Critical Power Evaluation with Consideration of Differences in Droplet Behavior due to Spacer Shape Influence.

Kanazawa, Toru; Nishida, Koji; Nagayoshi, Takuji; Kawasaki, Terufumi; Yokomizo, Osamu

doi:10.3327/jnst.32.301

Cited by 4 publications

(5 citation statements)

References 5 publications

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“…15(b) means the transport rate of liquid droplets to the liquid film on the rod surface. Kanazawa et al's analysis with inviscid fluid assumption showed the peak of the deposition rate in the spacer region, 8) whereas the present analysis in which the fluid is treated as the viscid fluid showed the peak downstream from the spacer. These droplet deposition characteristics obtained from the present analysis correspond to the experimental result conducted by Nishida et al…”

Section: Spacer Effectcontrasting

confidence: 44%

“…The basic concept in the present code to analyze the spacer effect is the same as that in the HIJET-AFIMA code. 8) However, the steam in an annular-dispersed flow region is assumed to be inviscid in the HIJET-AFIMA code, whereas it is treated as a viscid fluid in the present code. Parallel computation is applied only to the calculation of the spacer effect, since the other analytical steps do not require parallel computation because of very short computation time even with a single processor.…”

Section: Code Structurementioning

confidence: 99%

“…In addition, the inviscid fluid assumption was applied in their method to calculate flow behavior around spacers. In their calculation, the peak of droplet deposition rate on a liquid film appeared in the spacer region, 8) although the experiments showed that a liquid film thickened downstream from spacers.…”

Section: Introductionmentioning

confidence: 99%

“…However, an empirical parameter, which was expressed as "the model constant" in their paper, must be decided by referring to critical power experimental results. A method taking account of the spacer effect without any tuning parameters was developed by Kanazawa et al 8) In their method, critical power was calculated with the SILFEED code in which the spacer effect was incorporated as a parameter which was separately calculated with a different code called HIJET-AFIMA. Namely, they had to use two independent codes off-line to evaluate BWR critical power.…”

Section: Introductionmentioning

confidence: 99%

See 3 more Smart Citations

Critical Power Analysis with Mechanistic Models for Nuclear Fuel Bundles, (I)

Naitoh¹,

Ikeda²,

Nishida

et al. 2002

Journal of Nuclear Science and Technology

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The critical power analysis code for BWR fuel bundles, "CAPE-BWR", was developed. The objective of the development is to predict dryout phenomena of liquid film on fuel rod surfaces without tuning any parameters even for fuel bundle design improvements. The major features of the code are modular structure with mechanistic models and parallel computation. The calculation methods were divided into three steps: subchannel, liquid film flow and spacer effect analyses. The code was validated by the rod bundle test analyses. The overall comparison of calculated critical power with 166 measured data points showed −0.3% average difference with the standard deviation of 6.3%. The spatial domain decomposition method was applied for parallel computation of the spacer effect analysis. The parallelization efficiency was about 80%. The calculated dryout location agreed well with the measured one at the fullscale 8×8 bundle test. The code could trace the tendencies of the critical power depending on power distribution, spacer geometry and fluid conditions within a practical range of difference. From the calculation, difference of the critical power due to the spacer geometry was clarified to be caused by the difference of droplet deposition characteristics onto the liquid film.

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Section: Spacer Effectcontrasting

confidence: 44%

Section: Code Structurementioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

See 2 more Smart Citations

Critical Power Analysis with Mechanistic Models for Nuclear Fuel Bundles, (I)

Naitoh¹,

Ikeda²,

Nishida

et al. 2002

Journal of Nuclear Science and Technology

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“…Therefore, subchannel analysis is unable to predict the critical power considering the change in spacer geometry. Yamamoto et al [3] and Kanazawa et al [4] estimated a deposition enhancement coefficient due to the spacer by analyzing the behavior of droplets around the spacer using steam flow analysis. They applied the estimated coefficient to subchannel analysis and were relatively successful.…”

Section: Introductionmentioning

confidence: 99%

Prediction method of critical power by film flow rate measurement and subchannel analysis

Akiba

Takamasa

Morooka

2005

Heat Trans. Asian Res.

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This paper presents a method that can estimate the critical power of boiling water reactors, BWRs, with regard to spacer geometry. The current experimental method for estimating the critical power for BWR design requires many trained experts and expensive facilities to conduct the experiments. In the present method, the liquid film flow rate of adiabatic gas-liquid two-phase flow and a subchannel analysis of the actual BWR flow condition are measured experimentally and analyzed. In the experiment, deposition enhancement coefficients of three spacer geometries-a ferrule, an egg-crate, and a ferrule spacer with twisted tape (CYCLONE spacer)-were estimated by measuring the liquid film flow rate of air-water two-phase flow flowing up in a vertical square (4×4) rod bundle that simulated the rod bundle of a BWR. Using these coefficients, the critical powers for bundles using each type of spacer geometry were calculated in the subchannel analysis. This method was validated using previous critical power data in the actual BWR flow condition. The critical powers predicted by this method agreed well with those of the experimental data. The result confirmed the effectiveness of this experiment-simulation combined method, as well as the advantage over current experimental methods in terms of human and facility costs.

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A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

Chandraker

Vijayan

Sinha

et al. 2012

JPES

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The critical power corresponding to the Critical Heat Flux (CHF) or dryout condition is an important design parameter for the evaluation of safety margins in a nuclear fuel bundle. The empirical approaches for the prediction of CHF in a rod bundle are highly geometric specific and proprietary in nature. The critical power experiments are very expensive and technically challenging owing to the stringent simulation requirements for the rod bundle tests involving radial and axial power profiles. In view of this, the mechanistic approach has gained momentum in the thermal hydraulic community. The Liquid Film Dryout (LFD) in an annular flow is the mechanism of CHF under BWR conditions and the dryout modeling has been found to predict the CHF quite accurately for a tubular geometry. The successful extension of the mechanistic model of dryout to the rod bundle application is vital for the evaluation of critical power in the rod bundle. The present work proposes the uniform film flow approach around the rod by analyzing individual film of the subchannel bounded by rods with different heat fluxes resulting in different film flow rates around a rod and subsequently distributing the varying film flow rates of a rod to arrive at the uniform film flow rate as it has been found that the liquid film has a strong tendency to be uniform around the rod. The FIDOM-Rod code developed for the dryout prediction in BWR assemblies provides detailed solution of the multiple liquid films in a subchannel. The approach of uniform film flow rate around the rod simplifies the liquid film cross flow modeling and was found to provide dryout prediction with a good accuracy when compared with the experimental data of 16, 19 and 37 rod bundles under BWR conditions. The critical power has been predicted for a newly designed 54 rod bundle of the Advanced Heavy Water Reactor (AHWR). The selected constitutive models for the droplet entrainment and deposition rates validated for the dryout in tube were also found to perform well for the rod bundle under BWR conditions.

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Critical Power Evaluation with Consideration of Differences in Droplet Behavior due to Spacer Shape Influence.

Cited by 4 publications

References 5 publications

Critical Power Analysis with Mechanistic Models for Nuclear Fuel Bundles, (I)

Critical Power Analysis with Mechanistic Models for Nuclear Fuel Bundles, (I)

Prediction method of critical power by film flow rate measurement and subchannel analysis

A Mechanistic Approach for the Prediction of Critical Power in BWR Fuel Bundles

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