• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.6-16
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• 2017

This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU)-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700) fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the $2{\times}2$ checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

• Nuclear Engineering and Technology
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• v.28 no.4
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• pp.373-378
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• 1996

The CANFLEX fuel bundle has been developed by KAERI/AECL jointly to facilitate the use of various fuel cycles in CANDU-6 reactor. The structural analysis of the fuel bundles by hydraulic drag force is performed to evaluate the fuel integrity during the refuelling service. The present analysis method is newly developed for the structural integrity valuation by studying FEM modelling for the fuel bundles in a fuel channel. As compared the results of the mechanical strength test the displacement value of endplate given by analysis results shoo6 to be good agreement within 15% under the maximum design drag load. As the results of analysis, it is shown to keep the structural integrity of CANFLEX fuel bundles under hydraulic drag load during the refuelling service.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.337-338
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• 2005

• Nuclear Engineering and Technology
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• v.22 no.4
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• pp.389-409
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• 1990

As a principal design of advanced CANDU fuel bundle, CANDU-KF39, CANDU-KF40 and CANDU-KF43 fuel bundles were proposed and evaluated with respect to the operating conditions of the CANDU-6 reactor of Wolsung Unit-1. From the results, the advanced fuel bundles show to be improved economical and technical benefits compared with the current 37-element bundle. Especially, it was appeared that the KF-39 fuel bundle has more benefits of the safety, technical and economical aspects of Wolsung Unit-1 rather than those of the KF-40 and KF-43 fuel bundles.

• Transactions of the Korean Society of Mechanical Engineers
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• v.10 no.1
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• pp.7-14
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• 1986

To predict the fuel clad temperature during the reflooding phase of a LOCA, one may need a knowledge of reflood heat tranfer mechanism in a rod bundle. For this purpose reflooding experiments have been carried out with an electrically heated 3*3 rod bundle. Using the method for the determination of local heat transfer coefficient from the measured wall temperature the parametric effects of coolant flow rate, initial wall temperature, coolant subcooling and heat generation rate on the propagation of rewetting front were investigated. Prediction of the wall temperature histories for these experiments was discussed using REFLUX code with modification of the rewetting temperature correlation. Through this modification, better agreement between experiment and prediction was obtained.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.15-20
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• 1996

The nonlinear dynamic analyses were performed by newly developing an appropriate impact modelling for the evaluation of the CANFLEX fuel bundle structural integrity during the refuelling period. The initial load under the refuelling condition is considered as initial velocity at impact incident, and the impact of one bundle contacted another bundle for at short time is studied by performing several dynamic analysis method. The impact analysis shows to predict an appropriate velocity and acceleration profile according to load time history for two bundles impact.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.486-487
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.1008-1013
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• 1995

The CANFLEX fuel bundle has been developed by KAERI/AECL jointly to facilitate the use of various fuel cycles in CANDU-6 reactor. As one of the design evaluations, the structural analysis of the fuel bundles by hydraulic drag force is performed to evaluate the fuel integrity in the period of the refuelling in CANDU-6. The structural integrity is evaluated by FEM modelling for the complicated bundles configuration in channel. It is noted that the present analysis method is newly developed for the structural integrity evaluation. The analysis results show that the fuel bundle is shown to keep its structural integrity during the refuelling.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2005.11a
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• pp.503-506
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• 2005

The PWR Nuclear Fuel assembly consists of more than 250 fuel rods that are supported by leaf springs in the cells of more than 10 Spacer Grids (SG) along the rod length. Since it is not easy to conduct mechanical tests on a full-scale model basis, the small-scaled rod bundle (5$\times$5) is generally used for various performance tests during the development stage. As one of the small-scaled tests, a flow test should be carried out in order to verify the performance of the spacer grid like the coolant mixing performance and to obtain the Flow-Induced Vibration (FIV) characteristics of the rod bundle over the specified flow range. A vibration test should be also performed to obtain the modal parameters of the bundle prior to the flow test. In this study, we want to develop the estimation procedure of the damping ratio for the small scaled test bundle. For the damping factor of the rod bundle and the grid case at the first vibration mode, as one of the vibration tests, a so-called pluck testing has been performed in air as a preliminary test prior to in-flow damping measurement test. Logarithmic decrement method is used for calculation of the damping ratio. Estimated damping ratio of the rod bundle is about 0.7% with reasonable error of 2% for the previous results. Nonlinear behavior of the rod bundle might be stem mainly Iron the rod-grid support configuration.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.52-57
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• 1996

The coolant void reactivity is positive for CANDU reactor loaded with DUPIC fuel which has more fissile content compared to natural uranium. A parametric study was done to reduce the void reactivity of the fuel bundle and loss in discharge burnup was estimated. It is observed that the burnable absorbers like gadolinium, boron, europium are not able to keep the reduction in void reactivity uniform throughout fuel burnup. Dysprosium and erbium can keep the void reactivity reduction uniform throughout. fuel burnup but toss in discharge burnup for erbium case is more compared to that of dysprosium case.