• Nuclear Engineering and Technology
• /
• v.54 no.4
• /
• pp.1355-1362
• /
• 2022

Nuclear fuel, including its fresh state, must be handled safely due to its critical and hazardous nature. Under normal transport conditions, several interactions take place among different components, such as transport cask used for loading the nuclear fuel and tie-down structure to attach with the vehicle. To ensure structural integrity of the nuclear fuel, vibrations and impacts transmitted from the vehicle must be sufficiently reduced. Therefore, in this study, we conducted two transportation tests from Daejeon to Kijang in Korea to verify the vehicle-induced vibrational characteristics of the KJRR-F fresh nuclear fuel when transported under normal transport conditions. The speed and location of the vehicle were obtained via GPS, and the accelerations between the vehicle and the KJRR-F fresh nuclear fuel were measured. Additionally, using the acceleration results, a structural analysis was conducted to confirm the structural integrity of the nuclear fuel under the most severe conditions during normal transport.

• 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.

• 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.

• Proceedings of the Computational Structural Engineering Institute Conference
• /
• 2011.04a
• /
• pp.735-738
• /
• 2011

본 논문에서는 LS-DYNA를 이용하여 원자력 운반용기의 탄자충격에 대한 해석을 수행하였다. 문명의 발달과 더불어 원자력 발전소는 많이 생겼으며 미래에도 유망하다. 원자력 발전소에서 발생되는 사용 후 핵연료에는 환경이나 사람들에게 유해한 방사성 물질이 포함되어 있기 때문에 이를 운반하는 운반용기에 대한 구조적 안전성 확보가 필요하다. 운반용기의 이동과정에서 여러 가지 사고가 날 수 있으므로 이에 대한 대비가 필요하다. 해석에는 운반플라스크와 컨테이너가 사용되었다. 운반플라스크 안에 컨테이너가 들어가 있는 형상을 갖는데 이 부분에 탄자 충격을 가하고, 이 때 운반용기에서 받는 충격량과 변화에 대해 관찰하였다. 탄자도 실제 상황과 비슷하게 하기 위해 보편적으로 사용되는 k-2 소총에 들어가는 것으로 사용하였다. 이를 통하여 운반용기에 탄자충격이 가해졌을 때 구조적 안전성을 평가하였다.

• 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 Radioactive Waste Society Conference
• /
• 2018.11a
• /
• pp.137-138
• /
• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2018.05a
• /
• pp.88-89
• /
• 2018

• Nuclear Engineering and Technology
• /
• v.25 no.2
• /
• pp.248-254
• /
• 1993

The purpose of this study is to evaluate the thermal safety for dry transport of a shipping cask. Analysis condition was based on an ambient temperature of 38$^{\circ}C$ for normal heat condition. The cask was designed to carry 4PWR spent fuel assemblies with a burnup of 38,000 MWD/MTU and 3 years of cooling time. Thermal analysis was carried out by using the COBRA-SFS code. The fuel cavity was considered to be filled with air, nitrogen or helium gas for dry transport. The results of analysis showed that the maximum temperatures of fuel rod cladding in air and helium cavity would be 277$^{\circ}C$ and 226$^{\circ}C$, respectively, for 3 years of cooling time. These values were less than the specified temperature to maintain the thermal integrity of fuel assembly for dry transport.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2006.11a
• /
• pp.349-350
• /
• 2006

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2006.11a
• /
• pp.351-352
• /
• 2006