[en] In accordance with the amendment of Nuclear Safety Act in 2015 and the amendment of regulation on technical standards for nuclear power facilities in 2016, accident management is legislated and the range of accident management is classified design basis accident (DBA), multiple failure accident, extreme hazards and severe accident. This paper describes the preliminary evaluation of ISLOCA for OPR1000 nuclear power plants. Preliminary evaluation of interfacing system LOCA for OPR1000 is accomplished. In the system connected to the reactor coolant system pressure boundary, it was examined whether it meets design requirements of interfacing system LOCA. We also reviewed design improvements against interfacing system LOCA vulnerabilities.

[en] Hydrogen is one of an important energy for future hydrogen economy and noted as the alternative energy substituting oil consumption. Domestic hydrogen demands are expected to increase from 1,600 ton in 2015 year to 7.77million ton in 2040 year. Current hydrogen production has primarily accomplished with steam methane reforming (SMR). But for large-scale hydrogen production, we should develop an advanced hydrogen production technologies that avoid consumption of fossil fuels and emissions of greenhouse gases. Therefore, we are studying large-scale hydrogen production basis technology by using a nuclear power plant such as a very high temperature reactor and high temperature steam electrolysis. In this paper, KEPRI's development status and future strategy for nuclear hydrogen production technology will be described and proposed

[en] As a part of the project 'development of hydrogen production technologies by high temperature electrolysis using very high temperature reactor', we have developed an electrolyzer model for high temperature steam electrolysis (HTSE) system and carried out some preliminary estimations on the effects of heat recovery on the HTSE hydrogen production system. To produce massive hydrogen by using nuclear energy, the HTSE process is one of the promising technologies with sulfur-iodine and hybrid sulfur process. The HTSE produces hydrogen through electrochemical reaction within the solid oxide electrolysis cell (SOEC), which is a reverse reaction of solid oxide fuel cell (SOFC). The HTSE system generally operates in the temperature range of 700∼900 .deg. C. Advantages of HTSE hydrogen production are (a) clean hydrogen production from water without carbon oxide emission, (b) synergy effect due to using the current SOFC technology and (c) higher thermal efficiency of system when it is coupled nuclear reactor. Since the HTSE system operates over 700 .deg. C, the use of heat recovery is an important consideration for higher efficiency. In this paper, four different heat recovery configurations for the HTSE system have been investigated and estimated

[en] The new code is named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). As a part of code validation effort, an analysis of the ATLAS SGTR(Steam Generator Tube Rupture) experiment using SPACE code has been performed. The SGTR-HL-05 experiment is a multiple U-tube SGTR. The calculated results using the SPACE code are compared with those from the experiment. The ATLAS SGTR-HL-05 experiment, which is five U-tube rupture test, was simulated using the SPACE code. The calculated results were compared with those from the experiment. The comparisons of break flow rate and accumulated break flow mass show good agreement with the experiment. The SPACE code is capable of predicting physical phenomena occurred during ATLAS SGTR-HL-05 experiment

[en] Since the Fukushima accident, the concern has increased internationally about the disaster and the severe accident. In particular, the importance of severe accidents prevention and mitigation has been highlighted. KHNP is pushing for the development of integrated safety analysis codes applicable to multiple failure accident. It is necessary to the extension development of a code for apply the multiple failure accident to the SPACE which is developed for thermal analysis of domestic PWR. In order to apply the SPACE code to multiple failure accident, the PIRT(Phenomena Identification and Ranking Table) has to develop considering the physical phenomena expected in multiple failure accident. It developed a major thermal-hydraulic phenomenon PIRT for ATWS accidents for expanding the SPACE code to apply to the design extended conditions. The major scenarios and the major thermal-hydraulic phenomenon of the system/structure and the component were derived through the developing the PIRT. PIRT was able to derive the thermal-hydraulic model needed to expand the SPACE code.

[en] KHNP is pushing for the development of integrated safety analysis codes applicable to multiple failure accident. It is necessary to the extension development of a code for apply the multiple failure accident to the SPACE which is developed for thermal analysis of domestic PWR. In order to apply the SPACE code to multiple failure accident, the PIRT(Phenomena Identification and Ranking Table) has to develop considering the physical phenomena expected in multiple failure accident. It developed a major thermal-hydraulic phenomenon PIRT for MSGTR accidents for expanding the SPACE code to apply to the design extended conditions. The major scenarios and the major thermal-hydraulic phenomenon of the system/structure and the component were derived through the developing the PIRT. PIRT was able to derive the thermal-hydraulic model needed to expand the SPACE code.

[en] The preliminary design project of the Advanced Power Reactor 1000 (APR1000) has been performed by Korea Electric Power Corp. (KEPCO) and Korea Hydraulic and Nuclear Power Co. (KHNP) since the end of 2009. The APR1000 has been developed to implement accumulated operational experience and advanced safety features (ADFs) in the Optimized Power Reactor 1000 (OPR1000) plant to meet the requirements of Generation III+ nuclear power plants. As a design basis accident (DBA) analysis, a Non-Loss of Coolant Accident (Non-LOCA) analysis has been performed to confirm the performance of the structures, systems, and components (SSCs) under a wide spectrum of anticipated initial conditions and assumptions

[en] Korea Electric Power Research Institute(KEPRI) has been developing safety analysis methodology for nonloss of coolant accident(non-LOCA) analysis of Korea Standard Nuclear Power Plant(KSNP). This new methodology uses RETRAN as main safety analysis code. The standard RETRAN nodalization of KSNP includes a detailed 14-node steam generator secondary modeling and main steam safety valves(MSSV) with accumulation and blowdown characteristics. However, some transients, such as steam generator tube rupture(SGTR), does not require detailed steam generator secondary side modeling. In this paper, as a comparative study, simplified RETRAN model is used for SGTR analysis and the results are compared with results from more complex RETRAN model. Simplified RETRAN model include 2-node steam generator secondary side model, simplified MSSV opening-closing characteristics and simplified ruptured tube break flow model. Reference plant for analysis is Ulchin unit 3/4. With simplified model, similar results can be obtained with less computing time

[en] As a part of the unified safety analysis computer code development project funded by the Ministry of Commerce, Industry and Energy(MOCIE), Korea Electric Power Research Institute(KEPRI) has been developing the methodology of new safety analysis for Korea Standard Nuclear Power Plants(KSNP). The pressure and enthalpy of primary system is affected by the leakage owing to tube rupture and a thermal-hydraulic behavior in the reactor vessel is important in the light of reactor heat removal and recirculation of loop flow. To review effects for one channel or split core model in the SGTR accident, we considered two types of core model and analyzed using RETRAN-3D

[en] Korea Electric Power Research Institute(KEPRI) has been developed safety analysis methodology for nonloss of coolant accident(Non-LOCA) analysis of Optimized Power Rector 1000(OPR1000, previously KSNP). The methodology has been developed using RETRAN code of Electric Power Research Institute(EPRI) as a system analysis code and named Korea Non-LOCA Analysis Package(KNAP). Steam Generator Tube Rupture(SGTR) accident is one of the decrease in reactor coolant system inventory events and the results are typically described in the safety analysis report(SAR) chapter 15.X. KEPRI has been analyzed OPR1000 SGTR accident analysis as a part of the unified safety analysis computer code development project and applied the methodology to Advanced Power Reactor 1400(APR1400) to confirm the feasibility of that. APR1400 has been designed to generate about 1,400MWe with advanced features for greatly enhanced safety and economics goals. The SGTR analysis in APR1400 Standard Safety Analysis Report(SSAR) is simulated by CESEC-III code of Combustion Engineering(CE). In this study, to estimate the feasibility of the KNAP methodology and code system, SGTR accident is analyzed using RETRAN code and it is compared those from APR1400 SSAR