• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.921-928
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• 2013

A future nuclear energy system is being developed at Korea Atomic Energy Research Institute (KAERI), the system involves a Sodium Fast Reactor (SFR) linked with the pyro-process. The pyro-process produces a source material to fabricate a SFR fuel rod. Therefore, an isotopic fissile content assay is very important for fuel rod safety and SFR economics. A new technology for an analysis of isotopic fissile content has been proposed using a lead slowing down spectrometer (LSDS). The new technology has several features for a fissile analysis from spent fuel: direct isotopic fissile assay, no background interference, and no requirement from burnup history information. Several calculations were done on the designed spectrometer geometry: detection sensitivity, neutron energy spectrum analysis, neutron fission characteristics, self shielding analysis, and neutron production mechanism. The spectrum was well organized even at low neutron energy and the threshold fission chamber was a proper choice to get prompt fast fission neutrons. The characteristic fission signature was obtained in slowing down neutron energy from each fissile isotope. Another application of LSDS is for an optimum design of the spent fuel storage, maximization of the burnup credit and provision of the burnup code correction factor. Additionally, an isotopic fissile content assay will contribute to an increase in transparency and credibility for the utilization of spent fuel nuclear material, as internationally demanded.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2000.11a
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• pp.639-642
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• 2000

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2009.06a
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• pp.757-758
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• 2009

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2009.10a
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• pp.413-414
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• 2009

• 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.

• Korean Journal of Computational Design and Engineering
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• v.5 no.4
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• pp.327-335
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• 2000

The spent fuel rod extracting machine is automatically operated in high radioactive environment, so high reliability of operation is required. In this paper, to enhance the reliability of this machine by providing a close monitoring capability. a real time graphic simulation method is suggested. This method utilizes conventional IGRIP (Interactive Graphics Robot Instruction Program) 3D graphic simulation tool to visualize and simulate the 3D graphic model of this machine. Also, the dedicated protocol is defined for transmission of the operational data of the machine. The real time graphic simulation is realized by developing the socket module between a graphic workstation and a machine control computer through the TCP/IP network and by dividing the 3D graphic simulation GSL(Graphic Simulation Language) program as a small sized sub routine. The suggested method is implemented while automatically operating the rod extracting machine. The result of implementation shows that the real time 3D graphic simulation is well synchronized with the actual machine according to the operational data.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4499-4513
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• 2022

Spent fuel behavior of dry storage was simulated in a continuous state from steady-state operation by modifying FRAPCON-4.0 to incorporate spent fuel-specific fuel behavior models. Spent fuel behavior of a typical PWR was compared with that of NuScale Power Module (NPM^TM). Current PWR discharge burnup (60 MWd/kgU) gives a sufficient margin to the hoop stress limit of 90 MPa. Most hydrogen precipitation occurs in the first 50 years of dry storage, thereby no extra phenomenological safety factor is identified for extended dry storage up to 100 years. Regulation for spent fuel management can be significantly alleviated for LWR-based SMRs. Hydride embrittlement safety criterion is irrelevant to NuScale spent fuels; they have sufficiently lower plenum pressure and hydrogen contents compared to those of PWRs. Cladding creep out during dry storage reduces the subchannel area with burnup. The most deformed cladding outer diameter after 100 years of dry storage is found to be 9.64 mm for discharge burnup of 70 MWd/kgU. It may deteriorate heat transfer of dry storage by increasing flow resistance and decreasing the view factor of radiative heat transfer. Self-regulated by decreasing rod internal pressure with opening gap, cladding creep out closely reaches the saturated point after ~50 years of dry storage.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.31-39
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• 2013

Nuclear fuel can be damaged under unexpected circumstances in a nuclear reactor. Fuel rod failure can be occurred due to debris fretting or excessive hydriding or PCI (Pellet-to-clad Interaction) etc. It is important to identify the causes of such failed fuel rods for the safe operation of nuclear power plants. If a fuel rod failure occurs during the operation of a nuclear power plant, the coolant water is contaminated by leaked fission products, and in some case the power level of the plant may be lowered or the operation stopped. In addition, all spent fuels must be transferred to a dry storage. But failed fuel can not be transferred to a dry storage. Therefore, the purpose of this study is to develop a system which is capable of inspecting whether the spent fuel in the storage pool is failed or not. The sipping technology is to analyze the leakage of fission products in state of gas and liquid. The failed fuel inspection system with gamma analyzer has successfully demonstrated that the system is enough to find the failed fuel at Wolsong plant.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.365-365
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• 2001

• Journal of Korea Foundry Society
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• v.20 no.2
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• pp.80-88
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• 2000

Continuous casting equipment was designed to cast the metallic uranium rods, and a thermal analysis was carried out to calculate the temperature and solidification profiles. Fluid flow and heat transfer analysis model including the effects of phase change was used to simulate the continuous casting process by finite volume method. In the design of continuous casting equipment, the casting speed, pouring temperature and cooling conditions should be considered as significant factors. In this study, the effects of casting speed, pouring temperature, and air gap between the uranium and mold were investigate. The results represented that the temperature and solidification profiles of continuous casting equipment varied with the casting speed, pouring temperature, and air gap.