Evaluation of spent fuel transport cask from the radiological point of view
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A. Abdelhady
Abstract
As a perspective plan to transport the spent fuel of material testing of research reactor (MTR) from the temporary storage to permanent storage, choosing an adequate cask is very important to ensure the safety precautions during the transport process. Latin America cask is one of the transportation cask types may be chosen to transport the spent fuel elements where it was designed to transport the irradiated fuel for MTR and the TRIGA research reactors. Therefore, it must be evaluated from the neutronic, radiological, and thermal points of view. The cask has internal diameter of 60 cm which make it possible to content 21 of spent fuel elements of MTR. This study aims to evaluate dose rate distribution around the cask after loading with 21 of MTR spent fuel elements which have been stored for 5-years as a minimal decay time in the temporary storage. For this, MCNP5 code was used to determine the dose rate in the radial and axial directions around the cask. The results show that the dose rates at the cask surface and at 200 cm from the surface are lower than the permissible transportation limits.
Kurzfassung
Für den Transport des abgebrannten Brennstoffs des Materialprüf-Forschungsreaktors (MTR) vom Zwischenlager ins Endlager ist die Wahl eines geeigneten Behälters sehr wichtig, um die Sicherheitsvorkehrungen während des Transportvorgangs zu gewährleisten. Der lateinamerikanische Behälter ist einer der Transportbehältertypen, die für den Transport der abgebrannten Brennelemente gewählt werden können, da er für den Transport des bestrahlten Brennstoffs für MTR und die TRIGA-Forschungsreaktoren konzipiert wurde. Dies muss unter neutronischen, radiologischen und thermischen Gesichtspunkten bewertet werden. Der Behälter hat einen Innendurchmesser von 60 cm, was es ermöglicht, 21 abgebrannte Brennelemente der MTR zu enthalten. Ziel dieser Studie ist es, die Dosisleistungsverteilung um den Behälter herum nach der Beladung mit 21 abgebrannten MTR-Brennelementen, die 5 Jahre lang als minimale Zerfallszeit im Zwischenlager gelagert wurden, zu bewerten. Dazu wurde mit dem MCNP5-Code die Dosisleistung in radialer und axialer Richtung um den Behälter herum bestimmt. Die Ergebnisse zeigen, dass die Dosisleistungen an der Behälteroberfläche und bei 200 cm von der Oberfläche niedriger sind als die zulässigen Transportgrenzen.
References
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© 2019, Carl Hanser Verlag, München
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Artikel in diesem Heft
- Technical Contributions/Fachbeiträge
- The simulation and study of ELAP event with URG and FLEX mitigation strategies for PWR by using TRACE code
- Experimental study on penetration characteristic of submerged steam jet in quiescent water
- Research on multi-scale simulation model for single-phase water pipe networks
- Determination of radiological source term of CHASHMA-1 NPP during LOCA
- Assessment of the health damage costs of radionuclides releases from Muğla provinces lignite-fired power plants
- Natural radioactive risk assessment in top soil and possible health effect in Minim and Martap villages, Cameroon: using radioactive risk index and statistical analysis
- ATWS severe accident analysis in the loss of flow scenario using the MELCOR code in Bushehr nuclear Power Plant
- Evaluation of spent fuel transport cask from the radiological point of view
