[en] This paper describes the design progress for data management and communication networks to be co-operated as subsystems in KALIMER MMIS. Main functions and design bases are being established and validated for functional modules of these subsystems. Real-time data acquisition and signal validation, databases, and data logging have been designed as each functional module of data management while data interfaces of communication networks have been designed with the system information from Top-Tier Requirements for KALIMER MMIS. The conceptual design shall be refined through the iterative and detailed one

[en] A feasibility study, which standard PC hardware and Real-Time Linux are applied to real-time computer simulation of software for a nuclear simulator, is presented in this paper. The feasibility prototype was established with the existing software in the Compact Nuclear Simulator (CNS). Throughout the real-time implementation in the feasibility prototype, we has identified that the approach can enable the computer-based predictive simulation to be approached, due to both the remarkable improvement in real-time performance and the less efforts for real-time implementation under standard PC hardware and Real-Time Linux envrionments

[en] In CPC(Core Protection Calculator) of CE-type nuclear power plants, the core axial power distribution is calculated to evaluate the safety-related parameters. The accuracy of the CPC axial power distribution highly depends on the quality of the so called shape annealing matrix(SAM). Currently, SAM is determined by using data measured during startup test and used throughout the entire cycle. An issue concerned with SAM is that it is fairly sensitive to measurements and thus the fidelity of SAM is not guaranteed for all cycles. In this paper, a novel method to determine a high-performance SAM (HPSAM) is proposed, where both measured and simulated data are used in determining SAM

[en] This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place

[en] Programmable Logic Device (PLD), especially Complex PLD (CPLD) or Field Programmable Logic Array (FPGA), has been growing in interest in nuclear Instrumentation and Control (I and C) applications. PLD has been applied to replace an obsolete analog device or old-fashioned microprocessor, or to develop digital controller, subsystem or overall system on hardware aspects. This is the main reason why the PLD-based I and C design provides higher flexibility than the analog-based one, and the PLD-based I and C systems shows better real-time performance than the processor-based I and C systems. Due to the development of the PLD-based I and C systems, their nuclear qualification has been issued in the nuclear industry. Verification and Validation (V and V) is one of necessary qualification activities when a Hardware Description Language (HDL) is used to implement functions of the PLD-based I and C systems. The life cycle V and V process, described in this paper, has been defined as satisfying the nuclear V and V requirements, and it has been applied to verify Correctness, Completeness, and Consistency (3C) among design outputs in a safety-grade programmable logic controller and a safety-critical data communication system. Especially, software engineering techniques such as the Fagan Inspection, formal verification, simulated verification and automated testing have been defined for the life cycle V and V tasks of behavioral, structural, and physical design in VHDL

[en] The preliminary standard design of Advanced Power Reactor 1000 (APR1000), which uses a two-loop 1000 MWe pressurized water reactor (PWR), has been developed. The APR1000 incorporates a variety of ADFs (advanced design features) from the based model of the OPR1000. 24-month cycle operation and 30% core loading of MOX (Mixed Oxide, PuO₂-UO₂) are major ADFs of APR1000, and they have been analyzed to confirm their feasibility in this study. Through the analysis, it has been confirmed that the APR1000 core designs for 24-month cycle operation and 1/3 MOX fuel loading are effective to meet the requirements on core design criteria

[en] This study was approached to reuse source programs for a nuclear simulator under PC with Open Source Software(OSS) and to extend its applicability. Source programs in the Compact Nuclear Simulator(CNS), which has been operated for institutional research and training in KAERI, were reused and implemented for Linux-PC environment with the aim of supporting the study. PC with 500 MHz processor and Linux 2.2.5-22 kernel were utilized for the reuse implementation and it was investigated for some applications, through the functional testing for its main functions as interfaced with compact control panels in the current CNS. Development and upgrade of small-scale simulators, establishment of process simulation for PC, and development of prototype predictive simulation, can effectively be enabled with the experience though the reuse implementation was limited to port only CNS programs for PC with Linux

[en] To improve the computational accuracy of core axial power shapes in COLSS (Core Operating Limit Supervisory System) of ABB-CE reactors, a new method using extra 4 pseudo-detector signals to evaluate axial power shapes was proposed and tested for Younggwang Nuclear Unit (YGN) 3 cycle 3 and YGN 4 cycle 4. To find optimal correlation between each pseudo-detector signal and 5 real detector signals, the Alternating Conditional Expectation (ACE) algorithm was used. And the conventional Fourier fitting method was adopted to calculate 20-node axial power shapes with 9-detector information. To verify the usefulness of new method, a total of 3462 axial power shapes per each cycle produced by ROCS (Reactor Operation and Control Simulation) code were recalculated by different axial power shape reconstruction methods. The results were compared with those of the existing Fourier fitting method and stochastic method using the ACE algorithm. The average Root Means Square (RMS) error and average of axial peaking, DFZ, error of the proposed 9-detector method shows about 50% and 70 reduction, respectively, relative to the existing 5-detecotor method. Because the proposed 9-detector method, compared with the stochastic power prediction method, has no restriction on expanding from 20-node shape to 40- or 50-node axial power shape, it may be an useful method for precise reconstructing of axial power shapes when only 5-detector information are available

[en] CNS(Compact Nuclear Simulator) was developed at the end of 1980s, and have been used as training simulator for staffs of KAERI during 10 years. The operator panel interface cards and the graphic interface cards were designed with special purpose only for CNS. As these interface cards were worn out for 10 years, it was very difficult to get spare parts and to repair them. And the interface cards were damaged by over current happened by shortage of lamp in the operator panel. To solve these problem, the project 'Improvement of Compact Nuclear Simulator' was started from 1997. This paper only introduces about the improvement of computer complex and interface system

[en] In process instrumentation systems of such as nuclear plants, response time information is very important in most temperature transient measurements. Generally the response time of thermocouples is measured at a laboratory by using a plunge method. However, it is not easy to use the plunge testing method when a response time measurement of an installed thermocouple is required. A measurement system was developed to measure the response time of a thermocouple installed in a process by using the Loop Current Step Response(LCSR) testing method. This device heats a thermocouple by providing an electrical current, and then it measures the thermocouple output as the temperature of the thermocouple measurement junction returns to ambient temperature. The time constant of the thermocouple is determined from the transient curve of the thermocouple output indicating the temperature difference between the reference junction and measurement junction of the thermocouple. The device is designed to heat a middle point to reduce the temperature error caused by residual heat of thermocouple wire