• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.11a
• /
• pp.357-362
• /
• 1996

LOFT LBLOCA test, LP-02-6 was analyzed using RELAP5/MOD3.2. It has a distinguished thermal-hydraulic phenomenon of a positive bottom-up core flow in tile blowdown phase. A modified nodalization which is based on that used in LP-LB-1 calculation by Lubbesmeyer was used in the calculation. RELAP5/MOD3.2 predicted overall system hydraulic behavior relatively well. However, the bottom-up quenching in the middle part of the core was not predicted sufficiently. It was demonstrated also that the peak cladding temperature can be predicted well by adjusting a discharge coefficient. But more improvements are needed in order to apply this code to actual plants with less user dependency.

• Nuclear Engineering and Technology
• /
• v.52 no.8
• /
• pp.1724-1731
• /
• 2020

Low cycle fatigue tests on the hydrogenated welding seam of Zr-Sn-Nb alloy at room temperature and 360 ℃ had been carried out by using the funnel-shaped flat specimens. The relationships between nominal stress & strain directly measured across the funnel and local stress & strain at the root of the funnel are given by considering cyclic plasticity correction. The results show that the fatigue resistance of welding seam at room temperature is only slightly better than that at 360 ℃. Probabilistic fatigue life curves are obtained by using a two-parameter power function.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.713-718
• /
• 1998

Henry-Fauske critical flow model was incorporated into TRAC-PF1 to correct some errors in the original TRAC-PFI critical flow model. Henry-Fouske mode1 was numerically implemented and tested against steady-state steam-water experimental data. The model was incorporated into TRAC-PFI and code assessment against Marviken Critical Flow Tests 15 and 24 was carried out. Calculations using RELAP5/MOD3 were also made for comparison. Ten cases were calculated each test and sensitivity study on nodalization as well as critical flow or model was performed Stand-alone numerical model test and code assessment were done for verification and validation of code modification. Calculation results show that the modified version of TRAC-PF1 has a capability to model critical flow correctly in various conditions.

• Nuclear Engineering and Technology
• /
• v.45 no.3
• /
• pp.361-366
• /
• 2013

An imaging technique to visualize the internal defects in a plate-type nuclear fuel specimen was developed by using an active optical interferometer for a nondestructive quality inspection. A periodic thermal wave having a sinusoidal intensity pattern induced a periodical strain variation for the specimen. The varying strain image was acquired using an optical laser interferometer. The strain distribution over the internal defects will be distorted in an acquired strain image because a part of the thermal wave will be reflected from these defects during propagation. In this paper, internal defects were efficiently visualized by sequentially accumulating the extracted defect components. The experimental results confirmed that the developed visualization system can be a valuable tool to detect the internal defects in plate-type nuclear fuel.

• Nuclear Engineering and Technology
• /
• v.50 no.2
• /
• pp.218-222
• /
• 2018

An oxide-dispersion-strengthened (ODS) layer was formed on Zircaloy-4 tubes by a laser beam scanning process to increase mechanical strength. Laser beam was used to scan the yttrium oxide ($Y_2O_3$)-coated Zircaloy-4 tube to induce the penetration of $Y_2O_3$ particles into Zircaloy-4. Laser surface treatment resulted in the formation of an ODS layer as well as microstructural phase transformation at the surface of the tube. The mechanical strength of Zircaloy-4 increased with the formation of the ODS layer. The ring-tensile strength of Zircaloy-4 increased from 790 to 870 MPa at room temperature, from 500 to 575 MPa at $380^{\circ}C$, and from 385 to 470 MPa at $500^{\circ}C$. Strengthening became more effective as the test temperature increased. It was noted that brittle fracture occurred at room temperature, which was not observed at elevated temperatures. Resistance to dynamic high-temperature bursting improved. The burst temperature increased from 760 to $830^{\circ}C$ at a heating rate of $5^{\circ}C/s$ and internal pressure of 8.3 MPa. The burst opening was also smaller than those in fresh Zircaloy-4 tubes. This method is expected to enhance the safety of Zr fuel cladding tubes owing to the improvement of their mechanical properties.

• Nuclear Engineering and Technology
• /
• v.52 no.9
• /
• pp.1881-1895
• /
• 2020

The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performed against MCNP6 to confirm the suitability of MCS for the criticality and depletion analysis of the MTR. Then, the dependence of the effective neutron multiplication factor to the number of axial and radial depletion cells adopted in the fuel plates is performed with MCS in order to determine the minimum spatial segmentation of the fuel plates. Monte Carlo depletion results with 37,800 depletion cells are provided by MCS within acceptable calculation time and memory usage. The results show that at least 7 axial meshes per fuel plate are required to reach the same precision as the reference calculation whereas no significant differences are observed when modeling 1 or 10 radial meshes per fuel plate. This study demonstrates that MCS can address the need for Monte Carlo codes capable of providing reference solutions to complex reactor depletion problems with refined meshes for fuel management and research reactor applications.

• Nuclear Engineering and Technology
• /
• v.29 no.2
• /
• pp.175-180
• /
• 1997

It has been known that eater-side corrosion of fuel rods in nuclear reactor is accompanied with the metallic loss of wall thickness and hydrogen pickup in the fuel dadding tube. The fuel dad corrosion is one of the major factors to be controlled to maintain the fuel integrity during reactor operation. An oxide later thickness measuring device equipped with ECT probe system was developed by KAERI, and whose performance test was carried out in NDT(Non-destructive Test) hot-cell or PIE(Post Irradiation Examination) Facility. At first, the calibration/performance test was executed for the unirradiated standard specimen rod fabricated with several kinds of plastic thin films whose thickness ore predetermined, and the result of which showed a good precision within 10% of discrepancy. And then, hot test us peformed for the irradiated fuel rod selectively extracted from J44 fuel assembly discharged from Kori Unit-2. The data obtained with this device were compared with the metallographic result obtained from destructive examination in PIEF hot-cell on the same fuel rod to verify the validity of the measurement data.

• Nuclear Engineering and Technology
• /
• v.43 no.1
• /
• pp.75-82
• /
• 2011

The neutron multiplication factor in spent fuel storage racks, in which a stainless steel plate encloses a fuel assembly, was evaluated according to the variation of distance between the fuel assembly and stainless steel plate, as well as the pitch. The stainless steel plate position with the lowest multiplication factor on each pitch consistently appeared as 6mm or 9mm away from the outmost surface of the fuel assembly. Because the stainless steel plate has a thermal neutron absorption cross section, its ability to absorb neutrons can work best only if it is installed at the position where thermal neutrons can be gathered most easily. Therefore, the stainless steel plate position should not be too close or too far away from the fuel assembly, but it should be kept a pertinent distance from the fuel assembly.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.732-733
• /
• 2004

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.05a
• /
• pp.401-402
• /
• 2005