[en] The local instantaneous balance equation reads as, ∂/∂t(ρ_kψ_k)+∇·(ρ_kψ_ku_k)=-∇·J_k+ρ_kψ_k (1) where ρ_k,ψ_k,u_k,J_k and φ_k are the density, property of extensive characteristics, velocity, flux and source of k-phase, respectively. Nomenclatures for other variables are found in. Table-1 shows the field variables. e_k,q_k,ρ_k,γ_k,g_k,q_k,M_k,E_k are internal energy, heat flux, pressure, vaporization, gravity and internal heat rate, interfacial momentum and energy sources respectively. is the stress tensor and is decomposed into pressure and shear. The time averaged balance equation for any property ψ_k of k-phase can be presented as, ∂/∂t(ρ_kψ_k)+∇·(ρ_kψ_kU_k)=-∇·J_k+ρ_kφ_k+I_k. I_k≡-I/ΔtΣ_j{I/u_ni[(ρ_kψ_k)n_k·(u_k-u_i)-(n_k·J_k)]} (2) where the bar over a quantity indicates the time-averaging operation.is the surface normal vector in this paper. Considering the time-fluctuating terms, the time-averaged balance equation can be represented by using weighted mean variables as, (J_k^T: turbulent effectsTkJ): ∂/∂t(ρ_kψ_k)+∇·(ρ_kψ_ku_k)=-∇·(J_k+J_k^T)+ρ_kφ_k+I_k

[en] The Shutdown Cooling System (SCS) removes core decay heat during the planned plant shutdown or after the accident. A computer code such as DESCENT used by Combustion Engineering or RHRCOOL used by Westinghouse, is utilized to analyze the capacity and performance of the SCS for the system design of new plant and the replacement/repair of SCS heat exchanger of the operating reactors. These codes include approximated correlations for heat exchangers for the tube side flow ratio, total heat transfer coefficient, and the balance of the resistance constant calculated by the heat exchanger design codes, such as HTRI or HTFS. HTRI or HTFS does not have the capability to simulate the transient conditions of SCS. In this study, the SCS performance analysis and capacity evaluation (SPACE) code is developed to evaluate the total heat transfer coefficient for the heat exchanger as well as to analyze the SCS cooldown performance

[en] An Anticipated Transients without Scram (ATWS) is an anticipated operational occurrence accompanied by a failure of the reactor to trip when required. ATWS events are of concern since, under certain conditions (e.g., additional component and/or system failures) these could lead to unacceptable consequences up to and including core melt and release of radioactivity to the environment. The major concern of the ATWS derives from the consequences of the expected high primary system pressure, which is characteristic of this transient. RETRAN-3D is a best-estimate thermal-hydraulic transient code and is used as a system analysis code in the development of a non-LOCA safety analysis method for the future application to Optimized Power Reactor (OPR1000) design. A lot of efforts are now being made to investigate the applicability of the RETRAN-3D code especially to Non- LOCA analysis, by comparing the analysis results with those from the current licensing code, CESEC-III. The comparative simulations of Steam Line Break (SLB) and Locked Rotor event already showed that the RETRAN-3D code is applicable to the analysis of Non- LOCA events. However, ATWS analysis of OPR1000 using RETRAN-3D has not been performed. In this paper, detailed thermal hydraulic analyses for a loss of main feedwater event assuming that the DPS is not available in addition to the failure of RPS were performed using the RETRAN-3D code. To investigate the applicability of the RETRAN-3D to ATWS analysis of OPR 1000, the calculation results were also compared with those of CESEC-III

[en] Natural circulation is a heat removal process whereby Reactor Coolant System (RCS) flow is driven by temperature and density differences in the RCS fluid between the core and steam generators. The combination of core heat addition and steam generator heat removal would cause continuous flow to develop through the RCS and should provide enough heat removal to adequately cool the core. Natural Circulation Cooldown (NCC) capability was evaluated to ensure the safe shutdown function of the power up rated Kori 3 and 4 and Yonggwang 1 and 2 Nuclear Power Plants (NPPs). The evaluation was performed for the duration for normal operation to the conditions under which the initiation of the Residual Heat Removal (RHR) is permitted in accordance with the initiation of the Residual Heat Removal (RHR) is permitted in accordance with the US NRC Branch Technical Position (BTP) Reactor Systems Branch (RSB) 5-1. BTP RSB 5-1 requires the use of only safety-grade equipment and the assumptions of the concurrent loss of offsite power with a single failure. Steam Generator (SG) Power-Operated Relief Valve (PORV), Pressurizer PORV and Auxiliary Feedwater Pump were used as safety-grade means of the RCS cooldown and depressurization for NCC analysis of Kori 3 and 4 and Yonggwang 1 and 2 NPPs. The evaluation of NCC capability was done using a computer code, CENTS. The analysis result showed that the amount of safety-grade auxiliary feedwater required to cool the RCS down to RHR entry concluded that the power up rated Kori 3 and 4 and Yonggwang 1 and 2 NPPs can be cooled and de pressurized to RHR entry conditions by the natural circulation in conformance with BTP RSB 5-1 requirements

[en] The SPACE structured/staggered mesh system is based on the orthogonal hexahedral shape of cell and its surrounding faces. All of the geometric quantities are described in terms of cell volume, centroid, face area, and face center, so that 1-D pipe, 3-D Cartesian, and cylindrical mesh systems can be expressed in the same manner. In this paper, simple algebraic ways to generate the component-wise block meshes are described. In addition, a methodology to construct complex system by connecting the already generated block meshes is demonstrated

[en] In order to develop the Korean best-estimate safety analysis code for nuclear power plants, SPACE, several mesh systems and numerical schemes, such as structured/staggered, unstructured/ collocated meshes, semi-implicit, and nearly implicit numerical schemes, have been tried so far. In the earlier versions of SPACE, however, the numerical solution schemes using the structured/staggered meshes and unstructured/ collocated meshes were separately developed to two different versions of hydraulic solver: the staggered version of hydraulic solver and the collocated one. In this paper, these two different hydraulic solvers are merged into a unified version. With this new version of SPACE, users can divide the entire calculation domain into several sub-domains which have different types of mesh from each other. When modeling a reactor system, for example, users can apply the unstructured/collocated meshes to the reactor vessel head, and the structured/staggered meshes to the other part of the reactor system. In the following sections, the mesh system, numerical solution schemes, and code structure of the unified version are briefly described. Finally one of the application results is presented

[en] The control systems for the nuclear steam supply system (NSSS) make the nuclear power plant (NPP) operate efficiently under transient conditions as well as steady state conditions. The control performance of the NSSS control systems depends on their own control algorithms and relevant parameters such as gain values, time constants and so on. Currently, the values of these parameters are determined based on sensitivity studies using system simulation computer code and engineering judgments. In this study, the optimization for the parameters of the feedwater control system (FWCS) is performed to provide it with a better control performance. The optimization objective is to minimize deviation of the steam generator (SG) water level during transients. In this case, analytic objective function does not exist and the responses to input can be evaluated by computer simulations only. Therefore the simulation optimization methodology is used and the response surface methodology (RSM) is adopted as the simulation optimization algorithm. As a result, the control performance of the FWCS is remarkably improved

[en] The risk-informed regulation and application (RIR and RIA) based on probabilistic safety assessment (PSA) has been discussed and is under review for the operation, maintenance and design. In the future, the new approach for the design might be applied especially to the loss-of coolant accident (LOCA). The branch line connected to the main RCS piping is considered as the largest piping to credit the break. This break size is called as a transition break size (TBS). These new trends in the safety analysis of nuclear power plants reflect the need to eliminate the over conservatism in design concept and to secure economical efficiency as well as acceptable safety margin. In the design based on the RIR, the TBS LOCA will be the limiting accident for the emergency core cooling system (ECCS) or the safety injection system (SIS) design instead of a large break LOCA (LBLOCA). In this paper, a study on the TBS LOCA is performed for the Younggwang Nuclear Power Plant Units 5 and 6 (YGN 5 and 6), typical plants of OPR1000, for ECCS design optimization by comparing the results of the LBLOCA. This study was conducted as a part of the improvement in plant operation and design technology. The LBLOCA and TBS LOCA analyses were performed for the RCP discharge leg break changing the safety injection (SI) flow rate. The results are compared with those of LBLOCA in terms of peak clad temperature (PCT) to verify the validity of ECCS design changes

[en] Anticipated Transients Without Scram (ATWS) would occur on failure of all the control and shutdown assemblies to insert into the core following an automatic reactor trip. The major concern of the ATWS derives from consequences of the high primary system pressure which is the characteristic of the transients. According to section 2.4 of YVL guides which are Finnish regulations for safety of nuclear power plants (NPP), the acceptance criterion for the ATWS analysis is that the pressure of the protected item does not exceed a pressure limit that is 1.3 times the design pressure. The main purpose of this paper is to assess its impact on the APR1400 preliminarily, for Europe regulatory environments by applying European Utility Requirements (EUR) for Light Water Reactor Nuclear Power Plants

[en] A test version of the two-fluid program has been developed by extending the PISO algorithm. Unlike the conventional industry two-fluid codes, such as, RELAP5 and TRAC, this scheme does not need to develop a pressure matrix. Instead, it adopts the iterative procedure to implement the implicitness of the pressure. In this paper, a brief introduction to the numerical scheme will be presented. Then, its application to bubble column simulation will be described. Some concluding remarks will be followed