Spent fuel transportation cask under accidental fire conditions: Numerical analysis of gas transport phenomena affecting heat transfer in shielding materials

Zinet, Matthieu; Ghazal, Roula; Issard, H.; Bardon, Olivier

doi:10.1016/j.pnucene.2019.103045

Cited by 5 publications

(3 citation statements)

References 17 publications

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“…An analytical-numerical 2020 JINST 15 T11004 method for finding the EBF of a linear source within a cylindrical shield was adopted before [4] and it would be interesting to use a Monte Carlo method as an alternative, taking advantage of its larger versatility. The spent fuel rods in nuclear reactors [16][17][18][19][20][21][22][23][24][25][26] and brachytherapy seeds [27,28] can be referred as a good approximation to linear sources although the spent fuel rods are known to be a volumetric cylindrical gamma source. In order to simplify the case in this study, they are treated as a linear gamma source.…”

Section: Introductionmentioning

confidence: 99%

“…For this reason, an appropriate shielding is clearly needed for their transport and storage. Many previous works have been done to build a cylindrical shield for the spent fuel rods in terms of calculating the dose [16][17][18][19][20][21][22][23], but in this study, we investigated the cylindrical shield in terms of calculating the EBF through the Monte Carlo method. For verification, we calculated the total mass attenuation coefficients and mass energy-absorption coefficients of concrete using the MCNPX code and we validated as such the uncollided photon flux with theory.…”

Section: Introductionmentioning

confidence: 99%

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A Monte Carlo study on the gamma-ray buildup factors for the linear sources embedded in a cylindrical shield

2020

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While the studies on the photon buildup factors for point gamma sources have been widely made, studies for linear gamma sources, for which the spent nuclear fuel rods and/or brachytherapy seeds could be a relatively good approximation, have become extremely scarce. Buildup factors for linear gamma sources are computed here by means of the MCNPX Monte Carlo code. The exposure buildup factors were calculated for isotropic linear gamma sources with different lengths (1, 10, and 100 cm) and also for different energies (0.1, 0.5, 1, 5, and 10 MeV) in a cylindrical concrete shield up to depths of 10 mean free paths. A number of 45 simulations were done to reach the appropriate results for the linear gamma sources and 5 more simulations were performed for point gamma sources to be compared with the linear sources of photons at the same energies. For validation, we calculated the total mass attenuation coefficients, mass energy absorption coefficients, and uncollided photon flux for some photon energy in concrete and we showed that the simulation results have a good agreement with NIST-XCOM database and theory. Finally, comparisons are shown for the exposure buildup factors of the linear source in terms of the gamma-rays energy, of the length of the linear source and of the cylindrical shield, which can be useful in the design of a suitable shield.

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Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

A Monte Carlo study on the gamma-ray buildup factors for the linear sources embedded in a cylindrical shield

2020

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“…Therefore, the performance of the thermal conductivity of the cask in the transport process is vital. At present, the thermal analysis of the second generation advanced fuel assembly (AFA-2G) and the third generation advanced fuel _____________ * Corresponding author, e-mail: tianshujian@ncwu.edu.cn assembly (AFA-3G) PWR fuel assembly has been studied internationally [4][5][6]. Wix and Hohnstreiter [7] analyzed and studied the convection coefficient under the premise of ignoring the internal convection.…”

Section: Introductionmentioning

confidence: 99%

Numerical simulation of decay heat scattering out of AP1000 nuclear reactor spent fuel cask based on finite element method

Tian

et al. 2020

THERM SCI

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The heat transfer performance of the spent fuel transport cask is inseparably related to the safety of the whole reprocessing system. In this study, we carried out the thermal analysis on the NAC-STC transport cask for AP1000 spent fuel assembly to evaluate the thermal performance of transport cask by the finite element method software ANSYS. A computational dynamics model was developed to study the temperature distribution inside the transport cask and on the surface of the cask. The effectiveness of the numerical calculation is demonstrated by comparing with the theoretical results. The results show that transport cask can reach steady-state during transportation, and the highest temperature in the case is 328?, which is below the maximum safety limit of 400?. Besides, the temperature of the fuel element baskets, sealing ring, photon shielding layer and neutron shielding layer in the cask are all within the safety limit.

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Copper neutron transport libraries validation by means of a 252Cf standard neutron source

Schulc

Košťál

Novák

et al. 2021

Nuclear Engineering and Technology

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Spent fuel transportation cask under accidental fire conditions: Numerical analysis of gas transport phenomena affecting heat transfer in shielding materials

Cited by 5 publications

References 17 publications

A Monte Carlo study on the gamma-ray buildup factors for the linear sources embedded in a cylindrical shield

A Monte Carlo study on the gamma-ray buildup factors for the linear sources embedded in a cylindrical shield

Numerical simulation of decay heat scattering out of AP1000 nuclear reactor spent fuel cask based on finite element method

Copper neutron transport libraries validation by means of a 252Cf standard neutron source

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