[en] This research has been performed to introduce a concept of supercritical pressure water cooled reactor(SCWR) in Korea The area of research includes core conceptual design, evaluation of candidate fuel, fluid systems conceptual design with mechanical consideration, preparation of safety analysis code, and construction of supercritical pressure heat transfer test facility, SPHINX, and preliminary test. As a result of the research, a set of tools for the reactor core design has been developed and the conceptual core design with solid moderator was proposed. The direct thermodynamic cycle has been studied to find a optimum design. The safety analysis code has also been adapted to supercritical pressure condition. A supercritical pressure CO2 heat transfer test facility has been constructed and preliminary test proved the facility works as expected. The result of this project will be good basis for the participation in the international collaboration under GIF GEN-IV program and next 5-year mid and long term nuclear research program of MOST. The heat transfer test loop, SPHINX, completed as a result of this project may be used for the power cycle study as well as further heat transfer study for the various geometries

[en] The KNGR is to install a Fluidic Device at the bottom of the inner space of the SIT (Safety Injection Tank) to control the flow rate of safety injection coolant from SIT during LBLOCA. During the past two years, a scale model test to obtain the required flow characteristics of the device under the KNGR specific conditions has been performed using the experience and existing facility of AEA Technology (UK) with appropriate modifications. The performance verification test is to be performed this year to obtain optimum characteristics and design data of full size fluidic device. The purpose of the model test was to check the feasibility of developing the device and to produce a generic flow characteristic data. The test was performed in approximately 1/7 scale in terms of flow rate with full height and pressure. This report presents the details of system performance requirements for the device, design procedure for the fluidic device to be used, test facility and test method. The time dependent flow, pressure and Euler number are presented as characteristics curves and the most stable and the most effective flow control characteristic parameters were recommended through the evaluation. A method to predict the size of the fluidic device is presented. And a sizing algorithm, which can be used to conveniently determine the major geometric data of the device for various operating conditions, and a FORTRAN program to produce the prediction of performance curves have been developed. (author). 32 refs., 15 tabs., 47 figs

[en] International and domestic status of development activities of nuclear fusion energy technologies are analyzed and summarized. From these results a verifiable R and D strategy is derived which allows purposeful and successful participation in the ITER project and thus enables a domestic technological basis of the commercialization of nuclear fusion energy. A 45-year, three-stage plan is proposed with a detailed plan for the 10-year, 1st stage where a conceptual design of a Korean demonstration fusion power plant (KDEMO) will be developed as well as its key component designs such as breeder blanket

[en] International collaboration for the development of future innovative energy system is being activated for preparing the future nuclear renaissance and enhancing the competitiveness of nuclear energy technology. Exemplary programs of these international movement are GIF, the Generation IV International Forum, being driven by 10 advanced nuclear technology holding countries including Korea and U.S., and the INPRO of IAEA. Korea has been actively participated in these two programs as an original signatory of the GIF Charter and a regular member of INPRO. Parallel with these programs, Korea and U.S. have cooperated in the INERI, International Nuclear Energy Research Initiative programs, since 2001 under the bilateral agreement between two governments. Korea is going to start international collaboration for Gen IV development in the near future, and members of the INPRO have urged Korea to join the Case study of INPRO. KAERI as a leading research institute of Korea's nuclear R and D, should play a key role in developing future nuclear energy system, drawing-up a plan for effective performance is needed. Therefore, in this study, we are aiming at investigation of status of future innovative nuclear energy system development and setup of the strategic plan for effective Participation of KAERI in 'Advanced nuclear technology development project'. Contents and scope of the study for successful achievement are as follows; - Investigation and analysis of international and domestic trends related to GIF, IAEA INPRO, INERI between Korea and U.S. - Investigation and analysis of international and domestic R and D related to Gen IV - Support for international activities related to the INPRO and planning for effective engagement in the project - Development of draft executive national plan for participation in international R and D projects on Gen IV. In this study, results were mainly made in four fields. First, this study surveyed the international cooperation on future innovative nuclear energy system and the present progress of GIF, I-NERI and INPRO. Second, international R and D and technological state relating Generation IV nuclear energy system and future energy system requirement of IAEA INPRO were analyzed. Recommended R and D fields in technology road map for Gen IV, the R and D and the state-of-the-art of major countries were described. Third, we suggested the participation plan on the INPRO for the KAERI, and perform technical review and consultatory support. Finally, draft executive national plan for participation in international R and D projects on Gen IV was developed considering the national nuclear energy R and D and KAERI's. Drafted plan was made in Technical Committee composed of national experts concerned to nuclear technology, and the objective of national plan is development of Gen IV systems as main future electricity production system and related proliferation resistant fuel cycle technologies through international cooperation. Strategically, in the basis of collaboration in GIF international R and D, and in accordance with the 'choice and concentration' principle for efficient use of domestic resources, 3 major field of priority were selected, they are ; 1) Sodium-cooled fast reactor core and system design, 2) Proliferation resistant fuel cycle and advanced material technology, and 3) Key source technologies for new concept Gen IV systems. Planning report for these is appended next to main report. This study can furnish the base of the effective conduction of KAERI in international cooperation R and D for Gen IV, and contribute to the acquirement of future nuclear key source technology. The result of this study could be reflected in mending the KAERI's long term R and D policy and future management plan

[en] The program on Thermal-Hydraulic Evaluation by Testing and Analysis (THETA) for the development and licensing of the new design features in the APR1400 (Advanced Power Reactor-1400) is briefly introduced with a presentation on the research motivation and typical results of the separate effect tests and analyses of the major design features. The first part deals with multi-dimensional phenomena related to the safety analysis of the APR1400. One research area is related to the multi-dimensional behavior of the Safety Injection (SI) water in a reactor pressure vessel downcomer that uses a direct vessel injection type of SI system. The other area is associated with the condensation of steam jets and the resultant thermal mixing in a water pool; these phenomena are relevant to the depressurization of a Reactor Coolant System (RCS). The second part describes our efforts to develop new components for safety enhancements, such as a fluidic device as a passive SI flow controller and a sparger to depressurize the RCS. This work contributes to an understanding of the new thermal-hydraulic phenomena that are relevant to advanced reactor system designs; it also improves the prediction capabilities of analysis tools for multi-dimensional flow behavior, especially in complicated geometries

[en] The accuracy of the specification of requirements of a digital system is of prime importance to the acceptance and success of the system. The development, use, and regulation of computer systems in nuclear reactor Instrumentation and Control (I and C) systems to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Korean Next Generation Reactor (KNGR) Software Safety Verification and Validation (SSVV) Task, Korea Atomic Energy Research Institute, which investigates different aspects of computer software in reactor I and C systems, and describes the engineering procedures for developing such a software. The purpose of this guideline is to give the software safety evaluator the trail map between the code and standards layer and the design methodology and documents layer for the software important to safety in nuclear power plants. Recently, the requirements specification of safety-critical software systems and safety analysis of them are being recognized as one of the important issues in the software life cycle, and being developed new regulatory positions and standards by the regulatory and the standardization organizations such as IAEA, IEC, and IEEE. We presented the procedure for evaluating the software requirements specifications of the KNGR protection systems. We believe it can be useful for both licenser and licensee to conduct an evaluation of the safety in the requirements phase of developing the software. The guideline consists of the requirements engineering for software of KNGR protection systems in chapter 1, the evaluation checklist of software requirements specification in chapter2.3, and the safety evaluation procedure of KNGR software requirements specification in chapter 2.4

[en] This report describes on technical contents related cyber training system construct on KAERI Nuclear Training Center, and on using cases of cyber education in domestic and foreign countries. Also realtime training system through the internet and cyber training management system for atomic fields is developed. All users including trainee, course managers and lecturers can use new technical for create new paradigm

[en] An analytical study was performed to investigate the physical phenomena on the bubble behavior during the air discharge into the water pool through the sparger. The simplified numerical model was developed to simulate bubble behavior and the wall response induced by the pressure field. One case of the ABB-ATOM's condensation test was simulated with the condition established based on the test data. The bubble behavior and the pressure response on the wall were analyzed and the major parameters influence the bubble behavior and the pressure response were evaluated. Also, the applicability study of FLUENT code on the analysis of the bubble behavior and pressure response was performed. The select case of the ABB-ATOM's test data was simulated with the Volume of Fluid model of the FLUENT code. (author). 24 refs., 3 tabs., 35 figs

[en] The Korean Next Generation Reactor (KNGR) adopted an advanced design feature of safety depressurization system(SDS) to rapidly de pressurize the reactor coolant system(RCS) in case of beyond design basis events of severe accidents, or a highly unlikely event of a total loss of feedwater (TLOFW) to both steam generators. Two design approaches were considered for the KNGR SDS design. The use of bleed valves similar to those of ABB-CE's system 80+ is design option 1, while in design option 2, the Power Operated Safety Relief valve (POSRV) is considered to provide the combined function of overpressure protection and rapid depressurization. The purpose of this report is to investigate the feasibility of adoption of French SebimPOSRVs for KNGR SDS (design option 2). This report provides the methodology to analyze the TLOFW event with Sebim valves and presents the results of thermal hydraulic analyses using a best-estimate version CEFLASH-4AS/REM for the TLOFW event with feed and bleed. The analyses were performed using a preliminary KNGR design data. For design option 2, if the operator opens two out of the three Sebim valves in conjunction with the four HPSI pumps before a hot leg saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. The results of the present investigation demonstrate that the two design options are both feasible to mitigate the consequences of the TLOFW event with a sufficient margin. (Author). 22 refs., 3 tabs., 19 figs

[en] The Korean Next Generation Reactor (KNGR) adopts 4-train Direct Vessel Injection (DVI) configuration and injects the safety injection water directly into the downcomer through the 8.5'' DVI nozzle. Thus, the thermal hydraulic phenomena such as ECCS mixing and bypass are expected to be different from those observed in the cold leg injection. In order to investigate the realistic injection phenomena and modify the analysis code developed in the basis of cold leg injection, thermal hydraulic test with the performance evaluation is required. Preliminarily, the sequence of events and major thermal hydraulic phenomena during the small break LOCA for KNGR are identified from the analysis results calculated by the CEFLASH-4AS/REM. It is shown from the analysis results that the major transient behaviors including the core mixture level are largely affected by the downcomer modeling. Therefore, to investigate the proper thermal hydraulic phenomena occurring in the downcomer with limited budget and time, the separate effects test focusing on this region is considered to be effective and the conceptual test facility based on this recommended. For this test facility the test initial and boundary conditions are developed using the CEFLASH-4AS/REM analysis results that will be used as input for the preliminary test requirements. The final test requirements will be developed through the further discussions with the test performance group. (Author). 10 refs., 18 tabs., 4 figs