• Journal of the Korean Society of Safety
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• v.30 no.5
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• pp.123-130
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• 2015

Wolsong Unit 2,3&4 has been performing a containment integrity test during power operation. This test could impact to the safe operation during test. If an accident occurs during pressure dropping phase, reactor trip can be delayed because of the increased pressure difference which causes a time delay to reach the trip set-point. On the contrary, if an accident occurs during pressure increasing phase, reactor trip could be accelerated because the pressure difference to the trip set-point decrease. Point Lepreau nuclear power plant, which installed GCLMS (Gross Containment Leakage Monitoring System) in 1990, has discontinued the test since 1992 due to these adverse effects. Therefore, we evaluated the risk to obviate the GCLMS test based on PWR's ILRT (Integrated Leak Rate Test) extension methodologies. The results demonstrate that risk increase rate is not high in case of performing only ILRT test at every 5 years instead of doing GCLMS test at every 1.5 years. In addition, the result shows that GCLMS test can be removed on a risk-informed perspective since risk increasement is in acceptable area of regulatory acceptance criteria.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1015-1024
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• 2009

The presence of a non-condensable gas can considerably reduce the level of condensation heat transfer. The non-condensable gas effect is a primary concern in some passive systems used in advanced design concepts, such as the Passive Residual Heat Removal System (PRHRS) of the System-integrated Modular Advanced ReacTor (SMART) and the Passive Containment Cooling System (PCCS) of the Simplified Boiling Water Reactor (SBWR). This study examined the capability of the Multi-dimensional Analysis of Reactor Safety (MARS) code to predict condensation heat transfer in a vertical tube containing a non-condensable gas. Five experiments were simulated to evaluate the MARS code. The results of the simulations showed that the MARS code overestimated the condensation heat transfer coefficient compared to the experimental data. In particular, in small-diameter cases, the MARS predictions showed significant differences from the measured data, and the condensation heat transfer coefficient behavior along the tube did not match the experimental data. A new method for calculating condensation heat transfer coefficient was incorporated in MARS that considers the interfacial shear stress as well as flow condition determination criterion. The predictions were improved by using the new condensation model.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.413-422
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• 2014

Presented in this study are the variations of microstructures and mechanical properties with tempering and Post-Weld Heat Treatment (PWHT) conditions for SA508 Gr.4N steel used as Reactor Pressure Vessel (RPV) material. The blocks of model alloy were austenitized at the conventional temperature of $880^{\circ}C$ then tempered and post-weld heat treated at four different conditions. The hardness and yield strength decrease with increased tempering and PWHT temperatures, but impact toughness is significantly improved, especially in the specimens tempered at $630^{\circ}C$. The sample tempered at $630^{\circ}C$ with PWHT at $610^{\circ}C$ shows optimum mechanical properties in hardness, strength, and toughness, excluding only the transition property in the low temperature region. The microstructural observation and quantitative analysis of carbide size distribution show that the variations of mechanical properties are caused by the under-tempering and carbide coarsening which occurred during the heat treatment process. The introduction of PWHT results in the deterioration of the ductile-brittle transition property by an increase of coarse carbides controlling cleavage initiation, especially in the tempered state at $630^{\circ}C$.

• Journal of Mechanical Science and Technology
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• v.17 no.7
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• pp.947-957
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• 2003

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle. In order to complete the integrity evaluation of pressure tube, expert knowledge, iterative calculation procedures and a lot of input data are required. More over, results of integrity assessment may be different according to the evaluation method. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached database, was developed. The present system was built on the basis of 3D FEM results, ASME Sec. XI, and Fitness For Service Guidelines for CANDU pressure tubes issued by the AECL (Atomic Energy Canada Limited). The present system also covers the delayed hydride cracking and the blister evaluation, which are the characteristics of pressure tube integrity evaluation. In order to verify the present system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Nuclear Engineering and Technology
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• v.33 no.2
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• pp.209-224
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• 2001

The SSC-K system analysis code is under development at the Korea Atomic Energy Research Institute (KAERI) as a part of the KALIMER project. The SSC-K code is being used as the principal tool for analyzing a variety of off-normal conditions or accidents of the preliminary KALIMER design. The SSC-K code features a multiple-channel core representation coupled with a point kinetics model with reactivity feedback. It provides a detailed, one-dimensional thermal-hydraulic simulation of the primary and secondary sodium coolant circuits, as well as the balance-of-plant steam/water circuit. Recently a two-dimensional hot pool model was incorporated into SSC-K for analysis of thermal stratification phenomena in the hot pool. In addition, SSC-K contains detailed models for the passive decay heat removal system and a generalized plant control system. The SSC-K code has also been applied to the computational engine for an interactive simulation of the KALIMER plant. This paper presents an overview of the recent activities concerned with SSC-K code model development This paper focuses on both descriptions of the newly adopted thermal hydraulic and neutronic models, and applications to KALIMER analyses for typical anticipated transients without scram.

• Nuclear Engineering and Technology
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• v.32 no.1
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• pp.57-66
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• 2000

The RELAP5 code has been developed for best-estimate simulation of transients and accidents for pressurized water reactors and their associated systems, but it has not been fully assessed for those of CANDU reactors. However, a previous study suggested that the RELAP5 code could be applicable to simulate the transients and accidents for CANDU reactors. Nevertheless, it is indicated that there are some works to be resolved, such as modeling of headers and multi-channel simulation for the reactor core, etc. Therefore, this study has been initiated with an aim to identify the code applicability for all the postulated transients and accidents in CANDU reactors. In the present study, the small inlet header break experiment (B8604) in the RD-14 test facility was simulated with RELAP5/MOD3.2 code. The RELAP5 results were also compared with both experimental data and those of CATHENA analyses performed by AECL and the analyses demonstrated the code's capability to predict major . phenomena occurring in the transient with sufficient accuracy for both Qualitative and quantitative viewpoint However, some discrepancies in the depressurization of the primary heat transport system after the break and the consequent time delay of the major phenomena were also observed.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.817-824
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• 2014

To ensure the safety of research reactors, the water level must be maintained above the required height. When a pipe ruptures, the siphon phenomenon causes continuous loss of coolant until the hydraulic head is removed. To protect the reactor core from this kind of accident, a siphon breaker has been suggested as a passive safety device. This study mainly focused on two variables: the size of the pipe rupture and the timing of air entrainment. In this study, the size of the pipe rupture was increased to the guillotine break case. There was a region in which a larger pipe rupture did not need a larger siphon breaker, and the water flow rate was related to the size of the pipe rupture and affected the residual water quantity. The timing of air entrainment was predicted to influence residual water level. However, the residual water level was not affected by the timing of air entrainment. The experimental cases, which showed the characteristic of partical sweep-out mode in the separation of siphon breaking phenomenon [2], showed almost same trend of physical properties.

• Nuclear Engineering and Technology
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• v.37 no.3
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• pp.245-258
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• 2005

This paper addresses two types of uncertainty: stochastic uncertainty and subjective uncertainty in probabilistic accident consequence assessments. The off-site consequence assessment code OSCAAR has been applied to uncertainty and sensitivity analyses on the individual risks of early fatality and latent cancer fatality in the population outside the plant boundary due to a severe accident. A new stratified meteorological sampling scheme was successfully implemented into the trajectory model for atmospheric dispersion and the statistical variability of the probability distributions of the consequence was examined. A total of 65 uncertain input parameters was considered and 128 runs of OSCAAR with 144 meteorological sequences were performed in the parameter uncertainty analysis. The study provided the range of uncertainty for the expected values of individual risks of early and latent cancer fatality close to the site. In the sensitivity analyses, the correlation/regression measures were useful for identifying those input parameters whose uncertainty makes an important contribution to the overall uncertainty for the consequence. This could provide valuable insights into areas for further research aiming at reducing the uncertainties.

• Journal of the Korean Society of Safety
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• v.14 no.1
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• pp.73-77
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• 1999

In this study, $SO_2$ and $NO_2$ reduction have been investigated by using coil type plasma reactor. The experiments have been carried out changing discharge power, gas flow rate frequency and electrode style to obtain the decomposition rate. Decomposition rates of $SO_2$ and $NO_2$ were obtained 20~98% at gas flow rate 100ml/min~1,000ml/min and discharge power 5~25w respectively. The energy efficiency is very good at the high frequency power. The decomposition rate of $SO_2$ for 5kHz power supply is only 90%, but for 10kHz power supply is very high, more than 98% for 15w. The decomposition rate is increasing according to the residence time or the power consumption of the discharge. About 15W discharge power for 17$cm^2$ reactor is necessary to obtain the decomposition rate of $SO_2$ and $NO_2$ of more than 85% or 98%. From these experiments, the consumption power of the decomposition rate of 98% in 300ppm $NO_2$ gas in nitrogen gas proved to be 18W and 300ppm $SO_2$ gas to be 15w.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.3181-3188
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• 2008

The supercritical $CO_2$ Brayton cycle energy conversion system is presented as a promising alternative to the present Rankine cycle. The principal advantage of the S-$CO_2$ gas is a good efficiency at a modest temperature and a compact size of its components. The S-$CO_2$ Brayton cycle coupled to a SFR also excludes the possibilities of a SWR (Sodium-Water Reaction) which is a major safety-related event, so that the safety of a SFR can be improved. KAERI is conducting a feasibility study for the supercritical carbon dioxide (S-$CO_2$) Brayton cycle power conversion system coupled to KALIMER(Korea Advanced LIquid MEtal Reactor). The purpose of this research is to develop S-$CO_2$ Brayton cycle energy conversion systems and evaluate their performance when they are coupled to advanced nuclear reactor concepts of the type under investigation in the Generation IV Nuclear Energy Systems. This paper contains the research overview of the S-$CO_2$ Brayton cycle coupled to KALIMER-600 as an alternative energy conversion system.