• Proceedings of the KSME Conference
• /
• 2004.04a
• /
• pp.905-910
• /
• 2004

Vibration characteristics of the FIV test loop for the Flow-Induced Vibration(FIV) study of a PWR partial(5x5) fuel assembly are investigated by the Finite Element(FE) analysis and the modal test. For the FE analysis, 3-D beam element is used for the pipes and the test section and mass element used for the valves and flanges. The 'U' restrainer stiffness determined by numerical simulation is used for the FE model. The result of the FE analysis is compared with that of the modal test. The higher mode similarity between the test and analysis is observed in a few low modes. After that, the mode similarity reduce as the mode goes high. It is concluded that the first to the third vibration modes are observed at the lower parts of the 6 inches restoring line, followed by a local mode at the test section, and the natural frequencies of the modes are 22.4 Hz, 26.0 Hz, 27.5 Hz and 31.4 Hz.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 2000.11a
• /
• pp.413-416
• /
• 2000

The spent fuel skeleton is processed in the cutting processing after compacting. If the cutting length is processed in the same interval length. The spent fuel skeleton is stayed on the connection of bottom nozzle and guide tube. In the case, because the compressive stress is loaded along the length, the guide tube is generated the buckling stress and the deforming. But the deformed guide tube interrupted the guide tube inserted through compressive room. therefore, it is experimented for the optimum buckling stress and the preventing of guide deformed. This paper is predicted the all over buckling stress of the spent fuel skeleton by using experiment. The guide of Spent fuel skeleton have buckling characteristics of the medium column. The experiment and analysis is conducted by the comparing among the equation of Euler, Johnson and Engresser. The fittest one of method is Engresser equation.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2006.05a
• /
• pp.250-253
• /
• 2006

The in-pile Section (IPS) is subjected to flow-induced vibration(FIV) due to the flow of the primary coolant and then the structural integrity. The in-pile Section (IPS) of 3-pin Fuel Test Loop(FTL) shall be installed in the vortical hole call IR1 of HANARO reactor core. In order to verify the velocity and displacement both the inside region of IPS at the annular region of IPS, the vibration was measured by varing the flow rate on both regions. The displacements of fuel assembly in the in-pile Section (IPS) were found to be lower than the values of allowable design criteria.

• Journal of Sensor Science and Technology
• /
• v.23 no.4
• /
• pp.245-250
• /
• 2014

When the charged particle travels in transparent medium with a velocity greater than that of light in the same medium, the electromagnetic field close to the particle polarizes the medium along its path, and then the electrons in the atoms follow the waveform of the pulse which is called as Cerenkov light or radiation. This type of radiation can be easily observed in a spent fuel storage pit. In optical fibers, the Cerenkov light also can be generated due to their dielectric components. Accordingly, the radiation-induced light signals can be obtained using optical fibers without any scintillating material. In this study, to measure the intensities of Cerenkov radiation induced by gamma-rays, we have fabricated the fiber-optic Cerenkov radiation sensor system using silica optical fibers, plastic optical fibers, multi-anode photomultiplier tubes, simulated spent fuel assembly and a scanning system. To characterize the Cerenkov radiation generated in optical fibers, the intensities of Cerenkov radiation generated in the silica and plastic optical fibers were measured. Also, we measured the longitudinal distribution of gamma rays emitted from the Ir-192 isotope by using the fiber-optic Cerenkov radiation sensor system and simulated spent fuel assembly.

• Nuclear Engineering and Technology
• /
• v.24 no.3
• /
• pp.263-273
• /
• 1992

The study on the velocity distribution and the pressure drop characteristic of the nuclear fuel assembly is of importance for the thermal hydraulic design and safety analysis. The purpose of this experimental study is to investigate the hydraulic mixing behind the different kinds of spacer grids in the now or rod bundles. In this study, the detailed hydraulic characteristics in subchannels of 5$\times$5 PWR(Pressurized Water Reactor) rod bundles were measured using one-component He-Ne LDV(Laser Doppler Velocimeter). Measurements of the axial velocity, turbulent intensities and pressure drops were peformed Lateral velocity, turbulent intensities and Reynolds shear stress were also measured by adjust-ing LDV alignment. Friction factors in rod bundles and loss coefficients for spacer grids were evaluated from the measured pressure drops. Hydraulic mixing performance for different kinds of spacer grids could be investigated by estimating the turbulent cross-flow mixing rates between neighboring subchannels.

• Journal of the Korean Society of Radiology
• /
• v.11 no.7
• /
• pp.571-578
• /
• 2017

Bubbles are generated by the boiling of the cooling water when an accident occurs in the reactor and then in order to measure the void fraction, the Optical Fiber Probe(OFP) and optical camera are used in thermal hydraulic safety research. However, such an optical method is not suitable for measuring the void fraction in a $17{\times}17$ array of fuel rods due to the geometrical limitations. This study was conducted as a preliminary study using x-ray system and various phantoms before applying to rod bundles. Through radiographic and tomographic experiments, the tube voltage of the x-ray generator was 130 kVp and the tube current was 1 mA. In addition, it is possible to measure the hole of 1mm in size visually through the bubble resolution phantom, and it is confirmed that the contrast is relatively decreased in the inside of the freon in the case of the contrast evaluation using the road phantom. However, we could obtain good image without distortion when reconstructing the image. Bubble generation phantom experiments were used to confirm the flow direction of the bubbles and to acquire tomography images. The image J tool was used to measure the void fraction of 18 % for a single tomography image. This study has carried out previous researches for the measurement of the bubble rate around the nuclear fuel and could be used as a basic research for continuous research.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.16 no.2
• /
• pp.233-242
• /
• 2018

To overcome the low mechanical strength and corrosion behavior of a carbon steel canister at high temperature condition of a deep borehole, SiC ceramics were studied as an alternative material for the disposal canister. In this paper, a design concept for a SiC canister, along with an outer stainless steel container, was proposed, and its manufacturing feasibility was tested by fabricating several 1/3 scale canisters. The proposed canister can contain one PWR assembly. The outer container was also prepared for the string formation of SiC canisters. Thermal conductivity was measured for the SiC canister. The canister had a good thermal conductivity of above $70W{\cdot}m^{-1}{\cdot}K^{-1}$ at $100^{\circ}C$. The structural stability was checked under KURT environment, and it was found that the SiC ceramics did not exhibit any change for the 3 year corrosion test at $70^{\circ}C$. Therefore, it was concluded that SiC ceramics could be a good alternative to carbon steel in application to deep borehole disposal canisters.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.8 no.4
• /
• pp.347-353
• /
• 2010

Source terms of metal waste comprising a spent fuel assembly are relatively important when the spent fuel is pyroprocessed, because cesium, strontium, and transuranics are not a concern any more in the aspect of source term of permanent disposal. In this study, characteristics of radiation source terms for each structural component in spent fuel assembly was analyzed by using ORIGEN-S with a assumption that 10 metric tons of uranium is pyroprocessed. At first, mass and volume for each structural component of the fuel assembly were calculated in detail. Activation cross section library was generated by using KENO-VI/ORIGEN-S module for top-end piece and bottom-end piece, because those are located at outer core with different neutron spectrum compared to that of inner core. As a result, values of radioactivity, decay heat, and hazard index were reveled to be $1.40{\times}10^{15}$ Bequerels, 236 Watts, $4.34{\times}10^9m^3$-water, respectively, at 10 years after discharge. Those values correspond to 0.7 %, 1.1 %, 0.1 %, respectively, compared to that of spent fuel. Inconel 718 grid plate was shown to be the most important component in the all aspects of radioactivity, decay heat, and hazard index although the mass occupies only 1 % of the total. It was also shown that if the Inconel 718 grid plate is managed separately, the radioactivity and hazard index of metal waste could be decreased to 20~45 % and 30~45 %, respectively. As a whole, decay heat of metal waste was shown to be negligible in the aspect of disposal system design, while the radioactivity and hazard index are important.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2012.04a
• /
• pp.911-916
• /
• 2012

It is important to identify dynamic characteristics of nuclear fuel components. Since the fuel always exposed to turbulent flow, the dynamic contact between grids and rods is one of the fuel failure modes. The dynamic behavior of grids in nuclear fuels is quite complex, since two pairs of spring support are placed in the limited space. The strap in a cell has single spring and double dimples and this paper focuses on investigation of the grid strap(Test Fuel Strap, TFS) vibration in one cell. To identify the grid strap vibration, modal analysis of the strap is performed using Finite Element Method (FEM). Modal testing on a $5{\times}5$ grid structure without rods is performed. The modal testing results are compared to analytic results. In addition, random test considering rod effect is performed about a $5{\times}5$ grid with rods under real contact condition in the air. Finally, the strap vibration of a $5{\times}5$ fuel bundle in INvestigation of Flow INduced vIbraTion(INFINIT) facility is measured in real fluid velocity condition without heating. It is shown that modal frequencies from the test are almost equal to those peak frequencies in the INFINIT test.

• Nuclear Engineering and Technology
• /
• v.13 no.1
• /
• pp.1-11
• /
• 1981

The KINS (KAERI-Improved Nodal Simulation) program, a three-dimensional nodal simulation code for pressurized water reactors, has been developed and benchmarked against the first cycle of the Kori-1 reactor. The KINS program is based on the computational model used in FLARE code and has been modified to represent the PWR characteristics more explicitly. The critical boron concentration and three-dimensional power distribution at the beginning of life hot zero power have been calculated and compared with the operating data. A three-dimensional depletion calculation at the intervals of 1000 MWD/MTU turnup steps has been performed. As the result of comparison, our calculation is shown to be in excellent agreement with the operating data. It is displayed that, incorporated with the computing time, the KINS program is an effective and powerful tool for PWR core management.