• Proceedings of the Korean Society of Medical Physics Conference
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• 1999.11a
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• pp.58-61
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.443.1-443
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• 2002

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.471-481
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• 2003

In order to evaluate the microstructural characteristics of irradiated fuel claddings (Zircaloy-4), two irradiated specimens having different burnups (18 GWD/MTU and 42 GWD/MTU) were prepared from the G23-M4 fuel rods, which were loaded in Kori Unit 1 for 4 fuel cycles. The oxide thickness, hydride morphology and hardness change were characterized by an optical microscope and a micro-hardness tester after preparing the irradiated specimens in the PIE (Post Irradiation Examination) facilities. The dislocation loops and the amorphous transformation of precipitates induced by the neutron irradiation in the nuclear plant were also examined by a TEM. As the burnup increased from 18 GWD/MTU to 42 GWD/MTU, the oxide thickness increased from $6.0{\mu}m\;to\;25.3{\mu}m$ and the contents of hydrogen pick-up in the Zr matrix also greatly increased. In the comparison of the hardness of the unirradiated fuel cladding, the amount of hardness increase was nearly $9.3\%\;and\;24.2\%$ for the irradiated samples of 18 and 42 GWD/MTU, respectively. Both -type dislocation loops and -type dislocation components were observed simultaneously in the irradiated specimens and the densities of the dislocation was increased by increasing the burnup. The precipitates in both the irradiated specimens were amorphously transformed by the neutron irradiation and the trend of the amorphous transformation of the precipitates was enhanced at a higher burnup.

• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.219-222
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• 2013

Alloy 690 has been selected as a steam generator tubing material for SMART owing to a near immunity to primary water stress corrosion cracking. The steam generators of SMART are faced with a neutron flux due to the integrated arrangement inside a reactor vessel, and thus it is important to know the irradiation effects of the thermal conductivity of Alloy 690. Alloy 690 was irradiated at HANARO to fluences of (0.7-28) ${\times}10^{19}n/cm^2$ (E>0.1MeV) at $250^{\circ}C$, and its thermal conductivity was measured using the laser-flash equipment in the IMEF. The thermal conductivity of Alloy 690 was dependent on temperature, and it was a good fit to the Smith-Palmer equation, which modified the Wiedemann-Franz law. The irradiation at $250^{\circ}C$ did not degrade the thermal conductivity of Alloy 690, and even showed a small increase (1%) at fluences of (0.7~28) ${\times}10^{19}n/cm^2$ (E>0.1MeV).

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.58-67
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• 2020

This study presents cyclic stress-strain and tensile test results at room temperature (RT) and 316℃ using cold-worked TP304 stainless steel base and weld metals. By comparing the cyclic hardening/softening behavior and failure cycle of cold-worked materials with those of irradiated austenitic stainless steels, the feasibility of simulating the irradiation effect on cyclic deformation and failure behaviors of TP304 stainless steel base and weld metals was investigated. It was found that, in the absence of strain-induced martensite trasformation, cold-working could properly simulate the change in cyclic hardening/softening behavior of TP304 stainless steel base and weld metals due to neutron irradiation. It was also recognized that cold-working could adequately simulate the reduction in failure cycles of TP304 stainless steel base and weld metals due to neutron irradition in the low-cycle fatigue region.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.114-122
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• 2024

Irradiation-induced defects and the precipitates in the wrapper material of the Indian Fast Breeder Test Reactor (FBTR), SS 316 are analyzed using the synchrotron source-based Angle Dispersive X-Ray Diffraction (ADXRD) technique with X-rays of energy 17.185 keV (wavelength ~0.72146 Å). The differences and similarities in the high displacement damage samples as a function of dpa (displacement per atom) and dpa rate in the range of 2.9 × 10^-7- 9 × 10^-7 dpa/s are studied. Ferrite and M₂₃C₆ are commonly observed in the present set of high displacement damage 40-74 dpa SS 316 samples irradiated at temperatures in the range of 400-483 ℃. Also, the dislocation density has increased as a function of the irradiation dose. The X-ray diffraction peak profile parameters quantified such as peak shift and asymmetry show that the irradiation-induced defects are sensitive to the dpa rate-irradiation temperature combinations. The increase in yield strength as a function of displacement damage is also correlated to the dislocation density.

• Nuclear Engineering and Technology
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• v.34 no.6
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• pp.566-573
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• 2002

Neutron activation analysis based on $textsc{k}$$_{o}$-standardization method＃ ($textsc{k}$o-NAA) is Com as one of the most remarkable progresses of the NAA with advantages of experimental simplicity, high accuracy, excellent flexibility with respect to irradiation and counting conditions, and suitability for computerization. This study was carried out to determine the reactor neutron spectrum parameters, i.e. $\alpha$ and f as the main factors of irradiation quality at NAA ＃1 irradiation hole on HANARO research reactor, to evaluate peak detection efficiency of the gamma-ray spectrometer for the use in the $textsc{k}$$_{o}$ experiments and to compare the measured concentration results with the certified values of some SRMs applying the experimentally determined to-parameters.ers.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2505-2510
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• 2008

A nuclear fuel test loop (after below, FTL) is installed in IR1 of an irradiation hole in HANARO for testing neutron irradiation characteristics and thermo hydraulic characteristics of a fuel loaded in a light water power reactor or a heavy water power reactor. When HANARO is normally operated, the fuel loaded in the irradiation hole has a nuclear reaction heat generated by a neutron irradiation. To remove the generated heat and to maintain an operation condition of the test fuel, a main cooling water system (MCWS) is installed in the OPS of the FTL. This paper describes the cold function test results of the MCWS. It was confirmed through the test results that the system met the design requirements under a cold operation condition.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.323-331
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• 2015

Zirconium alloys are widely used in nuclear reactors as structural materials. During the operation, they are exposed to fast neutrons. Ion irradiation is used to simulate the damage introduced by neutron irradiation. In this article, we briefly review the neutron irradiation damage of zirconium alloys, then summarize the effect of ion irradiation on microstructural evolution, mechanical and corrosion properties, and their relationships. The microstructure components consist of dislocation loops, second phase precipitates, and gas bubbles. The microstructure parameters are also included such as domain size and microstrain determined by X-ray diffraction and the S-parameter determined by positron annihilation. Understanding the relationships of microstructure and properties is necessary for developing new advanced materials with higher irradiation tolerance.

• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2006.09b
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• pp.1161-1162
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• 2006

In order to overcome the recrystallization embrittlement and irradiation embrittlement of Mo, which are major problems for its fusion applications, internally nitrided Mo alloys were prepared by a novel multi-step internal nitriding. Neutron irradiation was performed in the Japan Material Testing Reactor (JMTR). After irradiation, nitrided Mo alloys exhibited $\iota$ ower ductile-brittle transition temperature than irradiated TZM. These results suggested that multi-step internal nitriding was effective to the improvement in the embrittlement by irradiation. Transmission electron microscope observation revealed that TiN particles precipitated by nitriding acted as a sink for irradiation-induced defects.