• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.1
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• pp.65-72
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• 2008

The purpose of the HLW deep geological disposal is to isolate and to delay the radioactive material release to human beings and the environment for a long time so that the toxicity does not affect to the environment. The main requirements for the HLW repository design is to keep the buffer temperature below $100\;^{\circ}C$ in order to maintain its integrity. So the cooling time of spent fuels discharged from the nuclear power plant is the key consideration factors for efficiency and economic feasibility of the repository. The disposal tunnel/disposal hole spacing, the disposal area and thermal capacity required for the deep geological repository layout which satisfies the temperature requirement of the disposal system is analyzed to set the optimized spent fuels cooling time. To do this, based on the reference disposal concept, thermal stability analyses of the disposal system have been performed and the derived results have been compared by setting the spent fuels cooling time and the disposal tunnel/disposal hole spacing in various ways. From these results, desirable spent fuels cooling time in view of disposal area is derived. The results shows that the time reaching the maximum temperature within the design limit of the temperature in the disposal site is likely shortened as the cooling time of spent fuels becomes short. Also it seems that the temperature-rising and-dropping patterns in the disposal site are of smoothly varying form as the cooling time of spent fuels becomes long. In addition, it is revealed that a desirable cooling time of spent fuels is approximately 40-50 years when spent fuels are supposedly disposed in the deep geological disposal site with its structural scale under consideration in this study.

• Journal of Radiation Protection and Research
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• v.21 no.4
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• pp.313-327
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• 1996

As an at-reactor(AR) storage method o( spent fuel, there are horizontal concrete module type, metal storage cask type, concrete storage cask type, dual purpose (transportation and storage) cask type and multi-purpose (transportation, storage and disposal) cask type. All other types except multi-purpose one have been already used for AR dry storage of spent fuels after obtaining operation license in various foreign countries. Also the development of multi-purpose type has been continued for operation license. In America, Japan, Germany, Canada, Spain, Switzerland, and Czech Republic, etc., AR dry storage facilities are under operation or on propulsion, and spent fuels are transported to interim storage facility or reprocessing plant after dry storage at reactor temporarily. At Wolsung site, in case of Korea, concrete silo type has already been introduced, and it is believed to be inevitable to store spent fuels at reactor temporarily, considering the reality that storage capacity of spent fuel is approaching to the limit in some nuclear power plants. In this report, the system characteristics, design requirements, technical standards and status of AR storage system, which is suitable for domestic site such as Kori, have been studied. In most cases, the licensed period of storage cask is limited up to 20 years and the integrity of material and maintenance of leaktightness are required during the whole service life.

• Analytical Science and Technology
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• v.11 no.6
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• pp.421-428
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• 1998

A method has been studied to separate Zr from various fission products in PWR spent nuclear fuels. A solution containing metal ions in place of radioactive fission products was prepared. The Zr was separated with 5 M HCl followed by eluting metal ions such as Ce, Nd, Cs, Rb, Ba, Sr, Ru, Rh, Pd, Ag and Cd with 12 M HCl on Dowex $1{\times}8$, anion exchange resin. The recovery of Zr was more than 95%. The purification of Zr was carried out on anion exchange resin, Dowex $1{\times}8$, in 5 M HCl in order to remove Mo causing isobaric effect during mass spectrometry. The method was applied to separate Zr from a spent PWR fuel. From mass spectrometric measurement, the purified Zr portion was not showed the isobars from other elements such as Mo and Sr.

• Journal of Energy Engineering
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• v.13 no.4
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• pp.259-266
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• 2004

Specific heat mechanism of oxide fuel is contributed by lattice vibration, dilatation, conduction electron and defect and excess specific heat. Model of oxide fuel for specific heat consists of specific heat at constant pressure term, dilatation specific heat term and defect specific heat term. In this study experimental and published data on the specific heats of oxide nuclear fuels have been reviewed and analyzed to recommend the best fitting model. The oxide fuels considered in this paper were UO$_2$, mixed (U, Pu) oxides and spent fuel. The specific heat data of spent fuel has been replaced by that of simulated fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.239-245
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• 2010

There are 11,811 LWR spent fuels stored at reactor sites, as of 2009. Source terms based on reference spent fuel which represents entire spent fuels with bounding values in the aspect of source term has been applied to a design of nuclear installations, instead of those which are generated by weighting respective source term for each spent fuel. Simplified regression models to estimate total decay heat, radioactivity, and ingestion hazard index for spent fuel from Westinghouse-type reactors were developed in this study, because it can be used as a fundamental model for weighting source term for respective spent fuel to exclude conservativeness in source terms. It was found that the estimated source terms agreed with calculated value from ORIGEN-ARP within 5%. It was also found that the conservativeness could be excluded if the weight source terms were used as reference source term in the design. Therefore, it is expected that the developed regression model could be widely used in the conceptual design process of nuclear facilities related with storage and disposal of spent nuclear fuel.

• 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.20 no.1
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• pp.33-41
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• 2022

Liquid Bi pool is a candidate electrode for an electrometallurgical process in the molten LiCl-KCl eutectic to treat the spent nuclear fuels from nuclear power plants. The electrochemical behavior of Bi³⁺ ions and the electrode reaction on liquid Bi pool were investigated with the cyclic voltammetry in an environment with or without BiCl₃ in the molten LiCl-KCl eutectic. Experimental results showed that two redox reactions of Bi³⁺ on inert W electrode and the shift of cathodic peak potentials of Li⁺ and Bi³⁺ on liquid Bi pool electrode in molten LiCl-KCl eutectic. It is confirmed that the redox reaction of lithium with respect to the liquid Bi pool electrode would occur in a wide range of potentials in molten LiCl-KCl eutectic. The obtained data will be used to design the electrometallurgical process for treating actinide and lanthanide from the spent nuclear fuels and to understand the electrochemical reactions of actinide and lanthanide at liquid Bi pool electrode in the molten LiCl-KCl eutectic.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2005.10a
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• pp.1055-1058
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• 2005

In the ACPF(Advanced spent nuclear fuel Conditioning Process Facility), the spent fuel pellets which are highly radioactive materials are separated with its clad and are fed into the next conditioning process. For this, at the other facility called PIEF(Post Irradiation Examination Facility) a spent fuel rod, 3.5 m long, is cut by 25 cm long which is suitable length fur the decladding process. These rod-cuts are packed into the capsule and are moved to the ACPF. Once the capsule is unloaded in the ACPF, the rod-cut is taken out one-by-one from the capsule and installed on the decladding device. In these processes, the crushed spent fuel pellet can be scattered inside the facilities and thus it contaminate the hot cell. In this paper, we developed the specially designed capsule which prevents the pellets scattering and remarkably reduces the leading and unloading time of the rod-cuts.

• Nuclear Engineering and Technology
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• v.10 no.1
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• pp.26-36
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• 1978

Numerical computations of radioactivities and decay energies in a spent fuel have been carried out for designing of a hot cell. Optimum wall and window thicknesses that can preserve spent fuel rods for experimental purposes are estimated with burnup rate of 33,000 MWD/T(U) which is nearly maximum from a pressurized water reactor such as the Go-Ri Unit 1. Before putting the spent fuels into a hot cell, it is assumed for thickness estimates of shield materials that they are cooled in a storage tay for several lime intervals. Considered are various types of shield materials through which changing the distances from a source to an observation point is also made.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1553-1561
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• 2024

Nondestructive radiography using cosmic ray muons has been used for decades to monitor nuclear reactor and spent nuclear fuel storage. Because nuclear fuel assemblies are highly dense and large, typical radiation probes such as x-rays cannot penetrate these target imaging objects. Although cosmic ray muons are highly penetrative for nuclear fuels as a result of their relatively high energy, the wide application of muon tomography is limited because of naturally low cosmic ray muon flux. This work presents a new image reconstruction algorithm to maximize the utility of cosmic ray muon in tomography applications. Muon momentum information is used to improve imaging resolution, as well as muon scattering angle. In this work, a new convolution was introduced known as M-value, which is a mathematical integration of two measured quantities: scattering angle and momentum. It captures the objects' quantity and density in a way that is easy to use with image reconstruction algorithms. The results demonstrate how to reconstruct images when muon momentum measurements are included in a typical muon scattering tomography algorithm. Using M-value improves muon tomography image resolution by replacing the scattering angle value without increasing computation costs. This new algorithm is projected to be a standard nondestructive radiography technique for spent nuclear fuel and nuclear material management.