• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.336-343
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• 1998

An iteration method has been developed for determining crack growth and fracture resistance cure (J-R curve) from the load versus load-line displacement record only. In this method, the hardening curve, the load versus displacement curve at a given crack length, is assumed to be a power-law function, where the exponent varies with the crack length. The exponent is determined by an iterative calculation method with the assumption that the exponent varies linearly with the load-line displacement. The proposed method was applied to the static J-R tests using compact tension(CT) specimens, a three-point bend (TPB) specimen, and a cracked round bar (CRB) specimen as well as it was applied to the quasi-dynamic J-R tests using CT specimens. The J-R curves determined by the proposed method were compared with those obtained by the conventional testing methodologies. The results showed that the J-R curves could be determined directly by the proposed iteration method with sufficient accuracy in the specimens from SA508, SA533, and SA516 pressure vessel steels and SA312 Type 347 stainless steel.

• Journal of Mechanical Science and Technology
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• v.18 no.12
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• pp.2158-2173
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• 2004

In this paper, the higher order terms in the crack tip stress fields are investigated macroscopically for more realistic assessment of structural material behaviors. For reactor pressure vessel material of A533B ferritic steel, effects of crack size and temperature have been evaluated using 3-point SENB specimens through a series of finite element analyses, tensile tests and fracture toughness tests. The T-stress, Q-parameter and q-parameter as well as the K and J-integral are calculated and mutual relationships are investigated also. Based on the evaluation, it has proven that the effect of crack size from standard length (a/W=0.53) to shallow length (a/W=0.11) is remarkable whilst the effect of temperature from -20$^{\circ}C$ to -60$^{\circ}C$ is negligible. Finally, the cleavage fracture toughness loci as a function of the promising Q-parameter or q-parameter are developed using specific test results as well as finite element analysis results, which can be applicable for structural integrity evaluation considering constraint effects.

• Journal of the Korean Society for Nondestructive Testing
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• v.17 no.1
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• pp.31-40
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• 1997

원자로 압력용기 재료의 중성자 조사 취화 문제는 원자력발전소의 안전성 및 수명 관리에 가장 중대 한 영향을 미친다. 재료의 조사 취화를 평가하기 위하여 수행하고 있는 충격 및 인장시험 같은 파괴적 시험 결과는 석출물 크기나 분포, 전위 밀도 등, 재료 자체의 조직학적 특성에 좌우되므로 한정된 시편을 이용한 평가에는 많은 불확실성이 존재하게 된다. 따라서 이와 같은 문제점을 해결하기 위하여 비파괴기술을 이용한 조사 취화 평가에 대한 많은 연구가 진행되고 있다. 현재 원자로 압력용기 재료의 조사 취화에 따른 미세 조직 변화를 분석하기 위하여 응용되고 있는 비파괴기술로는 전기, 자기, 전자기, 초음파 및 경도측정법 등이 있으나 비파괴피험 결과와 미세조직의 변화, 기계적 성질 및 취화 정도 등과의 상관 관계를 정립해야만 기존 파괴적 시험의 대체가 가능하게 된다. 따라서 현재까지 수행되고 있는 여러 비파괴기술을 이용한 조사 취화 평가 연구결과를 비교 분석하여 보다 실현 가능성 있는 비파괴기술을 검토하였다.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1174-1183
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• 2016

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.

• 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.

• 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.

• 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.

• Journal of the Mineralogical Society of Korea
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• v.16 no.2
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• pp.125-133
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• 2003

Saponite was synthesized from talc by hydrothermal method. The starting material was prepared by adding ($NO_3$)$Al_3$$.$$9H_2$O and Mg($NO_3$)$_2$$.$$6H_2$O solution to the talc powder. which was previously activated in air at 800 $^{\circ}C$ together with $Na_2$$CO_3$. The alkalinity of the solution was controlled by $NH_4$OH solution. The autoclaving was carried out in the closed stainless steel vessel (about 1 liter) for 40 hours under the pressure of 25 kgf/$\textrm{cm}^2$ at $ 230^{\circ}C$ The characterization of the reaction product shows that saponite was crystallized successfully. After the experimental results, pressure was not sensitive parameter in the range of 25 ∼ 75 kgf/$\textrm{cm}^2$, but longer reaction time results in better crystallinity.