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

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

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.34 no.9
• /
• pp.1175-1183
• /
• 2010

A spacer grid assembly is one of the most important structural components in a Light Water Reactor(LWR) nuclear fuel assembly. In the case of the Zircaloy spacer grid assembly, the primary design consideration is to ensure that lateral dynamic crush strength of the spacer grid assembly is sufficient to resist design basis loads and thereby prevent seismic accidents, without a significant increase in the hydraulic head loss for the reactor coolant in the reactor core. In this study, factors affecting the lateral dynamic crush strength of a spacer grid assembly were analyzed by performing lateral dynamic crush tests and finite element analyses. Further, an effective and economical method to enhance the lateral dynamic crush strength of the spacer grid assembly is proposed.

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

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

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
• /
• 2003.11a
• /
• pp.313-323
• /
• 2003

The sliding wear behaviors of Zircaloy-4 nuclear fuel rod were investigated using two support springs with convex and concave shapes in room temperature air and water. The main focus is to compare the wear behavior of various test variables such as slip amplitude, environment, contact contours with different spring shape and a number of cycles. The results indicated that wear volume and maximum wear depth increased with slip amplitude in both air and water, but their trends tended to change according to the spring shapes and test environments. In air condition, the wear volume was controlled by wear debris behavior generated on worn surface. As a result, final wear volume and maximum wear depth decreased if a ratio of protruded wear volume to worn area $(D_p)$ would be saturated to specific value. This is because wear particle layer could accommodate large strain by accumulating and transforming wear particle layer. However, in water condition, metal-to metal contact was more dominant and wear volume was greatly affected by changed mechanical behavior between contact surfaces since wear debris should be generated after repeated plastic deformation and fracture. After wear test, worn surfaces were examined using optical microscope and SEM and details of wear mechanism were discussed using a ratio of wear volume to worn area $(D_e)$ at each test condition.

• Nuclear Engineering and Technology
• /
• v.42 no.3
• /
• pp.249-258
• /
• 2010

Recently, many utilities have considered interim dry storage of spent nuclear fuel as an option for increasing spent fuel storage capacity. Foreign nuclear regulatory committees have provided some regulatory and licensing requirements for relatively low- and medium-burned spent fuel with respect to the prevention of spent fuel degradation during transportation and interim dry storage. In the present study, the effect of cladding creep and hydride distribution on spent fuel degradation is reviewed and performance tests with high-burned Zircaloy-4 and advanced Zr alloy spent fuel are proposed to investigate the effect of burnup and cladding materials on the current regulatory and licensing requirements. Creep tests were also performed to investigate the effect of temperature and tensile hoop stress on hydride reorientation and subsequently to examine the temperature and stress limits against cladding material failure. It is found that the spent fuel failure is mainly caused by cladding creep rupture combined with mechanical strength degradation and hydride reorientation. Hydride reorientation from the circumferential to radial direction may reduce the critical stress intensity that accelerates radial crack propagation. The results of cladding creep tests at $400^{\circ}C$ and 130MPa hoop stress performed in this study indicate that hydride reorientation may occur between 2.6% to 7.0% strain in tube diameter with a hydrogen content range of 40-120ppm. Therefore, it is concluded that hydride re-orientation behaviour is strongly correlated with the cladding creep-induced strain, which varies as functions of temperature and stress acting on the cladding.

• Nuclear Engineering and Technology
• /
• v.25 no.3
• /
• pp.381-393
• /
• 1993

This paper presents the experimental results on flooding of countercurrent flow in vertical multirod channels, which consists of falling water film and upward air flow. In particular, the effects of spacer grids, with and without mixing vane, and of blockage in the multirod bundle on the behaviour of flooding were investigated. The 5$\times$5 zircaloy tube bundle was used for the test section. The comparison of previous analytical models and empirical correlations with present data on flooding showed that the existing models and correlations predict much higher flooding curves. The spacer grid causes the lower flooding air flow rate to compare with the bare rod bundle. However, the mixing spacer grids need a higher flooding air flow rate for a constant liquid flow rate than the spacer grids without mixing vanes. The bundle containing blockages has the highest flooding air flow rate among the bundles with spacer grids and blockages. Empirical flooding correlations for the three types of test section have been made.

• Nuclear Engineering and Technology
• /
• v.40 no.1
• /
• pp.21-36
• /
• 2008

High bum-up fuel technology has been developed through a national R&D program, which covers key technology areas such as claddings, $UO_2$ pellets, spacer grids, performance code, and fuel assembly tests. New cladding alloys were developed through alloy designs, tube fabrication, out-of-pile test and in-reactor test. The new Zr-Nb tubes are found to be much better in their corrosion resistance and creep strength than the Zircaloy-4 tube, owing to an optimized composition and heat treatment of the new Zr-Nb alloys. A new fabrication technology for large grain $UO_2$ pellets was developed using various uranium oxide seeds and a micro-doping of Al. The uranium oxide seeds, which were added to $UO_2$ powder, were prepared by oxidizing and heat-treating scrap $UO_2$ pellets. A $UO_2$ pellet containing tungsten channels was fabricated for a thermal conductivity enhancement. For the fuel performance analysis, new high burnup models were developed and implemented in a code. This code was verified by an international database and our own database. The developed spacer grid has two features of contoured contact spring and hybrid mixing vanes. Mechanical and hydraulic tests showed that the spacer grid is superior in its rodsupporting, wear resistance and CHF performance. Finally, fuel assembly test technology was also developed. Facilities for mechanical and thermal hydraulic tests were constructed and are now in operation. Several achievements are to be utilized soon by the Korea Nuclear Fuel and thereby contribute to the economy and safety of PWR fuel in Korea

• Nuclear Engineering and Technology
• /
• v.53 no.1
• /
• pp.178-187
• /
• 2021

Dual-cooled annular fuel allows a significant increase in power density while maintaining or improving safety margins. However, the dual-cooled design brings much higher Zircaloy charge in reactor core, which could cause a great threaten of hydrogen explosion during severe accidents. Hence, an innovative fuel combined dual-cooled annular geometry and SiC cladding was proposed for the first time in this study. Capabilities of fuel design and behavior simulation were developed for this new fuel by the upgrade of FROBA-ANNULAR code. Considering characteristics of both SiC cladding and dual-cooled annular geometry, the basic fuel design was proposed and preliminary proved to be feasible. After that, a design optimization study was conducted, and the optimal values of as-fabricated plenum pressure and gas gap sizes were obtained. Finally, the performance simulation of the new fuel was carried out with the full consideration of realistic operation conditions. Results indicate that in addition to possessing advantages of both dual-cooled annular fuel and accident tolerant cladding at the same time, this innovative fuel could overcome the brittle failure issue of SiC induced by pellet-cladding interaction.