This paper aims to evaluate the mechanical integrity for Spent Nuclear Fuel (SNF) cladding under lateral loads during transportation. The evaluation process requires a conservative consideration of the degradation conditions of SNF cladding, especially the hydride effect, which reduces the ductility of the cladding. The dynamic forces occurring during the drop event are pinch force, axial force and bending moment. Among those forces, axial force and bending moment can induce transverse tearing of cladding. Our assessment of 14 × 14 PWR SNF was performed using finite element analysis considering SNF characteristics. We also considered the probabilistic procedures with a Monte Carlo method and a reliability evaluation. The evaluation results revealed that there was no probability of damage under normal conditions, and that under accident conditions the probability was small for transverse failure mode.

This study developed an analytical methodology for the mechanical integrity of spent nuclear fuel (SNF) cladding tubes under external pinch loads during transportation, with reference to the failure mode specified in the relevant guidelines. Special consideration was given to the degraded characteristics of SNF during dry storage, including oxide and hydride contents and orientations. The developed framework reflected a composite cladding model of elastic and plastic analysis approaches and correlation equations related to the mechanical parameters. The established models were employed for modeling the finite elements by coding their physical behaviors. A mechanical integrity evaluation of 14 × 14 PWR SNF was performed using this system. To ensure that the damage criteria met the applicable legal requirements, stress-strain analysis results were separated into elastic and plastic regions with the concept of strain energy, considering both normal and hypothetical accident conditions. Probabilistic procedures using Monte Carlo simulations and reliability evaluations were included. The evaluation results showed no probability of damage under the normal conditions, whereas there were small but considerably low probabilities under accident conditions. These results indicate that the proposed approach is a reliable predictor of SNF mechanical integrity.

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.

The nuclear fuel cladding temperatures of the HANARO fuel test loop have been calculated by MARS code for the large break loss-of-coolant accidents. Conservative method was used for the analysis of the loss-of-coolant accidents. Consequently, the maximum peak cladding temperature was predicted as 1235K, which was lower than the design limit temperature (1477K) of nuclear fuels for the HANARO fuel test loop. This means that the cooling capability of the emergency cooling water system for the HANARO fuel test loop is sufficient for the large break loss-of-coolant accidents.

The effect of β-heat treatment on th microstructure, mechanical properties and texture in the nuclear fuel cladding of Zircaloy-4 tubes was chosen at 1000, 1100 and 1200˚C, and the tubes were heat-treated by a high frequency vacuum induction furnace. Morphology of the second phase particles and α-grain of as-received tubes were markedly changed by heat treatment. The average sizes of second phase particles of as-received and β-heat treated tubes were 0.1μm and 0.076μm, respectively. However, the average sizes of second phase particles were not much changed in the β-heated temperatures. With increasing heat treatment temperatures, the 0.2% yield strength and the hoop strength were decreased because of changes in preferred orientation as will as α-plate width. Heat treated Zircaloy-4 tubes exhibited texture changes but the preferred orientation of grains still remained.

Zr합금에서 V, Sb의 함랗 변화가 Zr 합금의 부식 특성에 미치는 영향을 조사하기 위해 V, Sb함량을 각각 0.1, 0.2, 0.4wt.% 변화시킨 6종의 합금을 제조하고 360˚C 물 분위기에서 100일 동안 부식실험을 수행하였다. V이 0.2, 0.4wt% 첨가시편에서는 부식 속도의 천이 현상이 관찰되지 않았으나 0.1wt.% V 첨가 시편의 경우 10일 이후부터 무게증가량이 급격히 증가하는 부식 속도 가속 현상이 발생하였다. V 첨가량이 증가할수록 내식성이 증가하였으며 0.4wt.%V 첨가 합금이 가장 우수한 내식성을 보였다. Sb가 첨가된 삼원계 합금에서는 0.1, 0.4wt.%Sb 첨가 시편의 경우 초기부터 급격히 부식이 가속되는 현상이 발생하였으며 Sb 첨가량이 증가할수록 무게증가량이 감소하다가 다시 증가하여 0.2wt.% Sb 첨가에서 최소 무게증가량을 보였다. V, Sb함량이 증가할수록 석출물의 크기와 석출물의 부피 분율이 증가하는 경향을 보였으며 석출물의 크기가 0.11-0.13μm의 석출물 크기를 가질 때 가장 우수한 내식성을 보였다. 부식특성과 석출물 크기와의 관계로부터 적절한 크기의 석출물은 음극반응에서 전자의 전도를 제어하고 안정한 산화막 미세구조를 유지하는데 중요한 역할을 한다고 사료된다.