Masatoshi Kureta

Critical Power Correlation for Axially Uniformly Heated Tight-Lattice Bundles

Abstract Critical power experiments were carried out, and the critical power correlation for axially uniformly heated tight bundles has been derived based on the present experimental data and data sets measured by the Bettis Atomic Power Laboratory. The shape of the test section simulates the fuel assembly of the reduced-moderation water reactor (RMWR), which is a water-cooled breeder reactor with a core of the tight triangular fuel rod arrangement. The obtained correlation covers the following conditions: channel geometry (triangular arrangement bundle of 7 to 20 rods, 6.6 to 12.3 mm in rod diameter, 1.0- to 2.3-mm gap between rods, 1.37 to 1.8 m in heated length), mass velocity of 100 to 2500 kg/(m2s), inlet quality of –0.2 to 0, pressure of 2 to 8.5 MPa, and radial peaking factor of 0.98 to 1.5, which include uniform, center-peak, and liner transverse heat flux distribution data. An excellent agreement was obtained between the developed correlation and data (371 points) within an error of ±4.6%.

Study on point of net vapor generation by neutron radiography in subcooled boiling flow along narrow rectangular channels with short heated length

Abstract Point of net vapor generation (PNVG) was investigated based on the void fraction dataset obtained by high-frame-rate neutron radiography. The test channels used in the experiment were rectangular channels heated from one side with channel gap of 3 and 5 mm, channel width of 30 mm, and heated length of 100 mm. In this study, we discuss on (1) the determination of the instantaneous and time-averaged PNVG, (2) the effects of system parameters on PNVG, (3) the applicability of existing PNVG correlations to the channel with a short heated length, and (4) the effect of the PNVG in critical heat flux (CHF) model. The following results were obtained: (a) the effects of system parameters on the thermal equilibrium quality at the PNVG were small under the present conditions, (b) existing PNVG correlations tended to underestimate the thermal equilibrium quality at the PNVG in the channel with a short heated length, and (c) the prediction accuracy of Katto’s CHF model could be improved significantly by using the accurate PNVG.

Void fraction measurement in subcooled-boiling flow using high-frame-rate neutron radiography

Abstract A high-frame-rate neutron radiography (NR) technique was applied to measure the void fraction distribution in forced-convective subcooled-boiling flow. The focus was experimental technique and error estimation of the high-frame-rate NR. The results of void fraction measurement in the boiling flow were described. Measurement errors on instantaneous and time-averaged void fractions were evaluated experimentally and analytically. Measurement errors were within 18 and 2% for instantaneous void fraction (measurement time is 0.89 ms), and time-averaged void fraction, respectively. The void fraction distribution of subcooled boiling was measured using atmospheric-pressure water in rectangular channels with channel width 30 mm, heated length 100 mm, channel gap 3 and 5 mm, inlet water subcooling from 10 to 30 K, and mass velocity ranging from 240 to 2000 kg/(m2·s). One side of the channel was heated homogeneously. Instantaneous void fraction and time-averaged void fraction distribution were measured parametrically. The effects of flow parameters on void fraction were investigated.

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.

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

Development of Quantitative Analytical Procedures on Two-Phase Flow in Tight-Lattice Fuel Bundles for Reduced-Moderation Light-Water Reactors

R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) has started at Japan Atomic Energy Research Institute (JAERI) in collaboration with power companies, reactor vendors, universities since 2002. The RMWR is a light water reactor which a higher conversion ratio more than one can be expected. In order to attain this higher conversion ratio, triangular tight-lattice fuel bundles which gap spacing between each fuel rod is around 1 mm are required. As for the thermal design of the RMWR core, conventional analytical methods are no good because the conventional composition equations can not predict the RMWR core with high accuracy. Then, development of new quantitative analytical procedures was carried out. Those analytical procedures are constructed by model experiments and advanced two-phase flow analysis codes. This paper describes the results of the model experiments and analytical results with the developed analysis codes.Copyright

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.