• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2002-2010
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• 2024

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

• Korean Journal of Optics and Photonics
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• v.13 no.4
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• pp.301-307
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• 2002

The propagation characteristic of an optical waveguide was investigated by measuring with a near-field scanning optical microscope (NSOM) the evanescent field formed at the neighbor of its core-cladding interface. For this purpose, the NSOM system was developed specially as a form of Photon scanning tunneling microscope. The evanescent field distributions of several channel waveguides were measured at the wavelength of 1550 ㎚, and the usefulness of the system was verified by comparing experimental results with simulation results. In particular, the interference phenomena of the guided modes during their propagation along a multimode channel waveguide could be observed directly from the measured evanescent field distribution.

• Journal of computational fluids engineering
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• v.10 no.3 s.30
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• pp.9-18
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• 2005

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by the active reaction of fuel-cladding and steam in the reactor pressure vessel and released with steam into the containment. In order to mitigate hydrogen hazards possibly occurred in the NPP containment, hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) designed in Korea specifies 26 passive autocatalytic recombiners and 10 igniters installed in the containment for the hydrogen mitigation. in this study, the analysis of the hydrogen and steam behavior during a total lose of feed water (TLOFW) accident in the APR1400 containment has been conducted by using the CFD code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released in the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type dampers at the IRWST vents which are operated depending on the pressure difference between inside and outside of the IRWST. it was found that the flaps strongly affects the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and transition from deflagration to detonation (DDT) were evaluated by using Sigma-Lambda criteria. Numerical results indicate the DDT possibility could be heavily reduced in the IRWST compartment when the flaps are installed.

• Nuclear Engineering and Technology
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• v.24 no.1
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• pp.52-62
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• 1992

Radioactive nuclides deposited on out-of-core surface after the radiation in the core by the transport of corrosion products (CRUD) through the primary coolant system in PWR which is the major plant type in Korea, are leading sources of radiation exposure to plant maintenance personnel. Thus, the optimal chemistry operation method is required for the reduction of radiation exposure by the corrosion products. This study analysed the actual water chemistry operation data of four operating domestic PWRs. And in order to evaluate the coolant chemistry operation data, a computer code which can calculate the activity buildup in the various chemistry conditions of PWR coolant was employed. Through the analysis of comparison between the activity buildup of actual water chemistry operation mode and that of assumed Elevated Li operation mode calculated by the computer code, it was found that the out-of core radioactivity can be reduced by diminishing the deposition of corrosion products on the core in case that the Elevated Li operation mode is applied to the coolant chemistry operation of PWR. And the higher coolant pH operation was shown to have the advantage of the reduction of out-of-core activity buildup if the integrity of system structural materials and fuel cladding is guaranteed.

• Composites Research
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• v.25 no.6
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• pp.211-216
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• 2012

FBG sensor peaks could be split due to polarization by shear strain, when the fiber optic sensors embedded or attached to the structure. For the fiber optic sensor packages, sensor grating has to be protected from shear strains. Also, pretension has to be applied to the sensor because compressive strain must be measured. Without pretension of sensor, the sensor does not show any change of signal until it is stretched. In order to mesure compressive and tensile strains, two fixing point and prestressed sensor need. In the fixing point, just holding the optical fiber cause slip between core and cladding in the fiber. A Fixation method of prestressed FBG sensors fixed with partially stripped fibers was developed. The sensor package has the prestress controllable fixtures at the fixing points. Prestress to the sensor imposed by controlling the two fixed points with bolts and nuts make it easy to measure compressive strain as well as tensile strain. The fiber optic sensor packages applied to the actual structure and the structural monitoring system using the package can be applied to safety through surveillance.

• Wind and Structures
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• v.27 no.3
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• pp.163-173
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• 2018

The design wind pressure for low-rise buildings in the ASCE 7-10 is defined by procedures that are categorized into the Main Wind Force-Resisting System (MWFRS) and the Components and Cladding (C&C). Some of these procedures were originally developed based on steel portal frames of industrial buildings, while the residential structures are a completely different structural system, most of which are designed as low-rise light-frame wood constructions. The purpose of this study is to discuss the rationality (or irrationality) of the extension of the wind loads calculated by the ASCE 7-10 to the light-frame wood residential buildings that represent the most vulnerable structures under extreme wind conditions. To serve this purpose, the same approach as used in the development of Chapter 28 of the ASCE 7-10 that envelops peak responses is adopted in the present study. Database-assisted design (DAD) methodology is used by applying the dynamic wind loads from Louisiana State University (LSU) database on a typical residential building model to assess the applicability of the standard by comparing the induced responses. Rather than the postulated critical member demands on the industrial building such as the bending moments at the knee, the maximum values at the critical points for wood frame buildings under wind loads are used as indicators for the comparison. Then, the critical members are identified through these indicators in terms of the displacement or the uplift force at connections and roof envelope. As a result, some situations for each of the ASCE 7 procedures yielding unconservative wind loads on the typical low-rise residential building are identified.

• Nuclear Engineering and Technology
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• v.41 no.10
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• pp.1263-1274
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• 2009

Several experimental tests to simulate a reflood phase of a cold-leg LBLOCA of the APR1400 have been performed using the ATLAS facility. This paper describes the related experimental results with respect to the thermal-hydraulic behavior in the core and the system-core interactions during the reflood phase of the cold-leg LBLOCA conditions. The present descriptions will be focused on the LB-CL-09, LB-CL-11, LB-CL-14, and LB-CL-15 tests performed using the ATLAS. The LB-CL-09 is an integral effect test with conservative boundary condition; the LB-CL-11 and -14 are integral effect tests with realistic boundary conditions, and the LB-CL-15 is a separated effect test. The objectives of these tests are to investigate the thermal-hydraulic behavior during an entire reflood phase and to provide reliable experimental data for validating the LBLOCA analysis methodology for the APR1400. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results for the LBLOCA scenario of the APR1400. The ECC water flow rate from the safety injection tanks and the decay heat were simulated from the start of the reflood phase. The simulated core power was controlled to be 1.2 times that of the ANS-73 decay heat curve for LB-CL-09 and 1.02 times that of the ANS-79 decay curve for LB-CL-11, -14, and -15. The simulated ECC water flow rate from the high pressure safety injection pump was 0.32 kg/s. The present experimental data showed that the cladding temperature behavior is closely related to the collapsed water level in the core and the downcomer.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.169-182
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• 2014

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

• Membrane Journal
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• v.7 no.4
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• pp.167-174
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• 1997

The concentration of silica is required to meet a certain level because silica affects fuel and materials integrity by forming a zeolite layer on fuel cladding surfaces. When the established Feed and Bleed method is employed, nuclear waste increase and the corresponding amount of boric acid is constantly consumed. This study concentrates on minimizing the amount of nuclear waste and consumption of boric acid. Using five different membranes, operating conditions such as temperatur, feed water flow rate, boric acid recovery and silica removal rate were examined. A silica-selective removal system was designed based on the above optimization procedures. Three-stage system was designed with two characteristically different membranes so that it could correspond with the different situation easily. Compared to the pevious results of the Feed and Bleed method, the current method showed that the amount of nuclear waste was reduced to 7%, and the consumption of boric acid to 15.7%.

• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.97-109
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• 2020

On March 11, 2011, an earthquake followed by a tsunami caused an extended station blackout (SBO) at the Fukushima Dai-ichi NPP Units. The accident was initiated by a total loss of both onsite and offsite electrical power resulting in the loss of the ultimate heat sink for several days, and a consequent core melt in some units where proper mitigation strategies could not be implemented in a timely fashion. To enhance the plant's coping capability, the Diverse and Flexible Strategies (FLEX) were proposed to append the Emergency Operation Procedures (EOPs) by relying on portable equipment as an additional line of defense. To assess the success window of FLEX strategies, all sources of uncertainties need to be considered, using a physics-based model or system code. This necessitates conducting a large number of simulations to reflect all potential variations in initial, boundary, and design conditions as well as thermophysical properties, empirical models, and scenario uncertainties. Alternatively, data-driven models may provide a fast tool to predict the success window of FLEX strategies given the underlying uncertainties. This paper explores the applicability of Artificial Intelligence (AI) to identify the success window of FLEX strategy for extended SBO. The developed model can be trained and validated using data produced by the lumped parameter thermal-hydraulic code, MARS-KS, as best estimate system code loosely coupled with Dakota for uncertainty quantification. A Systems Engineering (SE) approach is used to plan and manage the process of using AI to predict the success window of FLEX strategies under extended SBO conditions.