[en] The KOREA regulatory body revised the “Nuclear Safety Regulations ” in June 2015, the severe accident management has been included in the existing design basis accident management. Currently, KHNP is pushing for the development of integrated safety analysis codes applicable to multiple failure accident. It is necessary to extend the SPACE code which is developed for thermal hydraulic analysis of domestic PWR, to multiple failure accidents. In order to apply the SPACE code to multiple failure accident, the PIRT (Phenomena Identification and Ranking Table) has to be developed considering the physical phenomena expected in multiple failure accident. In this paper, “Loss of Residual Heat Removal accident During Mid-Loop Operation” is selected for development of PIRT (Phenomena Identification and Ranking Table) which is considered physical phenomena expected in multiple failure accident. Major thermal-hydraulic phenomenon PIRT for loss of RHR accident during mid-Loop operation for expanding the SPACE code to apply to the design extended conditions is developed.

[en] Monitoring samples attached to neutron shield wall of outer core are consists of impact sample, tensile sample and temperature monitor. The temperature monitor samples are examined to confirm the operation conditions whether the capsule would be in the thermal environment between 579 .deg. F and 590 .deg. F or not. Each temperature monitor having the 579 .deg. F of melting point in surveillance capsule of YGN unit 1 is laid on upper and lower

[en] The code has been Verification (V and V) matrix prepared for the SPACE code as the second phase of the development. In this study, CEA withdrawal at power accident has been simulated using the SPACE code as one aspect of the V and V work. The results from this simulation were compared with results of the RETRAN code which was used to approval of methodology for safety analysis of OPR1000 from regulatory committee. The KNAP methodology is applied to OPR1000 CEA withdrawal at power accident analysis and the results are compared with those mentioned in OPR1000 results of RETRAN code. Although there is some difference in peak temperature and SG pressure, the results from RETRAN calculation show similar trends. Through this evaluation of a OPR1000 CEA withdrawal at power accident analysis using the SPACE code, it is concluded that the SPACE code has the capability to predict the system response caused by a CEA withdrawal at power accident

[en] The Korea Nuclear Hydro and Nuclear Power Co. (KHNP) has developed a safety analysis code, called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) by collaborative works with other Korean nuclear industries. The SPACE is a general-purpose best-estimated two-phase three-field thermal-hydraulic analysis code to analyze the safety and performance of pressurized water reactors (PWRs). The SPACE code has sufficient functions and capabilities to replace outdated vendor supplied codes and to be used for the safety analysis of operating PWRs and the design of advanced reactors. As a result of the second phase of the SPACE code development project, the 2.14 version of the code was released through the successive various V and V works using integral loop test data or plant operating data. In this study, the ATLAS main steam-line break (MSLB) test, SLB-GB-01, was simulated as a V and V work. The results were compared with the measured data. The ATALS MSLB test, SLB-GB-01, was simulated using the SPACE code. The results were compared with experimental data. Through the simulation, it was concluded that the SPACE code can effectively simulate MSLB accidents

[en] The Korea nuclear industry has developed a best estimated two phase three field thermal hydraulic analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plants), for safety analysis and design of a PWR (Pressurized Water Reactor). As the first phase, the demo version of the SPACE code was released in March 2010. The code has been verified and improved according to the Verification and Validation (V and V) matrix prepared for the SPACE code as the second phase of the development. In this study, PKL III G3.1 experiment has been simulated using the SPACE code as one aspect of the V and V work. The results from this experiment were compared with tests of the SPACE and MARS codes

[en] In this study, SGTR event of Hanul Unit 4 occurred in 2002 has been analyzed using the SPACE code as one aspect of the V and V work. The results from this work were compared with simulation of the SPACE codes. The SGTR of Hanul Unit 4 has been simulated for the SPACE code V and V. The results have been compared with those of the plant and RELAP5. Throughout the evaluation of SGTR of Hanul Unit 4 using the SPACE code, it is concluded that the SPACE code has a capability to predict the behavior of plant and thermal-hydraulic response caused by SGTR event. The Korea nuclear industry has developed a best estimated two-phase three-field thermal-hydraulic analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plants), for safety analysis and design of a PWR (Pressurized Water Reactor). As the first phase, the demo version of the SPACE code was released in March 2010. The code has been verified and improved according to the Verification and Validation (V and V) matrix prepared for the SPACE code as the second phase of the development

[en] Korea Electric Power Research Institute has launched a project to develop an in-house non-loss-of-coolant-accident analysis methodology to overcome the hardships caused by the narrow analytical scopes of existing methodologies. Prior to the development, some safety analysis codes were reviewed, and RETRAN-3D and VIPRE-01 were chosen as the base codes. The codes have been modified to improve the analytical capabilities required to analyze the nuclear power plants in Korea. The methodologies of the vendors and the Electric Power Research Institute have been reviewed, and some documents of foreign utilities have been used to compensate for the insufficiencies. For the next step, a draft methodology for pressurized water reactors has been developed and modified to apply to Westinghouse-type plants in Korea. To verify the feasibility of the methodology, some events of Yonggwang Units 1 and 2 have been analyzed from the standpoints of reactor coolant system pressure and the departure from nucleate boiling ratio. The results of the analyses show trends similar to those of the Final Safety Analysis Report

[en] The Korean nuclear industry is developing a thermalhydraulic analysis code for safety analysis of pressurized water reactors (PWRs). The new code is called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). The SPACE code adopts advanced physical modeling of two-phase flows, mainly two-fluid, three-field models that comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or non-structured meshes. The programming language for the SPACE code is C++ for object-oriented code architecture. The SPACE code will replace outdated vendor supplied codes and will be used for the safety analysis of operating PWRs and the design of advanced reactors. In this paper, the SGTR (Steam Generator Tube Rupture) experiment data (double-ended guillotine break of a single U-tube was simulated) in ATLAS have been simulated using the SPACE code as part of the V and V work. The results were compared with those of experiments and other code simulations

[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] In Korea, the nuclear industries such as fuel manufacturer, the architect engineer and the utility, have been using the methodologies and codes of vendors, such as Westinghouse(WH), Combustion Engineering, for the safety analyses of nuclear power plants. Consequently the industries have kept up the many organizations to operate the methodologies and to maintain the codes for each vendor. It may occur difficulty to improve the safety analyses efficiency and technology related. So, the necessity another of methodologies and code systems applicable to Non- LOCA, beyond design basis accident and performance analyses for all types of pressurized water reactor(PWR) has been raised. Due to the above reason, the Korea Electric Power Research Institute(KEPRI) had decided to develop the new safety analysis code system for Korea Standard Nuclear Power Plants in Korea. As the first requirement, the best-estimate codes were required for applicable wider application area and realistic behavior prediction of power plants with various and sophisticated functions. After the investigation for few candidates, RETRAN-3D has been chosen as a system analysis code. As a part of the feasibility estimation for the methodology and code system, CRD(Control Rod Drop) accident which an event of Non-LOCA accidents for Uljin units 3 and 4 and Yonggwang 1 and 2 was selected to verify the feasibility of the methodology using the RETRAN-3D. In this paper, RETRAN DNB Model and CETOP DNB Model were analyzed by using comparative method