• Journal of Radiation Protection and Research
• /
• v.18 no.2
• /
• pp.27-35
• /
• 1993

A shielding analysis of the shipping cask for transporting the CANDU spent fuel bundles has been studied. Radiation source term has been calculated on spent fuel with burn-up of 7,800 MWD/MTU and 5 years cooling time by ORIGEN2 code. The shielding calculation for the cask capable of transporting 378 bundles of CANDU spent fuel has been made by use of 1-D ANISN and 2-D DOT 4.2 codes. As a result of analysis, the optimum shield thickness of cask was obtained. And it is proved that the safety in radiation shielding under normal transport and hypothetical accident conditions is confirmed to satisfy the allowable values specified in IAEA Safety Series No. 6 and the Korean Atomic Law.

• Nuclear Engineering and Technology
• /
• v.54 no.10
• /
• pp.3766-3777
• /
• 2022

It is essential to ensure the safety of spent nuclear fuel (SNF) transport cask in drop situation that is included in transport accident scenarios. The safety of the drop situation is affected by the impact absorption performance of impact limiters. Therefore, when designing an impact limiter, the uncertainty in the material properties that affect the impact absorption performance must be considered. In this study, the material properties of the wood inside the impact limiter were selected as the variables for a parametric study. The sensitivity analysis of the drop response of the SNF transport cask with impact limiter was performed. The minimum wood strength required to prevent a direct collision between the cask and floor was derived from the analysis results. In addition, the plastic strain response was analyzed and strain-based evaluation was performed. Based on this result, the critical values of wood properties that change the impact dynamic characteristics were investigated. Finally, the optimal material properties of wood were obtained to secure the structural safety of the SNF transport cask. The results of this study can contribute to the development of SNF transport cask, thereby ensuring safety in transport accident conditions.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.4
• /
• pp.375-387
• /
• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2001.11b
• /
• pp.766-771
• /
• 2001

The package used to transport radioactive materials, which is called by cask, must be safe under normal and hypothetical accident conditions. These requirements for the cask design must be verified through test or finite element analysis. Since the cost for FE analysis is less than one for test. the verification by FE analysis is mainly used. But due to the complexity of mechanical behaviors. the results depends on how users apply the codes and it can cause severe errors during analysis. In this paper, finite element analysis is carried out for the 9 meters free drop and the puncture condition of the hypothetical accident conditions using LS-DYNA3D and ABAQUS/Explicit. We have investigated the analyzing technique for the free drop impact test of the cask and found several vulnerable cases to errors. The analyzed results were compared with each other. We have suggested a reliable and relatively simple analysis technique for the drop test of spent nuclear fuel casks.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2001.10a
• /
• pp.272-272
• /
• 2001

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.10a
• /
• pp.310.1-310
• /
• 2002

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2018.11a
• /
• pp.137-138
• /
• 2018

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.3073-3084
• /
• 2022

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2004.06a
• /
• pp.345-345
• /
• 2004

The spent fuel transportation casks have used as one of the most essential component in the nuclear industry. And, the number of the cask has been significantly increased in recent years. While the bulk amount of spent fuel in the world is still kept in the storage pool, the number of countries which have chosen the advantages of dual purpose cask for transportation and storage is rapidly increasing. The technical experience in the area of spent fuel transportation cask operation and maintenance for long period is also available and will be well utilized also in storage casks. The increasing use of casks for dual and multiple purposes raises an issue of long term consideration by international standardization. Accordingly IAEA is providing a regulatory requirements and guidelines as an effort for this standardization.

• Nuclear Engineering and Technology
• /
• v.20 no.1
• /
• pp.65-70
• /
• 1988

A Shielding analysis of the shipping cask designed conceptually, of which shielding material are lead and resin, for containing 4 PWR spent fuel assemblies, has been made with the help of a computer code, ANISN. The shielding materials being used in the cask have been selected and arranged to minimize cask weight while maintaining an overall shielding effectiveness. Radiation source terms have been calculated by means of ORIGIN-2 code under the assumptions of 38,000 MWD/MTU burnup and 3-year cooling time. A calculation of gamma-ray and neutron dose rates on the cask surface and 1m from the surface has been done. It is revealed that the total dose rates under the normal transport and hypothetical accident conditions meet the standards specified.