• Korean Journal of Materials Research
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• v.15 no.2
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• pp.126-133
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• 2005

Mechanical properties of zirconium cladding were evaluated by ring tension test to simulate Reactivity-Initiated Accident (RIA) at high burnup situation as an out-reactor test. Zircaloy-4 cladding was hydrided up to 1000 ppm as well as oxidized up to $100\;{\mu}m$ to simulate high-burnup situation. After simulated high-burnup treatment, ring tension test was carried out from 0.01 to 1/sec to correlate with actual RIA event. The results showed that ductility and circumferential toughness decreased with the hydrogen content and oxide thickness. Hydride generated inside cladding acted as brittle failure. Oxygen influenced cladding tube by the reduction of load bearing area, oxygen embrittlement, and thermal aging. Correlation between in-reactor RIA parameter like fuel enthalpy and out-reactor toughness was performed and showed a reasonable result.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.324-329
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• 2010

Fuel rods using in nuclear power plants consist of uranium dioxide pellets enclosed in zirconium alloy(zircaloy) tubes. It is vitally important for the pellet surface to remain free from pits, cracks and chipping defects after it is loaded into the tubes to prevent local hot spots during reactor operation. This paper investigates the feasibility study for detecting surface flaws of pellets contained within nuclear fuel rod through X-ray tomography simulation. Reconstructed images used by parallel and fan-beam filtered back projection method were presented and confirmed the accessibility between simulation data and MPS(missing pellet surface) image data.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.423-424
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• 2005

The performance of HANA claddings was evaluated in out-of-pile conditions. All the performance test results revealed that HANA claddings were superior to the reference claddings such as Zircaloy-4 and A-cladding. Corrosion resistance was improved by 60 to 70% compared to the commercial claddings. Creep, burst, tensile, LOCA, wear and microstructural properties were shown to be as good as the commercial claddings.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3379-3397
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• 2021

In this study, a multi-physics modeling method was developed to analyze a nuclear fuel rod's thermo-mechanical behavior especially for high temperature anisotropic creep deformation during ballooning and burst occurring in Loss of Coolant Accident (LOCA). Based on transient heat transfer and nonlinear mechanical analysis, the present work newly incorporated the nuclear fuel rod's special characteristics which include gap heat transfer, temperature and burnup dependent material properties, and especially for high temperature creep with material anisotropy. The proposed method was tested through various benchmark analyses and showed good agreements with analytical solutions. From the validation study with a cladding burst experiment which postulates the LOCA scenario, it was shown that the present development could predict the ballooning and burst behaviors accurately and showed the capability to predict anisotropic creep behavior during the LOCA. Moreover, in order to verify the anisotropic creep methodology proposed in this study, the comparison between modeling and experiment was made with isotropic material assumption. It was found that the present methodology with anisotropic creep could predict ballooning and burst more accurately and showed more realistic behavior of the cladding.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.357-364
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• 2022

The safety of a KTC-360 transport cask, a large-capacity pressurized heavy-water reactor transport cask that transports CANDU spent nuclear fuel discharged from the reactor after burning in a pressurized heavy-water reactor, must be demonstrated under the normal transport and accident conditions specified under transport cask regulations. To confirm the thermal integrity of this cask under normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62℃, indicating that such casks must be transported separately. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC-360 cask can be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel were 446℃ lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fire accident conditions.

• Journal of the Korean Ceramic Society
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• v.48 no.1
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• pp.69-73
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• 2011

A $ZrO_2$ thin film as a corrosion protective layer was deposited on Zircaloy-4 (Z-4) clad material using $N_2O$ as a reactive gas by RF reactive magnetron sputtering at room temperature. The Z-4 substrate was located in plasma or out of plasma during the $ZrO_2$ deposition process to investigate mechanical and corrosive properties for the plasma immersion. Tetragonal and monoclinic phases were existed in $ZrO_2$ thin film immersed in plasma. We observed that a grain size of the $ZrO_2$ thin film immersed in plasma state is larger than that of the $ZrO_2$ thin film out of plasma state. In addition, the corrosive property of the $ZrO_2$ thin films in the plasma was characterized using the weight gains of Z-4 after the corrosion test. Compared with the $ZrO_2$ thin film immersed out of plasma, the weight gains of $ZrO_2$ thin film immersed in plasma were larger. These results indicate that the $ZrO_2$ film with the tetragonal phase in the $ZrO_2$ can protect the Z-4 from corrosive phenomena.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.508-519
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• 2020

During transients or accidents, the reactor core is uncovered, and droplets entrained above the quench front collides with the uncovered fuel rod surface. Droplet impact cooling can reduce the peak cladding temperature. Besides zirconium-based cladding, versatile accidental tolerant fuel (ATF) claddings, including FeCrAl, have been proposed to increase the accident coping time. In order to investigate the effect of surface properties on droplet impact cooling of cladding surfaces, the droplet impact phenomena are photographed on the FeCrAl and zircaloy-4 (Zr-4) surfaces under different conditions. On the oxidized FeCrAl surface, the Leidenfrost phenomenon is not observed even when the surface temperature is as high as 550 ℃ with We > 30. Comparison of the impact behaviors observed on different materials shows that nucleate and transition boiling is more intensive on surfaces with larger thermal conductivity. The Leidenfrost point temperature (LPT) decreases with the solid thermal effusivity (${\sqrt{k{\rho}C_p}}$). However, the CHF temperature is relatively insensitive to the surface oxidation and Weber number. Droplet spreading diameter is analyzed quantitatively in the film boiling stage. Based on the energy balance a correlation is proposed for droplet maximum spreading factor. A mechanistic model is also developed for the LPT based on homogeneous nucleation theory.

• Proceedings of the Korean Institute of Surface Engineering Conference
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• 1999.10a
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• pp.105-105
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• 1999

핵연료 피복관은 핵연료에서 방사성 핵분열생성물의 방출을 저지하는 가장 뚱요한 방어막인데, 현재 지르칼로이 4가 피복관의 재료로 사용되고 있다. 사용후 핵연료는 원자력발전소내 습식 저장조에 저장되고 있으나, 지속적인 관리와 장소확보의 용이 성으로 인해 건식 저장조를 사용하는 추세에 있다. 본 연구에선 건식 저장조에 장 기간 저장되는 핵연료 피복관에 주변 환경으로부터 오염될 수 있는 소금기나 기름 등이 지르칼로이의 공기중 산화에 미치는 영향의 존재를 밝히려 한다. 현재 고리 원자력발전소에서 사용중인 핵연료 피복관을 1cm정도 높이로 자르고, 피복관 표면 을 ASTM -G2-88 방법으로 처리한 후 산화실험을 수행하였다. 산화정도는 간헐적 (intermittent) 방법을 사용하여 시편의 무게를 측정하여 구하였으며, 산화온도는 $400-500^{\circ}C$로 하였다. 소금이 흡착이 된 경우, 산화 속도는 흡착이 안된 시편보다 가속되었으며, 거의 이차법칙을 따르고 있다. 산화막 위의 흡착물의 영향을 알아보기 위해, 지르칼로이를 $500^{\circ}C$ 수증기에 $5g/m^2$ 두께로 산화시킨 후, 다시 산화실험을 수행하였다. 사용한 흡착물은 LiF, NaF, KF, NaCI 이다. 흡착물들은 산화를 대체로 가속시켰으며, NaF, KF, NaCI 순으로 그 영향력이 컸다. 그러나, LiF는 산화에 전혀 영향을 미치지 않았다. SIMS를 사용하여 각 시편의 두께에 따른 흡착물의 분포 를 알아보았다. 음이온(CI, F)과 양이온(Na, Li, K)이 산화막과 금속 경계면까지 관 찰되었으며, 음이온과 양이온의 분포는 대게 동일하였다. LiF의 경우 산화막에서 이들의 농도가 급격히 떨어지고 있음을 알 수 있었다. 산화막 내에서 이들 흡착물의 확산이 산화속도 가속의 원인이며 이들 흡착물중 CI과 F는 산화막과 금속 겸계면 에서 새로 생성되는 산화막의 강도에 영향을 미쳐, 일찍 미세균열을 만들기 시작하여 산화를 가속시키는 것으로 판단된다.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.742-754
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• 2020

Zircaloy cladding oxidation is an important phenomenon for both design basis accident and severe accidents, because it results in cladding embrittlement and rapid fuel temperature escalation. For this reason during the last decade, many experts have been conducting experiments to identify the oxidation phenomena that occur under design basis accidents and to develop mathematical analysis models. However, since the study of design extension conditions (DEC) is relatively insufficient, it is essential to develop and validate a physical and mathematical model simulating the oxidation of the cladding material at high temperatures. In this study, the QUENCH-05 and -06 experiments were utilized to develop the best-fitted oxidation model and to validate the SPACE code modified with it under the design extension condition. It is found out that the cladding temperature and oxidation thickness predicted by the Cathcart-Pawel oxidation model at low temperature (T < 1853 K) and Urbanic-Heidrick at high temperature (T > 1853 K) were in excellent agreement with the data of the QUENCH experiments. For 'LOCA without SI' (Safety Injection) accidents, which should be considered in design extension conditions, it has been performed the evaluation of the operator action time to prevent core melting for the APR1400 plant using the modified SPACE. For the 'LBLOCA without SI' and 'SBLOCA without SI' accidents, it has been performed that sensitivity analysis for the operator action time in terms of the number of SIT (Safety Injection Tank), the recovery number of the SIP (Safety Injection Pump), and the break sizes for the SBLOCA. Also, with the extended acceptance criteria, it has been evaluated the available operator action time margin and the power margin. It is confirmed that the power can be enabled to uprate about 12% through best-estimate calculations.

• Korean Journal of Materials Research
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• v.27 no.10
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• pp.563-568
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• 2017

In this study, the coating of an Al-Cr layer on the surface of a Zircaloy-4 alloy was carried out through plasma pretreatment coating and a laser surface melting process. Two different conditions for laser treatment, severe or minimal surface melting of the Zr alloy substrate, were applied to form the final coating. When there was significant surface melting of the Zr alloy, the solidification microstructure of the newly formed coating layer was mainly composed of needle-shaped $Al_3Zr$, Al(Cr) and $Al_7Cr$ phases. On the other hand, the solidification microstructure of the coating layer was mainly composed of Al(Cr) and $Al_7Cr$ phases when there was minimal surface melting of Zr base in the laser process. However, when the coating was maintained at $1100^{\circ}C$ for 2 hours, significant inter-diffusion occurred between the phases in the coating. As a result, the upper part of the coating layer was observed to mainly consist of $Al_3Zr$ and $Al_8Cr_5$ phases, regardless of the laser treatment conditions.