• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3709-3715
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• 2023

It is an important work to measure the dimensions of underwater spent fuel assemblies in the nuclear power industry during the overhaul, to judging whether the spent fuel assemblies can continue to be used. In this paper, a three dimensional reconstruction method for underwater spent fuel assemblies of nuclear reactor based on linear structured light is proposed, and the topography and size measurement was carried out based on the reconstructed 3D model. Multiple linear structured light sensors are used to obtain contour size data, and the shape data of the whole spent fuel assembly can be collected by one-dimensional scanning motion. In this paper, we also presented a corrected model to correct the measurement error introduced by lead-glass and water is corrected. Then, we set up an underwater measurement system for spent fuel assembly based on this method. Finally, an underwater measurement experiment is carried out to verify the 3D reconstruction ability and measurement ability of the system, and the measurement error is less than ±0.05 mm.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.386-395
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• 2004

The Korean fuel cycle scenario has been modeled by using the dynamic analysis method. For once-through fuel cycle model, the nuclear power plant construction plan was considered, and the nuclear demand growth rate from the year 2016 was assumed to be 1%. After setup the once-thorough fuel cycle model, the DUPIC and fast reactor scenarios were modeled to investigate the environmental effect of each fuel cycle. Through the calculation of the amount of spent fuel, and the amounts of plutonium and minor actinides were estimated and compared to those of the once-through fuel cycle. The results of the once-through fuel cycle shows that the demand grows to 64 GWe and the total amount of the spent fuel would be 100 kt in the year 2100, while the total spent fuel can be reduced by 50% when the DUPIC scenario is implemented

• Journal of Powder Materials
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• v.7 no.3
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• pp.123-130
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• 2000

The simulated DUPIC fuel provides a convenient way to investigate fuel properties and behaviours such as thermal conductivity, thermal expansion, fission gas release, leaching and so on. Several pellets simulating the composition and microstructure of the DUPIC fuel were fabricated from resintering powder through the OREOX process of the simulated spent fuel pellets, which were prepared from the mixture of stable forms of constituent nuclides. This study describes the powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using the simulated spent fuel. The homogeneity of additives in the powder was observed after attrition milling. The microstructure of the simulated spent fuel was in agreement with the previous studies. The densities and the grain size of simulated DUPIC fuel was pellets are higher than those of simulated spent fuel pellets. Small metallic precipitates and oxide precipitates were observed on matrix grain boundaries.

• Nuclear Engineering and Technology
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• v.40 no.6
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• pp.517-526
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• 2008

The widespread concern regarding the management of spent fuel that mainly contributes to nuclear waste has led to the development of the sodium-cooled fast reactor (SFR) as one of the most promising future types of reactors at both national and international levels. Various reactor deployment scenarios with SFR introductions with different conversion ratios in the existing PWR-dominant nuclear fleet have been assessed to optimize the SFR deployment strategy to replace PWRs with the view toward a reduction in the level of spent fuel as well as efficient uranium utilization through its reuse in a closed fuel cycle. An efficient reactor deployment strategy with the SFR introduction starting in 2040 has been drawn based on an SFR deployment strategy in which burners are deployed prior to breakeven reactors to reduce the amount of PWR spent fuel substantially at the early deployment stage. The PWR spent fuel disposal is reduced in this way by 98% and the cumulative uranium demand for PWRs to 2100 is projected to be 445 ktU, implying a uranium savings of 115 ktU. The SFR mix ratio in the nuclear fleet near the year 2100 is estimated to be approximately 35-40%. PWRs will remain as a main power reactor type until 2100 and SFRs will support waste minimization and fuel utilization.

• Nuclear Engineering and Technology
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• v.42 no.3
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• pp.231-236
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• 2010

The growth in global energy demand and the increased recognition of the impacts of carbon dioxide emissions from fossil fuel plants have aroused a renewed interest on nuclear energy. Many countries are looking afresh at building more nuclear power stations to deal with the twin problems of global warming and the need for more generating capacity. Many in the nuclear community are also anticipating a significant growth of new nuclear generation in the coming decades. If there is a nuclear renaissance, will the expansion of nuclear power be compatible with global non-proliferation and security? or will it add to the environmental burden from the large inventory of spent nuclear fuel already produced in existing nuclear power reactors? We learn from past peaceful nuclear activities that significant concerns associated with nuclear proliferation and spent-fuel management have resulted in a decrease in public acceptance for nuclear power in many countries. The terrorist attack in the United States (US) on September 11, 2001 also raised concern for security and worry that nuclear materials may fall into the wrong hands. As we increase the use of nuclear power, we must simultaneously reduce the proliferation, security and environmental risks in managing spent-fuel below where they are today.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1333-1344
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• 2019

An integrated approach of model simplification for high burnup spent nuclear fuel is proposed based on material calibration using optimization. The spent fuel rods are simplified into a beam with a homogenous isotropic material. The proposed approach of model simplification is applied to fuel rods with two kinds of interfacial configurations between the fuel pellets and cladding. The differences among the generated models and the effects of interfacial bonding efficiency are discussed. The strategy of model simplification adopted in this work is to force the simplified beam model of spent fuel rods to possess the same compliance and failure characteristics under critical loads as those that result in the failure of detailed fuel rod models. It is envisioned that the simplified model would enable the assessment of fuel rod failure through an assembly-level analysis, without resorting to a refined model for an individual fuel rod. The effective material properties of the simplified beam model were successfully identified using the integrated optimization process. The feasibility of using the developed simplified beam models in dynamic impact simulations for a horizontal drop condition is examined, and discussions are provided.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.717-723
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• 2018

Thermal analysis of accident conditions is an important problem during safety assessment of the dry spent nuclear fuel storage facilities. Thermal aspects of accident conditions with channel blockage of ventilated storage containers are considered in this article. Analysis of flow structure inside ventilated containers is carried out by numerical simulation. The main mechanisms of heat and mass transfer, which take part in spent nuclear fuel cooling, were detected. Classification of accidents on the basis of their influence on the maximum temperatures inside storage casks is proposed.

• Journal of Radiation Protection and Research
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• v.34 no.3
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• pp.144-150
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• 2009

This study examines how much the radiation dose rate around it varies if a crack occurs on the spent nuclear fuel rod. The spent nuclear fuel rod to be examined is that of Kori unit 3&4. The source terms are evaluated using the ORIGEN-ARP that is part of the version 5.1 of the SCALE package. The radiation dose rate is assessed using the TORT. To check if the structure of a fuel rod is appropriately modeled in the TORT calculation, the calculation results by the TORT are compared with those by the ANISN for the same case. From the code simulation, it is known that if a crack occurs on the spent nuclear fuel rod, the neutron dose rate varies depending on what material is the crack filled with, but the gamma dose rate varies irrespective of type of the material that the crack is filled with.

• Proceedings of the Korea Concrete Institute Conference
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• 2004.11a
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• pp.381-384
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• 2004

Delaying and objection for the construction of storage spent-fuel disposal has prompted to consider expanding on-site storage of spent reactor fuel since it can eliminate the need for costly and difficult shipping and control of the spent fuel completely under the direction of the owner-utility. The dry storage unit developed in Canada can accommodate Korea heavy water reactor fuel elements and become a candidate for the Korean market. In this paper, finite element analyses were carried out in order to investigate the structural behavior of the nuclear spent fuel dry storage system, which is subjected to impact loads such as collision of a truck load and dropping of flask under the irregular operation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.2
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• pp.255-258
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• 2022

Three government bodies, that is, the Ministry of Science and ICT (MSIT), Ministry of Trade, Industry, and Energy (MOTIE), and Nuclear Safety and Security (NSSC), jointly established the Institute for Korea Spent Nuclear Fuel (iKSNF) in December 2020 to secure the management technologies for spent nuclear fuel (SNF). The objective of iKSNF is to successfully conduct the long-term research and development program of the 「Development of Core Technologies to Ensure Safety of Spent Nuclear Fuel Storage and Disposal System」. Our program, known as the first multi-ministry program in the nuclear field of Korea, mainly focuses on developing core technologies required for the long-term management of SNF, including those for safe storage and deep geological disposal of SNF. The program comprises three subprograms and seven key projects covering the storage, disposal, and regulatory sectors of SNF management. Our program will last from 2021 through 2029, with a budget of approximately four billion USD sponsored by MSIT, MOTIE, and NSSC.