• Nuclear Engineering and Technology
• /
• v.49 no.8
• /
• pp.1617-1628
• /
• 2017

In the passive residual heat removal system of a molten salt reactor, one of the residual heat removal methods is to use the thimble-type heat transfer elements of the drain salt tank to remove the residual heat of fuel salts. An experimental loop is designed and built with a single heat transfer element to analyze the heat transfer and flow characteristics. In this research, the influence of the size of a three-layer thimble-type heat transfer element on the heat transfer rate is analyzed. Two methods are used to obtain the heat transfer rate, and a difference of results between methods is approximately 5%. The gas gap width between the thimble and the bayonet has a large effect on the heat transfer rate. As the gas gap width increases from 1.0 mm to 11.0 mm, the heat transfer rate decreases from 5.2 kW to 1.6 kW. In addition, a natural circulation startup process is described in this paper. Finally, flashing natural circulation instability has been observed in this thimble-type heat transfer element.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 2006.10a
• /
• pp.363-364
• /
• 2006

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

• Journal of the Korean Society for Nondestructive Testing
• /
• v.10 no.2
• /
• pp.15-22
• /
• 1990

Since in-core flux thimble tube wear the due to flow-induced vibration could degrade the integrity of nuclear reactor, the effective detection and interpretation of the wear is important. In order to establish an inspection technique for thimble tubes, an eddy current experiment was performed to determine the optimum test frequency, defect sensitivity and evaluation accuracy. Eddy current probes were designed and fabricated with a theory. Specimens with artificial defects were fabricated using electro discharge machining method. The results from inspection technique developed and on-site inspection showed good applicability.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.10a
• /
• pp.611-616
• /
• 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
• /
• 1996.05b
• /
• pp.166-172
• /
• 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
• /
• 2005.10b
• /
• pp.225-227
• /
• 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.

• Journal of Radiation Protection and Research
• /
• v.47 no.3
• /
• pp.152-157
• /
• 2022

Background: The hemi-body electron beam irradiation (HBIe^-) technique has been proposed for the treatment of mycosis fungoides. It spares healthy skin using an electron shield. However, shielding electrons is complicated owing to electron scattering effects. In this study, we developed a thimble-like head bolus shield that surrounds the patient's entire head to prevent irradiation of the head during HBIe^-. Materials and Methods: The feasibility of a thimble-like head bolus shield was evaluated using a simplified Geant4 Monte Carlo (MC) simulation. Subsequently, the head bolus was manufactured using a three-dimensional (3D) printed mold and Ecoflex 00-30 silicone. The fabricated head bolus was experimentally validated by measuring the dose to the Rando phantom using a metal-oxide-semiconductor field-effect transistor (MOSFET) detector with clinical configuration of HBIe^-. Results and Discussion: The thimble-like head bolus reduced the electron fluence by 2% compared with that without a shield in the MC simulations. In addition, an improvement in fluence degradation outside the head shield was observed. In the experimental validation using the inhouse-developed bolus shield, this head bolus reduced the electron dose to approximately 2.5% of the prescribed dose. Conclusion: A thimble-like head bolus shield for the HBIe^- technique was developed and validated in this study. This bolus effectively spares healthy skin without underdosage in the region of the target skin in HBIe^-.

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

• Nuclear Engineering and Technology
• /
• v.2 no.2
• /
• pp.67-72
• /
• 1970

The measurements of the fast neutron flux and its spectrum have been carried out by the threshold detectors in the central thimble of TRIGA Mark-II reactor operating at 250 KW. The following reactions have been employed for these measurements, viz : Ni$^{58}$ (n, p) Co$^{58}$ ； $Mg^{24}$ (n, p) Na$^{24}$ ； $Al^{27}$ (n, $\alpha$) Na$^{24}$ . From the activation data the fast neutron spectrum were calculated by CDC-3600 computer making use of two semi-empirical methods. It has been verified that the validity of assumption of a fission spectrum in the central thimble exists only above 1 to 2 Mev energy level. With this spectrum, a fast neutron flux in the range of 1 $\times$ 10$^{12}$ n/$\textrm{cm}^2$-sec above the energy of 2.6 Mev was observed in the central thimble of TRIGA MARK-II reactor.