• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.16 no.12S
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• pp.1292-1298
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• 2003

In this paper, dc reactor lot the inductive high-Tc superconducting fault current limiter (SFCL) was optimally designed by finite element method(FEM). The Prototype high-Tc do reactor was manufactured and compared to the results of design. This dc reactor consists of 4∼stacked double pancake coils which are wounded with Bi-2223 wire coated with SUS315L. Kapton tape is used for the insulation of turn to turn and layer to layer. Each pancake is connected in series by soldering Finally, optimal design and manufacture method lot the dc reactor is suggested in this paper. Through the comparison of result of optimal design and experimental result of prototype high-Tc superconducting dc reactor, reliance on the design of the high-Tc dc reactor tot the 1.2 kV/80 A SFCL is proved.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.1
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• pp.1-5
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• 2023

This paper describes a newly-developed analysis methodology in comprehensive vibration assessment program (CVAP) of reactor internals to develop a valid-prototype for the design of nuclear power plants. The new analysis methodology developed in this study will be confirmed through a scale model testing (SMT). Based on the measurements obtained from dynamic pressure transducers in the SMT, a new non-dimensional equation is developed to apply the forcing functions at reactor internals for the prototype. In addition to the new non-dimensional equation, a computational fluid dynamics(CFD) is used to develop the application of the hydraulic loads at reactor internals for the prototype.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1071-1082
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• 2016

Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the invessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.952-964
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• 2016

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium-water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium-water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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• pp.57-64
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• 2010

PHE(Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR(Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute established the gas loop for the performance test of components, which are used in the VHTR, and they manufactured a PHE prototype to be tested in the loop. In this study, as part of the high-temperature structural-integrity evaluation of the PHE prototype, which is scheduled to be tested in the gas loop, we carried out high-temperature structural-analysis modeling, thermal analysis, and thermal expansion analysis of the PHE prototype. The results obtained in this study will be used to design the performance test setup for the PHE prototype.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.776-783
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• 2020

As the refueling of a sodium-cooled fast reactor is conducted by rotating part of the reactor head without opening it, the monitoring of existing obstacles that can disturb the rotation of the reactor head is one of the most important issues. This paper deals with the ultrasonic ranging technique that directly monitors the existence of possible obstacles located in a lateral gap between the upper internal structure and the reactor core in a prototype generation IV sodium-cooled fast reactor (PGSFR). A 10 m long plate-type ultrasonic waveguide sensor, whose feasibility has been successfully demonstrated through preliminary tests, was employed for the ultrasonic ranging technique. The design of the sensor's wave radiating section was modified to improve the radiation performance, and the radiated field was investigated through beam profile measurements. A test facility simulating the lower part of the upper internal structure and the upper part of the reactor core with the same shapes and sizes as those in the PGSFR was newly constructed. Several under-water performance tests were then carried out at room temperature to investigate the applicability of the developed ranging technique using the plate-type ultrasonic waveguide sensor with the actual geometry of the PGSFR's internal structures.

• Proceedings of the Korea Institute of Applied Superconductivity and Cryogenics Conference
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• 2003.02a
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• pp.213-215
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• 2003

This paper deals with the characteristics of a prototype solenoid for basic design of DC reactor type superconducting fault current limiter (SFCL). The prototype high-Tc Super-conducting (HTS) solenoid was manufactured with 4 stacked Bi-2223 tape. The critical currents were measured with respect to the number of stacks. In order to test the safety of HTS solenoid in quenched state, the transport tests of AC over-current were performed. These experimental results could be applied to the basic design of HTS DC reactor for SFCL effectively.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.922-929
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• 2019

This paper presents the drop performance test of the control rod assembly which is one of the main components strongly related to the safety of the prototype generation IV sodium-cooled fast reactor. To investigate the drop performance, a real-sized control rod assembly that was recently modified based on the drop analysis results was newly fabricated, and several free drop tests under different flow rate conditions were carried out. Then the results were compared with those obtained from the previous tests conducted on the conceptually designed control rod assembly to demonstrate the improvement in performance. Moreover, the drop performance tests under several types and magnitudes of seismic loadings were also conducted to investigate the effect of the seismic loading on the drop performance of the modified control rod assembly. The results showed that the effects of the type and magnitude of the seismic loading on the drop performance of the modified control rod assembly were not significant. Also, the drop time requirement was successfully satisfied, even under the seismic loading conditions.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.351-359
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• 2016

The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

• Proceedings of the KIEE Conference
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• 2002.07d
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• pp.2182-2184
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• 2002

In this paper we describe a prototype Control Rod Control System(CRCS). The CRCS controls the motion of the full length rod drive mechanisms in response to signals from the Reactor Operator and the Reactor Control System. Each mechanism belongs to either Shutdown Banks or Control Banks. The CRCS also provides information regarding the rod motion, rod position, and the status of the Rod Control System. The prototype CRCS will be used to obtain the requirements for detailed design of a full-scale CRCS.