• Nuclear Engineering and Technology
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• 제50권2호
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• pp.218-222
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• 2018

An oxide-dispersion-strengthened (ODS) layer was formed on Zircaloy-4 tubes by a laser beam scanning process to increase mechanical strength. Laser beam was used to scan the yttrium oxide ($Y_2O_3$)-coated Zircaloy-4 tube to induce the penetration of $Y_2O_3$ particles into Zircaloy-4. Laser surface treatment resulted in the formation of an ODS layer as well as microstructural phase transformation at the surface of the tube. The mechanical strength of Zircaloy-4 increased with the formation of the ODS layer. The ring-tensile strength of Zircaloy-4 increased from 790 to 870 MPa at room temperature, from 500 to 575 MPa at $380^{\circ}C$, and from 385 to 470 MPa at $500^{\circ}C$. Strengthening became more effective as the test temperature increased. It was noted that brittle fracture occurred at room temperature, which was not observed at elevated temperatures. Resistance to dynamic high-temperature bursting improved. The burst temperature increased from 760 to $830^{\circ}C$ at a heating rate of $5^{\circ}C/s$ and internal pressure of 8.3 MPa. The burst opening was also smaller than those in fresh Zircaloy-4 tubes. This method is expected to enhance the safety of Zr fuel cladding tubes owing to the improvement of their mechanical properties.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1138-1147
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• 2018

Delayed hydride cracking (DHC) is an important failure mechanism for Zircaloy tubes in the demanding environment of nuclear reactors. The threshold stress intensity factor, $K_{IH}$, and critical hydride length, $l_C$, are important parameters to evaluate DHC. Theoretical models of them are developed for Zircaloy tubes undergoing non-homogenous temperature loading, with new stress distributions ahead of the crack tip and thermal stresses involved. A new stress distribution in the plastic zone ahead of the crack tip is proposed according to the fracture mechanics theory of second-order estimate of plastic zone size. The developed models with fewer fitting parameters are validated with the experimental results for $K_{IH}$ and $l_C$. The research results for radial cracking cases indicate that a better agreement for $K_{IH}$ can be achieved; the negative axial thermal stresses can lessen $K_{IH}$ and enlarge the critical hydride length, so its effect should be considered in the safety evaluation and constraint design for fuel rods; the critical hydride length $l_C$ changes slightly in a certain range of stress intensity factors, which interprets the phenomenon that the DHC velocity varies slowly in the steady crack growth stage. Besides, the sensitivity analysis of model parameters demonstrates that an increase in yield strength of zircaloy will result in a decrease in the critical hydride length $l_C$, and $K_{IH}$ will firstly decrease and then have a trend to increase with the yield strength of Zircaloy; higher fracture strength of hydrided zircaloy will lead to very high values of threshold stress intensity factor and critical hydride length at higher temperatures, which might be the main mechanism of crack arrest for some Zircaloy materials.

• 한국재료학회:학술대회논문집
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• 한국재료학회 2001년도 춘계학술발표강연 및 논문개요집
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• pp.47-47
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• 2001

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2003년도 춘계학술발표대회 요약집
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• pp.231.2-231.2
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• 2003

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.278-278
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• 1999

• 한국재료학회:학술대회논문집
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• 한국재료학회 2000년도 추계학술발표강연 및 논문개요집
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• pp.67-67
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• 2000

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 추계학술발표회요약집
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• pp.248.2-248
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• 1999

• 대한기계학회논문집A
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• 제35권8호
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• pp.889-897
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• 2011

본 연구에서는 핵연료 지지격자체의 재료인 Zircaloy-4 와 Zirlo 판재의 이론적 성형한계 예측모델을 제시했다. 먼저 인장시험 및 이방성시험으로 응력-변형률곡선과 이방성계수를 획득했으며, NUMISHEET 96을 따르는 돔장출시험으로 두 재료의 실험적 성형한계도들을 얻었다. 이론적 성형한계도는 성형한계모델과 항복조건의 영향을 받는다. Swift 확산네킹이론, Marciniak-Kuczynski 의 재료결함 모델, Storen-Rice 의 정점이론을 이용해 부변형률이 양인 구간에서의 성형한계 곡선을 구했으며, 부변형률이 음인 구간에는 Hill 의 국부네킹 이론을 적용했다. 또한 재료이방성을 고려하기 위해 Hill 48, Hosford 79 항복조건을 사용 했다. Swift 확산네킹모델 (Hill 48 항복조건 적용)과 Hill 모델은 각각 변형률비가 양과 음인 영역에 대해 Zircaloy-4의 성형한계도를 비교적 정확히 예측하며, Zirlo의 성형한계도는 Hosford 79 항복조건 (a = 8)을 적용한 Storen-Rice 모델로 나타낼 수 있다.

• Corrosion Science and Technology
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• 제6권4호
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• pp.170-176
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• 2007

Iodine-induced stress corrosion cracking (ISCC) properties and the associated ISCC process of Zircaloy-4 and an Nb-containing advanced nuclear fuel cladding were evaluated. An internal pressurization test with a pre-cracked specimen was performed with a stress-relieved (SR) or recrystallized (RX) microstructure at $350^{\circ}C$, in an iodine environment. The results showed that the $K_{ISCC}$ of the SR and RX Zircaloy-4 claddings were 3.3 and 4.8MPa\;m^{0.5}, respectively. And the crack propagation rate of the RX Zircaloy-4 was 10 times lower than that of the SR one. The chemical effect of iodine on the crack propagation rate was very high, which was increased $10^4$ times by iodine addition. Main factor affecting on the micro-crack nucleation was a pitting formation and its agglomeration along the grain boundary. However, this pitting formation on the grain-boundary was suppressed in the case of an Nb addition, which resulted in an increase of the ISCC resistance when compared to Zircaloy-4. Crack initiation and propagation mechanisms of fuel claddings were proposed by a grain boundary pitting model and a pitting assisted slip cleavage model and they showed reasonable results.

• Nuclear Engineering and Technology
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• 제46권3호
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• pp.423-430
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• 2014

An in-pile performance test of HANA claddings was conducted at up to 67 GWD/MTU in the Halden research reactor in Norway over a 6.5 year period. Four types of HANA claddings (HANA-3, HANA-4, HANA-5, and HANA-6) and a reference Zircaloy-4 cladding were used for the in-pile test. The evaluation parameters of the HANA claddings were the corrosion behavior, dimensional changes, hydrogen uptake, and tensile strength after the claddings were tested under the simulated operation conditions of a Korean commercial reactor. The oxide thickness ranged from 15 to 37 mm at a high flux region in the test rods, and all HANA claddings showed corrosion resistance superior to the Zircaloy-4 cladding. The creep-down rate of all HANA claddings was lower than that of the Zircaloy-4 cladding. In addition, the hydrogen content of the HANA claddings ranged from 54 to 96 wppm at the high heat flux region of the test rods, whereas the hydrogen content of the Zircaloy-4 cladding was 119 wppm. The tensile strength of the HANA and Zircaloy-4 claddings was similarly increased when compared to the un-irradiated claddings owing to the radiation-induced hardening.