• Nuclear Engineering and Technology
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• v.35 no.2
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• pp.91-107
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• 2003

Since DBA(Design Basis Accidents) has been studied rather separately from SA(Severe Accidents) in the conventional nuclear reactor safety analysis, the thermal hydraulics during transition between DBA and SA has not been identified so much as each accident itself. Thus, in this study, the thermal hydraulic behavior from DBA to the commencement of SA has been experimentally and analytically investigated for the long-term cooling phase of LB-LOCA(Large-Break Loss-of-Coolant Accident). Experiments were conducted for both cases of the loop seal open and closed in an integral test loop, named as SNUF (Seoul National University Facility), which was scaled down to l/6.4 in length and 1/178 in area of the APR1400 (Advanced Power Reactor 1400MWe). The core mixture level was a main measured value since it took major role in the fuel heat-up rate, the location of fuel melting initiation and the channel blockage by melting material during SA. Experimental results were compared to MAAP4.03 to assess its model of calculating the core mixture level. MAAP4.03 overestimates the core two- phase mixture level because sweep-out and spill-over and the measures to simulate the status of loop seal are not included, which is against the conservatism. Thus, it is recommended that MAAP4.03 should be improved to simulate the thermal hydraulic phenomena, such as sweep-out, spill-over and the status of loop seal.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.327-336
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• 2007

The influence of density differences on the mixing of the primary loop inventory and the Emergency Core Cooling (ECC) water in the downcomer of a Pressurised Water Reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. ROCOM is a 1:5 scaled model of a German PWR, and has been designed for coolant mixing studies. It is equipped with advanced instrumentation, which delivers high-resolution information for temperature or boron concentration fields. This paper presents a ROCOM experiment in which water with higher density was injected into a cold leg of the reactor model. Wire-mesh sensors measuring the tracer concentration were installed in the cold leg and upper and lower part of the downcomer. The experiment was run with 5% of the design flow rate in one loop and 10% density difference between the ECC and loop water especially for the validation of the Computational Fluid Dynamics (CFD) software ANSYS CFX. A mesh with two million control volumes was used for the calculations. The effects of turbulence on the mean flow were modelled with a Reynolds stress turbulence model. The results of the experiment and of the numerical calculations show that mixing is dominated by buoyancy effects: At higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this circumferential propagation. Therefore, density effects play an important role during natural convection with ECC injection in PWRs. ANSYS CFX was able to predict the observed flow patterns and mixing phenomena quite well.

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.489-504
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• 2013

Software safety for nuclear reactor protection systems (RPSs) is the most important requirement for the obtainment of permission for operation and export from government authorities, which is why it should be managed with well-experienced software development processes. The RPS software is typically modeled with function block diagrams (FBDs) in the design phase, and then mechanically translated into C programs in the implementation phase, which is finally compiled into executable machine codes and loaded on RPS hardware - PLC (Programmable Logic Controller). Whereas C Compilers are fully-verified COTS (Commercial Off-The-Shelf) software, translators from FBDs to C programs are provided by PLC vendors. Long-term experience, experiments and simulations have validated their correctness and function safety. This paper proposes a behavior-preserving translation from FBD design to C implementation for RPS software. It includes two sets of translation algorithms and rules as well as a prototype translator. We used an example of RPS software in a Korean nuclear power plant to demonstrate the correctness and effectiveness of the proposed translation.

• Nuclear Engineering and Technology
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• v.49 no.3
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• pp.517-524
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• 2017

Cyber security is an important issue in the field of nuclear engineering because nuclear facilities use digital equipment and digital systems that can lead to serious hazards in the event of an accident. Regulatory agencies worldwide have announced guidelines for cyber security related to nuclear issues, including U.S. NRC Regulatory Guide 5.71. It is important to evaluate cyber security risk in accordance with these regulatory guides. In this study, we propose a cyber security risk evaluation model for nuclear instrumentation and control systems using a Bayesian network and event trees. As it is difficult to perform penetration tests on the systems, the evaluation model can inform research on cyber threats to cyber security systems for nuclear facilities through the use of prior and posterior information and backpropagation calculations. Furthermore, we suggest a methodology for the application of analytical results from the Bayesian network model to an event tree model, which is a probabilistic safety assessment method. The proposed method will provide insight into safety and cyber security risks.

• Korean Journal of Hazardous Materials
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• v.6 no.2
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• pp.95-104
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• 2018

Gas-liquid bubble column reactors are extensively used in industrial processes. A detailed knowledge of bubble size distribution is needed for determining the mass transfer in gas-liquid film. Experimental data on bubble size distribution and liquid-side mass transfer coefficient($k_L$) were used to calculate the estimated time to saturation in bubble column reactor. Also, the gas flux was evaluated to the liquid-side mass transfer coefficient($k_L$) and solubility data for hydrogen sulfide($H_2S$) and chlorine($Cl_2$) absorption into water. Simulation results show that $H_2S$ absorption time to 50 % of saturation concentrations are 611 sec and 1,329 sec when bubble diameters are 0.5 mm and 4.5 mm, while absorbing 1 % $H_2S$ gas. In case of $Cl_2$, absorption time range 657 to 1,400 sec when bubble size range 0.5 mm to 4.5 mm, while absorbing 1 % $Cl_2$ gas. Calculated simulation results can be used in the design of emergency relief bubble reactors.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2454-2463
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• 2021

The objective of this research is to the use of americium (AmO₂) as a burnable absorber effectively instead of conventional gadolinium (Gd₂O₃) for VVER-1200 reactor by analyzing its impacts on reactivity, power peaking factor (PPF), safety factor, and quality of the spent fuel. The assembly is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library for finding the optimum amount and effective way of using AmO₂ as a burnable absorber. From these studies, it is found that AmO₂ can decrease the excess reactivity like Gd₂O₃ without changing the criticality life span and enrichment of ²³⁵U. A homogeneous mixture of the 0.20% AmO₂+ 4.95% enriched UO₂ fuel rod (model MF-4) decreases the PPF than the reference assembly. The use of AmO₂+UO₂ in the integral burnable absorber (IBA) rod or the outer layer could also decrease the PPF up to 10 GWd/t but increases rapidly after 30 GWd/t, which could be a safety threat. The fuel temperature coefficient and void coefficient of the model MF-4 are the same as the reference assembly. In addition, 22% of initially loaded Am are burning effectively and contributing to the power production.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1279-1288
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• 2019

The vortex-type fluidic diode (FD) is a key safety component for inherent safety in various advanced reactors such as the sodium fast reactor (SFR) and the molten salt reactor (MSR). In this study, topology optimization is conducted to optimize the design of the vortex-type fluidic diode. The optimization domain is simplified to 2-dimensional geometry for a tangential port and chamber. As a result, a design with a circular chamber and a restrictor at the tangential port is obtained. To verify the new design, experimental study and computational fluid dynamics (CFD) analysis were conducted for inlet Reynolds numbers between 2000 and 6000. However, the results show that the performance of the new design is no better than the original reference design. To analyze the cause of this result, detailed analysis is performed on the velocity and pressure field using flow visualization experiments and 3-D CFD analysis. The results show that the discrepancy between the optimization results in 2-D and the experimental results in 3-D originated from exclusion of an important pressure loss contributor in the optimization process. This study also concludes that the junction design of the axial port and chamber offers potential for improvement of fluidic diode performance.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1765-1775
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• 2019

Multiple spurious operations (MSOs) mean multiple fire induced circuit faults causing an undesired operation of one or more systems or components. The Nuclear Energy Institute (NEI) of the United States published NEI 00-01 as guidelines for solving MSOs. And this guideline includes MSO scenarios of pressurized water reactor (PWR) and boiling water reactor (BWR). Nuclear power plant operators in U.S. analyzed MSOs under MSO scenarios included in NEI 00-01 and operators of PWRs in Korea also analyzed MSOs under the scenarios of NEI 00-01. As there are no pressurized heavy water reactors (PHWRs) in the United States, MSO scenarios of PHWRs are not included in the NEI 00-01 and any feasible scenarios have not been developed. This paper developed MSO scenarios which can be applied to PHWRs by reviewing the 63 MSO scenarios included in NEI 00-01. This study found that seven scenarios out of the 63 MSO scenarios can be applied and three more scenarios need to be developed.

• Earthquakes and Structures
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• v.16 no.1
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• pp.41-54
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• 2019

Reinforced Concrete (RC) shear walls are widely used in Nuclear power plants as effective lateral force resisting elements of the structure and these may experience nonlinear behavior for higher earthquake demand. Short shear walls of aspect ratio less than 1.5 generally experience combined shear flexure interaction. This paper presents the results of the displacement-controlled experiments performed on six RC short shear walls with varying aspect ratios (1, 1.25 and 1.5) for monotonic and reversed quasi-static cyclic loading. Simulation of the shear walls is then carried out by Finite element modeling and also by macro modeling considering the coupled shear and flexure behaviour. The shear response is estimated by softened truss theory using the concrete model given by Vecchio and Collins (1994) with a modification in softening part of the model and flexure response is estimated using moment curvature relationship. The accuracy of modeling is validated by comparing the simulated response with experimental one. Moreover, based on the experimental work a multi-linear hysteretic model is proposed for short shear walls. Finally ultimate load, drift, ductility, stiffness reduction and failure pattern of the shear walls are studied in details and hysteretic energy dissipation along with damage index are evaluated.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.501-509
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• 2019

In this paper a body-centered cubic(BCC) crystal plasticity model based on microscopic dislocation mechanism is introduced and numerically implemented. The model is coupled with irradiation effect via tracking dislocation loop evolution on each slip system. On the basis of the model, uniaxial tensile tests of unirradiated and irradiated RPV steel(take Chinese A508-3 as an example) at different temperatures are simulated, and the simulation results agree well with the experimental results. Furthermore, crystal plasticity damage is introduced into the model. Then the damage behavior before and after irradiation is studied using the model. The results indicate that the model is an effective tool to study the effect of irradiation and temperature on the mechanical properties and damage behavior.