Abstract
As a part of Spent Nuclear Fuel (SNF) management, it is important to maintain the cladding integrity to prevent the potential release of hazardous radioactive materials. Accordingly, assessing the integrity of SNF under the generated shock and vibration loads during normal conditions is essential for safe transportation. In this study, failure criteria for two zirconium alloys were derived by utilizing data from tensile property tests, mechanical model calculations, and machine learning estimations. Integrity assessment under normal and postulated transportation conditions was also performed via finite element analyses. Loading conditions were generated by real time-acceleration data collected during transportation tests in the Republic of Korea. Shock and vibration loads were applied to a simplified light-water reactor assembly. Furthermore, postulated transportation conditions and pre-cracked cladding were assumed for more challenging conditions. Consequently, a comprehensive integrity assessment of SNF cladding under shock and vibration loads for both free-defect and defective (pre-cracked) cladding was conducted. In each case, the resultant maximum principal strains, stress intensity amplitudes, and stress intensity factors were significantly smaller than the strain-based failure criteria, S-N curve, and fracture toughness, respectively. Therefore, the integrity of SNF cladding under transportation conditions remained intact even when assuming further challenging postulated transportation load conditions.
| Original language | English |
|---|---|
| Article number | 113125 |
| Journal | Nuclear Engineering and Design |
| Volume | 421 |
| DOIs | |
| State | Published - May 2024 |
| Externally published | Yes |
Bibliographical note
Publisher Copyright:© 2024 Elsevier B.V.
UN SDGs
This output contributes to the following UN Sustainable Development Goals (SDGs)
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SDG 12 Responsible Consumption and Production
Keywords
- Finite element analysis
- Hydrogen embrittlement
- Machine learning
- Spent nuclear fuel
- Zirconium alloy
ASJC Scopus subject areas
- Nuclear and High Energy Physics
- General Materials Science
- Nuclear Energy and Engineering
- Safety, Risk, Reliability and Quality
- Waste Management and Disposal
- Mechanical Engineering
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