• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.66-66
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• 2002

Under Atomic Energy Society of Japan (AESJ) and Japan Atomic Energy Research Institute (JAERO, a Working Party on Reactor Physics of Accelerator-Driven System (ADS-WP) has been set since March 1999 to review and investigate special subjects related to reactor physics research of Accelerator-Driven System (ADS). In the ADSWP, the extensive and aggressive activity is being made by 25 professional members in the field of reactor physics in Japan. The ADS is now studying three subjects related to subcriticality of ADS; (1) calculation accuracy of sub criticality on ADS, (2) critical safety issues of ADS, and (3) theoretical review of subcriticality and its measurement methods. This paper describes two topics related to the subjects (1) and (2); one is an analysis of maximum reactivity potentially inserted to a subcritical core and the other is a benchmark proposal for checking calculation accuracy of sub criticality on ADS. The full specification of the calculation benchmark will be supplied by June 2002. Researchers from overseas, especially from Korea, are welcome to join this benchmark

• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.641-654
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• 2014

A new computer code, named CORONA (Core Reliable Optimization and thermo-fluid Network Analysis), was developed for the core thermo-fluid analysis of a prismatic gas cooled reactor. The CORONA code is targeted for whole-core thermo-fluid analysis of a prismatic gas cooled reactor, with fast computation and reasonable accuracy. In order to achieve this target, the development of CORONA focused on (1) an efficient numerical method, (2) efficient grid generation, and (3) parallel computation. The key idea for the efficient numerical method of CORONA is to solve a three-dimensional solid heat conduction equation combined with one-dimensional fluid flow network equations. The typical difficulties in generating computational grids for a whole core analysis were overcome by using a basic unit cell concept. A fast calculation was finally achieved by a block-wise parallel computation method. The objective of the present paper is to summarize the motivation and strategy, numerical approaches, verification and validation, parallel computation, and perspective of the CORONA code.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.65 no.11
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• pp.1848-1853
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• 2016

Control Rod Position Indicator in nuclear reactor vessel has developed for small reactor in Korea. Because of severe environment in reactor vessel, target of this study is to develop the suitable position indicator. In this study, solenoid type position indicator made of Mineral Insulated Cable(MI Cable) was introduced to adapt in severe environment. And inductance of the solenoid was used to indicate the rod position for high precision. But problem of this concept is that a linear slope of inductance is changed by temperature effect. To resolve this problem, two sensing coils were introduced for temperature compensation. A role of the sensing coil is to make reference linear equation about certain temperature. To confirm this concept, also, inductance of solenoid and sensing coils were measured at room and high temperature (${\sim}300^{\circ}C$). The results of measurement show that the position error of sensing coil between room and high temperature was about 2%. But it was identified that this error was resulted from insufficient test environment (temperature error between solenoid and sensing coils was about 2% at high temperature condition). Therefore, solenoid type position indicator shows that it is very suitable in reactor vessel as a high precision rod position indicator.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.289-290
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• 2017

• 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.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.34 no.4
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• pp.365-373
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• 2010

A chemical reactor model (CRM) was developed for a jet stirred reactor (JSR) to predict the emission of exhaust such as NOx. In this study, a two-PSR model was chosen as the chemical reactor model for the JSR. The predictions of NO formation in lean premixed methane-air combustion in the JSR were carried out by using CHEMKIN and GRI 3.0 methane-air combustion mechanism which include the four NO formation mechanisms. The calculated results were compared with Rutar's experimental data for the validation of the model. The effects of important parameters on NO formation and the contributions of the four NO pathways were investigated. In the flame region, the major pathway is the prompt mechanism, and in the post flame region, the major pathway is the Zelodovich mechanism. Under the lean premixed condition, the N2O mechanism is the important pathway in both flame and postflame regions.

• Transactions of the Korean hydrogen and new energy society
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• v.26 no.5
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• pp.445-452
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• 2015

A tubular packed bed reactor for the steam-$CO_2$ combined reforming of natural gas to produce the synthesis gas of a target $H_2/CO$ ratio 2.0 was simulated. The effects of the reactor dimension, the feed gas composition, and the gas feeding temperature upon the possibility of coke formation across the catalyst bed were investigated. For this purpose, 2-dimensional heterogeneous reactor model was used to determine the local gas concentrations and temperatures over the catalyst bed. The thermodynamic potential distribution of coke formation was determined by comparing the extent of reaction with the equilibrium constant given by the reaction, $CH_4+2CO{\Leftrightarrow}3C+2H_2O$. The simulation showed that catalysts packed in the central region nearer the entrance of the reactor were more prone to coking because of the regional characteristics of lower temperature, lower concentration of $H_2O$, and higher concentration of CO. With the higher feeding temperature, the feed gas composition of the increased $H_2O$ and correspondingly decreased $CO_2$, or the decrease in the reactor diameter, the volume fraction of the catalyst bed subsequent to coking could be diminished. Throughout the simulation, reactor dimension and reaction condition for coking-free operation were suggested.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.423-430
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• 2014

An in-pile performance test of HANA claddings was conducted at up to 67 GWD/MTU in the Halden research reactor in Norway over a 6.5 year period. Four types of HANA claddings (HANA-3, HANA-4, HANA-5, and HANA-6) and a reference Zircaloy-4 cladding were used for the in-pile test. The evaluation parameters of the HANA claddings were the corrosion behavior, dimensional changes, hydrogen uptake, and tensile strength after the claddings were tested under the simulated operation conditions of a Korean commercial reactor. The oxide thickness ranged from 15 to 37 mm at a high flux region in the test rods, and all HANA claddings showed corrosion resistance superior to the Zircaloy-4 cladding. The creep-down rate of all HANA claddings was lower than that of the Zircaloy-4 cladding. In addition, the hydrogen content of the HANA claddings ranged from 54 to 96 wppm at the high heat flux region of the test rods, whereas the hydrogen content of the Zircaloy-4 cladding was 119 wppm. The tensile strength of the HANA and Zircaloy-4 claddings was similarly increased when compared to the un-irradiated claddings owing to the radiation-induced hardening.

• Journal of computational fluids engineering
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• v.21 no.1
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• pp.19-29
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• 2016

A Prototype Gen-IV Sodium-cooled Fast Reactor which is one of the $4^{th}$ generation nuclear reactors is in development by Korea Atomic Energy Research Institute. The reactor is composed of four main fluid systems which are categorized by its functions, i.e., Primary Heat Transport System, Intermediate Heat Transport System, Decay Heat Removal System and Sodium-Water Reaction Pressure Relief System. The coolant of the reactor is liquid sodium and sodium-to-sodium heat exchangers are installed at the interfaces between two fluid systems, Intermediate Heat Exchangers between the Primary Heat Transport System and the Intermediate Heat Transport System and Decay Heat Exchangers between the Primary Heat Transport System and the Decay Heat Removal System. For the design and performance analysis of the Intermediate Heat Exchanger and the Decay Heat Exchanger, a computer code was written during previous step of research. In this work, the computer code named "SHXSA" has been validated preliminarily by computational fluid dynamics simulations.