Fiber laser welding has been developed for precise welding of small and complicate components assembled on the nuclear fuel irradiation test rig. In this research, laser welding characteristics of STS316L, the main material of nuclear fuel test rig, have been studied. Several welding experiments were carried out in lap welding of the tube and the end cap made of STS316L. Process variables such as non-focal length, shield gas, laser frequency and power are optimized and compared with the results of Zircaloy-4.
Zr 합금의 재결정 거동 및 미세조직 변화에 미치는 열처리 온도 및 시간의 영향의 조사하기위하여 순수 Zr과 Zircaloy-4, Zr-0.88n-0.4Nb-0.4Fe-0.2Cu, Zr-1Nb 합금을 냉간가공한 후 400˚C~900˚C에서 각각 30분~5000분 동안 열처리하였다. 열처리 온도에 따른 Zr합금의 경도, 미세조직 및 석출물 특성을 미소경도기, 광학 현미경 및 투과 전자 현미경을 이용하여 조사하였다. 냉간 가공채는 400˚C에서 600˚C 범위에서 재결정이 일어났는 데 합금원소가 증가함에 따라 재결정온도가 상승했고 결정립 성장이 억제되었다. 그리고 합금원소 증가에 따른 경도증가 영향이 재결정 이후에도 지속되었다. 열처리 온도 및 시간에 비례하여 재결정 이후 결정립 크기는 증가한 반면 경도변화는 상대적으로 미미하였다. Fe나 Cu가 Zr에 첨가될 경우 회복중 경도증가가 수반되는데, 이는 회복중 생성과 관련이 있는 것으로 사료된다.
The effect of β-heat treatment on th microstructure, mechanical properties and texture in the nuclear fuel cladding of Zircaloy-4 tubes was chosen at 1000, 1100 and 1200˚C, and the tubes were heat-treated by a high frequency vacuum induction furnace. Morphology of the second phase particles and α-grain of as-received tubes were markedly changed by heat treatment. The average sizes of second phase particles of as-received and β-heat treated tubes were 0.1μm and 0.076μm, respectively. However, the average sizes of second phase particles were not much changed in the β-heated temperatures. With increasing heat treatment temperatures, the 0.2% yield strength and the hoop strength were decreased because of changes in preferred orientation as will as α-plate width. Heat treated Zircaloy-4 tubes exhibited texture changes but the preferred orientation of grains still remained.
Zircaloy-4와 Zr-2.5Nb 합금의 부식에 미치는 냉각속도와 소둔온도의영향을 조사하기 위해서 여러 가지 방법으로 열처리된 시편에 대해서 autoclave 부식시험을 실시하였다. 냉각속도의 영향을 조사하기 위해서 시편을 1050˚C에서 30분 가열 후 염빙수냉, 수냉, 유냉, 공냉, 노냉의 방법에 의해 열처리하였으며, 소둔온도의 영향을 조사하기 위해서 α온도, α+β온도, β온도구역에서 열처리하였다. 500˚C부식시험 결과, Zircaloy-4합금에서는 nodule형 부식이 발생되는 반면에 Zr-2.5Nb 합금에서는 nodule형 부식이 발생되지 않았다. Zirfcaloy-4 합금에서는 nodule형 부식이 발생되는 반면에 Zr-2.5Nb 합금에서는 nodule형 부식이 발생되지 않았다. Zircaloy-4합금은 냉각속도가 빠를수록 내식성이 증가하는 반면에 Zr-2.5Nb합금은 냉각속도가 빠를수록 내식성이 감소하는 경향을 보였다. 또한 소둔온도가 증가할수록 Zr-2.5Nb 합금의 내식성은 감소하는 결과를 보였다. Zircaloy-4의 내식성은 Fe, Cr 원소의 기지내 분포와 석출물의 분포에 의해 지배를 받으며 Zr-2.5Nb 합금의 내식성은 기지조직내의 Nb 농도와 β-Nb상에 의해 지배를 받는 것으로 사료된다.
LiOH-H3BO3 용액중에서의 Zircaloy-4 핵연료 피복관의 부식가속과 억제현상을 조사하고 이러한 부식특성에 미치는 Li 및 B의 영향을 해석하기 위하여, 여러 조건의 LiOH-H3BO3</TEX> 용액을 사용하여 350˚C, 165bar의 고온, 고압 조건에서 Zircaloy-4 피복관의 노외 부식시험을 수행하였다. 원전 수화학 모의조건에 대응되는 용액 중에서의 부식속도의 천이는 물 분위기에서 보다 빨리 발생되고 천이후 물 분위기와 거의 유사한 부식속도를 나타내는 천이적 후의 부식거동을 보였다. 한편 pH의 변화는 부식특성에 큰 영향을 미치지 않았다. 부식가속과 억제 모의실험으로부터, 산화막내로 침투하는 Li의 양이 용액중 Li 농도에 크게 의존하며, Li 농도가 일정하게 정해진 용액의 경우 B 첨가에 관계없이 산화막내에 일정량의 Li이 농축될수 있다는 가정을 제시하였다. 또한 B 첨가에 의한 부식억제가 B 또는 B-(OH) 화합물의 산화막내 Li 침투 억제에 의한 것이 아니라 일들에 의해 산화막내로 산화성 성분의 이동이 억제되는데 기인할 수 있음을 제시하였다. 부식가속 개시점에 대응되는 산화막 두께측정 결과와 용액내 Li 농도간의 관계로부터, 용액중 Li 농도가 높을수록 부식가속이 얇은 산화막 두께에서 시작됨을 알았다. 특히 노내조건에서의 핵연료 피복관의 부식가속이 산화막내 Li 농축에 의해 일어나는 부식특성으로 해석될 수 있음을 보였다.
가압 경수로 핵연료의 중성자 조사 조건에서 Zircaloy피복관의 3축방향으로의 변동거동은 집합도 계수에 따른 크립 이방성고 조사성장 이방성을 통하여 분석될 수 있다. 이러한 크립과 조사성장의 이방성이 Zircaloy피복관의 각 축방향 변형율에 미치는 영향을 평가할 수 있는 방법론이 제시되었다. 연소 후 측정된 KOFA Zircaloy-4피복관의 변형율과 핵연료 성능 분석 코드의예측치를 토대로 하여 각 축방향 변형율을 계산한 결과 KOFA Aircaloy-4 피복관의 원주방향 변형은 크립에 의해 주로 일어난 반면, 피복관의 길이방향 변형은 조사성장에 의하여 일어났으나 낮은 조사량에서는 크립의 영향도 상당히 큰것으로 나타났다.
국산 핵연료에 사용되는 KOFA Zircaloy-4피복관의 조사성장 거동을 평가하고 제조 공정이 서로 다른 Siemens사 피복관의 조사성장거동과 비교하기 위하여 고리 2호기에 장전된 핵연료 피복관의 조사성장이 측정되었다. KOFA Zircaloy-4피복관은 최종 열처리시의 부분 재결정화로 인하여 fully annealed Zircaloy피복관고 Siemens사 피복관의 측정된 조사성장율이 차이는 제조공정의 차이에 기인한 피복관 집합도 계수의 차이로서 설명할 수 있었다. 고리 2호기 국산핵연료에서 측정된 자료를 이용하여 KOFA Zircaloy-4 피복관의 2단계 조사성장 모델이 유도되었는데 향후 측정자료가 많이 축적되면 유도된 모델의 정확성이 보다 명확하게 검증될 수 있을 것이다.
The hydride reorientation (HR) of used nuclear fuel cladding after operation affects the integrity during intermediate and disposal storage, as well as the handling processes associated with transportation and storage. In particular, during dry storage, which is an intermediate storage method, the radial hydrogen redistributes into circumferential hydrogen, increasing the embrittlement of used nuclear fuel cladding. This hydride reorientation is influenced by various key factors such as circumferential stress (hoop stress) due to internal rod pressure, maximum temperature reached, cooling rate during storage, and the concentration of precipitated hydrogen during irradiation. To simulate long-term dry storage of used nuclear fuel, hydrogenated Zircaloy-4 cladding (CWSRA) specimens were used in hydride reorientation tests under various hoop stress conditions (70, 80, 90, and 110 MPa) for extended cooling periods (3 months, 6 months, and 12 months). After the hydride reorientation tests, the cladding’s offset strain (%) was evaluated through a ring compression test, a mechanical property test encompassing both ductility and brittleness. In this study, the offset deformation of the hydride reorientation specimens was compared and evaluated through ring tensile tests. In this study, the offset deformation values were compared and evaluated through ring tensile tests of the hydride reorientation test specimens. Hydrogen in zirconium cladding reduces ductility from a physical perspective and induces rapid plastic deformation. Generally, even in hydrogenated unirradiated cladding, it maintains a tensile strength of around 800 MPa at room temperature. However, high hydrogen content accelerates plastic deformation. In contrast, samples with radial hydrogen distribution exhibit fracture behavior in the elastic region below 500 MPa. This is attributed to the directional of radial hydrogen distribution. Specimens with a hydrogen concentration of 200 ppm fracture faster than those with hydrogen concentrations exceeding 400 ppm. This is believed to be due to the ease of reorientation of radial hydrogen in cladding with relatively low hydrogen content. Although the consistency of the test results is not ideal, ongoing research is needed to identify trends in hydride reorientation from a cladding perspective.
Zircaloy-4 is utillzed in nuclear fuel rod cladding due to its strength and corrosion resistance. However, it can undergo deformation over time, known as creep, which poses a safety risk in reactors. Furthermore, hydrogen absorption during reactor operation can alter its properties and affect creep rates. Previous research suggests a trend in which hydrogen concentration corelates unidirectionally with creep rates, either increasing or decreasing as the concentration rises. This trend can also be observed in EPRI’s creep model, EDF-CEA Model-3. However, recent literature has suggested that creep behavior may vary depending on the state of hydrogen presence. Therefore, it has become evident that creep behavior can be influenced not only by hydrogen concentration but also by the state of hydrogen presence, whether it is in a solid solution state or precipitated as hydrides. Our study aimed to compare creep behavior in specimens with hydrogen concentrations below and above solubility limits. We fabricated Zircaloy-4 plate specimens with varying hydrogen concentrations and conducted creep tests. The results revealed that specimens below the solubility limit exhibited decreasing creep rates as hydrogen concentration increased, while those above the limit displayed increasing creep rates. This investigation confirms that the state of hydrogen presence significantly impacts creep behavior within Zircaloy-4 cladding. As part of our additional research plans, we intend to conduct creep tests on the material based on its orientation, whether it is in the rolling direction (RD) or the transverse direction (TD). We also plan to perform creep tests on ring specimens. Additionally, for the ring specimens, we aim to evaluate how creep behavior differs between the cold-worked stress-relieved (CWSR) condition and the recrystallized annealed (RXA) condition achieved through high-temperature heat treatment.
In the process of spent fuel dry storage, which is an intermediate management method, it was found that hydrides in the circumferential direction rearranged into radial hydrides. Various factors, such as hoop stress, peak temperature, cooling rate during the storage period, and hydrogen concentration accumulated during the burnup process, significantly affect the susceptibility of spent fuel cladding. In recent studies based on the hydrogen solubility value of about 210 ppm corresponding to the peak temperature of 400°C, if the threshold stress decreases as the hydrogen concentration increases in the low hydrogen range under 210 ppm, the threshold stress increases as the hydrogen concentration increases in the low hydrogen range under 210 ppm. The fundamental cause of this trend is the diffusion of hydrogen into the high-stress region due to the stress gradient formed in the specimen, and hydrogen compounds which remain undissolved in the circumferential direction, even at the peak temperature, play a crucial role to determine the magnitude of the threshold stress. This study evaluated the behavior of hydride reorientation under various hoop stress conditions (70, 80, 90, and 110 MPa) using unirradiated Zircaloy-4(CWSRA) cladding tubes under long-term cooling conditions (3, 6, and 12 months). The results of analyzing the offset strain by hydrogen concentration for long-term cooling showed that specimens with low hydrogen concentration exhibited higher integrity than specimens with high hydrogen concentration at hoop stresses of 90 and 110 MPa. The HR test using irradiated fuel cladding showed that specimens with low hydrogen concentrations exhibited relatively higher susceptibility. To quantify these results, it is necessary to research further in detail by repeated tests.
The hydride reorientation (HR) of the post-irradiated nuclear fuel cladding after use affects the integrity of the spent nuclear fuel. During the dry storage process, which is an intermediate storage method, it was found that the hydride in the circumferential direction is rearranged into radial hydride, and this is believed to be due to factors such as hoop stress, peak temperature, accumulated hydrogen concentration, and cooling rate during the storage period. f(HR) = f(Tmax) + f(σH) + f(CH) + f(△T) + f(10Cy) + f(cooling rate) + ...... To simulate long-term dry storage of spent nuclear fuel, the hydride reorientation behavior was evaluated using unirradiated Zircaloy-4 (CWSRA) cladding with hydrogen charged under various hoop stresses (70, 80, 90, and 110 MPa) at long-term cooling periods (3, 6, and 12 months). Test results showed that as the cooling time increased, the sample with 90 MPa hoop stress at a maximum temperature of 400°C approached the ductility recommendation limit of 2%. In a 90 MPa hoop stress specimen with 3 months cooling period at peak temperature of 400°C, the offset strain was 4.24% at room temperature RCT, while it showed the result of 2.86% for the cooling period of 12 months. On the other hand, the specimen with hoop stress of 110 MPa and cooling period of 12 months showed result of 1.4%. The test results need to take into account errors in hydrogen charging and hydrogen analysis, and it is necessary to consider reproducibility through repeated tests. These results indicate the need for continued attention to the evaluation of the effects of hydride reorientation due to long-term cooling in the context of the integrity of spent fuel.
A long-term cooling effect on hydride reorientation of a cladding tube can affect the integrity of spent nuclear fuel transportation and long-term storage. In this study, experimental setup for investigating the degree of radial reorientation of hydrides in the circumferential direction during the long-term cooling was established. The experimental setup was designed to be simplified since the long-term evaluation requires a long term period such as 12, 18 and 24 months when the cladding tube specimen is gradually cooled down from 400°C to 100°C. For the test, hydrogen-charged specimens of 100 ppm, 200 ppm, and 500 ppm were prepared. The specimen was sealed with fixtures and check valve, and was pressurized up to 90 Mpa. To heat the specimen, a box-type furnace was used while the temperature of the specimen was measured from thermocouples attached to the specimen. After the heat treatment, the long-term cooling was performed by developing temperature control program to investigate several cooling rate conditions of the specimen. As a reference case, microstructure and brittle property of the hydrogen-charged specimens of 100 ppm, 200 ppm, and 500 ppm without the long-term cooling was observed. In the case of the hydrogen content, it was uniformly distributed in circumferential direction although it was non-uniform in the axial direction. In the case of the brittle property, a compression test was performed. For the future work, the microstructure and brittle property of the hydrogencharged specimens after the several long-cooling conditions were investigated. Then, the degree of radial reorientation of hydrides in the circumferential direction during the long-term cooling was studied.
The evaluation of the damage ratio of spent nuclear fuel is a very important intermediate variable for dry storage risk assessment, which requires an interdisciplinary and comprehensive investigation. It is known that the pinch load applied to the cladding can leaded to Mode-3 failure and the cladding becomes more vulnerable to this failure mode with the existence of radial hydrides and other forms of mechanical defects. In this study, the failure resistance of Zircaloy-4 cladding against the pinch load is investigated using numerical simulations assuming the existence of radial hydrides. The simulation model is based on the microscopic images of cladding. A pixel-based finite element model was created by separating the Zircaloy-4 and hydride using the image segmentation method. The image segmentation method uses a morphology operation basis, which is a preprocessing method through erosion operation after image expansion to enable normal segmentation by emphasizing pixels corresponding to hydrides. The segmented images are converted into a finite element model by assigning node and element numbers together with corresponding material properties. Using the generated hydride cladding finite element model, several numerical methods are investigated to simulate crack propagation and cladding failure under pinch load. Using extended finite element (XFEM) models the initiation and propagation of a discrete crack along an arbitrary, solution-dependent path can be simulated without the requirement of remeshing. The applicability of fracture mechanical parameters such as stress intensity, J-integral was also investigated.