Toru Nakatsuka

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.

Assessment of Applicability of Two-Fluid Model Code ACE-3D to Heat Transfer Test of Supercritical Water Flowing in an Annular Channel

A supercritical-water-cooled reactor (SCWR) is a high-temperature, high-pressure water cooled reactor that operates above the critical pressure of water. In order to perform efficiently the thermal design of the SCWR, it is important to assess the thermal hydraulics in rod bundles of the core. Experimental conditions of mockup tests, however, may be limited because of technical and financial reasons. Therefore, it is required to establish an analytical design technique that can extrapolate experimental data to various design conditions of the reactor. Japan Atomic Energy Agency (JAEA) has improved the three-dimensional two-fluid model analysis code ACE-3D, which was originally developed for the two-phase flow thermal hydraulics of light water reactors, to handle the thermal hydraulic properties of water in the supercritical region. In the present study, heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which were performed at JAEA, were analyzed with the improved ACE-3D to assess the prediction performance of the code. As a result, it was implied that the ACE-3D code is applicable to the prediction of wall temperatures of a single rod that simulates the fuel bundle geometry of the SCWR core.

Subchannel Analysis of CHF Experiments for Tight Lattice Core with COBRA-TF

Reduced-Moderation Water reactor (RMWR) is an innovative light water reactor developed by Japan Atomic Energy Research Institute (JAERI). The RMWR comprises tight-lattice fuel assemblies with gap clearance around 1.0 mm for reduction of the water volume ratio to achieve a high conversion ratio. It is important to evaluate the thermal margin of the tight-lattice core. Subchannel analyses are expected to be useful to prediction of critical heat flux (CHF) and to provide valuable information to supplement thermal hydraulic experiments. In the present study, to assess the applicability of subchannel analysis for tight-lattice cores, series of tight-lattice CHF experiments performed in JAERI were analyzed with COBRA-TF code. For the axially uniform heated tight-lattice rod bundle, COBRA-TF gives good prediction of critical power for mass velocity of around 500 kg/(ms), while it underestimates the critical power for lower mass velocity and overestimates for higher mass velocity. Predicted axial positions at BT corresponded with those of the experiments axially. However, the predicted subchannel position was outer channels and differed from the measured position. For the axially double-humped heated bundle, COBRA-TF gives good prediction of critical power for mass velocity of around 200 kg/(ms), and overestimates for higher mass velocity. It turned out that the two-phase multiplier of friction loss have a large influences on the flow distribution among the subchannels. To improve the calculation accuracy, it is required to predict precisely the flow distribution including the prediction of pressure distribution in a tight-lattice bundle system.