• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.2
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• pp.141-154
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• 2015

In general, conventional criticality analysis for spent fuel transport/storage systems have been performed based on the assumption of fresh fuel concerning the potential uncertainties from number density calculations of actinide nuclides and fission products in spent fuel. However, these evaluation methods cause financial losses due to an excessive criticality margin. In order to overcome this disadvantage, many studies have recently been conducted to design and commercialize a transportation and storage cask applied to the Burnup Credit (BUC). This study conducted an assessment to ensure criticality safety for reactor operating parameters, axial burn-up profiles and misload accident conditions, which are the factors that are likely to affect criticality safety when the BUC is applied to the dual-purpose cask under development at the KOrea RADioactive waste agency (KORAD). As a result, it was found that criticality resulting from specific power, changed substantially and relied on conditions of low enrichment and high burn-up. Considering the end effect in the case of high burn-up produced a positive-definite result. In particular, the increment of maximum effective multiplication factors due to misloading was 0.18467, confirming that misload is a factor that must be taken into account when applying the BUC. The results of this study may therefore be utilized as references in developing technologies to apply the BUC to domestic models and operational procedures or preventing any misload accidents during the process of spent fuel loading.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.10 no.12
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• pp.3748-3754
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• 2009

An integral head loss test in a test apparatus was conducted to simulate chemical effects on a head loss across a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). The test was conducted during 30 days in the condition of a short spray, a long spray, and no materials with chemical effects. The result exhibited that the head loss was affected on amounts of the exposed materials according to spray conditions. XRD analysis of the collected precipitates showed that the precipitates were phosphate compounds. Comparison of the head loss with dissolved species concentration showed that high increase rate of the head loss resulted from the corrosion of aluminum and zinc but slow increase rate of the head loss resulted from the precipitates induced by Si, Mg, and Ca from leaching reaction at NUKON and concrete after passivation of metal specimens.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.126-139
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• 1994

Best-estimate thermal-hydraulic codes, CATHARE2 V1.2 and RELAP5/MOD3, hate been assessed against the BETHSY 6.2 tc six-inch cold leg break loss-of-coolant accident (LOCA) test. Main objective is to analyze the overall capabilities of the two codes on physical phenomena of concern during the small break LOCA i.e. two-phase critical flow, depressurization, core water level de-pression, loop seal clearing, liquid holdup, etc. The calculation results show that the too codes predict well both in the occurrences and trends of major two-phase flow phenomena observed. Especially, the CATHARE2 calculations show better agreements with the experimental data. However, the two codes, in common, show some deviations in the predictions of loop seal clearing, collapsed core water level after the loop seal clearing, and accumulator injection behaviors. The discrepancies found from the comprision with the experimental data are larger in the RELAP5 results than in the CATHARE2. To analyze the deviations of the two code predictions in detail, several sensitivity calculations have been performed. In addition to the change of two-phase discharge coefficients for the break junction, fine nodalization and some corrections of the interphase drag term are made. For CATHARE2, the change of interphase drag force improves the mass distribution in the primary side. And the prediction of SG pressure is improved by the modification of boundary conditions. For RELAP5, any single input change doesn't improve the whole result and it is found that the interphase drag model has still large uncertainties.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.55-61
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• 2015

When a radiation detector is applied to the measurement of the radioactivity of high-level of radioactive materials or the rapid response to the nuclear accident, several collimators with the different inner radii should be prepared according to the level of dose rate. This makes the in-situ measurement impractical, because of the heavy weight of the collimator. In this study, an IRIS collimator was developed so as to have a function of controlling the inner radius, with the same method used in optical camera, to vary the attenuation ratio of radiation. The shutter was made to have the double tungsten layers with different phase angles to prevent the radiation from penetrating owing to the mechanical tolerance. The performance evaluation through the MCNP code was conducted by calculating the attenuation ratio according to the inner radius of the collimator. The attenuation ratio was marked on the outer scale ring of the collimator. It is expected that when a radiation detector with the IRIS collimator is used for the in-situ measurement, it can change the attenuation ratio of the incident photon to the detector without replacing the collimator.

• Journal of Radiation Protection and Research
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• v.38 no.2
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• pp.60-67
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• 2013

A diffusion coefficient is an important parameter in the prediction of atmospheric dispersion using a Gaussian plume model, and its modelling approach varies. In this study, dispersion coefficients recommended by the U. S. Nuclear Regulatory Commission's (U. S. NRC's) regulatory guide and the Canadian Nuclear Safety Commission's (CNSC's) regulatory guide, and used in probabilistic accident consequence analysis codes MACCS and MACCS2 have been investigated. Based on the atmospheric dispersion model for a hypothetical accidental release recommended by the U. S. NRC, its influence to atmospheric dispersion factor was discussed. It was found that diffusion coefficients are basically predicted from a Pasquill- Gifford curve, but various curve fitting equations are recommended or used. A lateral dispersion coefficient is corrected with consideration for the additional spread due to plume meandering in all models, however its modelling approach showed a distinctive difference. Moreover, a vertical dispersion coefficient is corrected with consideration for the additional plume spread due to surface roughness in all models, except for the U. S. NRC's recommendation. For a specified surface roughness, the atmospheric dispersion factors showed differences up to approximately 4 times depending on the modelling approach of a dispersion coefficient. For the same model, the atmospheric dispersion factors showed differences by 2 to 3 times depending on surface roughness.

• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.2
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• pp.191-198
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• 2008

Flow accelerated corrosion (FAC) phenomenon of low alloy carbon steels in nuclear power plant has been known as one of major degradation mechanisms. It has a potential to cause nuclear pipe rupture accident which may directly impact on the plant reliability and safety. Recently, the equipotential switching direct current potential drop (ES-DCPD) method has been developed, by the present authors, as a method to monitor wall loss in a piping. This method can rapidly monitor the thinning of piping, utilizing either the wide range monitoring (WiRM) or the narrow range monitoring (NaRM) technique. WiRM is a method to monitor wide range of straight piping, whereas NaRM focuses significantly on a narrow range such as an elbow. WiRM and NaRM can improve the reliability of the current FAC screening method that is based on computer modeling on fluid flow conditions. In this paper, the measurements by ES-DCPD are performed with signal sensitivity analyses in the laboratory environment for extended period and showed the viability of ES-DCPD for real plant applications.

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.137-149
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• 1996

The present study is to analyze an integral test, BETHSY test 6.9c, which represent loss of RURS accident under mid-loop operation. Both the pressurizer manway and the steam generator outlet plenum manway are opened as vent paths in order to prevent the system from pressurization by removing the steam generated in the core. The main purposes are to gain insights into the physical phenomena and identify sensitive parameters. Assessment of capability of CATHARE2 prediction can be established the effective recovery procedures using the code in an actual plant. Most of important physical phenomena in the experiment could be predicted by the CATHARE2 code. The peak pressure in the upper plenum is predicted higher than experimental value by 7 kPa since the differential pressure between the pressurizer and the surge line is overestimated. The timing of core uncovery is delayed by 500 seconds mainly due to discrepancy in the core void distribution. It is demonstrated that openings of the pressurizer manwey and the steam generator manway can prevent the core uncovery using only gravity feed injection. Although some disagreements are found in the detailed phenomena, the code prediction is considered reasonable for the overall system behaviors.

• Journal of Environmental Policy
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• v.13 no.4
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• pp.79-105
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• 2014

Natech risk is a type of complex disasters that natural hazards trigger technological disaster or industrial accidents. Research on Natech risk has been started from the mid-1990s in European countries and the Unites States, and drawn much more attention after the Fukushima nuclear accident caused by the 2011 East Japan earthquake. While early studies on Natech risk have focused on the causal natural hazards and possibility to occur, and the resulting spill of hazardous materials from the perspective of science and engineering, the recent research interests lie on effective Natech risk management. Especially, emphasizing the difference of Natech risk management from traditional disaster management, issues of uncertainty management, integration between natural disaster and technological disaster, and responsibility, has been drawn attention. In Korea, Natech risk has not been introduced as a research topic. Although some regulatory improvements have been made in nuclear safety and chemical Substance management after the Fukushima disaster, the potential impact of natural hazards in these areas has not been considered yet. It is necessary to raise the issues of Natech risk management in research and policy areas through active discussion and interdisciplinary approaches.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.2
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• pp.105-115
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• 2012

Necessity of demonstration test to evaluate the structural integrity of a basket for accident conditions arose during license approval procedure for the WSPP's dry storage facility named MACSTOR/KN-400. A drop test facility for demonstration was constructed in KAERI site and demonstration tests for basket drop were conducted. As the upper welding region of a loaded basket was collided with a dropped basket during the drop test, the welding in this region was fractured and leakage happened after the drop test. The enhancement of basket design was needed since the existing basket design was not able to satisfy the performance requirement. The directions for design modification were determined and six enhanced designs were derived based on these directions. Structural analyses and specimen tests for each enhanced design were conducted. By evaluating structural analysis results and test results, one among six enhanced designs was decided as a final design for revision. The final design was the one to reduce the height of central post of a basket and to decrease the impact velocity with a dropped basket. Test basket models were fabricated with accordance with the final enhanced design. Additional demonstration test was performed for this test model and all the performance requirements were satisfied.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.10 no.11
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• pp.3260-3268
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• 2009

A test apparatus has been fabricated to simulate chemical effect on head loss through a strainer in a pressurized water reactor (PWR) containment water pool after a loss of coolant accident (LOCA). Tests were conducted under condition of same ratio of strainer surface area to water volume between the test appratus and the containment sump. A series of tests have been performed to investigate the effects of spray, existence of calcium-silicate with tri-sodium phosphate (TSP), and composition of materials. The results showed that head loss across the chemical bed with even a small amount of calcium-silicate insulation instantaneously increased as soon as TSP was added to the test solution. Also, the head loss across the test screen is strongly affected by spray duration and is increased rapidly at the early stage, because of high dissolution and precipitation of aluminum and zinc. After passivation of aluminum and zinc by corrosion, the head loss increase is much slowed down and is mainly induced by materials such as calcium, silicon, and magnesium leached from NUKONTM and concrete. Furthermore, it is newly found that the spay buffer agent, tri-sodium phosphate, to form protective coating on the aluminum surface and reduce aluminum leaching is not effective for a large amount of aluminum and a long spray.