• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3260-3271
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• 2022

Experience shows that the solution to waste management in any national programme is lengthy and burdened with uncertainties. There are several uncertainties that contribute to the costs associated with spent fuel management. In this work, we have analysed the impact of the current nuclear data on the isotopic composition of the spent fuel and consequently their influence on the main spent fuel observables such as decay heat, activity, neutron multiplication factor, and neutron and photon source terms. Nuclear libraries based on the most general nuclear data ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0 and JEFF-3.3 are considered. A typical NPP Krško fuel assembly is analysed using the Monte Carlo code Serpent 2. The analysis considers burnup of up to 60 GWd/tU and cooling times of up to 100 years. The comparison of results showed significant differences, which should be taken into account when selecting the library and evaluating the uncertainty in determining the characteristics of the spent fuel.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.738-744
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• 2018

Fuel failure due to debris is a major cause of failure in pressurized water reactors. Fuel vendors have developed various filtering devices to reduce debris-induced failure and have evaluated filtering performance with their own test facilities and methods. Because of the different test facilities and methods, it is difficult to compare filtering performances objectively. This study presents an improved filtering test and an efficiency calculation method to fairly compare fuel-filtering efficiency regardless of the vendor's filtering features. To enhance the reliability of our evaluation, we established requirements for the test method and had a facility constructed according to the requirements. This article describes the debris specimens, the amount of debris, and the replicates for the proposed test method. A calculation method of comprehensive debris-filtering efficiency using a weighted mean is proposed. The test method was verified by repeated tests, and the tests were carried out using the PLUS7 and 17ACE7 test fuels to calculate the comprehensive debris-filtering efficiencies. The evaluation results revealed that the filtering performance of PLUS7 is better than that of 17ACE7. The proposed method can be used on any kind of debris-filtering devices and is appropriate for use as a standard.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.4
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• pp.435-446
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• 2021

This article examines the consequences of a significant spent fuel management decision or event in the United States, South Korea and Taiwan. For the United States, it is the financial impact of the Department of Energy's inability to take possession of spent fuel from commercial nuclear power companies beginning in 1998 as directed by Congress. For South Korea, it is the potential financial and socioeconomic impact of the successful construction, licensing and operation of a low and intermediate level waste disposal facility on the siting of a spent fuel/high level waste repository. For Taiwan, it is the operational impact of the Kuosheng 1 reactor running out of space in its spent fuel pool. From these, it draws six broad lessons other countries new to, or preparing for, nuclear energy production might take from these experiences. These include conservative planning, treating the back-end of the fuel cycle holistically and building trust through a step-by-step approach to waste disposal.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.1
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• pp.115-147
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• 2023

The Fukushima accident in 2011 revealed some major flaws in traditional nuclear fuel materials under accidental conditions. Thus, the focus of research has shifted toward "accident tolerant fuel" (ATF). The aim of this approach is to develop fuel material solutions that lead to improved reactor safety. The application of protective coatings on the surface of nuclear fuel cladding has been proposed as a near-term solution within the ATF framework. Many coating materials are being developed and evaluated. In this article, an overview of different zirconium-based alloys currently in use in the nuclear industry is provided, and their performances in normal and accidental conditions are discussed. Coating materials proposed by different institutions and organizations, their performances under different conditions simulating nuclear reactor environments are reviewed. The strengths and weaknesses of these coatings are highlighted, and the challenges addressed by different studies are summarized, providing a basis for future research. Finally, technologies and methods used to synthesize thin-film coatings are outlined.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.1
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• pp.119-126
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• 2022

This study provides technical information about the nuclear fuel handling process, which consists of various subprocesses starting from new fuel receipt to spent fuel shipment at a nuclear power plant and the design requirements of fuel handling equipment. The fuel handling system is an integrated system of equipment, tools, and procedures that allow refueling, handling and storage of fuel assemblies, which comprise the fuel handling process. The understanding and reaffirming of detailed code requirements are requested for application to the design of the fuel handling and storage facility. We reviewed the design requirements of the fuel handling equipment for its adequate cooling, prevention of criticality, its operability and maintainability, and for the prevention of fuel damage and radiological release. Furthermore, we discussed additional technical issues related to upgrading the current code requirements based on the modification of the fuel handling equipment. The suggested information provided in this paper would be beneficial to enhance the safety and the reliability of the fuel handling equipment during the handling of new and spent fuel.

• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.133-140
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• 2011

In this paper, a deformation measurement technology for nuclear fuel rods is proposed. The deformation measurement system includes a high-definition CMOS image sensor, a lens, a semiconductor laser line beam marker, and optical and mechanical accessories. The basic idea of the proposed deformation measurement system is to illuminate the outer surface of a fuel rod with a collimated laser line beam at an angle of 45 degrees or higher. For this method, it is assumed that a nuclear fuel rod and the optical axis of the image sensor for observing the rod are vertically composed. The relative motion of the fuel rod in the horizontal direction causes the illuminated laser line beam to move vertically along the surface of the fuel rod. The resulting change of the laser line beam position on the surface of the fuel rod is imaged as a parabolic beam in the high-definition CMOS image sensor. An ellipse model is then extracted from the parabolic beam pattern. The center coordinates of the ellipse model are taken as the feature of the deformed fuel rod. The vertical offset of the feature point of the nuclear fuel rod is derived based on the displacement of the offset in the horizontal direction. Based on the experimental results for a nuclear fuel rod sample with a formation of surface crud, an inspection resolution of 50 ${\mu}m$ is achieved using the proposed method. In terms of the degree of precision, this inspection resolution is an improvement of more than 300% from a 150 ${\mu}m$ resolution, which is the conventional measurement criteria required for the deformation of neutron irradiated fuel rods.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.365-369
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• 2022

Operating and decommissioning nuclear power plants generates radioactive waste. This radioactive waste can be categorized into several different levels, for example, low, intermediate, and high, according to the regulations. Currently, low and intermediate-level waste are stored in conventional 200-liter drums to be disposed. However, in Korea, the disposal of intermediate-level radioactive waste is virtually impossible as there are no available facilities. Furthermore, large-sized intermediate-level radioactive waste, such as reactor internals from decommissioning, need to be segmented into smaller sizes so they can be adequately stored in the conventional drums. This segmentation process requires additional costs and also produces secondary waste. Therefore, this paper suggests repurposing the no-longer-used spent nuclear fuel casks. The casks are larger in size than the conventional drums, thus requiring less segmentation of waste. Furthermore, the safety requirements of the spent nuclear fuel casks are severer than those of the drums. Hence, repurposed spent nuclear fuel casks could better address potential risks such as dropping, submerging, or a fire. In addition, the spent nuclear fuel casks need to be disposed in compliance with the regulations for low level radioactive waste. This cost may be avoided by repurposing the casks.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4731-4742
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• 2022

RBMK fuel assemblies differ from other LWR FA due to a specific arrangement of the fuel rods, the low enrichment, and the used burnable absorber - erbium. Therefore, there is a challenge to adapt modeling tools, developed for other LWR types, to solve RBMK problems. A set of 10 different depletion simulation schemes were tested to estimate the impact on reactivity and spent fuel composition of possible SCALE code options for the neutron transport modelling and the use of different nuclear data libraries. The simulations were performed using cross-section libraries based on both, VII.0 and VII.1, versions of ENDF/B nuclear data, and assuming continuous energy and multigroup simulation modes, standard and user-defined Dancoff factor values, and employing deterministic and Monte Carlo methods. The criticality analysis with burn-up credit was performed for the SFP loaded with RBMK-1500 FA. Spent fuel compositions were taken from each of 10 performed depletion simulations. The criticality of SFP is found to be overestimated by up to 0.08% in simulation cases using user-defined Dancoff factors comparing the results obtained using the continuous energy library (VII.1 version of ENDF/B nuclear data). It was shown that such discrepancy is determined by the higher U-235 and Pu-239 isotopes concentrations calculated.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3723-3740
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• 2021

The harsh conditions in the reactor affect the thermal and mechanical performance of the fuel plate heavily. Some in-pile behaviors, like fission-induced swelling, can cause a large deformation of fuel plate at very high burnup, which may even disturb the flow of coolant. In this research, the emphasis is put on the thermal expansion, fission-induced swelling, interaction layer (IL) growth, creep of the fuel meat, and plasticity of the cladding for the U₃Si₂/Al dispersion fuel plate. A detailed model of the fuel meat swelling is developed. Taking these in-pile behaviors into consideration, the three-dimensional large deformation incremental constitutive relations and stress update algorithms have been developed to study its thermal-mechanical performance under normal conditions using Abaqus. Results have shown that IL can effectively decrease the thermal conductivity of fuel meat. The high Mises stress region mainly locates at the interface between fuel meat and cladding, especially around the side edge of the interface. With irradiation time increasing, the stress in the fuel plate gets larger resulting from the growth of fuel meat swelling but then decreases under the effect of creep deformation. For the cladding, plasticity deformation does not occur within the irradiation time.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.129-130
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• 2018