• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.15-19
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• 2013

The auxiliary feedwater is an important to remove the heat from the reactor core when the main feedwater system is unavailable. In most initiating events in Probabilistic Safety Assessment(PSA), the operaton of this system is required to mitigate the accidents. For one of domestic nuclear power plants, a design change of a turbine-driven auxiliary feedwater pump(TD-AFWP), pipe, and valves in the auxiliary system is implemented due to the aging related deterioration by long time operation. This change includes the replacement of the TD-AFWP, the relocation of some valves for improving the system availability, a new cross-tie line, and the installation of manual valves for maintenance. The design modification affects the PSA because the system is critical to mitigate the accidents. In this paper, the safety effect of the change of the auxiliary feedwater system is assessed with regard to the PSA view point. The results demonstrate that this change can supply the auxiliary feedwater from the TD-AFWP in the accident with the motor-driven auxiliary feedwater pump(MD-AFWP) unavailable due to test or maintenance. In addition, the change of MOV's normal position from "close" to "open" can deliver the water to steam generator in the loss of offsite power(LOOP) event. Therefore, it is confirmed that the design change of the auxiliary feedwater system reduces the total core damage frequency(CDF).

• Journal of the Korean Magnetics Society
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• v.26 no.5
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• pp.154-158
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• 2016

A CRDM (Control Rod Drive Mechanism) is an electromagnetic device which drives a control rod assembly linearly to regulate the reactivity of a nuclear core. An RPIS (Rod Position Indication System) is used as a position indicator for a control rod assembly of a CRDM of SMART, and an RPIS consists of permanent magnets and reed switches. SMART is designed for the maximum coolant temperature of $350^{\circ}C$, and the permanent magnets are installed inside of the reactor. The reed switches and electrical circuit are installed outside of the reactor on the other hand. Test coil for a reed switch is test equipment for quality verification of a reed switch, and a test coil consists of a coil and core. In this study, magnetic property of test coil and permanent magnet on a reed switch is compared by using finite element electromagnetic simulation.

• The KSFM Journal of Fluid Machinery
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• v.17 no.6
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• pp.104-108
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• 2014

The recent Fukushima nuclear power plant accidents shows that the core make up at high RCS pressure condition is very important to prevent core melting. The core make up flow at high pressure condition should be driven by gravity force or passive forces because the AC-powered safety features are not available during a Station Black Out (SBO) accident. The reactor Coolant System (RCS) mass inventory is continuously decreased by releasing steam through the pressurizer safety valves after reactor trip during a SBO accident. The core will be melted down within 2~3 hours without core make up action by active or passive mode. In the new design concept of a Hybrid Safety Injection Tank (Hybrid SIT) both for low and high RCS pressure conditions, the low pressure nitrogen gas serves as a charging pressure for a LBLOCA injection mode, while the PZR high pressure steam provides an equalizing pressure for a high pressure injection mode such as a SBO accident. After the pressure equalizing process by battery driven initiation valve at a high pressure SBO condition, the Hybrid SIT injection water will be passively injected into the reactor downcomer by gravity head. The SBO simulation by MARS code show that the core makeup injection flow through the Hybrid SIT continued up to the SIT empty condition, and the core heatup is delayed as much.

• Journal of Energy Engineering
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• v.16 no.3
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• pp.103-112
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• 2007

Flood issue for nuclear power plants designed and built in 1970 is extremely severe for main steam header compartment and main feedwater line region of intermediate building and lower floor. A calculation for flood level at the main feedwater line isolation compartment is now performing by hand calculation. But, this methodology is quite conservative assumption. The goal of this study was to develop method to analyze flowrate using the RETRAN-3D computer code, and the developed method was applied to flood level analysis following main feedwater line break. As a result of analysis, flood level was low remarkably.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.5
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• pp.469-474
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• 2014

The high temperature gas-cooled reactor (HTR-10) is designed to produce electricity and hydrogen. Graphite is used as reflector, support structures, and a moderator in reactor core; it has good resistance to neutron and is a suitable material at high temperatures. Friction is generated in the graphite structures for the core reflector, support structures, and moderator because of vibration from the HTR-10 fuel cycle flow. In this study, the wear characteristics of the isotropic graphite IG-110 used in HTR-10 were evaluated. The reciprocating wear test was carried out for graphite against graphite. The effects of changes in the contact load and sliding speeds at room temperature and $400^{\circ}C$ on the coefficient of friction and specific wear rate were evaluated. The wear behavior of graphite IG-110 was evaluated based on the wear surfaces.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.338-344
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• 2010

A double rotation C-scanning system and a software program Under-Sodium MultiVIEW have been developed for the under-sodium viewing of a reactor core and in-vessel structures of a sodium-cooled fast reactor KALIMER-600. Double rotation C-scanning system has been designed and manufactured by the reproduction of double rotation plug of a reactor head in KALIMER-600. Hardware system which consists of a double rotating scanner, ultrasonic waveguide sensors, a high power ultrasonic pulser-receiver, a scanner driving module and a multi channel A/D board have been constructed. The functions of scanner control, image mapping and signal processing of Under-Sodium MultiVIEW program have been implemented by using a LabVIEW graphical programming language. The performance of Under-Sodium MultiVIEW program was verified by a double rotation C-scanning test in water.

• Proceedings of the Korea Society for Simulation Conference
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• 2000.11a
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• pp.61-66
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• 2000

The KEPRI has been upgrading the KNPEC(Korea Nuclear Power Education Center) #2 simulator, a replica of Yonggwang Unit 1 & 2, due to the outdated systems. The scenarios, such as the continuous load change, are selected to verify and validate the simulator, and the data required to V&V are generated with the best-estimated codes, RETRAN and MARS. The reactor coolant system and steam generator system are cut up into volumes and junctions for the accurate model of the scenarios, and other components and control systems are modeled. For the model the operation and design data of the plants is used and in some cases the data of Kori Unit 3 & 4 is used to fill up the lack of required data. The results of some selected analyses with the models are compared with the operating data of the plants to verify the models, and the analyses of the scenarios are carried out to generated the data for the V&V of the simulator

• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.90-96
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• 1984

Estimated dose rates in spent fuel pool storage with the extended fuel cycle core management were reviewed and compared with design limit after calculation with the aid of DLC-23/CASK(22 n, 18 g) nuclear data and ANISN code. Radioactivity and gamma spectrum within spent fuel assemblies were calculated with ORIGEN code by extended fuel cycle model. In the calculation of dose rate, the fuel pool geometry was assumed to be infinite slab. Also, composition materials and radiation source within assemblies which are being stored in pool storage were assumed to be uniformly distributed throughout all the assemblies. As a result of culculation of dose rate from stored assemblies and waterborne radionuclides in pool water, the calculated dose rates appear to be lower than design basis limit under normal condition as well as abnormal condition.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.31 no.1
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• pp.31-38
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• 2018

Earthquake disasters that exceed the design criteria can pose significant threats to nuclear facilities. Seismic probabilistic safety assessment(PSA) is a probabilistic way to quantify such risks. Accordingly, seismic PSA has been applied to domestic and overseas nuclear power plants, and the safety of nuclear power plants was evaluated and prepared against earthquake hazards. However, there were few examples where seismic PSA was applied in case of a research reactor with a relatively small size compared to nuclear power plants. Therefore, in this study, seismic PSA technique was applied to actually completed research reactor to analyze its safety. Also, based on these results, the optimization study on the seismic capacity of the system constituting the research reactor was carried out. As a result, the possibility of damage to the core caused by the earthquake hazard was quantified in the research reactor and its safety was confirmed. The optimization study showed that the optimal seismic capacity distribution was obtained to ensure maximum safety at a low cost compared with the current design. These results, in the future, can expect to be used as a quantitative indicator to effectively improve the safety of the research reactor with respect to earthquakes.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.276-284
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• 1993

The ABB-CE System-80 reactor power cutback system(RPCS) is designed to enable continuous operation of the reactor without trip in the events of the loss of one of the two main feedwater pumps and loss of load, and thus improves plant availability in a cost effective manner. In this study expansion of RPCS has been investigated for continuous reactor operation without trip in the event of an inward control element assembly(CEA) deviation including a single rod drop. Under the expanded function of RPCS the control system will provide a rapid core power reduction on demand by releasing CEAs to drop into the core and reduce the turbine power, if necessary, to follow the reactor power variation. This design feature which is included as the new design features to be incorporated in the ABB-CE System-80+ meets the EPRI advanced light water reactor(ALWR) requirements. For this study core analysis models of System-80+ have been developed to simulate the nuclear steam supply system(NSSS) response as well as the RPCS initiation of rapid CEA insertion. The results of this study demonstrate that the reactor trip can be avoided in the event of inward CEA deviation including a single rod drop by the RPCS initiation and thus the plant availability and capacity factor would be increased.