• Nuclear Engineering and Technology
• /
• v.45 no.5
• /
• pp.589-596
• /
• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.10a
• /
• pp.41-47
• /
• 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
• /
• v.28 no.3
• /
• pp.280-289
• /
• 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.

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

• Nuclear Engineering and Technology
• /
• v.26 no.2
• /
• pp.161-167
• /
• 1994

This paper presents a concept of a pressure tube-type water-cooled reactor without the emergency core cooling system. It adopts an innovative fuel channel design using metallic fuel matrix to improve heat transfer from fuel to moderator at loss of coolant cooling. The heat produced in the fuel is cooled by the coolant system during normal operation, but by the passive moderator system at loss of coolant cooling including the loss-of-coolant accident(LOCA). Simple analysis shows that the fuel channel temperature can be maintained within the permissible range for both normal operation and a complete LOCA.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.10a
• /
• pp.35-40
• /
• 1995

Considering the higher discharge burnup, lower channel refuelling rate, lower linear element rating(LER), lower coolant void reactivity and axial power shape, CANFLEX-RU fuel bundle is optimized for CANDU-6 by grading the fissile composition in the ring-wise of the bundle and by applying fuel management scheme appropriately. The fissile composition of the fuel bundle is graded as the recovered uranium (0.9 w/o U-235) in the outer and intermediate elements, depleted Uranium (0.2 w/o U-235) in the center element, natural uranium (0.71 w/o U-235) in the inner elements. Enrichment is not required for these fuel. The fissile composition is optimized by lattice calculation and by time-averaged reactor simulation. CANFLEX-RU optimized for CANDU-6 resulted to be the 15% lower channel refuelling rate, acceptable axial power profile and power envelope, 70% higher discharge burnup, 15% lower LER and not increase coolant void reactivity compared with the 37-element natural uranium bundle for CANDU-6.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.10a
• /
• pp.109-114
• /
• 1995

A preliminary study of the heterogeneity effect of spent P% fuel in CANDU was made using a reduced spent PWR fuel data base. The instantaneous core simulation has shown that the refueling ripple in the CANDU reactor is large if the spent PWR fuel is directly used. But the fuel heterogeneity effect can be reduced appreciably by blending spent PWR fuel with a small amount of fresh UO$_2$. The refueling simulation has shown that the operating margins of 6.0% and 8.7% are achievable for the peak channel and bundle powers, respectively, with the blended fuel.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.169-174
• /
• 1998

An 1200-day time-dependent fuel-management for the transition from 37-element fuel to CANFLEX-NU fuel in a CANDU 6 reactor has been simulated to show the compatibility of the CANFLEX-NU fuel with the reactor operation. The simulation calculation were carried out with the RFSP code, provided by cell averaged hel properties obtained from the POWDERPUFS-V code. The refueling scheme for both fuels was an eight bundle shift art a time. The simulation results show that the maximum channel and bundle powers were maintained below the licence limit of the CANDU 6. This indicates that the CANFLEX-NU fuel bundle is compatible with the CANDU 6 reactor operation during the transition period.

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.3085-3094
• /
• 2022

The design concept of a Small High-flux Multipurpose LBE(Lead Bismuth Eutectic) cooled Fast Reactor (SHMLFR) was proposed in the paper. The primary cooling system of the reactor is forced circulation, and the fuel element form is arc-plate loaded high enrichment MOX fuel. The core is cylindrical with a flux trap set in the center of the core, which can be used as an irradiation channel. According to the requirements of the core physical design, a series of physical design criteria and constraints were given, and the steady and transient parameters of the reactor were calculated and analyzed. Regarding the thermal and hydraulic phenomena of the reactor, a simplified model was used to conduct a preliminary analysis of the fuel plates at special positions, and the temperature field distribution of the fuel plate with the highest power density under different coolant flow rates was simulated. The results show that the various parameters of SHMLFR meet the requirements and design criteria of the physical design of the core and the thermal design of the reactor. This implies that the conceptual design of SHMLFR is feasible.

• Transactions of the Korean Society of Mechanical Engineers B
• /
• v.30 no.1 s.244
• /
• pp.24-31
• /
• 2006

Mixing vanes have been installed in the space grid of nuclear fuel rod bundle to improve turbulent heat transfer. Split mixing vanes induce the vortex flow in the cooling water to swirl in sub-channel of fuel assembly. But, The swirling flow decays rapidly so that the heat transfer enhancing effect limited to short length after the mixing vane. In thi present study, the large scale vortex flow(LSVF) is generated by rearranging the mixing vanes to the coordinated directions. This LSVF mixing vanes generate the most strong secondary flow vortices which maintain about 35 $D_H$ after the spacer grid. The streamwise vorticity generated by LSVF sustain two times more than that split mixing vane. Heat transfer in the rod bundle occurs greatly at the same direction to cross flow, and maximum temperature at the surface of bundle drops about 1.5K