• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.19-33
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• 2024

In nuclear thermal-hydraulic system codes, most correlations used for vertical pipes, under downward two-phase flow, have been developed considering small pipes or pool systems. This suggests that there could be uncertainties in applying the correlations to accident scenarios involving large vertical pipes owing to the difference in the characteristics of two-phase flows, or flow conditions, between large and small pipes. In this study, we modified the Multi-dimensional Analysis of Reactor Safety KINS Standard (MARS-KS) code using correlations, such as the drift-flux model and two-phase multiplier, developed in a plant-scale air-inflow experiment conducted for a pipe of diameter 600 mm under downward two-phase flow. The results were then analyzed and compared with those based on previous correlations developed for small pipes and pool conditions. The modified code indicated a good estimation performance in two plant-scale experiments with large pipes. For the siphon-breaking experiment, the maximum errors in water flow for modified and original codes were 2.2% and 30.3%, respectively. For the air-inflow accident experiment, the original code could not predict the trend of frictional pressure gradient in two-phase flow as / increased, while the modified MARS-KS code showed a good estimation performance of the gradient with maximum error of 3.5%.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.32 no.1
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• pp.17-27
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• 2019

In this paper, we study the applicability of Tuned Mass Damper(TMD) to improve seismic performance of piping system under earthquake loading. For this purpose, a mode analysis of the target pipeline is performed, and TMD installation locations are selected as important modes with relatively large mass participation ratio in each direction. In order to design the TMD at selected positions, each corresponding mode is replaced with a SDOF damped model, and accordingly the corresponding pipeline is converted into a 2-DOF system by considering the TMD as a SDOF damped model. Then, optimal design values of the TMD, which can minimize the dynamic amplification factor of the transformed 2-DOF system, are derived through GA optimization method. The proposed TMD design values are applied to the pipeline numerical model to analyze seismic performance with and without TMD installation. As a result of numerical analyses, it is confirmed that the directional acceleration responses, the maximum normal stresses and directional reaction forces of the pipeline system are reduced, quite a lot. The results of this study are expected to be used as basic information with respect to the improvement of the seismic performance of the piping system in the future.

• Journal of the Korean Operations Research and Management Science Society
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• v.30 no.2
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• pp.63-71
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• 2005

In the field of nuclear reactor safety study, common cause failures (CCFs) became significant contributors to system failure probability and core damage frequency in most Probabilistic risk assessments. However, it is hard to estimate the reliability of such a system, because of the dependency of components caused by CCFs. In order to analyze the system, we propose an analytic method that can find the parameters with lack of raw data. This study adopts the shock model in which the failure probability increases as the shock is cumulated. We use two-step Expectation and Maximization (EM) algorithm to find the unknown parameters. In order to verify the analysis result, we perform the simulation under same environment. This approach might be helpful to build the defensive strategy for the CCFs.

• Proceedings of the KIEE Conference
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• 2004.07d
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• pp.2298-2300
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• 2004

This paper does with the design. implementation, and test of a CRCS for nuclear power plants. Although CRCS is still classified into non-safety class, much attention on its reliability issue has been given so far because of its importance for the stable operation of the reactor in the plant. In terms of technical aspects, our system adopts a full-duplex configuration to enhance reliability in contrast to the existing systems that are all simplex.

• Journal of Welding and Joining
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• v.19 no.1
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• pp.88-94
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• 2001

DUPIC (Direct use of spent PWR fuel in CANDU reactors) nuclear fuel is a CANDU fuel fabricated remotely from spent PWR fuel materials in a hot cell. The soundness of the end closure welds of nuclear fuel elements is an important factor for the safety and performance of nuclear fuel. To evaluate the soundness of the end closure welds of DUPIC fuel element, a precise X-ray inspection system is developed using a micro-focus X-ray generator with an image intensifier and a real time camera system. The fuel elements made of Zircaloy-4 and stainless steel by an Nd:YAG laser welding and a TIG welding aye inspected by the developed inspection system. The soundness of the welds of the fuel elements was confirmed by the X-ray inspection process, and the irradiation test of DUPIC fuel elements has been successfully completed at the HANARO research reactor.

• Proceedings of the Korean Information Science Society Conference
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• 2001.04a
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• pp.586-588
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• 2001

원전보호계통(RPS: Reactor Protection System)은 사고 시 치명적 피해를 입을 수 있다는 점에서 안전에 대한 중요도가 가장 높은 Safety 1E class로 분류되며, 이러한 보호계통을 디지털 라이즈 하는데 있어서 높은 신뢰도에 대한 보장이 필요하다. 따라서 본 논문에서는 DPPS(Digital Plant Protection System) 내에서 작동하는 내장형 소프트웨어의 높은 신뢰성을 보장하기 위한 테스팅 방법론을 제시하고자 한다. DPPS에서 작동하는 내장형 소프트웨어를 테스트하기 위한 방법은 크게 두 가지로 나누어진다. 첫 번째 단계는 절차중심의 프로그램에서 객체를 추출하고 이를 이용하여 클래스를 추출하는 제공학의 단계이다. 두 번째 단계는 이러한 클래스들을 이용하여 레벨별 테스팅을 수행하기 위한 테스트 아이템을 추출하고, 추출된 테스트 아이템을 이용하여 테스트 케이스를 선정하는 단계이다. 이렇게 각 레벨별로 선정된 테스트 케이스를 이용하여 단위 테스팅, 통합 테스팅, 시스템 테스팅 이렇게 3단계의 레벨별 테스팅을 수행한다.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1635-1643
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• 2022

STPA (System-Theoretic Process Analysis) is a widely used safety analysis technique to identify UCAs (Unsafe Control Actions) resulting in potential losses. It is totally dependent on the experience and ability of analysts to construct an information model called Control Structures, upon which analysts try to identify unsafe controls between system components. This paper proposes a formal approach to support the manual identification of UCAs, effectively and systematically. It allows analysts to mechanically extract Process Model, an important element that makes up the Control Structures, from a formal requirements specification for a software controller. It then concisely constructs the contents of Context Tables, from which analysts can identify all relevant UCAs effectively, using a software fault tree analysis technique. The case study with a preliminary version of a Korean nuclear reactor protections system shows the proposed approach's effectiveness and applicability.

• KIPS Transactions on Computer and Communication Systems
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• v.2 no.12
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• pp.561-568
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• 2013

Instrumentation and control system in nuclear power plant performs protecting, controling and monitoring safety operation of Nuclear Power Plant. As cyber attack to the control equipment of instrumentation and control system can cause reactor shutdown and radiation release, it is required to design the instrumentation and control system considering cyber security in accordance with regulatory guides and industrial standards. In this paper, we proposed a design method of uni-directional communication structure which is required in the design of defense-in-depth model according to regulatory guides and industrial standards and we implemented a communication board with the proposed method. This communication board was tested in various test environments and test items and we concluded it can provide uni-directional communication structure required to design of defense-in-depth model against cyber attack by analyzing the results. The proposed method and implemented communication board were applied in the design of SMART (system-integrated modular advanced reactor) I&C (instrumentation and control) systems.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.1
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• pp.1-5
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• 2010

Acoustic emission(AE) has emerged as a powerful nondestructive tool to detect or monitor preexisting defects and leaks in the vessel structures. A Bragg grating based acoustic emission sensor system is developed. Various type of fiber Bragg grating sensor including the variable length of sensing part was fabricated and prototype sensor system was tested by using PZT pulser and pencil lead break sources. Two types of sensor attachment were used. First, the fiber Bragg grating sensor was attached fully to the surface using bonding agent. Second one is that one part of fiber was attached to the surface partly by bonding and the other part of fiber will be act as a cantilever. That is, the resonant frequency of the fiber Bragg grating sensor will depend on the length of sensing part. The final goal of the sensor system is to provide on-line monitoring of cracks or leaks in reactor vessel head penetration of nuclear power plants.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.731-744
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• 2013

Energy security is a topic of high importance to many countries throughout the world. Countries with access to vast energy supplies enjoy all of the economic and political benefits that come with controlling a highly sought after commodity. Given the desire to diversify away from fossil fuels due to rising environmental and economic concerns, there are limited technology options available for baseload electricity generation. Further complicating this issue is the desire for energy sources to be sustainable and globally scalable in addition to being economic and environmentally benign. Nuclear energy in its current form meets many but not all of these attributes. In order to address these limitations, TerraPower, LLC has developed the Traveling Wave Reactor (TWR) which is a near-term deployable and truly sustainable energy solution that is globally scalable for the indefinite future. The fast neutron spectrum allows up to a ~30-fold gain in fuel utilization efficiency when compared to conventional light water reactors utilizing enriched fuel. When compared to other fast reactors, TWRs represent the lowest cost alternative to enjoy the energy security benefits of an advanced nuclear fuel cycle without the associated proliferation concerns of chemical reprocessing. On a country level, this represents a significant savings in the energy generation infrastructure for several reasons 1) no reprocessing plants need to be built, 2) a reduced number of enrichment plants need to be built, 3) reduced waste production results in a lower repository capacity requirement and reduced waste transportation costs and 4) less uranium ore needs to be mined or purchased since natural or depleted uranium can be used directly as fuel. With advanced technological development and added cost, TWRs are also capable of reusing both their own used fuel and used fuel from LWRs, thereby eliminating the need for enrichment in the longer term and reducing the overall societal waste burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P) reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.