• Journal of the Korean Society for Heat Treatment
• /
• v.18 no.2
• /
• pp.112-118
• /
• 2005

Study was focused on the effect of the cooling rate on the embrittlement behavior of Zircaloy-4 cladding simulated Loss Of Coolant Accident (LOCA) environment. Claddings were oxidized at given temperature and given time followed by various water quenching in the range of $0.6^{\circ}C$ and $100^{\circ}C$ per second. Cladding failed after water quenching above the threshold oxidation. Threshold oxidation was decreased as the cooling rate increased, which is due to the matensite structure formed during fast cooling rate.

• Proceedings of the KWS Conference
• /
• 2006.05a
• /
• pp.39-41
• /
• 2006

The weld quality of spacer grids in Pressurized Water Reactors(PWR) fuel is extremely important for the fuel assembly performance in the nuclear renter. The spacer grid welds are currently evaluated mainly by the metallographic examination although it reveals only cross-points which are welded by the laser beam. This experiment is also to compare the weldability of Zircaloy-4 spacer grids using by the GTA and laser beam. The effect of node geometries of spacer grids for GTAW and LBW has been studied and optimum conditions of spacer grid welding have been found. Microstructures and micro-hardness of GTA and laser beam welded zones have been also compared.

• Proceedings of the KWS Conference
• /
• 2006.05a
• /
• pp.42-44
• /
• 2006

A spacer grid assembly, which is an interconnected array of slotted grid straps and is welded at the intersections to form an egg crate structure, is one of the main structural components of the nuclear fuel assembly for pressurized water reactors(PWRs). The spacer grid assembly is structurally required to have enough buckling strength under various kinds of lateral loads acting on the nuclear fuel assembly so as to keep the nuclear fuel assembly straight. To meet this requirement, it is necessary to weld the welding parts carefully and precisely. In this study, a series of welding tests were carried out to find an optimum welding condition. After examining and analyzing the specimens welded from the welding conditions, a recommendable laser welding condition was selected for the KAERI designed Zircaloy spacer grid assembly.

• Journal of Mechanical Science and Technology
• /
• v.15 no.9
• /
• pp.1274-1280
• /
• 2001

The fretting wear behavior of the contact between Zircaloy-4 tube and Inconel 600, which are used as the fuel rod cladding and grid, respectively, in PWR nuclear power plants was investigated in air. In the study, number of cycles, slip amplitude and normal load were selected as the main factors of fretting wear. The results indicated that wear increased with load, slip amplitude and number of cycles but was affected mainly by the slip amplitude. SEM micrographs revealed the characteristics of fretting wear features on the surface of the specimens such as stick, partial slip and gross slip which depended on the slip amplitude. It was found that fretting wear was caused by the crack generation along the stick-slip boundaries due to the accumulation of plastic flow at small slip amplitudes and by abrasive wear in the entire contact area at high slip amplitudes.

• Tribology and Lubricants
• /
• v.17 no.6
• /
• pp.453-458
• /
• 2001

The fretting wear of a nuclear fuel rod it a dangerous phenomenon. In this study, TiN coating was used to reduce the fretting wear of Zircaloy-4 tube, a nuclear fuel rod cladding material. TiN coating is probably one of the molt frequently and successfully used PVD coatings for the mitigation of fretting wear. The fretting tester was designed and manufactured for this experiment. The number of cycles, slip amplitude and normal load were selected as main factors of fretting wear. The results of this research showed that wear volume was improved 1.3∼3.2 times with TiN coating. The worn surfaces were observed by SEM. Wear mechanism at lower slip amplitude was the brittle cracks and rupture of TiN coating. However, adhesive and abrasive wear were mainly observed on most surfaces at higher slip amplitude.

• Nuclear Engineering and Technology
• /
• v.30 no.5
• /
• pp.396-402
• /
• 1998

Experiments on the steamside corrosion and hydrogen pick-up of Zircaloy-4 under high pressure up to 10.3MPa are carried out to estimate the pressure effects on the kinetics. Temperature and reaction time are determined to be 37$0^{\circ}C$ and 72hours for the pre-transition test and $700^{\circ}C$ and 210minutes for the post-transition test, respectively. Results show that under 10.3MPa pressure the oxidation reaction is 50% and 100% enhanced in the pre-and the post-transition regime, respectively. Total amount of hydrogen uptake in the reaction is proportionally increased as corrosion weight gain is elevated. However, pick-up fraction is not affected by the high pressure. The fraction is almost twice greater than that in the waterside corrosion. Edges in the specimens play a certain role in the enhancement, especially in the post-transition regime. To identify physical property changes of oxide film such as micro-cracks or micro-pores, careful and thorough examination must be needed with some special techniques.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1995.05b
• /
• pp.623-628
• /
• 1995

여러 가지 Zr합금에 대한 부식시험을 autoclave를 이용하여 물과 여러 가지 Li 용액에서 수행하였다. 합금은 11종의 신합금을 사용하였는데 크게 나누어 Zircaloy형 합금(ZrSnFeCr), ZrNbFeCr, ZrSnNbFeCr과 ZrFeCr 합금으로 대별되며, 비교평가를 위해 표준 Zircaloy-4 합금에 대해서도 부식시험을 수행하였다. 모든 합금에서 Li을 일정농도이상 첨가할 때 부식은 가속되는데, 부식은 Li의 농도가 2.2와 30 ppm 사이일 때 가속되기 시작한다. Li은 부식거동에 있어서 천이후 영역에서의 부식속도 보다는 천이시간과 무게 증가량에 더 영향을 끼치는 것으로 나타났다. 수소흡수율은 Li 농도와 합금에 따라서 강하게 영향을 받는 것으로 나타났으며, Li 농도가 30 ppm 이상에서는 Li 가속부식과 함께 Li가속 수소흡수현상이 나타났다. ZrSnFeCr합금들은 낮은 부식속도와 늦은 천이현상을 보이며 표준 Zircaloy보다 훨씬 우수한 부식저항성을 보인 반면에, 대부분의 Nb첨가 합금은 높은 부식속도와 빠른 천이 현상을 보였다.

• Nuclear Engineering and Technology
• /
• v.52 no.1
• /
• pp.135-147
• /
• 2020

A methodology for evaluation of mechanical and thermal effects of localized non-axisymmetric oxidation in zircaloy claddings on LWR fuel reliability is proposed. To this end, the basic capabilities of the FALCON fuel behaviour code are used. Examples of methodology application to adjustment of selected operational limits for modern BWR fuel rods, to capture effects of the excess local oxidation, are presented. Specifically, the limiting rod internal pressure for the onset of cladding lift-off is reduced, depending on initial excess oxidation spot sizes. Also, the power limits for Anticipated Operational Occurrences are adjusted, to preclude fuel melting and cladding failure due to PCMI and PCI-SCC in the affected fuel rods.

• Proceedings of the KWS Conference
• /
• 2001.10a
• /
• pp.65-68
• /
• 2001

Corrosion and tensile properties of zircaloy-4 cladding tube having a laser welding part in elevated temperature are studied to present the criterion of quality evaluation in nuclear reactor and to found the scientific basis of SCC, with laser welding method using by coupling up cladding tube to end cap. In the result of tensile test(40$0^{\circ}C$), the fracture is not happened in the welding part but base metal and the result of corrosion test(40$0^{\circ}C$ 1500psi steam), corrosion rate of the molten zone and PMZ is a little higher than the other zone

• Journal of Welding and Joining
• /
• v.30 no.3
• /
• pp.51-56
• /
• 2012

Different microstructures in the weld zone of a metal structure including a fusion zone and heat affected zone are formed as compared to the base material. Thus, the mechanical properties in the weld zone are different from those in the base material. As the basic data for reliably understanding the structural characteristics of welded nuclear material, the mechanical properties in the weld zone and base material for a Zircaloy-4 strap and Hastelloy${(R)}$-X alloy strap are measured using an instrumented indentation technique (IIT) in this study.