Bundles for Innovative Water Reactor for Flexible Fuel Cycle

Abstract A research and development project to investigate thermal-hydraulic performance in the tight-lattice rod bundles of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been in progress at Japan Atomic Energy Agency in collaboration with power companies, reactor vendors, and universities since 2002. The FLWR can realize favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burnup, and long operation cycle, based on matured light water reactor technologies. Mixed-oxide fuel assemblies with tight lattice arrangement are used because they increase the conversion ratio by reducing the moderation of neutrons. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. Information about the effects of the gap width and grid spacer configuration on the flow characteristics in the FLWR core is still insufficient. Thus, we are developing procedures for qualitative analysis of thermal-hydraulic performance of the FLWR core using an advanced numerical simulation technology. In this study, an advanced two-fluid model is developed to economize on the computing resources. In the model, interface structures larger than computational cells (such as liquid film) are simulated by the interface tracking method, and small bubbles and droplets are estimated by the two-fluid model. In this paper, we describe the outline of this model and the numerical simulations we performed to validate the model performance qualitatively.

Critical Power Correlation for Tight-Lattice Rod Bundles

Developing design correlation for the prediction of critical power in rod bundles is indispensable for R&D of Reduced-Moderation Water Reactor (RMWR) which adopts a triangular tight-lattice fuel rod configuration and axially double-humped-heated profile. In this research, critical power correlation for tight-lattice rod bundles is newly proposed using 7-rod axially uniform-heated data, 7-rod and 37-rod axially double-humped-heated data at Japan Atomic Energy Research Institute (JAERI). For comparatively high mass velocity region, the correlation is written in local critical heat flux-critical quality type. For low mass velocity region, it is written in critical quality-annular flow length type. The standard deviation of ECPR (Experimental Critical Power Ratio) to the whole JAERI data (694 data points) is 6%. The correlation is verified by Bettis Atomic Power Laboratory data (177 points, standard deviation: 7.7%). The correlation is confirmed being able to give good prediction for the effects of mass velocity, inlet temperature, pressure and heated equivalent diameter on critical power. The applicable range of the correlation is: Rod number lower than 37, rod gap from 1.0 to 2.29 mm, heated length from 1.26 to 1.8 m, mass velocity from 150 to 2,000 kg/m2·s and pressure from 2 to 11 MPa.

Carryover characteristic during reflood process in large scale separate effects tests

Abstract A study on carryover characteristic in the core and upper plenum during the reflood phase in PWR-LOCA was performed with the use of data from the slab core test facility (SCTF) having eight rod bundles scale. Void fraction distribution in the core was strongly related to quench propagation in rod bundles. A correlation for mass effluent rate out of core was derived from the void fraction distribution characteristic. The correlation was found to be widely applicable. On the other hand, the capture of entrained liquid in the upper plenum by structures and water pool is below 30% of the entrainment mass flow rate during most of the reflood phase and increase when the steam velocity decreases. Since entrainment rate into hot leg increases with increase of liquid flow out of the core, the reflood velocity should tend to be suppressed with time because of stronger steam binding effect.

ROSA/LSTF Tests and RELAP5 Posttest Analyses for PWR Safety System Using Steam Generator Secondary-Side Depressurization against Effects of Release of Nitrogen Gas Dissolved in Accumulator Water

Two tests related to a new safety system for a pressurized water reactor were performed with the ROSA/LSTF (rig of safety assessment/large scale test facility). The tests simulated cold leg small-break loss-of-coolant accidents with 2-inch diameter break using an early steam generator (SG) secondary-side depressurization with or without release of nitrogen gas dissolved in accumulator (ACC) water. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. The pressure difference between the primary and SG secondary sides after the actuation of ACC system was larger in the test with the dissolved gas release than that in the test without the dissolved gas release. No core uncovery and heatup took place because of the ACC coolant injection and two-phase natural circulation. Long-term core cooling was ensured by the actuation of low-pressure injection system. The RELAP5 code predicted most of the overall trends of the major thermal-hydraulic responses after adjusting a break discharge coefficient for two-phase discharge flow under the assumption of releasing all the dissolved gas at the vessel upper plenum.

Feasibility Study on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles for Reduced-Moderation Water Reactors

R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute (JAERI) in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R&D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R&D plan and describe the current status focused on an experimental study using large-scale (37-rod bundle) test facility. Steady-state critical power experiments are conducted with the test facility and the experimental data reveal the feasibility of RMWR.Copyright

A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA

A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation PWR can be maintained.Copyright

Numerical investigation of heat transfer enhancement phenomenon during the reflood phase of PWR-LOCA

The heat transfer in higher power bundles was enhanced in large-scale reflood tests at Japan Atomic Energy Research Institute. The heat transfer enhancement in the core under a radial power distrib...