• Journal of Korea Society of Industrial Information Systems
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• v.23 no.1
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• pp.31-41
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• 2018

This paper discusses a structure design and thermal analysis of cryogenic conduction cooling system for a high current HTS DC reactor. Dimensions of the conduction cooling system parts including HTS magnets, bobbin structures, current leads, support bars, and thermal exchangers were calculated and drawn using a 3D CAD program. A finite element method model was built for determining the optimal design parameters and analyzing the thermo-mechanical characteristics. The operating current and inductance of the reactor magnet were 1,500 A, 400 mH, respectively. The thermal load of the HTS DC reactor was analyzed for determining the cooling capacity of the cryo-cooler. Hence, we carried out the operating test of conduction cooling system of the 1st stage area with high current flow. The cooper bars was cooled down to 40 K and HTS leads operated stably. As a experiment result, the total heat load of the 1st stage area is 190 W. The study results can be effectively utilized for the design and fabrication of a commercial HTS DC reactor.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1037-1042
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• 2018

The present study aims to assess the excess induced reactivity in a Miniature Neutron Source Reactor (MNSR) for a Beyond Design Basis Accident (BDBA) scenario. The BDBA scenario as defined in the Safety Analysis Report (SAR) of the reactor involves sticking of the control rod and filling of the inner and outer irradiation sites with water. At the end of the MNSR core life, 10.95 cm of Beryllium is added to the top of the core as a reflector which affects some neutronic parameters such as effective delayed neutrons fraction (${\beta}_{eff}$), the reactivity worth of inner and outer irradiation sites that are filled with water and the reactivity worth of the control rod. Given those influences and changes, new neutronic calculations are required to be able to demonstrate the reactor safety. Therefore, a validated MCNPX model is used to calculate all neutronic parameters at the end of the reactor core life. The calculations show that the induced reactivity in the BDBA scenario increases at the end of core life to $7.90{\pm}0.01mk$ which is significantly higher than the induced reactivity of 6.80 mk given in the SAR of MNSR for the same scenario but at the beginning of the core's life. Also this value is 3.90 mk higher than the maximum allowable operational limit (i.e. 4.00 mk).

• Korean Chemical Engineering Research
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• v.59 no.3
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• pp.467-479
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• 2021

Abstract: A kinetic model was developed for the dehydration of ethanol to ethylene based on two parallel reaction pathways. Kinetic parameters were estimated by fitting experimental data of powder catalysts in a lab-scale test, and the effectiveness factor was determined using data from pellet-type catalysts in bench-scale experiments. The developed model was used to design a multitubular fixed-bed reactor (MTR) and an adiabatic reactor (AR) at a 10 ton per day scale. The two different reactor types resulted in different process configurations: the MTR consumed the ethanol completely and did not produce the reaction intermediate, diethyl ether (DEE), resulting in simple separation trains at the expense of high equipment cost for the reactor, whereas the AR required azeotropic distillation and cryogenic distillation to recycle the unreacted ethanol and to separate the undesired DEE, respectively. Quantitative analysis based on the equipment and annual energy costs showed that, despite high equipment cost of the reactor, the MTR process had the advantages of high productivity and simple separation trains, whereas the use of additional separation trains in the AR process increased both the total equipment cost and the annual energy cost per unit production rate.

• Structural Engineering and Mechanics
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• v.76 no.3
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• pp.367-378
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• 2020

This paper presents a new design concept for ocean nuclear power plants (ONPPs) using a tension leg platform (TLP). The system-integrated modular advanced reactor, which is one of the successful small modular reactors, is mounted for demonstration. The authors define the design requirements and parameters, modularize and rearrange the nuclear and other facilities, and propose a new total general arrangement. The most fundamental level of design results for the platform and tendon system are provided, and the construction procedure and safety features are discussed. The integrated passive safety system developed for the gravity based structure-type ONPP is also available in the TLP-type ONPP with minor modifications. The safety system fully utilizes the benefits of the ocean environment, and enhances the safety features of the proposed concept. For the verification of the design concept, hydrodynamic analyses are performed using the commercial software ANSYS AQWA with the Pierson-Moskowitz and JONSWAP wave spectra that represent various ocean environments and the results are discussed.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.293-298
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• 1997

The H$_{\infty}$robust controller fur the reactor power control system is designed by use of the mixed weight sensitivity. The system is configured into the typical two-port model with which the weight functions are augmented. Since the solution depends on the weighting functions and the problem is of non-convex, the genetic algorithm is used to determine the weighting functions. The cost function applied in the genetic algorithm permits the direct control of the power tracking performances. In addition, the actual operating constraints such as rod velocity and acceleration can be treated as design parameters. Compared with the conventional approach, the controller designed by the genetic algorithm results in the better performances with the realistic constraints. Also, it is found that the genetic algorithm could be used as an effective tool in the robust design. robust design.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1776-1783
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• 2019

To accurately analyze the accidents in nuclear reactors, a thermohydraulic-neutronic coupling calculation is required to solve fluid dynamics and nuclear reactor kinetics equations in fine cells simultaneously and evaluate the local effects of neutronic and thermohydraulic parameters on each other. In the present study, a 3D thermohydraulic-neutronic coupling model is developed, validated and then applied for Isfahan MNSR (Miniature Neutron Source reactor) safety analysis. The proposed model is developed using FLUENT software and user defined functions (UDF) are applied to simulate the neutronic behavior of MNSR. The validation of the proposed model is first evaluated using 1mk reactivity insertion experiment into Isfahan MNSR core. Then, the developed coupling code is applied for a design basis accident (DBA) scenario analysis with the insertion of maximum allowed cold core reactivity of 4 mk. The results show that the proposed model is able to predict the behavior of the reactor core under normal and accident conditions with a good accuracy.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.9 s.252
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• pp.850-857
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• 2006

Hydrogen production using current fueling facilities is essential for near-term applications of fuel cells. A preliminary process for developing a natural gas autothermal reforming (ATR) reactor for fuel cells is presented in this paper. A experimental reactor for methane ATR was constructed and used for characterization of Jin reactor. Temperature profiles of the reactor were observed, and reformed gas compositions were analyzed to evaluate efficiency, conversion and reaction heat with varying amounts of $O_2/CH_4$ at selected furnace temperature and $H_2O/CH_4$. The amount of $O_2/CH_4$ showed strong offsets on reactor temperature, efficiency and conversion indicating that $O_2/CH_4$ is a crucial operation condition. Operation conditions which result in thermal neutrality of ATR reactor system were determined for two cases of an ATR system based on the estimation of enthalpy difference between reactants of assumed inlet temperatures and the products from experimental results. The determined conditions for thermally neutral operations could be used for guidelines to design reformers and for determining the operation parameters of a self sustaining ATR reactor.

• Nuclear Engineering and Technology
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• v.42 no.6
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• pp.690-703
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• 2010

Sampling based uncertainty analysis was carried out to quantify uncertainty in predictions of best estimate code RELAP5/MOD3.2 for a thermal hydraulic test (10% hot leg break LOCA) performed in the Large Scale Test Facility (LSTF) as a part of an IAEA coordinated research project. The nodalisation of the test facility was qualified for both steady state and transient level by systematically applying the procedures led by uncertainty methodology based on accuracy extrapolation (UMAE); uncertainty analysis was carried out using the Latin hypercube sampling (LHS) method to evaluate uncertainty for ten input parameters. Sixteen output parameters were selected for uncertainty evaluation and uncertainty band between $5^{th}$ and $95^{th}$ percentile of the output parameters were evaluated. It was observed that the uncertainty band for the primary pressure during two phase blowdown is larger than that of the remaining period. Similarly, a larger uncertainty band is observed relating to accumulator injection flow during reflood phase. Importance analysis was also carried out and standard rank regression coefficients were computed to quantify the effect of each individual input parameter on output parameters. It was observed that the break discharge coefficient is the most important uncertain parameter relating to the prediction of all the primary side parameters and that the steam generator (SG) relief pressure setting is the most important parameter in predicting the SG secondary pressure.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.380-385
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• 1996

The genetic algorithm(GA) is applied to the design of the nuclear power control system. The reactor control system model is described in the LQR configuration. The LQR system order is increased to make the tracking system. The key parameters of the design are weighting matrices, and these are usually determined through numerous simulations in the conventional design. To determine the more objective and optimal weightings, the improved GA is applied. The results show that the weightings determined by the GA yield the better system responses than those obtained by tile conventional design method.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.235-241
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• 1995

YGN 3 and 4 are the nuclear power plants having System 80 characteristics with a rated thermal output of 2815 MWth and a nominal net electrical output of 1040 MWe. YGN 3 achieved commercial operation on March 31, 1995 and YGN 4 completed Power Ascension Test (PAT) at 20%, 50%, 80% and 100% power by September 23, 1995. YGN 3 and 4 design incorporates the Reactor Power Cutback System (RPCS) which reduces plant trips caused by Loss of Load (LOL)/Turbine Trip and Loss of One Main Feedwater Pump (LOMFWP). The key design objective of the RPCS is to improve overall plant availability and performance, while minimizing challenges to the plant safety systems. The RPCS is designed to rapidly reduce reactor power by dropping preselected Control Element Assemblies (CEAs) while other NSSS control systems maintain process parameters within acceptable ranges. Extensive RPCS related tests performed during the initial startup of YGN 4 demonstrated that the RPCS can maintain the reactor on-line without opening primary or secondary safety valves and without actuating the Engineered Safety Features Actuation System (ESFAS). It is expected that use of the RPCS at YGN will increase the overall availability of the units and reduce the number of challenges to plant safety systems.