• 대한기계학회논문집A
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• 제30권12호
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• pp.1526-1533
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• 2006

The aim of this research was to confirm the existence of the thermal creep degradation by hydrides of Zr-2.5Nb pressure tube materials. Small punch creep tests were performed to obtain the relationship between a creep displacement and a loading period at $300^{\circ}C$. A creep stress and a creep strain rate were also converted from the previous results. The creep material constants and the creep stress exponents at the different hydride contents were compared. Finally the hydrides of the axial and circumferential section were observed using OM, SEM and TEM. The following conclusions were made: 1) The degradation of the thermal creep by hydrides was existed and it strongly depended on the hydride contents. 2) As the hydride contents were increased, the creep stress exponents (m) were also increased. 3) Even though the hydride was not precipitated in 50 ppm materials at $300^{\circ}C$, the degradation of thermal creep was found. Therefore, it was believed that this phenomenon strongly related to the hydride precipitation at room temperature.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.1133-1138
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• 2003

The pressure tubes, which contain high temperature heavy water and fuel, are within the core of a CANDU nuclear reactor, and are thus subjected to high stresses, temperature gradient, and neutron flux. Further, it is well known that pressure tubes of cold-worked Zr-2.5Nb materials result in hydrogen diffusion, which create fully-hydrided regions (frequently called Blister). Thus a proper investigation of hydrogen diffusion within zirconium-alloy nuclear components, such as CANDU pressure tube and fuel channels is essential to predict the structural integrity of these components. In this respect, this paper presents numerical investigation of hydrogen diffusion to quantify the hydrogen concentration for blister growth of CANDU pressure tube. For this purpose, coupled temperature-hydrogen diffusion analyses are performed by means of two-dimensional finite element analysis. Comparison of predicted temperature field and blister with published test data shows good agreement.

• 대한기계학회논문집A
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• 제27권1호
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• pp.94-101
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• 2003

The pressure tubes made of cold-worked Zr-2.5Nb alloy are subjected to creep deformation during service period resulting in changes to their geometry such as longitudinal elongation, diameter increase and sagging. To evaluate integrity of them, information on the material creep property of the serviced tubes is essential. As one of the methods with which the creep property is directly measured from the serviced components, small punch(SP) creep testing has been considered as a substitute for the conventional uniaxial creep testing. In this study, applicability of the SP creep testing to Zr-2.5Nb pressure tube alloy was studied particularly by measuring the power law creep constants, A, n. The SP creep test has been successfully applied fur other high temperature materials which have isotropic behavior. Since the Zr-2.5Nb alloy has anisotropic property, applicability of the SP creep testing can be limited. Uniaxial creep tests and small punch creep tests were conducted with Zr-2.5Nb pressure tube alloy along with finite element analyses. Creep constants obtained by each test method are compared. It was argued that the SP creep test result gave results reflecting material properties of both directions. But the equations derived in the previous study for isotropic materials need to be modified. Discussions were made fur future research directions for application of the SP creep testing to Zr-2.5Nb tube alloy.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.252.2-252.2
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• 2003

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 요약집 춘계학술발표대회
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• pp.319-319
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• 2004

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.93-98
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• 2000

Unpredictable failures can occur due to the DHC (delayed hydride cracking) or the degradation of fracture toughness by hydride embrittlement in CANDU pressure tube which can result from the absorption of hydrogen or deuterium in the high temperature coolant. To investigate the hydride embrittlement of CANDU Zr-2.5Nb pressure tube, the transverse tensile test and the fracture toughness test were performed from room temperature to $300^{\circ}C$ using three different specimens which have an AR (As Received), 100, and 200 ppm hydrogen. As the amount of absorbed hydrogen was increased, the transverse yield strength and the ultimate tensile strength were also increased. In addition, as the test temperature became higher they were decreased linearly. While, at room temperature, the hydrogenbsorbed specimens represented the embrittlement which resulted in sudden decreasing of fracture toughness, the fracture characteristics became ductile such as AR specimen at high temperatures. Through the observation of fracture surface using SEM, it was found that the stress state of mixed mode could be related to the fissure which was believed to decrease the global fracture toughness.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(3)
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• pp.176-181
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• 1996

중수로 원자로에서 핵심 구조물인 압력관은 합금이 사용되고 있는데, 고온, 고압의 방사선 분위기에서 사용되기 때문에 가동중 변형을 받는다. 따라서 본 연구에서는 압력관의 건전성 평가기술을 확보하기 위하여 Zr-2.5wt.% Nb 압력관의 집합조직과 온도 변화에 따른 강도 특성을 조사하였다. 집합조직의 변화는 시편의 채취방향을 달리하여 네 가지의 집합조직 변화를 주었으며, 이들 시편에 대해 상온에서 부터 45$0^{\circ}C$까지 온도를 변화시가면서 인장시험을 실시하였다. 온도가 증가함에 따라 시편의 집합조직과 관계없이 네 종류의 시편 모두가 30$0^{\circ}C$까지는 강도가 직선적으로 감소하는 현상을 보이며, 30$0^{\circ}C$에서 37$0^{\circ}C$ 사이의 온도에서 강도가 약간 증가하는 현상을 보였다. 또한 37$0^{\circ}C$이상의 온도에서는 강도가 다시 직선적으로 감소하는 경향을 나타냈다. 이 같은 현상은 30$0^{\circ}C$ 까지는 프리즘 슬림(-slip)으로 전위가 움직이다가 30$0^{\circ}C$ 이상의 온도에서부터 37$0^{\circ}C$까지는 피라미드 슬림(-slip)계의 경쟁 상호작용에 의한 추가적인 응력이 필요하여 나타난 현상이라 사료된다. 온도의 변화에 따른 강도와 연신율의 curve에서 30$0^{\circ}C$와 37$0^{\circ}C$의 범위에서 plateaus region과 연신율이 최소값을 갖는 것은 동적변형시효(dynamic strain ageing)현상과 관련이 있는 것으로 생각되나 추가적인 연구가 필요하다. 집합도가 변함에 따라 인장강도는 T(transverse)>L(longitudinal)> 30。> 45 。순으로 나타났으며 이는 slip과 twin의 복합적인 작용에 의한 것으로 해석된다.

• 대한기계학회논문집A
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• 제28권2호
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• pp.189-195
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• 2004

The pressure tubes, which contain high temperature heavy water and fuel, are within the core of a CANDU nuclear reactor, and are thus subjected to high stresses, temperature gradient, and neutron flux. Further, it is well known that pressure tubes of cold-worked Zr-2.5Nb materials result in hydrogen diffusion, which create fully-hydrided regions (frequently called Blister). Thus a proper investigation of hydrogen diffusion within zirconium-alloy nuclear components, such as CANDU pressure tube and fuel channels is essential to predict the structural integrity of these components. In this respect, this paper presents numerical investigation of hydrogen diffusion to quantify the hydrogen concentration fur blister growth of CANDU pressure tube. For this purpose, coupled temperature-hydrogen diffusion analyses are performed by means of two-dimensional finite element analysis. Comparison of predicted temperature field and blister with published test data shows good agreement.

• 대한기계학회논문집A
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• 제25권3호
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• pp.510-519
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• 2001

The tensile and fracture toughness tests have been conducted to investigate the degradations of mechanical properties induced mainly by neutron irradiations in Zr-2.5Nb CANDU pressure tube materials operated in Wolsung Unit-1. the tests were performed at room, 150, 200, 250, 300 $\^{C}$ for the irradiated and unirradiated specimens in hot cell. The specimens were directly machined from the tube retaining original curvature using specially designed electric discharge machine(EDM). From the tensile tests of the irradiated specimens, it was found that tensile strength was increased and total elongation was decreased compared to those of the unirradiated ones. The active voltages in the fracture toughness tests for the irradiated showed the discontinuous abrupt increases caused by crack jumping in lower temperature. In the crack resistance curves we found the stable crack growth in the unirradiated, whereas the unstable and three crack growth stages in the irradiated specimens due to the accumulated irradiation defects. The various fracture characteristic values in the irradiated are remarkably lower than those of the unirradiated. Through the fractography, we found in the irradiated that smaller dimple and shorter fissures than the unirradiated, and that the fractured surface had three regions that were flat, transition and slant/shear area. These can explain the difference in the crack growth characteristic values of the irradiated and the unirradiated ones.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(2)
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• pp.153-158
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• 1997

Zr-2.5%Nb 합금에서 응력방향에 따른 DHC특성의 차이를 알아보고자 하였다. 판상의 CT시편을 이용하여 수소를 200 ppm 주입하고 응력을 압력관의 길이 방향으로 가하고 notch를 윈주방향으로 한 경우와 원주방향으로 응력을 가하고 notch를 길이 방향으로 한 경우의 균열전파속도를 측정하여 본 결과 길이 방향으로 응력을 가하였을 때 균열전파속도가 1/100 정도 감소하였으며, 균열발생을 위한 임계응력확대계수도 커짐을 알 수 있었다. 그리고 균열전파 방향도 원주방향으로 응력을 가하였을 때는 균열이 precrack을 따라 그대로 진행되었으나, 응력을 길이 방향으로 가하였을 때는 precrack을 따라 균열이 전파되지 못하고 균열분리 현상을 보였다 이것은 원래 모재가 보유하고 있는 집합조직과, 응력에의하여 수소화물이 재배열할 때 기존의 a상에서의 특정 방향 관계를 유지하여 석출함으로써 균열이 수소화물을 따라 전파됨이 원인인 것으로 생각된다. 응력을 원주방향으로 가하였을 때 균열주위에 수소화물이 길게 석출하지만, 응력을 길이 방향으로 기하였을 때는 수소화물이 20$\mu\textrm{m}$ 정도의 작은 크기로 분리된 균열과 같은 방향으로 분포하고 있음을 관찰하였다. 이로부터 집합조직을 개량함으로써 DHC저항성에 대한 효과를 얻을 수 있음을 확인 할 수 있었고 DHCV model에서 방향성을 수소화물의 재배열인자로부터 고려할 필요성이 있음을 알게 되었다.