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[en] The purpose of this report provides information related to the design of the Oregon State University Advanced Plant Experiment (APEX) test facility. Information provided in this report have been pulled from the following information sources: Reference 1: R. Nourgaliev and et.al, 'Summary Report on NGSAC (Next-Generation Safety Analysis Code) Development and Testing,' Idaho National Laboratory, 2011. Note that this is report has not been released as an external report. Reference 2: O. Stevens, Characterization of the Advanced Plant Experiment (APEX) Passive Residual Heat Removal System Heat Exchanger, Master Thesis, June 1996. Reference 3: J. Reyes, Jr., Q. Wu, and J. King, Jr., Scaling Assessment for the Design of the OSU APEX-1000 Test Facility, OSU-APEX-03001 (Rev. 0), May 2003. Reference 4: J. Reyes et al, Final Report of the NRC AP600 Research Conducted at Oregon State University, NUREG/CR-6641, July 1999. Reference 5: K. Welter et al, APEX-1000 Confirmatory Testing to Support AP1000 Design Certification (non-proprietary), NUREG-1826, August 2005.
[en] Manual opening of the POSRVs will result in the release of mass and energy to the SG compartment connected with the containment atmosphere, if no containment heat removal systems are available, the containment will pressurize. If RCS depressurization may result in a severe challenge, then the TSC may consider the impact unacceptable and may want to perform mitigative actions. In this study, the Negative-Impact associated with depressurizing the RCS during Station Blockout (SBO) is evaluated quantitatively by utilizing the Modular Accident Analysis Program (MAAP5) computer code that can analyze the containment pressure. The Negative-Impact associated with depressurizing the RCS has been evaluated during SBO. In this study, the analysis results for the RCS depressurization strategy after the CET exceeds 1200 .deg. F indicate that the containment pressure would maintain at a lower pressure than the severe challenge setpoint. But, since the containment pressure when the POSRVs are opened is expected to increase about 2021cmH_2O, the opening of the pressurizer POSRVs may result in a severe challenge to the plant for high containment pressure condition
[en] The Tower Cooling Water System (TC) is designed to reject the heat load generated by operating the accelerators and the utility facilities through the component cooling water (CCW) heat exchangers. The circulating water discharged from the circulating water pumps passes through the CCW heat exchangers, the Chiller condenser and the air compressor, and the heated circulating water is return to the cooling tower for the heat removal. In this study, The design of Tower Cooling Water System is changed as follows : At First, The quantity of cells is changed into six in order to operate the cooling tower accurately correspond with condition of each equipment of head loads. The fans of cooling tower are controlled by the signal of TEW installed in the latter parts of it. The type of circulation water pump is modified to centrifugal pump and debris filter system is deleted
[en] A neutronics benchmark specification was developed to support the analysis of SHRT-45R test, which was conducted in April 1986 using Experimental Breeder Reactor II (EBR-II) to demonstrate that a sodium-cooled fast reactor (SFR) with a sodium-bonded metallic fuel could be designed such that passive phenomena, as opposed to active electromechanical systems, are effective in protecting the reactor against the potentially adverse consequences of unprotected accidents.
[en] In the passive residual heat removal system of a molten salt reactor, one of the residual heat removal methods is to use the thimble-type heat transfer elements of the drain salt tank to remove the residual heat of fuel salts. An experimental loop is designed and built with a single heat transfer element to analyze the heat transfer and flow characteristics. In this research, the influence of the size of a three-layer thimble-type heat transfer element on the heat transfer rate is analyzed. Two methods are used to obtain the heat transfer rate, and a difference of results between methods is approximately 5%. The gas gap width between the thimble and the bayonet has a large effect on the heat transfer rate. As the gas gap width increases from 1.0 mm to 11.0 mm, the heat transfer rate decreases from 5.2 kW to 1.6 kW. In addition, a natural circulation startup process is described in this paper. Finally, flashing natural circulation instability has been observed in this thimble-type heat transfer element
[en] As part of a Department of National Fusion Research Institute (NFRI) Project, Seoul National University (SNU) is calculating pumping power of blanket system using various coolants. The main goal of this preliminary study is to estimate minimum coolant pumping powers for blanket heat removal and to compare them for evaluating thermal hydraulic performances. In this study, the calculations were conducted based on a plate type blanket design, which was newly proposed by NFRI. This study is basically consisted of three steps. First, candidates of coolant are selected. Second, calculating conditions are determined. Last, the pumping powers of various coolants are calculated and compared. The details are described below
[en] Two-phase closed thermosyphon (TPCT) is vertically oriented wickless heat pipe that has working fluid in the interior. The TPCT transports a large amount of heat from evaporator to condenser by phase change of working fluid, and the working fluid passively returns to evaporator by gravity. Due to these advantages of the TPCT, the TPCT is considered as method of PRHR (Passive Residual Heat Removal) system in nuclear system. Parametric studies have done to investigate the heat transfer characteristics of the TPCT. Different working fluids such as water, ethanol, methanol and acetone were used at various filling ratios and at different operating temperatures to find maximum heat transport capabilities of TPCT. Effect of heat transfer rate, filling ratio and aspect ratio were investigated. Inclined angle effect was investigated at several filling ratios and working fluids. This study is interested in silicon oil effect on the TPCT. To carry out the experiment, experimental apparatus is designed and manufactured. In design process, the TPCT operation limit is considered This study is interested in silicon oil effect on the TPCT. Experiments were carried out at three oil weight percent with three input power. Effect of oil on the TPCT is evaluated by inner wall temperature distribution and thermal resistance. In this study, silicon oil effect on TPCT was investigated. The TPCT was operated with several oil weight percent and input power. From experiment, overall, the silicon oil reduced evaporator thermal performance, but enhanced condenser thermal performance. However, the TPCT total thermal performance was reduced by 100 c St silicon oil
[en] The Gen-IV nuclear reactor should assure the economic improvement and enhanced safety to be commercial reactor but it is not so easy to satisfy the both goals at a time. Therefore, it is required to minimize the excessive conservatism of the safety systems and improve the economic competitiveness simultaneously. This study is on the system arrangement of a pool type sodium-cooled fast reactor with a large capacity at the preliminary conceptual design stage. System arrangement designs of the heat transport system, residual heat removal system and reactor building are carried out. Current system arrangement study is performed at the preliminary conceptual design level and thus the future work will be followed to obtain the feasibility for the design concept on the manufacturability of components and structures
[en] The loss of residual heat removal system during midloop operation was performed using RELAP5/MOD3 and RELAP5/Mod3.1. The objective of present analysis is to assess the code capability in predicting the major thermal-hydraulic behavior under the influence of non-condensable gas and the low pressure condition. The experiment was conducted at the Large Scale Test Facility (LSTF). The experiment involved a 5% cold leg break along with the loss of the residual heat removal system. The analysis results using RELAP5/MOD3 were not in good agreement with the experiment. Besides, the water property errors occurred too frequently from the 450 sec after the initiation of the transient. The analysis results using RELAP5/MOD3.1 predicted well the overall transient behavior until the loop seal clearing. However, due to the underprediction of steam condensation the results after the loop seal clearing differ from the experiment
[en] KEPRI is developing the defense-in-depth model ,which is called the ORION model, to monitor risk during the planned outage. The defense-in-depth is measured by number, redundancy and diversity of systems, structures, and components (SSCs) which are needed to mitigate challenges to the safety functions. To complete qualitative defense-in-depth model, also additional deterministic information such as thermal margin when a loss of residual heat removal event occurs during outage is required. The relationship between the expected decay heat levels and the time until RCS boiling, core uncovery, and core damage is described and the estimation result of the thermal margin during the specific refueling outage of reference plant is presented in this paper