• Journal of the Korean Society for Precision Engineering
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• v.22 no.3 s.168
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• pp.93-100
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• 2005

The top and bottom nozzles of PWR fuel assembly are connected by guide thimbles and an instrumentation tube that are connected with spacer grids. The fuel rods are inserted into the each cell of spacer grids. The loads acting on the fuel assembly are transmitted to the guide thimbles through the flow plate of top nozzle The axial loads applied to the fuel assembly are not equally distributed among the guide thimble due to the geometry of the top nozzle flow plate and spacer grid. In this study, the load concentration factors for the $17\times17$ fuel assembly were calculated. The analytical model fur the calculation of the load concentration factor of top nozzle flow plate was developed using ANSYS 5.6. The finite element analyses were performed using the model composed of top nozzle, guide thimble, and spacer grid. And, the analysis results were compared with the test results.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.197-204
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• 1994

In a PWR rod cluster control assembly(RCCA) for shutdown is released upon action of control rod drive mechanism and falls down through the guide thimble by its weight. Drop time and impact velocity of the RCCA are two key parameters with respect to reactivity insertion time and the mechanical integrity of fuel assembly. Therefore, the precise control of drop time and impact velocity is prerequisite to modifying the existing design features of the RCCA and guide thimble or newly designing them. During its falling down into the core, the RCCA is retarded by various forces acting on it such as fluid resistance caused by the RCCA movement, buoyance and mechanical friction caused by contacting inner surface of the guide thimble, etc. However, complicated coupling of the various forces makes it difficult to derive an analytical dynamic equation for the drop time and impact velocity. This paper deals with the development of a computer program containing an analytical dynamic equation applicable to the Korean Fuel Assembly(KOFA). The computer program is benchmarked with an available single control rod drop tests. Since the predicted values are in good agreement with the test results, the computer program developed in this paper can be employed to modify the exiting design features of the RCCA and guide thimble and to develope their new design features for advanced nuclear reactors.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.611-616
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• 1997

The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for 8 Westinghouse type PWR plants in Korea as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increases approximately by 1.2%. The resulting DNBR penalties can be covered with the current DNBR margin. Accident analyses are also investigated that the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.166-172
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• 1996

The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for Westinghouse type PWR plants as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increase approximately by 1.2%. The resulting DNBR penalties can be covered within the current DNBR margin. Accident analyses are also investigated and the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation.

• Proceedings of the KIEE Conference
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• 2005.10b
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• pp.225-227
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• 2005

The in-core neutron Flux Mapping System in a pressurized water reactor yields information on the neutron flux distribution in the reactor core at selected core locations by means of movable detectors. The obtained data are used to verify the reactor core design parameters. The detector cables run through guide tubes(thimbles), and typically thirty-six to fifty-eight thimbles are allocated in the reactor depending on the number of fuel assemblies. These thimbles are inserted into nuclear fuel assemblies through conduits connected from the bottom of the reactor vessel to a seal table. During the plant refueling outage period, the thimbles are withdrawn up to 4m from the seal table, the height of a nuclear fuel. In spite of their importance, however, the thimble handling work has been performed by only human operators. In addition, its efficiency is very low due to narrow working environments on the seal table, thereby resulting in the excessive radiation exposure of maintenance personnel. To solve these problems, a new thimble handling equipment for in-core flux mapping system was developed, and we confirmed its effectiveness through experiments.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.845-846
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• 2004

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.14 no.4
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• pp.1-7
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• 2000

In this paper, we study a cause of malfunction of switch to control drive motor in DFMS(Digital Flux Mapping System) which can measure incore neutron flux of the nuclear plant, and develope a method to solve this problem. DFMS has the type of generating contact signal by mechanical switch lever, which is operated whenever thimble detector inserted or withdrawed through thimble Guide Tube. However the characteristics of the lever tend to be changed by mechanical degrade or bad environment and the lever finally generates errotic contact signal. Therefore we installed electric coil ass'yin the outside of Guide Tube instead of mechanical switch assy's. In addition we applied resonance effect to control circuit and installed condenser in the input of power supply to protect noise and interference. After completion of this improvement, we tested this improved device repetitively under the various conditions. In conclusion, we identified the generation of the desired contact signal and the prevention of detector failure through plant surveillance test during normal plant operation.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.318-329
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• 2016

Reducing critical boron concentration in a commercial pressurized water reactor core offers many advantages in view of safety and economics. This paper presents a preliminary investigation of a reduced-boron pressurized water reactor core to achieve a clearly negative moderator temperature coefficient at hot zero power using the newly-proposed "Burnable absorber-Integrated Guide Thimble" (BigT) absorbers. The reference core is based on a commercial OPR1000 equilibrium configuration. The reduced-boron ORP1000 configuration was determined by simply replacing commercial gadolinia-based burnable absorbers with the optimized BigT-loaded design. The equilibrium cores in this study were directly searched via repetitive Monte Carlo depletion calculations until convergence. The results demonstrate that, with the same fuel management scheme as in the reference core, application of the BigT absorbers can effectively reduce the critical boron concentration at the beginning of cycle by about 65 ppm. More crucially, the analyses indicate promising potential of the reduced-boron OPR1000 core with the BigT absorbers, as its moderator temperature coefficient at the beginning of cycle is clearly more negative and all other vital neutronic parameters are within practical safety limits. All simulations were completed using the Monte Carlo Serpent code with the ENDF/B-VII.0 library.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.869-877
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• 2020

A 24-month Advanced Power Reactor 1400 (APR1400) core with a very-low-boron (VLB) concentration has been investigated for an inherently safe and high-performance PWR in this work. To develop a high-performance APR1400 which is able to do the passive frequency control operation, VLB feature is essential. In this paper, the centrally-shielded burnable absorber (CSBA) is utilized for an efficient VLB operation in the 24-month cycle APR1400 core. This innovative design of the VLB APR1400 core includes the optimization of burnable absorber and loading pattern as well as axial cutback for a 24-month cycle operation. In addition to CSBA, an Er-doped guide thimble is also introduced for partial management of the excess reactivity and local peaking factor. To improve the neutron economy of the core, two alternative radial reflectors are adopted in this study, which are SS-304 and ZrO₂. The core reactivity and power distributions for a 2-batch equilibrium cycle are analyzed and compared for each reflector design. Numerical results show that a VLB core can be successfully designed with 24-month cycle and the cycle length is improved significantly with the alternative reflectors. The neutronic analyses are performed using the Monte Carlo Serpent code and 3-D diffusion code COREDAX-2 with the ENDF/B-VII.1.