Critical Heat Flux in Tight Lattice Core of Reduced-Moderation Water Reactor (RMWR)

Kureta, Masatoshi; Araya, Fumimasa; Ohnuki, Akira; Ōkubo, Tsutomu; Iwamura, Takamichi; Akimoto, Hazime; Yamamoto, Kumiko; Okada, Hiroyuki

doi:10.1299/jsmepes.2002.8.575

Cited by 5 publications

(8 citation statements)

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“…Thereby, the local heat flux is considered being an important parameter in developing BT correlation, especially for axial non-uniform heated condition (For axial uniform heated condition, local heat flux is actually can be substituted by critical power, with only a coefficient needed in the substitution). Actually, although Arai correlation can give a good prediction in the present research for the data in the range 350 < G < 500 kg/m 2 s near P ex = 7.2 MPa, this success is considered being quite a coincidence, because the correlation under-predicts the data that are in the same range under axial uniform heated condition (5) . With using the thought of Arai correlation and developing correlation by doing revision to Arai correlation, we find it is very difficult to develop a new correlation that can give good predictions to both JAERI axial uniform and non-uniform heated data.…”

Section: Discussionmentioning

confidence: 46%

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Critical Power in 7-Rod Tight Lattice Bundle

Liu¹,

Kureta²,

Akimoto³

2004

JSME international journal. Ser. B, Fluids and thermal engineer

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The Reduced-Moderation Water Reactor (RMWR) has recently becomes of great concern. The RMWR is expected to promote the effective utilization of uranium recourse. The RMWR is based on water-cooled reactor technology, with achieved under lower core water volume and water flow rate. In comparison with the current light water reactors whose water-to-fuel volume ratio is about 2 -3, in the RMWR, this value is reduced to less than 0.5. Thereby, there is a need to research its cooling characteristics. Experimental research on critical power in tight lattice bundle that simulates the RMWR has been carried out in Japan Atomic Energy Research Institute (JAERI). The bundle consists one center rod and six peripheral rods. The 7 rods are arranged on a 14.3 mm equilateral triangular pitch. Each rod is 13 mm in outside diameter. An axial 12-step power distribution is employed to simulate the complicate heating condition in RMWR. Experiments are carried out under G = 100 -1 400 kg/m 2 s, P ex = 2 -8.5 MPa. Effects of mass velocity, inlet temperature, pressure, radial peaking factor and axial peaking factor on critical power and critical quality are discussed. Compared with axial uniform heating condition, the axial non-uniform heating condition causes an obvious decrease in critical quality. Arai correlation, which is the only correlation that has been optimized for tight lattice condition, is verified with the present experimental data. The correlation is found to be able to give reasonable prediction only around RMWR nominal operating condition.

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Section: Discussionmentioning

confidence: 46%

“…F r is radial peaking factor. Arai correlation has been reported a not-satisfied prediction accuracy to the critical power data carried in JAERI under axial uniform heated condition (5) . In following, we'll verify the Arai correlation for the data taken for the present axial non-uniform heated condition.…”

Section: Verification Of the Arai Correlationmentioning

confidence: 95%

Critical Power in 7-Rod Tight Lattice Bundle

Liu¹,

Kureta²,

Akimoto³

2004

JSME international journal. Ser. B, Fluids and thermal engineer

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“…Thereby, critical power experiments in tight-lattice rod bundles, for both axially uniform-heated and double-humpedheated condition, have been carried out at JAERI using 7-rod and 37-rod bundles over a relatively extensive flow condition. [3][4][5] Figure 3 shows the cross-sectional views of the JAERI 7-rod (for both axially uniform-heated and doublehumped-heated test sections) and 37-rod test sections. The descriptions to the 3 test sets are described in Table 1.…”

Section: Introductionmentioning

confidence: 99%

Critical Power Correlation for Tight-Lattice Rod Bundles

Liu

Kureta

Ohnuki

et al. 2005

Journal of Nuclear Science and Technology

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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 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/m 2 Ás and pressure from 2 to 11 MPa.

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Section: Introductionmentioning

confidence: 99%

“…In the present study, boiling transition (BT) prediction performance of the subchannel analysis code COBRA-TF improved by JAERI 7) was assessed for the series of tight-lattice CHF experiments performed in JAERI 5,8) . 5,8) CHF experiments for the tight-lattice rod bundles have been conducted. The aims of the experiments are (1) to make clear the parameter effects on critical power, (2) to evaluate the existing design correlation, (3) to develop the high accuracy correlation and (4) to verify the detailed numerical analysis code.…”

Section: Introductionmentioning

confidence: 99%

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

Nakatsuka

Kureta

Ōkubo

et al. 2002

The Proceedings of the JSME Annual Meeting

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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/(m 2 s), 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/(m 2 s), 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. Ch2 Ch3,5 Ch4 Elevation, z [m] 1.8 Calculated dryout Pex = 7.2 [MPa] G = 500 [kg/(m 2 s)] T in = 282.5 [ o C] Ch2 Ch3,5 Ch4 Elevation, z [m] 1.8 Pex = 7.2 [MPa] G = 500 [kg/(m 2 s)] T in = 282.5 [ o C]

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Critical Heat Flux in Tight Lattice Core of Reduced-Moderation Water Reactor (RMWR)

Cited by 5 publications

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Critical Power in 7-Rod Tight Lattice Bundle

Critical Power in 7-Rod Tight Lattice Bundle

Critical Power Correlation for Tight-Lattice Rod Bundles

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

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