Journal of Nuclear Science, Engineering and Technology (JONSAT)
In cooperation with the Iranian Nuclear Society

Safety assessment of the lead cask transfer of irradiated fuel and structural materials in drop accident conditions using ANSYS simulation

Document Type : Scientific Note

Authors

R. Gostariani 1
N. Eqra 2
J. Dabiri 1

1 Reactor and Nuclear Safety Research School, Nuclear Science and Technology Research Institute, AEOI, P.O.BOX: 14399-51113, Tehran – Iran

2 School of Mechanical Engineering, Sharif University of Technology, P.O.Box: 1458889694, Tehran – Iran

https://doi.org/10.24200/nst.2024.1542.2006
Abstract
The assessment of the irradiation behavior of nuclear fuel and structural materials is carried out by conducting post-irradiation examination at the hot cell facility. One of the challenges in performing this examination is the safe transfer of cask containing irradiated nuclear fuel and structural materials from the reactor to the hot cell. Therefore, in this study, the safety assessment of the lead cask transfer of a fuel plate, a zirconium fuel cladding, and oxide fuel pellets (UO2) against the drop accident was conducted using finite element analysis with ANSYS software. The drop of the cask from a height of 9 meters in vertical, horizontal, and inclined directions is one of the most common accidents defined in IAEA Safety Standards No. 6 (SSR-6). In this study, the yield stress of the materials inside the cask was extracted from the references and compared with the maximum stress resulting from the finite element simulation results. Therefore, the safety acceptance limit after an accident was presented as determining the safety factor for each component inside the cask. The simulation results showed that the cask contents are particularly susceptible to serious damage in the drop accident of a 9-meter, especially vertical and horizontal drop. Therefore, a corrective action was taken by incorporating a shock absorber made of polyurethane foam under and around the cask. Finally, a re-simulation with the presence of the shock absorber demonstrated the safety acceptance of the cask against the drop accident.

Highlights

  1. Bhandekar A, Pandit K, Dhotre M, Nagaraju P, Rath B, Mishra P, Kumar S, Dubey J.S, Mallik G.K, Singh J.L. New Hot Cell Facility for Post Irradiation Examination. BARe. 2015;19.

 

  1. Cole J.I, Jue J-F, Moore G.A. Fuel Design and Fabrication: Research Reactor Fuel. Idaho National Lab.(INL), Idaho Falls, ID (United States). 2021.

 

  1. Durazzo M, Umbehaun P, Torres W, Souza J, Silva D, Andrade D. Procedures for manufacturing an instrumented nuclear fuel element. Progress in Nuclear Energy. 2019;113:166-174.

 

  1. Lyons M, Boyle R, Davies J, Hazel V, Rowland T. UO2 properties affecting performance. Nuclear Engineering and Design. 1972;21(2):167-199.

 

  1. Alfa S, Nazar R, Sulaksono S, Fitriana R, Nagara N. Editors. Transfer cask design for plate type fuel elements of the Bandung Conversion Reactor. AIP Conference Proceedings. AIP Publishing. 2019.

 

  1. Regulations for the Safe Transport of Radioactive Material. Vienna: International Atomic Energy Agency. 2018.

 

  1. Kim K-S, Chung S-H, Kim J-S, Choi K-S, Yun H-D. Demonstration of structural performance of IP-2 packages by advanced analytical simulation and full-scale drop test. Nuclear engineering and design. 2010;240(3):639-655.

 

  1. Saito Y, Kishimoto J, Matsuoka T, Tamaki H, Kitada A. Containment integrity evaluation of MSF-type cask for interim storage and transport of PWR spent fuel. International Journal of Pressure Vessels and Piping. 2014;117:33-41.

 

  1. Hao Y, Wang J, Li Y, Wu B, Wang H, Ma T. Editors. Study on the structural evaluation and optimization of spent nuclear fuel cask. International Conference on Nuclear Engineering. American Society of Mechanical Engineers. 2021.

 

  1. Hao Y, Wang J, Wu B, Ma T, Wang H, Liu B. Study on the most severe condition for structural assessment of nuclear fuel transport package under impact loading. Progress in Nuclear Energy. 2022;153:104447.

 

  1. Kim S-P, Kim J, Sohn D, Kwon H, Shin M. Stress-based vs. Strain-based safety evaluations of spent nuclear fuel transport casks in energy-limited events. Nuclear Engineering and Design. 2019;355:110324.

 

  1. Rezaeian M, Kamali J, Ahmadi S.J. Thermal Evaluation of a Dual Purpose Cask for Bushehr Nuclear Reactor Spent Fuels Under Normal Conditions. J. of Nucl Sci. and Tech. 219;87(1):81-88 [In Persian].

 

  1. Sedigh Y, Azimfar S.A. Safety Analysis of Spent Fuel Transportation Cask of Bushehr Nuclear Power Plant through the Passing of Fire Tunnel with ANSYS®10.0. J. of Nucl Sci. and Tech. 2009;30(1):57-61 [In Persian].

 

  1. Wang X-S, Xu Y. Mechanical characterizations of the dispersion U3Si2-Al fuel plate with sandwich structure. Applied Composite Materials. 2003;10(3):159-167.

 

  1. Sengupta A.K, Bhagat R.K, Jarvis T, Banerjee J, Kutty T.R.G, Ravi K, D'Souza O.L, Keswani R, Nair M.R, Ramachandran R, Purusnotham D.S.C. Some important properties of simulated UO2 fuel. 1999.

 

  1. Rosinger H, Northwood D. The elastic properties of zirconium alloy fuel cladding and pressure tubing materials. Journal of Nuclear Materials. 1979;79(1):170-179.

 

  1. Hammad A, El-Mashri S, Nasr M. Mechanical properties of the Zr-1% Nb alloy at elevated temperatures. Journal of nuclear materials. 1992;186(2):166-176.

 

  1. Ding F, Liu T, Zhang H, Liu L, Li Y. Stress‐strain curves for polyurethane elastomers: A statistical assessment of constitutive models. Journal of Applied Polymer Science. 2021;138(39):51269.

 

  1. Wellman G. Transportation system impact limiter design using rigid polyurethane foam. Sandia National Labs. 1985.

 

  1. https://www.matweb.com/. Materials Properties Data.

Keywords

Nuclear fuel transfer cask
Irradiated structural materials
Safety assessment
9-meter drop accident
ANSYS software
  1. Bhandekar A, Pandit K, Dhotre M, Nagaraju P, Rath B, Mishra P, Kumar S, Dubey J.S, Mallik G.K, Singh J.L. New Hot Cell Facility for Post Irradiation Examination. BARe. 2015;19.

 

  1. Cole J.I, Jue J-F, Moore G.A. Fuel Design and Fabrication: Research Reactor Fuel. Idaho National Lab.(INL), Idaho Falls, ID (United States). 2021.

 

  1. Durazzo M, Umbehaun P, Torres W, Souza J, Silva D, Andrade D. Procedures for manufacturing an instrumented nuclear fuel element. Progress in Nuclear Energy. 2019;113:166-174.

 

  1. Lyons M, Boyle R, Davies J, Hazel V, Rowland T. UO2 properties affecting performance. Nuclear Engineering and Design. 1972;21(2):167-199.

 

  1. Alfa S, Nazar R, Sulaksono S, Fitriana R, Nagara N. Editors. Transfer cask design for plate type fuel elements of the Bandung Conversion Reactor. AIP Conference Proceedings. AIP Publishing. 2019.

 

  1. Regulations for the Safe Transport of Radioactive Material. Vienna: International Atomic Energy Agency. 2018.

 

  1. Kim K-S, Chung S-H, Kim J-S, Choi K-S, Yun H-D. Demonstration of structural performance of IP-2 packages by advanced analytical simulation and full-scale drop test. Nuclear engineering and design. 2010;240(3):639-655.

 

  1. Saito Y, Kishimoto J, Matsuoka T, Tamaki H, Kitada A. Containment integrity evaluation of MSF-type cask for interim storage and transport of PWR spent fuel. International Journal of Pressure Vessels and Piping. 2014;117:33-41.

 

  1. Hao Y, Wang J, Li Y, Wu B, Wang H, Ma T. Editors. Study on the structural evaluation and optimization of spent nuclear fuel cask. International Conference on Nuclear Engineering. American Society of Mechanical Engineers. 2021.

 

  1. Hao Y, Wang J, Wu B, Ma T, Wang H, Liu B. Study on the most severe condition for structural assessment of nuclear fuel transport package under impact loading. Progress in Nuclear Energy. 2022;153:104447.

 

  1. Kim S-P, Kim J, Sohn D, Kwon H, Shin M. Stress-based vs. Strain-based safety evaluations of spent nuclear fuel transport casks in energy-limited events. Nuclear Engineering and Design. 2019;355:110324.

 

  1. Rezaeian M, Kamali J, Ahmadi S.J. Thermal Evaluation of a Dual Purpose Cask for Bushehr Nuclear Reactor Spent Fuels Under Normal Conditions. J. of Nucl Sci. and Tech. 219;87(1):81-88 [In Persian].

 

  1. Sedigh Y, Azimfar S.A. Safety Analysis of Spent Fuel Transportation Cask of Bushehr Nuclear Power Plant through the Passing of Fire Tunnel with ANSYS®10.0. J. of Nucl Sci. and Tech. 2009;30(1):57-61 [In Persian].

 

  1. Wang X-S, Xu Y. Mechanical characterizations of the dispersion U3Si2-Al fuel plate with sandwich structure. Applied Composite Materials. 2003;10(3):159-167.

 

  1. Sengupta A.K, Bhagat R.K, Jarvis T, Banerjee J, Kutty T.R.G, Ravi K, D'Souza O.L, Keswani R, Nair M.R, Ramachandran R, Purusnotham D.S.C. Some important properties of simulated UO2 fuel. 1999.

 

  1. Rosinger H, Northwood D. The elastic properties of zirconium alloy fuel cladding and pressure tubing materials. Journal of Nuclear Materials. 1979;79(1):170-179.

 

  1. Hammad A, El-Mashri S, Nasr M. Mechanical properties of the Zr-1% Nb alloy at elevated temperatures. Journal of nuclear materials. 1992;186(2):166-176.

 

  1. Ding F, Liu T, Zhang H, Liu L, Li Y. Stress‐strain curves for polyurethane elastomers: A statistical assessment of constitutive models. Journal of Applied Polymer Science. 2021;138(39):51269.

 

  1. Wellman G. Transportation system impact limiter design using rigid polyurethane foam. Sandia National Labs. 1985.

 

  1. https://www.matweb.com/. Materials Properties Data.

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