• Nuclear Engineering and Technology
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• v.16 no.1
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• pp.11-17
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• 1984

Sample calculations were done for KORI-1 to develop a better understanding of what happens after very small LOCA ($\leq$0.05 ft$^2$). For a water-side break with the break size larger than 0.006 ft$^2$, fluid-loss through break exceeds the makeup. If the break size is larger than 0.008ft$^2$, decay heat can be completely removed through break. Based on these results, it was concluded that KORI-1 is fairly safe for the whole spectrum of sizes in very small LOCA. However, for the reactor with 900 MWe or 1200 MWe, a certain spectrum of sizes in very small LOCA should be carefully considered. In the accident sequence the transition from natural circulation to pool boiling or from pool boiling to natural circulation may be troublesome to the operator or in the safety analysis. Operator's intervention was discussed; primary pump shutoff, HPI pump shutoff, break isolation, and opening relief valve. It was proved that continuous operation of HPI pumps after shutdown will not threaten the integrity of the primary system.

• Proceedings of the SAREK Conference
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• 2007.11a
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• pp.305-309
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• 2007

FTL(Fuel Test Loop) is a facility that confirms performance of nuclear fuel at a similar irradiation condition with that of nuclear power plant. FTL construction work began on August, 2006 and ended on March, 2007. During Construction, ensuring the worker's safety was the top priority and installation of the FTL without hampering the integrity of the HANARO was the next one. The installation works were done successfully overcoming the difficulties such as on the limited space, on the radiation hazard inside the reactor pool, and finally on the shortening of the shut down period of the HANARO. The Commissioning of the FTL is to check the function and the performance of the equipment and the overall system as well. The FTL shall start operation with high burn up test fuels in early 2008 if the commissioning and licensing progress on schedule.

• Journal of KIISE:Software and Applications
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• v.27 no.2
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• pp.157-167
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• 2000

As the requirements of a software system become large and complex, it causes some problems such that requirements specification using formal methods becomes larger in its size and less understandable. In order to solve such problems, the concepts of modularity and object are adopted to specify the requirements. In addition, top-down and compositional approach to handle such requirements are also adopted. In our paper, we suggest an object-oriented Petri net, called HOONet, to hierarchically specify and verify the complex requirements by incorporating the concepts of modularity, object, abstraction and refinement into a formal method. Our HOONet method supports the incremental specification and verification of partially described or not yet fully analyzed requirements. We also show the applicability of our method by modeling and verifying the requirements of a reactor safety control system.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.17-36
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• 2010

In this study, we compare the mass release rates of radionuclides(1) from waste forms arising from the KIEP-21 pyroprocessing system with (2) those from the directly-disposed pressurized-water reactor spent fuel, to investigate the potential radiological and environmental impacts. In both cases, most actinides and their daughters have been observed to remain in the vicinity of waste packages as precipitates because of their low solubility. The effects of the waste-form alteration rate on the release of radionuclides from the engineered-barrier boundary have been found to be significant, especially for congruently released radionuclides. the total mass release rate of radionuclides from direct disposal concept is similar to those from the pyroprocessing disposal concept. While the mass release rates for most radionuclides would decrease to negligible levels due to radioactive decay while in the engineered barriers and the surrounding host rock in both cases even without assuming any dilution or dispersal mechanisms during their transport, significant mass release rates for three fission-product radionuclides, $^{129}I$, $^{79}Se$, and $^{36}Cl$, are observed at the 1,000-m location in the host rock. For these three radionuclides, we need to account for dilution/dispersal in the geosphere and the biosphere to confirm finally that the repository would achieve sufficient level of radiological safety. This can be done only after we have known where the repository site would by sited. the footprint of repository for the KIEP-21 system is about one tenth of those for the direct disposal.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2997-3009
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• 2023

The Passive containment Air-cooling System (PAS) can effectively remove the decay heat of the modular small nuclear reactor after an accident. The details of natural convection around the dome, which is a key part of PAS, were investigated numerically in the present study. The thermal dynamics around the dome were studied through the temperature, pressure and velocity contours and the streamlines. Additionally, the formation of the buoyant plume at the top of the dome was investigated. The results show that with the increase of Ra, the lift-off point moves toward the bottom of the dome, and the eddy under the buoyant plume grows larger gradually, which enhances the heat transfer. And the heat transfer along the dome surface with different truncation angles was investigated. As the angle increases, the heat transfer coefficient becomes stronger as well. Consequently, a newly developed heat transfer correlation considering the influence of truncation angle for the dome is proposed based on the simulated results. This study could provide a better understanding of natural convection around the dome of PAS and the proposed correlation could also offer more predictive value in the improvement of nuclear safety.

• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.5
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• pp.465-471
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• 2013

A steam generator in nuclear power plant is a heatexchager which is used to convert water into steam from heat produced in a nuclear reactor core, and the steam produced in steam generator is delivered to the turbine to generate electricity. Because of damage to steam generator tubing may impair its ability to adequately perform required safety functions in terms of both structural integrity and leakage integrity, eddy current testing is periodically performed to evaluate the integrity of tubes in steam generator. This assessment is normally performed during a reactor refueling outage. Currently, the eddy current testing for steam generator of nuclear power plant in Korea is performed in accordance with KEPIC & ASME Code requirements, the eddy current testing system is consists of remote data acquisition unit and data analysis program to evaluate the acquired data. The KEPIC & ASME Code require that the electrical properties of remote data acquisition unit, such as total harmonic distortion, input & output impedance, amplifier linearity & stability, phase linearity, bandwidth & demodulation filter response, analog-to-digital conversion, and channel crosstalk shall be measured in accordance with the KEPIC & ASME Code requirements. In this paper, the measurement requirements of electrical properties for eddy current testing instrument described in KEPIC & ASME Code are presented, and the measurement results of newly developed eddy current testing instrument by KHNP(Korea Hydro & Nuclear Power Co., LTD) are presented.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.28 no.5
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• pp.523-533
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• 2015

A reactor containment acts as a final barrier to prevent leakage of radioactive material due to the possible reactor accidents into external environment. Because of the functional importance of the containment building, the SIT(Structural Integrity Test) for containments shall be performed to evaluate the structural acceptability and demonstrate the quality of construction. An initial numerical analysis was performed to simulate the results obtained from the SIT for a prestressed concrete(PSC) structure. But the analysis results by the initial model expected smaller displacements than the measured ones by 30% at some locations. Accordingly, the research and development to improve the initial model to corelate the measured results of the SIT more properly have been performed. In this paper, the effects of the loss of concrete due to duct for tendons and the contact of duct and tendons in un-bonded tendon system are mainly evaluated based on the preliminary analysis results. In addition, the importances of the proper definition of mesh connectivity among structural elements of concrete, liner plates, rebars and tendons are discussed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.729-738
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• 2009

Given the evolution of High-Temperature Gas-cooled Reactor(HTGR) designs, the source terms for licensing must be developed. There are three potential source terms: fission products, actinides in the fuel and tritium in the coolant. It is necessary to provide first an inventory of the source terms under normal operations. An analysis of source terms has yet to be performed for HTGRs. The previous code, which can estimate the inventory of the source terms for LWRs, cannot be used for HTGRs because the general data of a typical neutron cross-section and flux has not been developed. Thus, this paper uses a combination of the MCNP, ORIGEN, and MONTETEBURNS codes for an estimation of the source terms. A method in which the HTR-10 core is constructed using the unit lattice of a body-centered cubic is developed for core modeling. Based on this modeling method by MCNP, the generation of fission products, actinides and tritium with an increase in the burnup ratio is simulated. The model developed by MCNP appears feasible through a comparison with models developed in previous studies. Continuous fuel management is divided into five periods for the feeding and discharging of fuel pebbles. This discrete fuel management scheme is employed using the MONTEBURNS code. Finally, the work is investigated for 22 isotope fission products of nuclides, 22 actinides in the core, and tritium in the coolant. The activities are mainly distributed within the range of $10^{15}{\sim}10^{17}$ Bq in the equilibrium core of HTR-10. The results appear to be highly probable, and they would be informative when the spent fuel of HTGRs is taken into account. The tritium inventory in the primary coolant is also taken into account without a helium purification system. This article can lay a foundation for future work on analyses of source terms as a platform for safety assessment in HTGRs.

• Transactions of the KSME C: Technology and Education
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• v.3 no.4
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• pp.257-264
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• 2015

To maintain sustainability of nuclear energy as an important energy source, both safety problem and Spent Nuclear Fuels(SNFs) problem should be solved. In case of Gen-IV reactors such as fast reactor, SNFs can be used as fuels by using fast neutrons. It can be a suitable treatment method of high-level waste in near future. Liquid metals such as Sodium or Lead-Bismuth Eutectic (LBE) can be possibly used as a coolant to use fast neutrons. In this paper, it was described that natural circulation parameter studies, design analyses, material selections and a completion of facilities. To develop a natural circulation facility, thermal hydraulic analyses were performed. Installation technique of liquid metal natural circulation were secured.