• Journal of the Korean Professional Engineers Association
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• v.45 no.3
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• pp.18-22
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• 2012

Main purpose of this paper is to describe what is radwastes and spent fuel of radioisotope and nuclear power plants. Although this term is well adapted among the people nowadays but it is necessary to know more clear definition in accordance with atomic energy promotion law. By various means of scientific knowledge and engineering techniques, we can shield harmful radiation from radwastes and spent fuel, also could know how we can take benefits from radioisotope applications for human health care and other applications. It is realized that problem of spent fuel is not easy task to solve at the moment because we have to consider many sides.

• 제어로봇시스템학회:학술대회논문집
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• 1991.10a
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• pp.911-916
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• 1991

Spent nuclear fuel is highly radioactive and requires remote operation due to radiation exposure. Motors which have been used in the highly radioactive environments are a DC type because of their easy implementation on control system. However there are some problems such as mechanical maintenance of brush and commutator, high cost, and heavy material control. AC servo motors are applied and tested on fuzzy and conventional control algorithms. Fuzzy logic controls of AC servo give adequate control accuracy and power for spent fuel handling in radioactive environments.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.549-555
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• 2017

A lead slowing-down spectrometer (LSDS) system is under development to analyze isotopic fissile content that is applicable to spent fuel and recycled material. The source neutron mechanism for efficient and effective generation was also determined. The source neutron interacts with a lead medium and produces continuous neutron energy, and this energy generates dominant fission at each fissile, below the unresolved resonance region. From the relationship between the induced fissile fission and the fast fission neutron detection, a mathematical assay model for an isotopic fissile material was set up. The assay model can be expanded for all fissile materials. The correction factor for self-shielding was defined in the fuel assay area. The corrected fission signature provides well-defined fission properties with an increase in the fissile content. The assay procedure was also established. The assay energy range is very important to take into account the prominent fission structure of each fissile material. Fission detection occurred according to the change of the Pu239 weight percent (wt%), but the content of U235 and Pu241 was fixed at 1 wt%. The assay result was obtained with 2~3% uncertainty for Pu239, depending on the amount of Pu239 in the fuel. The results show that LSDS is a very powerful technique to assay the isotopic fissile content in spent fuel and recycled materials for the reuse of fissile materials. Additionally, a LSDS is applicable during the optimum design of spent fuel storage facilities and their management. The isotopic fissile content assay will increase the transparency and credibility of spent fuel storage.

• 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.8 no.3
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• pp.239-245
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• 2010

There are 11,811 LWR spent fuels stored at reactor sites, as of 2009. Source terms based on reference spent fuel which represents entire spent fuels with bounding values in the aspect of source term has been applied to a design of nuclear installations, instead of those which are generated by weighting respective source term for each spent fuel. Simplified regression models to estimate total decay heat, radioactivity, and ingestion hazard index for spent fuel from Westinghouse-type reactors were developed in this study, because it can be used as a fundamental model for weighting source term for respective spent fuel to exclude conservativeness in source terms. It was found that the estimated source terms agreed with calculated value from ORIGEN-ARP within 5%. It was also found that the conservativeness could be excluded if the weight source terms were used as reference source term in the design. Therefore, it is expected that the developed regression model could be widely used in the conceptual design process of nuclear facilities related with storage and disposal of spent nuclear fuel.

• 제어로봇시스템학회:학술대회논문집
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• 2003.10a
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• pp.1571-1574
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• 2003

A shared compliant control scheme based on IMC is proposed for the position-force force reflecting control system. The controller of the slave manipulator is designed by IMC method for the open loop unstable plant. The compliant control is implemented by first order low pass filter. In the proposed scheme, the slave manipulator well tracks the position of the master manipulator in free space and the compliance of the slave manipulator is autonomously controlled in contact condition. The simulation results show that the excellence of the proposed controller.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.470-475
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• 1998

The marginal nuclear criticality analysis for the high density spent fuel storage at a PWR plant was carried out by using the HELIOS and CASMO-3 codes. More than 20 % of the calculated reactivity saving effect is observed in this analysis. This mainly comes from the adoption of some important fission products and B-10 in the criticality analysis. By taking burnup and boron credits, the high capacity of the spent fuel storage rack can be more fully utilized, reducing the space of storage. Larger storage for a given inventory of spent fuel should result in remarkable cost savings and mort importantly reduce the risks to the public and occupational workers.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.847-852
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• 1995

A shielding analysis for KSC-7, the shipping cask for transporting the 7 PWR spent fuel assemblies, has been carried out. Radiation source term has been calculated on spent fuel with burnup of 50,000 MWD/MTU and 1.5 years cooling time by ORIGEN2 code. The shielding calculation for the cask has been made by using MCNP4A code with continuous cross section data library from ENDF/B-V. As a result of neutron dose rate analysis, another shielding calculational model on spent fuel shipping cask was provided which is using the Monte Carlo method.

• 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.

• Bulletin of the Korean Chemical Society
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• v.20 no.1
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• pp.59-64
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• 1999

Based on the study of hydrolysis behaviour of U(Ⅵ) ion and major fission product metal ions such as Cs(Ⅰ), Ce(Ⅲ), Nd(Ⅲ), Mo(Ⅵ), Ru(Ⅱ), and ZR(Ⅳ) in the titration media, the performance of noncomplexing-alkalimetric titration method for the determination of free acid in the presence of these metal ions was investigated and its results were compared to those from the completing methods. The free acidities could be determined as low as 0.05 meq in uranium solutions in which the molar ratio of U(Ⅵ)/H+ was less than 5, when the end-point of titration was estimated by Gran plot. The biases in the determinations were less than 1% and about +3% respectively for 0.4 meq and 0.05 meq of free acid at the U(Vl)/H+ molar ratio of up to 5. Applicability of this method to the determination of free acid in spent fuel solutions was confirmed by the analysis of nitric acid content in simulated spent fuel solutions and in a real spent fuel solution.