• Proceedings of the Acoustical Society of Korea Conference
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• autumn
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• pp.231-234
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• 2000

In relation to a non-destructive evaluation of irradiation damage of micro-structure of interstitial, void and dislocation, the changes in the hysteresis loop and Barkhausen noise amplitude and the harmonics frequency due to neutron irradiation were measured and evaluated. The Mn-Mo-Ni low alloy steel of reactor pressure vessel was irradiated to a neutron fluence of $2.3\times10^{19}n/cm^2$ $(E\ge1MeV)$ at $288^{\circ}C.$The saturation magnetization of neutron irradiated metal did not change. Neutron irradiation caused the coercivity to increase, whereas susceptibility to decrease. The amplitude of Barkhausen noise parameters associated with the domain wall motion were decreased by neutron irradiation. The spectrum of Barkhausen noise was analyzed with an applied frequency of 4Hz and 8Hz, and a sampling time of 50 $\mu$ sec and 20 $\mu$ sec. The harmonic frequency of Joule effect shows 4Hz, 8Hz, 12Hz and 16Hz reflected from an unirradiated specimen. On the contrary, the harmonic frequency disappeared for the irradiated specimen. Harmonic frequency of induced voltage of sinusoidal magnetic field And Spectrum of Barkhausen noise on material is determined.

• Journal of Ocean Engineering and Technology
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• v.8 no.2
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• pp.173-184
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• 1994

원자력 압력용기강의 탄소성 파괴인성값 $J_IC$를 CT형 시험편을 이용하여 검토하였으며, 평활 시험편 및 측면홈 시험편의 두께는 각각 $B_O$=25.4mm, $B_N$=20.4mm 이다. 측면홈의 깊이는 19.7% 이며, 홈의 각도는 90 .deg.로 가공하였다. 탄소성 파괴인성시험은 ASTM E813-81과 JSME S001-81의 추천방법에 따라 실시하였다. 두 추천방법으로 실험한 결과 ASTM 방법에 의한 $J_IC$값이 과대평가됨으로써 부대조건에 만족되지 못하였지만 JSME방법은 만족되었다. 측면홈 시험편은 R고선법에 의한 ductile tearing의 결정이 평활 시험편보다 용이하였으며, 이에 따른 $J_IC$값의 정확성을 배가 할 수 있었다. 또한 임계 스트레치존 폭($SZW_C$)은 측면홈에 의한 높은 3축응력이 발생되어 평활시험편보다 적게 나타났으며, 이러한 복합적인 원인에 기인하여 스트레치존법에 의한 $J_IC$평가는 R곡선법에 의한 평가보다 약간 과대평가됨을 알 수 있었다.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1343-1354
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• 2022

Experiments were performed to measure the critical heat flux (CHF) on a vertical surface abutting a coarse packed bed of spherical particles. This geometry is representative of a CANDU reactor calandria tubesheet facing the end shield cavity during the in-vessel retention (IVR) phase of a severe accident. Deionized light water was used as the working fluid. Low carbon steel shielding balls with diameters ranging from 6.4 to 12.7 mm were used, allowing for the development of an empirical correlation of CHF as a function of shielding ball diameter. Previously published data is used to develop a more comprehensive empirical correlation accounting for the impacts of both shielding ball diameter and heating surface height. Tests using borosilicate shielding balls demonstrated that the dependence of CHF on shielding ball thermal conductivity is insignificant. The deposition of iron oxide particles transported from shielding balls to the heating surface is verified to increase CHF non-trivially. The results presented in this paper improve the state of the knowledge base permitting quantitative prediction of CHF in the CANDU end shield, refining our ability to assess the feasibility of IVR. The findings clarify the mechanisms governing CHF in this scenario, permitting identification of potential future research directions.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.2 s.173
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• pp.303-310
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• 2000

Fracture toughness of five different reactor pressure vessel steels was characterized in the transition temperature region by the ASTM E1921-97 standard method using Charpy-sized small specimens. T he predominant fracture mode of the tested steels was transgranular cleavage in the test conditions. A statistical analysis based on the Weibull distribution was applied to the interpretation of the scattered fracture toughness data. The size-dependence of the measured fracture toughness values was also well predicted by means of the Weibull probabilistic analysis. The measured fracture toughness transition curves followed the temperature-dependence of the ASTM master curve within the expected scatter bands. Therefore, the fracture toughness characteristics in the transition region could be described by a single parameter, so-called the reference temperature (T。), for a given steel. The determined reference temperatures of the tested materials could not be correlated with the conventional index temperatures from Charpy impact tests.

• Proceedings of the KSME Conference
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• 2000.04a
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• pp.43-48
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• 2000

In this study, a small punch creep(SP-Creep) test using miniaturized specimen$(10{\times}10{\times}0.5mm)$ has been described for the development of the newly semi-destructive creep test method for high temperature structural components such as headers and tubes of boiler turbine casino and rotor and reactor vessel. The SP-Creep testing technique has been applied to 2.25Cr-1Mo(STBA24) steel used widely as boiler tube material and the creep test temperature are varied at $550^{\circ}C{\sim}600^{\circ}C$. The overall deformations of SP-Creep curves are definitely depended with applied load and creep test temperature and show the creep behaviors of three steps like conventional uniaxial creep curves. The steady state creep rate${\delta}_{ss}$ of SP-Creep curve for miniaturized specimen increases with increasing creep temperature, but the exponential value with creep loading is decreased. The activation energy$(Q_{spc})$ during SP-Creep deformation with various test temperatures shows 605.7kJ/mol that is g.eater than 467.4kJ/mol reported in uniaxial creep test. This may be caused by the difference of stress states during creep deformation In two creep test. But from the experimental results, e.g. SP-Creep curve behaviors, the steady state creep rate${\delta}_{ss}$ with creep temperature, and the exponential value(n) with creep loading, it can be summarized that the SP-Creep test may be a useful test method to evaluate the creep properties of the heat resisting material.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.340-350
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• 2015

This paper proposes strain-based acceptance criteria for assessing plastic instability of the safety class 1 nuclear components made of ferritic steel during level D service loads. The strain-based criteria were proposed with two approaches: (1) a section average approach and (2) a critical location approach. Both approaches were based on the damage initiation point corresponding to the maximum load-carrying capability point instead of the fracture point via tensile tests and finite element analysis (FEA) for the notched specimen under uni-axial tensile loading. The two proposed criteria were reviewed from the viewpoint of design practice and philosophy to select a more appropriate criterion. As a result of the review, it was found that the section average approach is more appropriate than the critical location approach from the viewpoint of design practice and philosophy. Finally, the criterion based on the section average approach was applied to a simplified reactor pressure vessel (RPV) outlet nozzle subject to SSE loads. The application shows that the strain-based acceptance criteria can consider cumulative damages caused by the sequential loads unlike the stress-based acceptance criteria and can reduce the overconservatism of the stress-based acceptance criteria, which often occurs for level D service loads.

• Journal of the Korean Society for Nondestructive Testing
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• v.18 no.6
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• pp.477-483
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• 1998

Hysteresis loop, Barkhausen noise(BN), and hardness were measured in the neutron irradiated RPV steel for various fluence, irradiated dose up to $10^{18}n/cm^2$. The coercivity, remanence and maximum induction of neutron irradiated samples did not change significantly, but the BNA and BNE were decreased as the neutron irradiation increased. The changes of BNE and BNA were characterized by three stages with respect to neutron dose. The BNA and BNE were decreased with an increase of neutron dose to $10^{12}n/cm^2$, and remained nearly constant up to $10^{16}n/cm^2$, then were decreased rapidly with an increase of the neutron dose above $10^{16}n/cm^2$. On the other hand, the hardness was observed revesely with the change of BNA and BNE.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.246-258
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• 2016

The shield building of AP1000 was designed to protect the steel containment vessel of the nuclear reactor. Therefore, the safety and integrity must be ensured during the plant life in any conditions such as an earthquake. The aim of this paper is to study the effect of water in the water tank on the response of the AP1000 shield building when subjected to three-dimensional seismic ground acceleration. The smoothed particle hydrodynamics method (SPH) and finite element method (FEM) coupling method is used to numerically simulate the fluid and structure interaction (FSI) between water in the water tank and the AP1000 shield building. Then the grid convergence of FEM and SPH for the AP1000 shield building is analyzed. Next the modal analysis of the AP1000 shield building with various water levels (WLs) in the water tank is taken. Meanwhile, the pressure due to sloshing and oscillation of the water in the gravity drain water tank is studied. The influences of the height of water in the water tank on the time history of acceleration of the AP1000 shield building are discussed, as well as the distributions of amplification, acceleration, displacement, and stresses of the AP1000 shield building. Research on the relationship between the WLs in the water tank and the response spectrums of the structure are also taken. The results show that the high WL in the water tank can limit the vibration of the AP1000 shield building and can more efficiently dissipate the kinetic energy of the AP1000 shield building by fluid-structure interaction.

• Nuclear Engineering and Technology
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• v.22 no.4
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• pp.337-350
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• 1990

A post-irradiation annealing study was conducted with use of reactor pressure vessel(RPV) steel A533B Cl.1 base metal irradiated to a dose of 4.84$\times$10$^{18}$ n/$\textrm{cm}^2$ at about 38$0^{\circ}C$. Microhardness and positron annihilation (PA) methods were used to obtain better understanding of the recovery of radiation hardening. Isochronal anneal experiments indicated that two recovery processes occur during annealing of irradiated specimens. The first recovery process occurs in the temperature range of 280-3O5$^{\circ}C$, Michrohardness and positron annihilation (PA) methods were used to obtain better understanding of the recovery of radiation hardening. Isochronal anneal experiments indicated that two recovery processes occur during annealing of irradiated specimens. The first recovery process occurrs in the temperature range of 280-305$^{\circ}C$. The variations of Ip, Iw and R parameters indicated that the formation of vacancy clusters by vacancy agglomeration and the annihilation of monovacancies are the first recovery process. The second recovery process occurs in the range of 405-49$0^{\circ}C$ and positron annihilation parameters measured indicated that the dissolution of carbon atoms decorated around vacancy-type defects and possible precipitates, and the annihilation of monovacancies give rise to the second recovery process. It was further indicated that radiation anneal hardening (RAH) in the range of 305-405$^{\circ}C$ between the temperature ranges for the two processes occurs due to the formation of carbon-decorated vacancy clusters and precipitates. The activation energies, orders of reaction and other characteristics of recovery processes were determined by the Meechan-Brinkman method. The activation energy for the first recovery process was determined as 1.76 eV and that for the second recovery process as 2.00eV. These values are lower than those obtained by other workers. This difference may be attributed to the lower copper content of the RPV steel used in the present study. The order of reaction for the first recovery process was determined as 1.78, while that for the second recovery process as 1.67 Non-integer orders of reaction for recovery processes seem to be attributed to the fact that several mechanisms for the first order and the second order of reaction are compounded in one process. This result also supports for the above conclusions from measurements of PA parameters.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.140-148
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• 1994

A qualification test was performed for the iron removal chemical cleaning of the secondary side of nuclear steam generators at the selected temperature, 1$25^{\circ}C$, higher than the standard application temperature, 93$^{\circ}C$. The field cleaning condition for a nuclear unit was tested in a bench scale test loop including a SUS 316 stainless steel autoclave with one gallon capacity as a test vessel. The kinetics of sludge dissolution, corrosion of the secondary side materials and change of solvent chemistry were monitored. Test results indicated that more thorough cleaning was accomplished in less than half of the cleaning time required at 93$^{\circ}C$. And the total corrosions of the secondary side materials were found to be less than the values at 93$^{\circ}C$. While the solvent is recirculated and heated by an external chemical cleaning equipment for the conventional 93$^{\circ}C$ process, the secondary side is heated by the lateral heat of the primary coolant without the recirculation of the cleaning solution, and the solvent is mixed by vigorous boiling induced by periodic ventilation for the high temperature process. The requirement that the reactor coolant pumps should be running during the cleaning operation is the major disadvantage of the high temperature process which also should be considered when chemical cleaning is planned for steam generators under operation.