• Analytical Science and Technology
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• v.16 no.6
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• pp.431-437
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• 2003

Metalization yield of uranium oxide to uranium metal from lithium reduction process of spent pressurized water reactor (PWR) fuels was measured using thermogravimetric analyzer. A reduced metal produced in the process was divided into a solid and a powder part, and each metalization yield was measured. Metalization yield of the solid part was 90.7~95.9 wt%, and the powder being 77.8~71.5 wt% individually. Oxidation behaviour of the quartemary alloy was investigated to take data on the thermal oxidation stability necessary for the study on dry storage of the reduced metal. At $600{\sim}700^{\circ}C$, weight increments of alloy of Mo, Ru, Rh and Pd was 0.40~0.55 wt%. Phase change on the surface of the alloy was started at $750^{\circ}C$. In particular, Mo was rapidly oxidized and then the alloy lost 0.76~25.22 wt% in weight.

• Analytical Science and Technology
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• v.13 no.6
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• pp.781-788
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• 2000

Tellurium in simulated nuclear spent fuels (SIMFUEL) has been determined by hydride generation-inductively coupled plasma atomic emission spectrometry (HG-ICP-AES). Parameters such as concentrations of HCl and $NaBH_4$, flow rate of HCl and $NaBH_4$ were optimized and then the effects of U, Mo, Pd, Rh and Ru on the Te intensity were investigated. A thiourea as a masking agent was used to eliminate or minimize such interferences specially caused by palladium. Tellurium was measured by HG-ICP-AES and ICP-MS after separation of tellurium from SIMFUEL with cation exchange chromatography. The relative deviation between direct measurement and separation method was less than 6% based on the data by ICP-MS.

• The Journal of Engineering Geology
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• v.11 no.3
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• pp.255-267
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• 2001

A comparison study is performed to understand the coupling behavior of the thermal, hydraulic, and mechanical interactions in the vicinity of an underground radwaste repository, assumed to be located at a depth of 500 m, within a granitic rock mass with a 58$^{\circ}$ dipping fault passing through the roof-wall intersection of the repository cavern. The two dimensional universal distinct element code, UDEC is used for the analysis. The model includes a granitic rock meas, a canister with PWR spent fuels surrounded by the compacted bentonite inside a deposition hole, and the mixed bentonite backfilled in the rest of the space within a repository cavern. The coupling behavior of hydromechanical, thermomechanical, and thermohydromechanical interaction has been studied and compared. The effect of the time-dependent decaying heat, from the radioactive materials in PWR spent fuels, on the repository and its surroundings has been studied. A steady state flow algorithm is used for the hydraulic analysis.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.997-1001
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• 2016

Conductive ceramics are being developed to replace current Pt anodes in the electrolytic reduction of spent oxide fuels in pyroprocessing. While several conductive ceramics have shown promising electrochemical properties in small-scale experiments, their long-term stabilities have not yet been investigated. In this study, the chemical stability of conductive $La_{0.33}Sr_{0.67}MnO_3$ in $LiCl-Li_2O$ molten salt at $650^{\circ}C$ was investigated to examine its feasibility as an anode material. Dissolution of Sr at the anode surface led to structural collapse, thereby indicating that the lifetime of the $La_{0.33}Sr_{0.67}MnO_3$ anode is limited. The dissolution rate of Sr is likely to be influenced by the local environment around Sr in the perovskite framework.

• Nuclear Engineering and Technology
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• v.34 no.3
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• pp.187-201
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• 2002

The KN-12 cask is designed to transport 12 PWR spent nuclear fuels and to comply with the requirements of Korea Atomic Energy Act, IAEA Safety Standards Series No.57-1 and US 10 CFR Part 71 for a Type B(U)F package. It provides containment, radiation shielding, structural integrity, criticality control and heat removal for normal transport and hypothetical accident conditions. W.H 14$\times$14, 16$\times$16 and 17$\times$17 fuel assemblies with maximum allowable initial enrichment of 5.0 wt.%, maximum average burn-up of 50,000 MWD/MTU and minimum cooling time of 7 years being used in Korea will be loaded and subsequently transported under dry and wet conditions. A forged cylindrical cask body which constitutes the containment vessel is closed by a cask lid. Polyethylene rods for neutron shielding are arranged in two rows of longitudinal bore holes in the cask body wall. A fuel basket to accommodate up to 12 PWR fuel assemblies provides support of the fuels, control of criticality and a path to dissipate heat. Impact limiters to absorb the impact energy under the hypothetical accident conditions are attacked at the top and at the bottom side of the cask during transport. Handling weight loaded with water is 74.8 tons and transport weight loaded with water with the impact limiters is 84.3 tons. The cask will be licensed in accordance with Korea Atomic Energy Act 3nd fabricated in Korea in accordance with ASME B&PV Code Section 111, Division 3.

• 제어로봇시스템학회:학술대회논문집
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• 2003.10a
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• pp.1354-1358
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• 2003

The process equipment for handling high level radioactive materials like spent fuels is operated in a hot cell, due to high radioactivity. Thus, this equipment should be maintained and repaired optimally by a remotely operated manipulator. The master-slave manipulators(MSM) are widely used as a remote handling device in the hot cell. The equipment in the hot cell should be optimally placed within the workspace of the wall-mounted slave manipulator for the maintenance operation. But, because of the complexity in the hot cell, there would be some parts of the equipment that are not reached by the MSM. In this study, the maintenance process for these parts of the equipment is developed using virtual prototyping technology. To analyze the workspace of the maintenance device in the hot cell and to develop the maintenance processes for the process equipment, the virtual mock-up of the hot cell for the spent fuel handling process is implemented using IGRIP. For the implementation of the virtual mock-up, the parts of the equipment and maintenance devices such as the MSM and servo manipulator are modeled and assembled in 3-D graphics, and the appropriate kinematics are assigned. Also, the virtual workcell of the spent fuel management process is implemented in the graphical environment, which is the same as the real environment. Using this mock-up, the workspace of the manipulators in the hot cell and the operator's view through the wall-mounted lead glass are analyzed. Also, for the dedicated maintenance operation, the analyses for the detailed area of the end effectors in accordance with the slave manipulator's position and orientation are carried out. The parts of the equipment that are located outside of the MSM's workspace are specified and the maintenance process of the parts using the servo manipulator that is mounted in the hot cell is proposed. To monitor the process in the hot cell remotely, the virtual display system by a virtual camera in the virtual work cell is also proposed. And the graphic simulation using a virtual mock-up is performed to verify the proposed maintenance process. The maintenance process proposed in this study can be effectively used in the real hot cell operation and the implemented virtual mock-up can be used for analyzing the various hot cell operations and enhancing the reliability and safety of the spent fuel management.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.163-169
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• 2015

In many nuclear power plant sites in Korea, high density storage racks were installed in the spent fuel pool to expand the spent fuel storage capacity. Nevertheless, the capability of the Hanbit nuclear site will be saturated by 2024. Also, 10 NPPs will reach their design life expiration date by 2029. In the case of the US, SFPI (Spent Fuel Pool Island) operated temporarily as a spent fuel storage option before spent nuclear fuels were transported to an interim storage facility or a final disposal facility. As a spent fuel storage option after shutdown during decommissioning, the SFPI concept can be expected to have the following effects: reduced occupational exposure, lower cost of operation, strengthened safety, and so on. This paper presents a case study associated with the regulations, operating experiences, and systems of SFPI in the US. In conclusion, the following steps are recommended for applying SFPI during decommissioning in Korea: confirmation of design change scope of SFPI and expected final cost, the submission of a decommissioning plan which is reflected in SFPI improvement plans, safety assessment using PSR, application of an operating license change for design change, regulatory body review and approval, design change, inspection by the regulatory body, education and commissioning for SFPI, SFPI operation and periodic inspection, and dismantling of SFPI.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.261-267
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• 2004

Inventories, projections, and characteristics of spent nuclear fuel(SNF) generated from domestic nuclear power plants were updated to support high-level waste disposal system design. The historical and projected inventory by the end 2055 is expected to be 20,500 and 14,800MTU for PWR and CANDU spent nuclear fuel, respectively The ratio of quantity for TEX>$17{\times}17$ SNF was shown to be 0.6 as of 2003. The amount of TEX>$17{\times}17$ SNF, however, will be less than that of TEX>$16{\times}16$ KSFA after 2012, while the quantity of TEX>$16{\times}16$ KSFA will reach to 70% of the total spent fuels in the 2055. Average turnup of SNF revealed ~36GWD/MTU and ~40GWD/MTU for the period of 1994-1999 and 2000-2003, respectively. It is expected that the average burnup of SNF will exceed 45GWD/MTU at the end of 2000's. Therefore, it seems reasonable to use the TEX>$17{\times}17$ 4.5w/o, 45GWD/MTU as the Reference SNF at present state. The TEX>$16{\times}16$ KSFA 4.5w/o, 55GWD/MTU, however, should be Reference SNF after ~2010.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2002.05a
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• pp.840-843
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• 2002

The dry pulverizing/Mixing device is used to deal with the spent fuels for the safe disposal. The separated pellets from hulls by a slitting device are put and oxidized from UO$_2$ solid pellet to U$_3$O$\_$8/ powder in the device. The device have been developed based on a voloxidation method which is one of several dry de-cladding methods. We have benchmarked dry de-cladding methods, analyzed applicability to the advanced spent fuel management process, integrated and compared several configuration, and finally derived detailed specifications proper to requirements for the device. Also, thermal characteristics of the device such as thermal stress and strain have been analyzed by the commercial software, 1-DEAS, and the reliability of the results have been verified by the KOLAS(Korea Laboratory Accreditation Scheme). The UO$_2$ solid pellets are put in the device which has a capacity of 20 kgHM per a batch, heated up about 600$^{\circ}C$ in the air environment. Then, the UO$_2$ solid pellets are oxidized into the U$_3$O$\_$8/ powder, and the powder is collected in a special vessel. The device has been designed and developed as fellows: the multi-staged fine hole meshes are used to reduce the size of the powder gradually, heat and air(oxygen) are supplied continuously to reduce the reaction time, and slight vibration effect are applied to collect powder cling to the device.

• Nuclear Engineering and Technology
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• v.22 no.2
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• pp.83-94
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• 1990

The radioactivity in the spent fuel storage pool is calculated to ensure to maintain its concentration below the permissible limit, when the storage capacity of Uljin nuclear power plant unit 1&2 is extended from 9/3 to 32/3 core using consolidated fuels in maximum density rack (MDR). For this evalulation, two models to calculate the spent fuel pool activities on the continuous and intermittent operating its purification system are developed and these results compared, The results of above two cases show that the current water purification system can not guarantee the radioactivity concentration below the design limit, 5$\times$10$^{-4}$ $\mu$Ci/ml, for the extention to 32/3 core. Therefore, it has been concluded that a modification of the current purification system is necessary to extend the spent fuel storage capacity with the above method. The alternative way suggested in this study is to increase the number of cation bed demineralizers.