Hidesada Tamai

Pressure Drop Experiments using Tight-Lattice 37-Rod Bundles

In order to design a Reduced-Moderation Water Reactor (RMWR) core from a thermal-hydraulic point of view, an evaluation method on the pressure drop in a tight-lattice rod bundle is required. In this study, axial pressure drops in tight-lattice 37-rod bundles were measured under conditions of 2-9 MPa in exit pressure and 200-1,000 kg/(m2·s) in mass velocity. The measured pressure drops were compared with calculated ones by the evaluation method with the Martinelli-Nelsons correlation. The comparison shows that a single-phase friction factor can be applied not only to a circular tube but also to a tight-lattice bundle except for an extremely small gap width. Then two-phase friction loss is a dominant component and accounts for about 60% of the pressure drop under an RMWR nominal operating condition. The evaluation method can evaluate effects of the flow area configuration (rod number, rod diameter, gap width) and axial power distribution under a wide range of flow conditions, and it can yield a good prediction of the pressure drop in a tight-lattice bundle.

Critical Power Experiment with a Tight-Lattice 37-Rod Bundle

Since most of critical power or CHF data have been collected in tube, annulus, or BWR geometries under BWR flow conditions, critical power data for highly tight and triangular lattice bundles under low mass velocity are indispensable for thermal-hydraulic design of Reduced-Moderation Water Reactor. Large-scale thermal-hydraulic experiments which use a basic 37-rod bundle test section (rod diameter: 13.0 mm, gap width between rods: 1.3 mm) were therefore carried out in this study within range of 2-9 MPa in pressure and 150-1,000 kg/(m2·s) in mass velocity. Fundamental characteristics of boiling transition were investigated through effects of flow parameter on critical power and those of rod number. It was confirmed that the fundamental characteristics in 37-rod bundle are similar to those in 7- rod bundle and in case of the BWR geometry. The results of the transverse non-uniform power distribution test and subchannel analysis suggest that the critical power becomes higher when the transverse local quality distribution closes to uniform.

Gap Width Effect on Critical Power based on Tight-Lattice 37-Rod Bundle Experiments

Since most critical power data have been collected for tube, annulus, or BWR geometries under BWR flow conditions, there is a lack of the critical power data for very tight triangular lattice bundles under low mass velocity flow conditions that is indispensable for the thermal-hydraulic design of the Innovative Water Reactor for FLexible Fuel Cycle (FLWR). Large-scale thermal-hydraulic experiments using two tight-lattice 37-rod bundle test sections with 1.0 and 1.3 mm gaps respectively were therefore carried out within the range of 2–9 MPa in pressure and 150–1200 kg/m2 s in mass velocity. It was confirmed that the fundamental characteristics of the flow parameter impact on critical power are similar between 1.0 and 1.3 mm gaps. Then, the gap width effect was discussed using the relationship between critical quality and mass velocity. No significant differences are recognized under high mass velocity conditions (>700 kg/m2 s), whereas the critical quality in the 1.0 mm gap experiments tends to be lower under low mass velocity conditions (>700 kg/m2 s) and the difference is less than 10%.

Critical Power Characteristics in 37-rod Tight Lattice Bundles under Transient Conditions

Critical power characteristics in the postulated abnormal transient processes that may be possibly met in the operation of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) were investigated for the design of the FLWR core. Transient Boiling Transition (BT) tests were carried out using two sets of 37-rod tight lattice rod bundles (rod diameter: 13 mm; rod clearance: 1.3 mm or 1.0 mm) at Japan Atomic Energy Agency (JAEA) under the conditions covering the FLWR operating condition (Pex = 7:2 MPa, Tin = 556 K) for mass velocity G = 400-800 kg/(m2 s). For the postulated power increase and flow decrease transients, no obvious change of the critical power against the steady one was observed. The traditional quasi-steady characteristic was confirmed to be working for the postulated power increase and flow decrease transients. The experiments were analyzed with TRAC-BF1 code, where the JAEA newest critical power correlation for the tight lattice rod bundles was implemented for the BT judgment. The TRAC-BF1 code showed good prediction for the occurrence or the non occurrence of the BT and for the exact BT starting time. The traditional quasi-steady state prediction of the BT in transient process was confirmed to be applicable for the postulated abnormal transient processes in the tight lattice rod bundles.

Effect of Rod Bowing on Critical Power Based on Tight-Lattice 37-Rod Bundle Experiments

The confirmation of thermal-hydraulic performance is one of the most important R&D requirements for the design of the Innovative Water Reactor for FLexible Fuel Cycle (FLWR). Since the effect of rod bowing on critical power has not been determined yet due to the lack of experimental data, a large-scale thermal-hydraulic experiment using a tight-lattice 37-rod bundle test section with a bowed rod was carried out with pressure ranging from 2–9 MPa and mass velocity at 200–1000 kg/(m2s). It was confirmed that boiling transition (BT) occurs downstream of the rod contact point, and that the wall temperature trace during the BT follows the typical BT pattern of BWR. The critical power with a bowed rod is about 10% lower than that without rod bowing. The critical power increases monotonically with the increase in mass velocity, with the decrease in inlet water temperature, and with the decrease in exit pressure, and these trends are similar to those of the basic bundle without rod bowing. Thus, there is a negligible effect of rod bowing on the dependence of critical power on the mass velocity, the inlet temperature, and the exit pressure.

Pressure Propagation and Critical Flow (Compressible Fluid Flow)

In the previous chapters, discussion was devoted mainly to a fluid for which there is no need to consider the fluid compressibility. In this chapter, we consider flows in a wide range of temperatures and pressures and treat each gas as a compressible fluid. In such a case, the sound velocity plays a unique role, and the Mach number of the flow becomes an important parameter that determines the flow characteristics. In addition to air, a wide variety of gases, including steam and combustion gases, are considered here. Most of these gases are not ideal gases; however, since they may be regarded as approximating ideal gases due to the high temperature, the computational formulas for ideal gases are described, unless otherwise noted. This treatment is an approximation; however, it is sufficient to present qualitative features and provide theoretical results with good reliability in many cases. In addition, the flow is assumed to be steady. Another treatment must be considered when an unsteady flow has an important role; however, since that leads to a complicated analysis and significantly limits the generality of the conditions considered, no explanation is given here.

Condensation Heat Transfer

When vapor comes in contact with the surface of an object kept at a temperature lower than the saturation temperature, it condenses into a liquid with the release of latent heat. The heat transfer accompanied by condensation is called condensation heat transfer. The components that utilize condensation heat transfer include the steam turbine condenser in use for nuclear power plants and the refrigerator condenser.

Laminar Flow and Turbulent Flow (The Similarity Rule)

When considering the fluid motion, it is convenient to express the fundamental equations in a dimensionless form. For simplicity, we consider a one-dimensional flow of incompressible fluid that flows perpendicular to the direction of gravitational force. In this case, the Navier-Stokes equation is expressed as the following equation: n n

Boiling Heat Transfer and Critical Heat Flux

frac{partial u}{partial t}+ufrac{partial u}{partial x}=-frac{1}{rho}frac{partial p}{partial x}+nu frac{partial^2u}{partial {x}^2}