• Journal of the Society of Naval Architects of Korea
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• v.57 no.4
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• pp.241-253
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• 2020

The Small-Unit Riverine Craft (SURC) is a small high-speed vessel used by navies and marine corps in relatively shallow waterway environments, such as riverine areas or littoral coasts. In the past, SURCs have primarily been rigid-hulled inflatable boats constructed using composite materials such as glass fiber reinforced plastics. More recently, single-hull SURCs have been manufactured using aluminum for weight reduction. In this study, a Carbon Fiber Reinforced Polymer (CFRP) material was applied instead to examine its feasibility in the basic design of an SURC with a hull length of 10 m. The CFRP structural design was obtained using the properties of a marine CFRP laminate, determined in a previous study. Next, the designed CFRP SURC was modeled to confirm its functionality, then compared with existing aluminum SURCs, indicating that the CFRP SURC was 41.49 % lighter, reduced fuel consumption by 30 %, and could sail 50 NM further for every hour of engine operation. A method for reducing the high cost of carbon fiber was also proposed based on the adjustment of the carbon fiber content to provide the optimum strength where required. The data developed in this study can be used as a basis for further design of CFRP craft.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.19-26
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• 1988

The low leakage leading pattern has features as the placement of some fresh fuel assemblies in the core interior to reduce the neutron fluence on the pressure vessel and to enhance the neutron economics. But as fresh fuel assemblies are loaded in the core interior, the local power tends to exceed safety limit due to the high reactivity of the fresh assemblies. Therefore, a large number of burnable poisons must be utilized in a low leakage scheme to suppress the high assembly power as well as the excess reactivity. In this study the effects of burnable poisons are treated as a perturbation on the power distribution, and the 'Power Sensitivity Coefficient' concept is adopted. An application study is performed for cycle 1 of the Korea Nuclear Unit-7 (KNU-7) to justify the usefulness of the reverse depletion method coupled with the above concept. To obtain the optimal burnable poision distribution at the given burnup step, the linear programming technique is adopted. The result shows maximum 4.5% error in the amount of burnable poisons between the calculated and the reference values. It is concluded that the design methodology which consists of the reverse depletion, the power sensitivity coefficient concept, and the linear programming technique can be used to find the optimal turnable poison distribution.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.735-744
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• 2012

In order to quantify the flow distribution characteristics of APR+ reactor, a test was performed on a test facility, ACOP ($\underline{A}$PR+ $\underline{C}$ore Flow & $\underline{P}$ressure Test Facility), having a length scale of 1/5 referring to the prototype plant. The major parameters are core inlet flow and outlet pressure distribution and sectional pressure drops along the major flow path inside reactor vessel. To preserve the flow characteristics of prototype plant, the test facility was designed based on a preservation of major flow path geometry. An Euler number is considered as primary dimensionless parameter, which is conserved with a 1/40.9 of Reynolds number scaling ratio. ACOP simplifies each fuel assembly into a hydraulic simulator having the same axial flow resistance and lateral cross flow characteristics. In order to supply boundary condition to estimate thermal margins of the reactor, the distribution of inlet core flow and core exit pressure were measured in each of 257 fuel assembly simulators. In total, 584 points of static pressure and differential pressures were measured with a limited number of differential pressure transmitters by developing a sequential operation system of valves. In the current study, reactor flow characteristics under the balanced four-cold leg flow conditions at each of the cold legs were quantified, which is a part of the test matrix composing the APR+ flow distribution test program. The final identification of the reactor flow distribution was obtained by ensemble averaging 15 independent test data. The details of the design of the test facility, experiment, and data analysis are included in the current paper.

• Journal of the Korea Convergence Society
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• v.8 no.10
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• pp.185-191
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• 2017

Recently, for such reasons as its inexpensive price and eco-friendliness, LNG has been under the limelight as an alternative fuel for vessels and is expected to grow rapidly in the industry. However, the technology level of domestic shipbuilders in manufacturing the cryogenic pump designed to supply LNG for vessels is so low that design and manufacturing technology of core parts are in urgent need. Therefore, this study describes the stepwise development procedure of cryogenic submerged centrifugal pump for ship LNG supply system. And it aims to suggest practical and specific development methods of the pump by approaching the characteristics of each step and major development items from the standpoint of engineering and management.

• International Journal of Naval Architecture and Ocean Engineering
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• v.9 no.2
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• pp.185-198
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• 2017

The application of composites as an alternative material for marine steel hatch covers is the subject of this study. Two separate approaches are considered; weight reduction approach and strengthening approach. For both approaches Finite Element Analysis (FEA) was performed using ANSYS software. Critical design parameters of the composite hatch cover and FEA are discussed in details. Regarding the weight reduction approach; steel hatch covers of a bulk carrier were replaced by composite covers and a weight reduction of 44.32% was achieved leading to many benefits including fuel saving, Deadweight Increment and lower center of gravity of the vessel. For the strengthening approach; the foremost hatch cover was strengthened to withstand 150% of the load required by IACS for safer navigation while no change in weight was made between the steel and composite covers. Results show that both approaches are feasible and advantageous.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3966-3978
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• 2021

ISFRA (Integrated SFR Analysis Program for PSA) computer program has been developed for simulating the response of the PGSFR pool design with metal fuel during a severe accident. This paper describes validation of the ISFRA aerosol model against the Aerosol Behavior Code Validation and Evaluation (ABCOVE) experiments undertaken in 1980s for radionuclide transport within a SFR containment. ABCOVE AB5, AB6, and AB7 tests are simulated using the ISFRA aerosol model and the results are compared against the measured data as well as with the simulation results of the MELCOR severe accident code. It is revealed that the ISFRA prediction of single-component aerosols inside a vessel (AB5) is in good agreement with the experimental data as well as with the results of the aerosol model in MELCOR. Moreover, the ISFRA aerosol model can predict the "washout" phenomenon due to the interaction between two aerosol species (AB6) and two-component aerosols without strong mutual interference (AB7). Based on the theory review of the aerosol correlation technique, it is concluded that the ISFRA aerosol model can provide fast, stable calculations with reasonable accuracy for most of the cases unless the aerosol size distribution is strongly deformed from log-normal distribution.

• Journal of the Korean Society of Marine Environment & Safety
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• v.25 no.5
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• pp.627-633
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• 2019

This study aimed to predict the resistance and propulsion performance of a ship using computational fluid dynamics (CFD) and a database as well as establish an assessment method for the energy efficiency design index (EEDI) using the results. First, the total resistance of the studied ship is obtained using CFD. A flow analysis is conducted with the free surface and trim and sinkage using a commercial CFD code (STAR-CCM+). The effective power of the ship is assessed based on the CFD results. The quasi-propulsive efficiency is calculated from an empirical prediction equation using experimental data and similar material. Finally, a general calculation program for the EEDI is established based on the hydrodynamic results, ship information for principal particulars, conversion factor of $CO_2$ for fuels, and fuel consumption.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2756-2766
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• 2024

Small rod-controlled pressurized water reactors (PWR) are the ideal energy source for vessel propulsion, benefiting from their high reactivity control efficiency. Since the control rods (CRs) increase the complexity of reactivity control, this paper seeks to study the performance of CRs in small rod-controlled PWRs to extend the lifetime and reduce power offset due to CRs. This study investigates CR grouping, move-in/out strategies, and axially non-uniform design effects on core neutron physics metrics. These metrics include axial offset (AO), core lifetime (CL), fuel utilization (FU), and radial power peaking factor (R-PPF). To simulate the movement of the CRs, a "Critical-CR-burnup" function was developed in OpenMC. In CR designs, the CRs are grouped into three banks to study the simultaneous and prioritized move-in/out strategies. The results show CL extension from 590 effective full power days (EFPDs) to 638-698 EFPDs. A lower-worth prioritized strategy minimizes AO and the extremum values decrease from -0.69 and + 0.81 to -0.28 and + 0.51. Although an axially non-uniform CR design can improve AO at the beginning of cycle (BOC), considering the overall CR worth change is crucial, as a significant decrease can adversely impact axial power distribution during the middle of cycle (MOC).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.91-100
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• 2016

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

• 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.