• Proceedings of the Korean Society of Precision Engineering Conference
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• 2011.06a
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• pp.1213-1214
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• 2011

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.167-181
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• 2019

Currently, the development of evaluation technology for vibration and shock load characteristics and spent nuclear fuel structural integrity under normal conditions of transport is being conducted in the Republic of Korea. This is the first such research conducted in the Republic of Korea and, thus, previous international studies need to be investigated and will be referred to in the ongoing project. Before 2000, several studies related to measurement of vibration and shock loads on spent nuclear fuel were conducted in the US. US national research institutes conducted uniaxial fuel assembly shaker tests, concrete block tests, and multi-axis fuel assembly tests between 2009 and 2016. In 2017, multi-modal transportation tests including road, sea, and rail transport were also performed by research institutes from the US, Spain and the Republic of Korea. Therefore, test preparation procedures, acceleration and strain measurement results, and finite-element and multi-body dynamics analysis were investigated. Based on the measured strain data, the preliminary conclusion was obtained that the measured strain was too small to cause damage to spent nuclear fuel rods. However, this conclusion is a preliminary conclusion that only reviews part of the results; a detailed review is being conducted in the US. The investigation of international studies on spent nuclear fuel structural integrity tests considering vibration and shock loads under normal conditions of transport in the US will be useful data for the project being conducted in the Republic of Korea.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.284-292
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• 2019

For the continuous operation of a nuclear reactor, burnt fuel needs to be replaced with fresh fuel, where appropriate (ex-vessel) fuel handling is required. Particularly for the Sodium-cooled Fast Reactor (SFR) refueling, its process has unique challenges due to liquid sodium coolant. The ex-vessel spent fuel transportation should concern several design features such as the radiation shielding, decay-heat removal, and inert space separated from air. This paper proposes a new design optimization methodology of cask shielding to transport the spent fuel assembly in a prototype SFR for the first time. The Particle Swarm Optimization (PSO) algorithm had been applied to design trade-offs between shielding and cask weight. The cask is designed as a double-cylinder structure to block an inert sodium region from the air-cooling space. The PSO process yielded the optimum shielding thickness of 26 cm, considering the weight as well. To confirm the shielding performance, the radiation dose of spent fuel removed at its peak burnup and after 1-year cooling was calculated. Two different fuel positions located during transportation were also investigated to consider a functional disorder in a cask drive system. This study concludes the current cask design in normal operations is satisfactory in accordance with regulatory rules.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.91-100
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• 2005

Kyung-hee Thorium Fuel (KTF), a heterogeneous thorium-based seed and blanket design concept for pressurized light water reactors, is being studied as an alternative to enhance proliferation resistance and fuel cycle economics of PWRs. The proliferation resistance characteristics of the KTF assembly design were evaluated through parametric studies using neutronic performance indices such as Bare Critical Mass (BCM), Spontaneous Neutron Source rate (SNS), Thermal Generation rate (TG), and Radio-Toxicity. Also, Fissile Economic Index (FEI), a new index for gauging fuel cycle economy, was suggested and applied to optimize the KTF design. A core loaded with optimized KTF assemblies with a seed-to-blanket ratio of 1: 1 was tested at the Korea Next Generation Reactor (KNGR), ARP-1400. Core design characteristics for cycle length, power distribution, and power peaking were evaluated by HELIOS and MASTER code systems for nine reload cycles. The core calculation results show that the KTF assembly design has nearly the same neutronic performance as those of a conventional $UO_2$ fuel assembly. However, the power peaking factor is relatively higher than that of conventional PWRs as the maximum Fq is 2.69 at the M$9^{th}$ equilibrium cycle while the design limit is 2.58. In order to assess the economic potential of a heterogeneous thorium fuel core, the front-end fuel cycle costs as well as the spent fuel disposal costs were compared with those of a reference PWR fueled with $UO_2$. In the case of comprising back-end fuel cycle cost, the fuel cycle cost of APR-1400 with a KTF assembly is 4.99 mills/KWe-yr, which is lower than that (5.23 mills/KWe-yr) of a conventional PWR. Proliferation resistance potential, BCM, SNS, and TG of a heterogeneous thorium-fueled core are much higher than those of the $UO_2$ core. The once-through fuel cycle application of heterogeneous thorium fuel assemblies demonstrated good competitiveness relative to $UO_2$ in terms of economics.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2001.04a
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• pp.631-634
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• 2001

In this study, the graphic simulation system of multi joint manipulator is developed to analyze and optimize the remote handling processes for the spent fuel assembly. This system consists of a 3-D graphical modeling system, a device assembling system, and a motion simulation system. To analyze and optimize the processes involved in multi-joint manipulator operation such as NFBC transportation process and bottom nozzle removal process, the virtual work place is implemented using a computer graphic technology. This virtual workcell is exactly same as that of the real environment. This graphic simulation system of the multi-joint manipulator can be effectively used for designing the main processes and maintenance processes of the spent fuel management.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.504-511
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• 2018

In Lithuania, all spent nuclear fuel (SNF) resulted from the operation of the Ignalina Nuclear Power Plant (INPP), which had two Russian Acronym for "Channelized Large Power Reactor"-type reactors. After the final shutdown, the total amount of SNF at the INPP was approximately 22,000 fuel assemblies. All these assemblies will be stored for about 50 years and disposed of after that. The decision to shut down and decommission both reactors in Lithuania before termination of design period raises a significant challenge for the treatment of accumulated SNF. Therefore, various techniques and technologies for SNF management were developed and justified for that specific case, and a set of special equipment was installed at the INPP, the effectiveness of which was demonstrated during its operation. This article presents unique techniques related to the management of SNF adopted and commissioned at the INPP after its operation shutdown, namely fuel rod cladding leak tightness control system and special equipment for collection of possible spillage during handling of SNF assembly in the hot cell. The operational experience and measurement results of fuel rod cladding leak tightness control system are presented.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.383.2-383
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.306-306
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• 1998

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.67-67
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.349-349
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• 1999