• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.110-116
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• 2020

As the management plan for domestic spent nuclear fuel is delayed, the storage of the operating nuclear power plant is approaching saturation, and the Kori 1 Unit that has reached its end of operation life is preparing for the dismantling plan. The first stage of dismantling is the transfer of spent nuclear fuel stored in storage at plants. The spent fuel management process leads to temporary storage, interim storage, reprocessing and permanent disposal. In this paper, the technical issues to be considered when transporting spent fuel in this process are summarized. The spent fuels are treated as high-level radioactive waste and strictly managed according to international regulations. A series of integrity tests are performed to demonstrate that spent fuel can be safely stored for decades in a dry environment before being transferred to an intermediate storage facility. The safety of spent fuel transport container must be demonstrated under normal transport conditions and virtual accident conditions. IAEA international standards are commonly applied to the design of transport containers, licensing regulations and transport regulations worldwide. In addition, each country operates a physical protection system to reduce and respond to the threat of radioactive terrorism.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.4
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• pp.299-314
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• 2014

A lifting device is used to deal with transport cask for the transportation of spent fuels from nuclear power plants. This study performed theoretical analysis and numerical simulation to evaluate the structural integrity of the lifting device based on Nuclear Safety and Security Commission(NSSC) Notice No.2013-27 and US 10CFR Part 71 ${\S}71.45$. The results of theoretical analysis showed that the maximum stresses of all components were below the allowable values. This result confirmed that the lifting device was structurally safe during operation. The results of finite element analysis also showed that it was evaluated to satisfy the design criteria bothyielding and ultimate condition. All components have been shown to ensure the structural safety due to sufficient safety margins. In other words, the safety factor was 3 or more for the yielding condition and was 5 or more for the ultimate condition.

• 제어로봇시스템학회:학술대회논문집
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• 1996.10b
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• pp.1060-1063
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• 1996

Remote Cask Grappling and Lid Unbolting Device (RCGLUD) is developed as a dedicated device capable of performing complete procedure of handling nuclear spent fuel transport cask. Since RCGLUD is suspended to an overhead crane, its body should undergo prolonged vibration upon actuation in rotational direction and it becomes difficult to achieve precise grappling of the cask. Therefore, this paper presents an adaptation of input shaping technique to effectively suppress the rotational vibration of RCGLUD and achieve precise positioning in rotational direction. This technique has a practical merit in that it requires only the information on the system natural frequency and the damping ratio. Its performance is verified by both simulation and experimental studies, and revealed that the method is also insensitive to modeling error.

• Journal of Radiation Protection and Research
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• v.18 no.2
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• pp.47-59
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• 1993

The criticality of the shipping cask(KSC-7) for transportion of 7PWR spent fuel assemblies has been calculated and analysised on the basis of neutron transport theory. For criticality analysis, effects of the rod pitches, the fixed neutron absorbers(borated sus+boral) were considered. The effective multiplication factor has been calculated by KENO-Va, Mote Carlo method computer code, with the HANSEN-ROACH 16 group cross section set, which was made for personal computer system. The criticality for the KSC-7 cask was calculated in terms of the fresh fuel which was conservative for the aspects of nuclear critility. From the results of criticality analysis, the calculated Keff is proved to be lower than subcritical limit during normal transportation and under hypothetical accident condition. The maximum calculated criticalities of the KSC-7 were lower the safety criticality limit 1.0 recommended by US 10CFR71 both under normal and hypothetical accident condition. Also, to verify the KSC-7 criticality calculation results by using KENO-Va, it was carried out benchmark calculation with experimental data of B & W(Bobcock and Wilcox) company. From the 3s series of calculation of the KSC-7 cask and benchmark calculation, the cask was safely designed in nuclear criticality, respectively.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.18 no.1
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• pp.43-49
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• 2005

The package used to transport radioactive materials, which is called by the shipping 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 the one for test, the verification by FE analysis is mainly used. But due to the complexity of mechanical behaviors, the results depend on how users apply the codes and can cause severe errors during analysis. In this paper, finite element analysis is carried out for the 9 meters free drop condition of the hypothetical accident conditions using LS-DYNA3D and ABAQUS/Explicit. We have investigated the analyzing technique lot the free drop impact test of the cask and investigated several vulnerable cases. 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.

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

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.2
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• pp.193-204
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• 2023

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO₂ fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.1
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• pp.9-16
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• 2009

Transport package for radioactive material should be structurally safe under puncture drop condition and its safety should be verified by test and numerical analysis. Most finite element analyses for puncture drop have been performed without modeling the impact limiter since failure is occurred in the materials of the impact limiter. This paper presents a new modeling methodology, where an element is eroded in case that the material's failure criteria are reached at the element's integration point, to investigate the effect of the impact limiter in the puncture process. The effectiveness of the proposed scheme is shown through the puncture drop analysis of hotcell transport cask, which is under design in KAERI. The results show that about 80 percent of the total impact energy is absorbed due to the deformation of impact limiter. Using the present method the puncture drop can be analyzed more accurately, but it would give conservative results compared to the actual test condition.

• Nuclear Engineering and Technology
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• v.42 no.5
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• pp.582-589
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• 2010

As a part of the back-end fuel cycle, transportation of spent nuclear fuel (SNF) from nuclear power plants (NPPs) to a fuel storage facility is very important in establishing a nuclear fuel cycle. In Korea, the accumulated amount of SNF in the NPP pools is troublesome since the temporary storage facilities at these NPP pools are expected to be full of SNF within ten years. Therefore, Korea cannot help but plan for the construction of an interim storage facility to solve this problem in the near future. Especially, a decision on several factors, such as where the interim storage facility should be located, how many casks a transport ship can carry at a time and how many casks are initially required, affect the configuration of the transportation system. In order to analyze the various possible candidate scenarios, we assumed four cases for the interim storage facility location, three cases for the load capacity that a transport ship can carry and two cases for the total amount of casks used for transportation. First, this study considered the currently accumulated amount of SNF in Korea, and the amount of SNF generated from NPPs until all NPPs are shut down. Then, how much SNF per year must be transported from the NPPs to an interim storage facility was calculated during an assumed transportation period. Second, 24 candidate transportation scenarios were constructed by a combination of the decision factors. To construct viable yearly transportation schedules for the selected 24 scenarios, we created a spreadsheet program named TranScenario, which was developed by using MS EXCEL. TranScenario can help schedulers input shipping routes and allocate transportation casks. Also, TranScenario provides information on the cask distribution in the NPPs and in the interim storage facility automatically, by displaying it in real time according to the shipping routes, cask types and cask numbers that the user generates. Once a yearly transportation schedule is established, TranScenario provides some statistical information, such as the voyage time, the availability of the interim storage facility, the number of transported casks sent from the NPPs, and the number of transported casks received at the interim storage facility. By using this information, users can verify and validate a yearly transportation schedule. In this way, the 24 candidate scenarios could be constructed easily. Finally, these 24 scenarios were compared in terms of their operation cost.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.690-695
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• 2023

Recommendations are given to improve the efficiency of radiation protection of transport casks for SNF transportation. The attenuation of ${\gamma}$-quanta of long-lived isotopes ¹³⁴Cs, ^137mBa(¹³⁷Cs), ¹⁵⁴Eu and ⁶⁰Co by optimizing the thicknesses and arrangement of layers of Fe and Pb radiation shields of transport casks is studied. The fixed radiation shielding mass (fixed mass thickness) is chosen as the main optimization criterion. The effect of the placement order of Fe and Pb layers in a combined two-layer radiation shield with an equivalent thickness of 30 cm is studied in detail. It is shown that with the same mass thicknesses of the Fe and Pb layers, the placement of Fe in the first layer, and Pb - in the second one provides more than twofold attenuation of ${\gamma}$-quanta compared to the reverse placement: Pb - in the first layer, Fe - in the second. The increase in the efficiency of attenuation of ${\gamma}$-quanta for TC with combined shielding of Fe and Pb is shown to be achieved by designing the first layer of radiation shielding around the canister with SNF from Fe of the maximum possible thickness.