稠密バンドル用沸騰遷移相関式の改良(NP3 新型炉技術)

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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 with a gap of about 1 mm between rods. 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 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 with a gap of about 1 mm between rods. 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 low mass velocity region (< 300 kg/m^2s), the correlation is written in critical quality-annular flow length type. For high mass velocity region (> 300 kg/m^2s), it is written in local critical heat flux-critical quality 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: gap between rods from 1.0 to 2.29 mm, heated length from 1.26 to 1.8 m, mass velocity from 150 to 2000 kg/m^2s and pressure from 2 to 11 Mpa.
一般社団法人日本機械学会の論文
2004-06-22

稠密バンドル用沸騰遷移相関式の改良(NP3 新型炉技術)

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