• 한국안전학회지
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• 제30권5호
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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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• 제41권8호
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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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• 제41권8호
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• pp.989-994
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• 2009

We at Mitsubishi have lined up Gen-III+ solutions for a wide variety of global customers: ATMEA1 of the 1100MWe class, and an APWR with the largest capacity of 1700MWe. In this paper, we would like to introduce the APWR. With an increased requirement for nuclear power generation as an effective countermeasure against global warming, we have established the APWR plant, a large-capacity Mitsubishi standard reactor combining our accumulated experience and technology as an integrated PWR plant supplier. The APWR plant has achieved high reliability, safety and enhanced economy based on a technology that has been developed with the support of the government and utilities through improvement and standardization programs of light water reactors. Currently, Tsuruga Units 3 and 4, the first two APWRs, are undergoing licensing, while we are making efforts to obtain the standard design certification (DC) of US-APWR and preparing for the European Utility Requirements (EUR) compliance assessment of EU-APWR. Mitsubishi Heavy Industries, Ltd. (MHI) positions the APWR as a core technology that will contribute to the prevention of global warming and meet worldwide requirements.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1367-1379
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• 2020

The small modular sodium-cooled fast reactor (SMSFR) is an important component of Generation-IV reactors. The objective of this work is to improve the reactivity control in SMSFR by using innovative systems, including burnable poisons and optimized control rods. SMSFR with MOX fuel usually exhibits high burnup reactivity loss that leads to high excess reactivity and potential fuel melting in control rod withdrawal (CRW) accidents, which becomes an important constraint on the safety and economic efficiency of SMSFR. This work applies two types of burnable poisons in a SMSFR to reduce the excess reactivity. The first one homogenously loads minor actinides in the fuel. The second one combines absorber and moderators in specific assemblies. The influence of burnable poisons on the core characteristics is discussed and integrated into the analysis of CRW accidents. The results show that burnable poisons improve the safety performance of the core in a significant way. Burnable poisons also lessen the demand for the number, absorption ability, and insertion depth of control rods. Two optimized control rod designs with rare earth oxides (Eu₂O₃ and Gd₂O₃) and moderators are compared to the conventional design with natural boron carbide (B₄C). The optimized designs show improved neutronic and safety performance.

• 한국환경보건학회지
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• 제29권3호
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• pp.86-90
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• 2003

In this study, lots of methods have been studing to utilize energy and decrease contaminated effluents. There has been great progress on IGCC (Integrated gasification combined cycle) to reduce thermal energy losses. The following results have been conducted from desulfurization experiments using waste shell to remove H$_2$S. Fixed bed desulfurization experiments, to obtain basic data for scale-up was indicated. Oyster was the best among the various sorbents, like the results of TGA. Especially, H$_2$S removal efficiency of uncalcined oyster was the highest. When use oyster as desulfurization sorbents, calcination process was not needed. Thus, high desulfurization efficiency would be expected. Fixed bed reactor experiments were indicated particle size of sorbents. These had influenced on desulfurization capacity. As smaller particle size was found better desulfurization capacity. Large capacity difference was found between 0.613 mm and 0.335 mm. But, differences between 0.335 mm and 0.241 mm was relatively small. As bed temperature increased, H$_2$S removal capacity increased. Therefore, both particle size and bed temperature should be considered to remove H$_2$S by sorbents.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.725-736
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• 2014

It is very important to detect and identify small anomalies and component failures for the safe operation of complex and large-scale artifacts such as nuclear power plants. Each diagnostic technique has its own advantages and limitations. These facts inspire us not only to enhance the capability of diagnostic techniques but also to integrate the results of diagnostic subsystems in order to obtain more accurate diagnostic results. The article describes the outline of four diagnostic techniques developed for the condition monitoring of the fast breeder reactor "Monju". The techniques are (1) estimation technique of important state variables based on a physical model of the component, (2) a state identification technique by non-linear discrimination function applying SVM (Support Vector Machine), (3) a diagnostic technique applying WT (Wavelet Transformation) to detect changes in the characteristics of measurement signals, and (4) a state identification technique effectively using past cases. In addition, a hybrid diagnostic system in which a final diagnostic result is given by integrating the results from subsystems is introduced, where two sets of values called confidence values and trust values are used. A technique to determine the trust value is investigated under the condition that the confidence value is determined by each subsystem.

• 대한전기학회논문지:시스템및제어부문D
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• 제52권8호
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• pp.452-461
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• 2003

The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. A survey on the RSPT(Reed Switch Position Transmitter) type control rod position indicator system and its actual implementation in the existing nuclear power plants in Korea was performed first. The control rod position indicator having the high performance for SMART was developed on the basis of RSPT technology identified through the survey. The arrangement of permanent magnet and reed switches is the most important procedure in the design of control rod position indicator. The hysteresis of reed switches is one of the important factors in a repeat accuracy of control rod position indicator as well. This paper investigates efficiency of the magnetic flux concentrator and the hysteresis using FEM and verified differences in physicals characteristics by comparing the results of FEM and those of the experiment. As a result, it is shown that the characteristics of prototype control rod position indicator have a good agreement with the results of FEM.

• 한국안전학회지
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• 제26권4호
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• pp.7-13
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• 2011

A study on reliability estimation of sequential-ordered multiple failure modes, which are sequentially ordered between failure modes in a considering system, was performed. Especially, an approach to estimate the probabilities of failure modes has been proposed under an assumption that failure modes are mutually exclusive and sequentially ordered by only a critical variable. A feasibility of the proposed approach were studied by a practical example, which is a reliability estimation of passive safety systems for a probabilistic safety assessment(PSA) of a very high temperature reactor(VHTR) that is under development as a future nuclear system with enhanced safety features. It is difficult to define a robust failure state of this nuclear system because of its enhanced radiation release characteristics, so the new approach is a useful concept to estimate not only its safety but also a PSA. A feasibility study applied two failure modes(e.g., small and large release of radioactive materials) with considering the integrated behavior of this nuclear system. It is expected that the multiple release states for a practical estimation can be easily extended to the aforementioned example. It was found out that the proposed approach was a useful technique to cover the unfavorable features of this nuclear system as to performing a VHTR PSA.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3449-3459
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• 2021

With the increasing availability of high-performance computing (HPC) platforms, uncertainty quantification (UQ) and sensitivity analyses (SA) can be efficiently leveraged to optimize design parameters of complex engineering problems using modeling and simulation tools. The workflow involved in such studies heavily relies on HPC resources and hence requires pre-processing and post-processing capabilities of large amounts of data along with remote submission capabilities. The NEAMS Workbench addresses all aspects of the workflows involved in these studies by relying on a user-friendly graphical user interface and a python application program interface. This paper highlights the NEAMS Workbench capabilities by presenting a semiautomated coupling scheme between Dakota and any given package integrated with the NEAMS Workbench, yielding a simplified workflow for users. This new capability is demonstrated by running a SA of a turbulent flow in a pipe using the open-source Nek5000 CFD code. A total of 54 jobs were run on a HPC platform using the remote capabilities of the NEAMS Workbench. The results demonstrate that the semiautomated coupling scheme involving Dakota can be efficiently used for UQ and SA while keeping scripting tasks to a minimum for users. All input and output files used in this work are available in https://code.ornl.gov/neams-workbench/dakota-nek5000-study.

• 한국안전학회지
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• 제36권2호
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• pp.111-119
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• 2021

Multi-dimensional analysis of reactor safety-KINS standard (MARS-KS) is a thermal-hydraulic code to simulate multiple design basis accidents in reactors. The code has been essential to assess nuclear safety, but has mainly focused on light water reactors, which are in the majority in South Korea. Few previous studies considered pressurized heavy water reactor (PHWR) applications. To verify the code applicability for PHWRs, it is necessary to develop MARS-KS input decks under various transient conditions. This study proposes an input model to simulate small-break loss of coolant accidents for PHWRs. The input model includes major equipment and experimental conditions for test B9802. Calculation results for selected variables during steady-state closely follow test data within ±4%. We adopted the Henry-Fauske model to simulate break flow, with coefficients having similar trends to integrated break mass and trip time for the power supply. Transient calculation results for major thermal-hydraulic factors showed good agreement with experimental data, but further study is required to analyze heat transfer and void condensation inside steam generator u-tubes.