• 비파괴검사학회지
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• 제9권2호
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• pp.42-49
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• 1989

It has not been performed to inspect the underclad cracking in Korea nuclear plant since there is no Code Requirements for inspection. However, underclad cracks in nuclear pressure vessels were reported firstly in 1970. The objection of this study is to be established the ultrasonic inspection techniques for underclad cracking. The ultrasonic inspection of bimetalic stainless steel weld is very difficult by high attenuation and multiple scattering at weld surface and weld/base metal interface. The various inspection methods using $70^{\circ}$ refracted longitudinal wave, 50/70 tandem transducer, $45^{\circ}\;and\;60^{\circ}$ single shear wave are compared. Experiments on limited specimens applied same condition to nuclear pressure vessels shows that $70^{\circ}$ refracted longitudinal wave method is the best one for the detection of underclad cracks. 50/70 tandem transducer using SPOT(Satellite Pulse Observation Technique) is more effective for underclad crack sizing than other sizing methods.

• Structural Engineering and Mechanics
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• 제37권3호
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• pp.257-265
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• 2011

Plant-specific analyses of 5 types of domestic reactors in Korea are performed to assure the structural integrity of the reactor pressure vessel (RPV) during transients which are expected to initiate pressurized thermal shock (PTS) events. The failure probability of the RPV due to PTS is obtained by performing probabilistic fracture mechanics analysis. The through-wall cracking frequency is calculated and compared to the acceptance criterion. Considering the fluence at the end of life expected by surveillance test, the sufficient safety margin is expected for the structural integrity of all reactor pressure vessels except for the oldest one during the pressurized thermal shock events. If the flaw with aspect ratio of 1/12 is considered to eliminate the conservatism, the acceptance criteria is not exceeded for all plants until the fluence level of $8{\times}10^{19}\;n/cm^2$, generating sufficient margin beyond the design life.

• 대한토목학회논문집
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• 제29권6A호
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• pp.577-585
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• 2009

이 논문은 원전 격납건물의 극한내압능력 평가와 비선형해석을 수행하기 위하여 개발된 해석프로그램인 9절점 퇴화 쉘 유한요소에 대하여 기술하였다. 개발된 쉘 유한요소는 퇴화 고체기법과 구조물에서 발생하는 횡전단변형도를 고려하기 위하여 Reissner-Mindlin(RM)가정을 도입하였다. 콘크리트의 재료모델은 등가응력-등가변형률의 관계를 이용하여 콘크리트의 응력과 변형률의 수준을 결정하고, 콘크리트에 균열이 발생하면 부착응력을 고려하는 인장강성모델과 균열면에서의 전단전달 메카니즘 그리고 균열면에서 압축강도 감소모델 등으로 재료적 거동을 나타내었다. 또한 균열발생기준으로 압축-인장영역에는 Niwa가 제안한 응력포락선을 도입하였고, 인장-인장영역에는 Aoyagi-Yamada가 제안한 응력포락선을 사용하였다. 개발된 프로그램의 성능은 다양한 수치예제를 통하여 검증하였다. 검증예제 결과로부터 개발된 쉘 유한요소를 이용한 해석결과는 실험결과 또는 다른 해석결과와 유사한 결과를 도출하였다.

• Nuclear Engineering and Technology
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• 제27권2호
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• pp.226-233
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• 1995

CANDU 원자로에서 심각하게 대두되는 압력관 파손을 방지하기 위해 압력관의 두께를 증가시키는 방안이 연구되었다. 본 연구에서는 압력관 두께변화가 Zr-2.5Nb 압력관의 응력, 수소농도 및 수소지연파괴에 미치는 영향에 대해 연구를 수행하였다. 압력관 두께가 현재의 4.2 mm에서 5.2 mm로 증가할 경우에 압력관이 받는 응력과 발전소 가동중에 누적되는 중수소 흡수량은 19% 줄어드는 것으로 나타났으며, 압력관에 균열이 발생할 경우 발전소 냉각동안에 일어나는 균열 성장은 상당히 감소한다. 수소지연파괴는 압력관이 받는 응력과 누적되는 수소량에 비해 지배되는데 이와 같은 결과로부터 두꺼운 압력관은 수소지연파괴 관점에서 상당한 이점이 있는 것으로 평가되었다. 그러나 압력관 두께 증가는 수소지연파괴의 성장속도를 가속할수도 있으므로 앞으로 연구할 사항이다.

• 소성∙가공
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• 제20권2호
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• pp.173-178
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• 2011

Environmental Stress Cracking(ESC) is one of the most common causes of unexpected brittle failure of thermoplastic polymers. The exposure of polymers to liquid chemicals tends to accelerate the crazing process, initiating crazes at stresses that are much lower than the stress causing crazing in air. In this study, ESC of acrylonitirile butadiene styrene(ABS) was investigated as a function of the molding conditions such as injection velocity, packing pressure, and melt temperature. A constant strain was applied to the injection molded specimens through a 1.26% strain jig and a mixture of toluene and isopropyl alcohol was used as the liquid chemical. In order to examine the effects of the molding conditions on ESC, an experimental design method was adopted and it was found that the injection velocity was the dominant factor. In addition, predictions from numerical analyses were compared with the experimental results. It was found that the residual stress in the injection molded part was associated with the environmental stress cracking resistance (ESCR).

• Corrosion Science and Technology
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• 제3권1호
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• pp.30-33
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• 2004

It has been reported that some events of a rupture of seam generator tube have occurred in nuclear power plants around the world. Main causes of the leakage are from various types of corrosion in the steam generator(SG) tubings. Primary water stress corrosion cracking(PWSCC) of steam generator tubings have occurred in many tubes in Korean plant, and they were repaired using sleeves or plugs, In order to develop proper repair criteria, it is necessary to ascertain the leak behavior of the tubings. A high pressure leak and burst testing system was manufactured. Various types of Electro Discharged Machined (EDM) notches were developed on the SG tubes. Leak rate and burst pressure were measured on the tubes at room temperature. Burst pressure of the part through wall defected tubes depends on the defect depth, Water flow rates after the burst were independent of the t1aw types; tubes having 20 to 60 mm long EDM notches showed similar flow rates regardless of the defect depth. A fast pressurization rate gave the tube a lower burst pressure than the case of a slow pressurization.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 1997년도 특별강연 및 추계학술발표 개요집
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• pp.16-18
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• 1997

• 비파괴검사학회지
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• 제13권2호
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• pp.31-39
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• 1993

국내 원자력 압력 용기의 피복하부 결함에 대한 검사는 1970년대 초부터 이 피복하부 결함에 대한 보고가 되기 시작하였으나 적용 규격인 ASME Code의 요구사항이 아니므로 현재까지는 검사를 수행해 오지 않고 있다. 본 실험의 목적은 이러한 피복하부 결함을 검출하는데 적절한 초음파검사 방법의 조건을 찾는 것이다. 실험은 $70^{\circ}$굴절종파, 50/70 multibeam탐촉자, SLIC-50 탐촉자 등을 사용하여 원자력 압력용기의 피복 용접과 같은 조건하에서 피복용접 시킨 초음파보정 시험편과 demonstration시험편에 대하여 수행하였다. 실험 결과 피복하부 결함의 검출은 50/70 multibeam탐촉자가 효과적이었으면, 피복하부 결함의 길이 평가는 $70^{\circ}$굴절종파로 수행하는 것이 바람직한 것으로 나타났으며, 반면에 피복하부 결함의 깊이 측정은 SLIC-50 탐촉자를 사용하여 M-SPOT 방범과 M-PET 방법으로 평가하는 것이 가장 효과적인 것으로 사료된다.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.171-178
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• 2009

The rate of delayed hydride cracking (DHC), V, has been measured in cold-worked and stress-relieved Zircaloy-4 fuel cladding using the Pin-Loading Tension technique. At $250^{\circ}C$ the mean value of V from 69 specimens was $3.3({\pm}0.8)x10^{-8}$ m/s while the temperature dependence up to $275^{\circ}C$ was described by Aexp(-Q/RT), where Q is 48.3 kJ/mol. No cracking or cracking at very low rates was observed at higher temperatures. The fracture surface consisted of flat fracture with no striations. The results are compared with previous results on fuel cladding and pressure tubes.

• Journal of Welding and Joining
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• 제34권5호
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• pp.54-60
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• 2016

In order to quantitatively evaluate the solidification cracking susceptibility in laser welds of three types of austenitic stainless steels (type 310: A mode, type 316-A: AF mode, type 316-B: FA mode solidifications), the laser beam welding (LBW) transverse-Varestraint tests consisted of multi-mode fiber laser, welding robot and hydraulic pressure system were performed. As the welding speed increased from 1.67 to 40.0 mm/s, the solidification brittle temperature range (BTR) of laser welds for type 316 stainless steels enlarged (316-A: from 37 to 46 K, 316-B: from 14 to 40 K), while the BTR for type 310 stainless steel reduced from 146 to 120 K. In other words, it founds that solidification cracking susceptibility could not be simply mitigated through application of LBW process, and the BTR variation behavior is quite different upon solidification mode of austenitic stainless steels.