[en] The procedures may also be used as an aid for assessing and documenting the results of tests. The commissioning procedures should include information that specifies several items. Those are mainly (1) all the activities and performance parameters that are to be measured under specified steady state and transient conditions, (2) the requirements on performance, together with clearly stated acceptance criteria. The final phase of stage C commissioning is reactor performance test, which is to prove the integrated performance (neutron power calibration, Control Absorber Rod drop time, I and C functioning, Rod worth, Core heat removal with natural mechanism) and the safety of the research reactor at full power with fuel loaded. Commissioning procedure was developed to show the safety of the research reactor. Both indirect and direct indicators were selected to show that the safety is ensured: 1) indirect parameters which imply success of safety functions: power, flow, opening valves, system response as-designed; 2) direct parameters which shows no failure of safety functions: no meaningful increase in level of neutron in the cooling system. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters. A preliminary comparison to conservative estimation has shown that the nuclear reactor safety of the research reactor will be assured by verifying that the reactor power and the PCS flow rate are conservative

[en] The system codes used for the safety analysis of reactor facilities should accurately and reliably predict the transient phenomena for the facility. Therefore, it is necessary to have appropriate model and analysis capabilities for the purpose, the range, and the major thermal hydraulic phenomena of the facility. Many experiments were carried out to understand the thermal hydraulic characteristics of siphon break. Full-scale experiments have been performed by POSTECH in various sizes of pipe rupture and siphon breaker line. A series of experiments has been conducted by Idaho state university in an experimental facility consisting of relatively small size pipe. In this paper, SPACE3.0 and RELAP5/MOD3.3 were used to simulate the experimental data. Their calculation results were compared with the experimental data. The simulations of siphon break experiments were performed using SPACE3.0 and RELAP5/MOD3.3. In the 14-inch LOCA with a 4-inch siphon breaker, both SPACE3.0 and RELAP5/MOD3.3 calculate the pool level lower than the experiment when the siphon phenomenon is finished, and SPACE3.0 predicts much lower pool level than RELAP5/MOD3.3. Both SPACE3.0 and RELAP5/MOD3.3 well predict the POSTECH experiment with the 14-inch break size and 6-inch siphon breaker and the Idaho state university experiment.

[en] The thermal hydraulic analysis of a research reactor building becomes much more important during longterm cooling stage in loss of normal electric power if the building is designed as containment to fulfill the enhanced regulation requirements of radiological consequence. Since the existing containment analysis computer codes are oriented to the condition and validated for nuclear power plants, it is necessary that a computer code adequate for the research reactors is developed. The purpose of this paper is to identify the thermal hydraulic phenomena during the long-term cooling of a research reactor and to select the appropriate analysis models and to use it as the basic data for the development of a computer code for reactor building and pool cooling analysis. This paper consists of three steps, and a description of each step is as follows: - PIRT for identification of thermal-hydrodynamic phenomena during long-term cooling of research reactors - Literature review and selection of analysis models for the thermal hydraulic phenomena determined by PIRT - Development of the code structure and algorithm of a computer code for cooling analysis of reactor buildings and pool.

[en] Following the Three Mile Island Unit 2 (TMI-2) accident, the United States Nuclear Regulatory Commission (NRC) established the sets of requirements addressing their objective to improve the quality of operational information for dealing with emergency events in nuclear power plants. The Emergency Operating Guidelines (EOG) should be presented to provide technical information to prepare reactor-specific Emergency Operating Procedures (EOP) which cover operation during emergency events. EOG must provide guidance for both classes of emergencies. Thus, when a reactor trip occurs or should occur, the operators can refer to guidance which will provide a safe response whether or not a symptom set is identified: EOG written to treat specific symptoms are called event-based recovery guidelines (ERG); the EOG which provides guidance for undiagnosed events for which a reactor trip is required is called the Symptom-based Recovery Guidelines (SRGs). For helping to design and develop a set of EOG/EOP, a stencil in a sheet was proposed as an easy and intuitive tool to gather information of a research reactor related to safety/safety-related functions. With concept of safety functions, the stencil will be able to give a strong but very easy, straightforward, systematic, comprehensive tool to analyze the architecture of the reactor, to cover the whole SSC information

[en] The thermal hydraulic phenomena within the research reactors are identified and the ranking tables are prepared to cover the initiating events up to beyond design basis accidents. Multi-dimensional physics is also included for the use in the future. Proposed were the RRs PIRTs, which will be the bases for the requirements for the code extension and the V&V. A licensed system code, dedicated to research reactors under low temperature and low pressure, is under development based on the SPACE code, licensed as a domestic thermo-hydraulic system code for nuclear power plants under high temperature and high pressure in Korea. Phenomena Identification and Ranking Tables for the research reactors were developed by considering the characteristics of the research reactors such as the HARARO, JRTR, KJRR, and potentially expected research reactors.

[en] In this study, we focused on the implementation of an upwind method for a new 2- dimensional 2-fluid model including the surface tension effect in the momentum equations. This model consists of a complete set of 8 equations including 2-mass, 4-momentum, and 2-internal energy conservations having all real eigenvalues. Based on this equation system with upwind numerical method, we first make a pilot 2-dimensional code and then applied it to 2 benchmarks. The first benchmark is the shock tube problem, the result shows reasonable wave characteristics including void wave propagation. The second benchmark is a water faucet problem and the results are in good agreement with the analytical solutions

[en] A post-test simulation on a SBLOCA test for a PSV line break (SB-PSV-02) has been performed with the MARS-KS code to assess its simulation capability for the SBLOCA scenario of the SMART design. The SBLOCA scenario in the SMART design was well reproduced using the VISTA-ITL facility, which has a reduced-height, 1/1310-volume scaled test facility based on the design features of SMART, and the measured thermal-hydraulic data were properly simulated with the MARS-KS code. The present analysis results on SBLOCA can provide a good understanding on the thermal-hydraulic characteristics of the SBLOCA behavior of the SMART design. (authors)