• Nuclear Engineering and Technology
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• v.27 no.3
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• pp.358-373
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• 1995

The Heat Transport system loop stability of CANDU-6 reactor using the CANFLEX fuel bundle was studied. The Thermal-hydraulic behavior of CANFLEX fuel bundle is similar to the conventional 37-element fuel bundle since the reactor power and the frictional pressure drop through the fuel channel is almost the same each other, Mounter the CANFLEX fuel bundle gives higher critical channel power and more homogeneous enthalpy distributions in the subchannels than 37-element fuel bundle. The SOPHT modelling or the CANFLEX fuel bundle and the Reactor outlet Header(ROH) interconnection line was made and the stability analysis response of Wolsong-1 reactor with CANFLEX fuel bundle was obtained. Without the ROH interconnection line the Heat Transport system loop using 43-element fuel bundle is unstable like the current 37-element fuel bundle. With the ROH interconnection line, however, the Heat Transport system is stable within $\pm$1％ of nominal flow. In the Heat Transport system loop stability point of view for Wolsong-1 plant therefore, the CANFLEX fuel loading is considered to be acceptable.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1314-1321
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• 2019

In case of multiple failure scenario, such as LOCA with ECCS failure, the decay heat continues to raise the reactor core temperature, eventually leading to the core voiding. In such scenario the convective heat transfer becomes poor and the majority of the heat transfer from fuel bundle takes place by radiation mode. During this abnormal working condition, if the channel pressure is less than 1 MPa, the PT sags and come in contact with the CT. This results in high rate of heat transfer from contact location to moderator. The present paper aims to capture the temperature profile over a simulated nuclear channel during such scenario at a steady state temperature of $600^{\circ}C$ (Centre pin) at two different configurations of PT i.e. PT concentric with CT and PT contact with CT. The results showed that the bottom nodes of all the components (Fuel bundle, PT and CT) of the simulated channel was greatly influenced by the PT/CT contact. Moreover, higher temperature were observed at top nodes of the PT and outer pins of the fuel bundle. However, no significant variation in temperatures were obtained in fuel bundle and CT in concentric condition.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.41-47
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• 1995

A parametric study bas been performed for the various refueling schemes of CANDU 6 reactor loaded with reference DUPIC fuel. The optimum discharge burnup was determined such that the peak bundle power is minimized for the equilibrium core. Based on the results of instantaneous core calculation using patterned random age distributions, it was decided to perform the refueling simulations only for 2-bundle and 4-bundle shift refueling schemes. The 600 FPD simulation has shown that the operational margins of the channel and bundle power to the license limits are 7.9% and 17.1%, respectively, for 2-bundle shift refueling scheme. The 4-bundle shift refueling scheme also satisfies the license limits and the operational margins of the channel and bundle power are 7.1% and 9.8%, respectively. The result of refueling simulation indicate the possibility of using reference DUPIC fuel in current CANDU 6 reactor.

• Nuclear Engineering and Technology
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• v.30 no.4
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• pp.328-341
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• 1998

This paper presents in detail the hydraulic characteristic measurements using LDV(Laser Doppler Velocimetry) in subchannels of a HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids and has a cylindrical configuration. Axial velocity and turbulent intensity were measured. The effects of the spacer grids on the turbulent flow were investigated using the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of the fuel bundle and the loss coefficients for the spacer grids were estimated from the measured pressure drops. The turbulent thermal mixing phenomena were discussed.

• Nuclear Engineering and Technology
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• v.34 no.2
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• pp.176-185
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• 2002

A 4-bundle shift refuelling method of CANDU6 F/H (Fuel Handling) System is analyzed to assess the operational flexibility and capacity of F/H system. The current 8-bundle shift refuelling scheme requires to refuel eight fuel bundles from a single fuel channel, and to refuel 14 fuel channels in a week on average assuming that the reactor is in a steady state. The analysis showed that the 4-bundle shift refuelling method increases F/M (Fuelling Machine) duty cycle and operator load. However, the fuellin’g method change from the 8- to 4-bundle shift refuelling ill not require additional team of operators. A marginal increase in the maintenance cost may be resulted in by the change of fuelling method and the increase of fatigue usage factors requires some components to be replaced during the FM lifetime.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.501-506
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• 1997

This paper presents the hydraulic characteristics measured by using LDV(Laser Doppler Velocimetry) in subchannels of a HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops fer each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regard ins the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.479-490
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• 2005

The thermalhydraulic performance of a CANDU-6 reactor loaded with various CANFLEX fuel bundles is evaluated by the NUCIRC code, which is incorporated with recent models of pressure drop and critical heat flux (CHF) predictions based on high-pressure steam-water tests for the CANFLEX bundle as well as a 37-element bundle. The distributions of channel flow rate, channel exit quality, critical channel power (CCP), and critical power ratio (CPR) for the CANFLEX bundles (with natural or recycled uranium fuel) in the CANDU-6 reactor fuel channel are calculated by the code. The effects of axial and radial heat flux on CCP are evaluated by assuming that the recycled uranium fuel (CANFLEX-RU) has the same geometric data as the natural uranium fuel bundle (CANFLEX-NU), but a different power distribution due to different fuel composition and refueling scheme. In addition, the effects of pressure tube creep and bearing-pad height are examined by comparing various results of uncrept, and $3.3\%\;and\;5.1\%$ crept channels loaded with CANFLEX bundles with 1.4 mm or 1.7 mm high bearing-pads with those of the 37-element bundle. The distributions of the channel flow rate and CCP for the CANFLEX-NU or -RU bundle show a typical trend for a CANDU-6 reactor channel, and the CPRs are maintained above at least 1.444 (NU) or 1.455 (RU) in the uncrept channel. The enhanced CHF of the CANFLEX bundle (particularly with 1.7mm height bearing-pads) produces a higher thermal margin and considerably less sensitivity to CCP reduction due to the pressure tube creep than the 37-element bundle. The CCP enhancement due to the raised bearing-pads is estimated to be about $3\%\~5\%$ for the CANFLEX-NU and $2\%\~6\%$ for the CANFLEX-RU bundle, respectively.

• Nuclear Engineering and Technology
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• v.41 no.7
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• pp.969-978
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• 2009

The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced $CANDU^{(R)1}$ reactor fuel bundle. Based primarily on the $CANFLEX^{(R)2}$ fuel bundle, several geometry changes (such as element sizes and pitch-circle diameters of various element rings) were examined to optimize the dryout power and pressure-drop performances of the new fuel bundle. An experiment was performed to obtain dryout power measurements for verification of the ASSERT-PV code predictions. It was carried out using an electrically heated, Refrigerant-134a cooled, fuel bundle string simulator. The axial power profile of the simulator was uniform, while the radial power profile of the element rings was varied simulating profiles in bundles with various fuel compositions and burn-ups. Dryout power measurements are predicted closely using the ASSERT-PV code, particularly at low flows and low pressures, but are overpredicted at high flows and high pressures. The majority of data shows that dryout powers are underpredicted at low inlet-fluid temperatures but overpredicted at high inlet-fluid temperatures.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.651-656
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• 1997

Fission product release (FPR) assessment for End Fitting Failure (EFF) in CANDU reactor loaded with CANFLEX-natural uranium (NU) fuel bundles has been peformed. The predicted results are compared with those for the reactor loaded with standard 37-element bundles. The total channel I-131 release at the end of transient for EFF accident is calculated to be 380.8 TBq and 602.9 TBq for the CANFLEX bundle and standard bundle channel cases, respectively. They are 4.9% and 7.9% of the total inventory, respectively. The lower total releases of the CANFLEX bundle O6 channel are attributed to the lower initial fuel temperatures caused by the lower linear element power of the CANFLEX bundle compared with the standard bundle.

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.280-289
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• 1996

The pressure drop of twelve fuel bundle string in the CANDU-6 fuel channel is equal to the sum of the eleven junction pressure losses, the bundle string entrance and exit pressure losses, the skin friction pressure loss, and other appendage pressure losses, where the junction loss is dependent on the bundle end faces and angular alignments of the junctions. The results of the single junction pressure drop tests in a short rig show that the most probable pressure drop of the eleven junctions was analytically equal to the eleven times of average pressure drop of all the possible single junction pressure drops, and also that the largest and smallest junction pressure drops across the eleven junctions probably occurred only with BA and BB type junctions, respectively, where A and B denote the bundle end sides with an end-plates on which a company monogram is stamped and unstamped, respectively.