• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2019년도 춘계학술논문요약집
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• pp.65-66
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• 2019

• Nuclear Engineering and Technology
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• 제49권7호
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• pp.1472-1482
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• 2017

Zirconium alloy cladding tube specimens were irradiated at $380^{\circ}C$ up to a fast neutron fluence of $7.5{\times}10^{24}n/m^2$ in a research reactor to investigate the effect of neutron irradiation on hydride reorientation and mechanical property degradation. Cool-down tests from $400^{\circ}C$ to $200^{\circ}C$ under 150 MPa tensile hoop stress were performed. These tests indicate that the irradiated specimens generated a smaller radial hydride fraction than did the unirradiated specimens and that higher hydrogen content generated a smaller radial hydride fraction. The irradiated specimens of 500 ppm-H showed smaller ultimate tensile strength and plastic strain than those characteristics of the 250 ppm-H specimens. This mechanical property degradation caused by neutron irradiation can be explained by tensile hoop stress-induced microcrack formation on the hydrides in the irradiation-damaged matrix and subsequent microcrack propagation along the hydrides and/or through the matrix.

• 대한기계학회논문집A
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• 제35권3호
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• pp.281-287
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• 2011

얇은 관 탄성좌굴 공식의 불확실성을 고려하기 위해, 공식을 구성하는 파라미터인 튜브재료의 탄성계수, 푸아송 비, 튜브 두께 및 지름의 불확실성을 분석하였다. 본 연구는 원자로에서 연소되는 핵연료봉과 같이 사용 중 함몰을 엄격히 방지하고 있는 얇은 관의 설계신뢰도를 향상시키는 데에 중요하다. 분석 방법은 각각의 파라미터가 변화할 수 있는 범위를 충분히 포함할 수 있는 최소의 탄성좌굴 안전율을 구하고 이를 선형적으로 합하여 최종의 최소안전율을 구하였다. 최소 안전율에 가장 큰 영향을 미치는 파라미터는 관의 두께로 나타났다. 두께가 얇을수록 더 큰 최소안전율이 필요하며 예로 적용한 지르코늄 합금관의 경우, 두께가 0.254 와 0.87 mm 일 때 최소안전율은 각각 1.547 과 3.487 로 나타났다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.995-996
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• 2005

• 비파괴검사학회지
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• 제23권4호
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• pp.334-341
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• 2003

중수로 압력관이 주위를 둘러싸고 있는 칼란드리아관과 접촉될 경우, 압력관의 내면과 외면의 온도차로 인하여 수소(중수소)의 열 확산이 발생하며 결과적으로 압력관 외면에 수소화물 블리스터가 형성된다. 수소화물 블리스터는 음향학적으로 지르코늄 매질과 연속성을 가지기 때문에 일반적인 초음파 검사법으로는 탐지하기가 어렵다. 지르코늄 압력관 외면에 발생한 작은 수소화물 블리스터를 압력관 내면에서 탐지하기 위하여 초음파 모드 변환 및 속도비 방법을 개발하였다. 정적인 열확산 실험 장치를 사용하여 압력관 외면에 수소화물 블리스터를 성장시켰다. 종파 에코의 비행시간과 모드 변환된 반사 횡파 에코의 비행시간을 측정하여 종파 대 횡파 속도비를 계산하였으며 이를 속도비를 수정된 등고선 표현 방식으로 나타냈다. 초음파 속도비 방법이 일반적인 종파 비행시간방법보다 수소화물 블리스터 탐지 감도가 우수하며 블리스터 형상화 측면에서도 실제 형상과 유사하게 재현하고 있음을 알 수 있었다. 또한 중수로 압력관 초음파 검사사양과 동일하게 최적화 조건에서 수소화물 블리스터 탐지한계는 보수적인 관점에서 압력관 외면에 나타나는 크기를 기준으로 약 $500{\mu}m$로 평가되었다.

• 한국재료학회지
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• 제29권8호
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• pp.483-490
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• 2019

Currently, the Korean nuclear industry uses ZIRLO as material for nuclear fuel cladding(zirconium alloy). KEPCO Nuclear Fuel is in the process of developing a HANA alloy to enable domestic production of cladding. Cladding manufacture involves multistage heat treatments and pickling processes, the latter of which is vital for the removal of defects and impurities on the cladding surface. SMUT that forms on the cladding surface during such pickling process is a source of surface defects during heat treatment and post-treatment processes if not removed. This study analyzes ZIRLO, HANA-4, and HANA-6 alloy claddings to extensively study the SEM/EDS, XRD, and particle size characteristics of SMUT, which are second phase particles that are formed on the cladding surface during pickling processes. Using the analysis results, this study observes SMUT formation characteristics according to Nb concentration in Zr alloys during the washing process following the pickling process. In addition, this study observes SMUT removal characteristics on cladding surfaces according to concentrations of nitric acid and hydrofluoric acid in the acid solution.

• 공업화학
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• 제10권8호
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• pp.1099-1103
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• 1999

침지코팅법에 의해 지르코늄이 혼성된 폴리카르보실란(PZC)을 알루미나 지지체 위에 코팅한 후 573~823 K에서 열분해하여 무기 복합막을 제조하였으며 $1{\mu}m$의 두께를 갖는 균일한 막을 얻을 수 있었다. 무기 복합막의 기체 투과 시험은 투과기체를 He, $N_2$, $CO_2$, $O_2$로 하고 투과온도 범위를 303~423 K에서 수행하였다. 투과온도가 증가할수록 기체 투과계수와 분리계수는 증가하였다. 이러한 현상을 통해 PZC 복합막에 대해서 기체투과흐름은 activated 확산 현상을 나타내고 있음을 확인하였으며 특히 $CO_2$의 경우에 $N_2$에 대한 분리계수는 4.9로 가장 큰 값을 보였다.

• Nuclear Engineering and Technology
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• 제40권2호
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• pp.139-146
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• 2008

This paper presents two of the fuel channel designs being considered for the CANDU-SCWR, a pressure-tube type supercritical water cooled reactor. The first is an insulated pressure tube design. The pressure tube is thermally insulated from the hot coolant by a porous ceramic insulator. Each pressure tube is in direct contact with the moderator, which operates at an average temperature of about $80^{\circ}C$. The low temperature allows zirconium alloys to be used. A perforated metal liner protects the insulator from being damaged by the fuel bundles and erosion by the coolant. The coolant pressure is transmitted through the perforated metal liner and insulator and applied directly to the pressure tube. The second is a re-entrant design. The fuel channel consists of two concentric tubes, and a calandria tube that separates them from the moderator. The coolant enters between the annulus of the two concentric fuel channel tubes, then exits the fuel channel through the inner tube, where the fuel bundles reside. The outer tube bears the coolant pressure and its temperature will be the same as the coolant inlet temperature, ${\sim}350^{\circ}C$. Advantages and disadvantages of these designs and the material requirements are discussed.

• Nuclear Engineering and Technology
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• 제47권6호
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• pp.746-755
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• 2015

To investigate the effect of hydrogen and oxygen contents on hydride reorientations during cool-down processes, zirconium-niobium cladding tube specimens were hydrogen-charged before some specimens were oxidized, resulting in 250 ppm and 500 ppm hydrogen-charged specimens containing no oxide and an oxide thickness of $0.38{\mu}m$ at each surface. The nonoxidized and oxidized hydrogen-charged specimens were heated up to $400^{\circ}C$ and then cooled down to room temperature at cooling rates of $0.3^{\circ}C/min$ and $8.0^{\circ}C/min$ under a tensile hoop stress of 150 MPa. The lower hydrogen contents and the slower cooling rate generated a larger fraction of radial hydrides, a longer radial hydride length, and a lower ultimate tensile strength and plastic elongation. In addition, the oxidized specimens generated a smaller fraction of radial hydrides and a lower ultimate tensile strength and plastic elongation than the nonoxidized specimens. This may be due to: a solubility difference between room temperature and $400^{\circ}C$; an oxygen-induced increase in hydrogen solubility and radial hydride nucleation energy; high temperature residence time during the cool-down; or undissolved circumferential hydrides at $400^{\circ}C$.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.834-839
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• 2001

Calandria tube wrapping each pressure tube is one of the key structural components of CANDU reactor(Calandria) which is consisted of many pressure tubes containing nuclear fuel assemblies. As the Calandria tube(made of zirconium alloy) is sagging due to its thermal and irradiation creep during the plant operation, it possibly contacts with liquid injection nozzle crossing beneath the Calandria tube, which subsequently results in difficulties on the safe operation. It is therefore necessary to check the gap for the confirmation of no contacts between the two tubes, Calandria tube and liquid injection tube, with a proper measure during the life of plant. In this study, an ultrasonic measurement method was selected among several methods investigated. The ultrasonic device being developed for the measurement of the gap was introduced and its preliminary performance test results were presented here. The gap between LIN and CT at site was measured using by this ultrasonic device at site.