Kori Nuclear Power Plant Unit Research Articles

Simplified Assessment Methodology of Radionuclide Contamination for Subsurface Soil

Determining whether to release a site after decommissioning a nuclear facility should be preceded by an environmental impact assessment of the exposure radiation dose according to the radionuclides in the soil. Currently, in Korea, various evaluation methodologies and decommissioning technologies are being studied for the first decommissioning of nuclear power plants, starting with Kori Nuclear Power Plant Unit (Kori-1), which is based on the “Multi-Agency Radiation Survey and Site Investigation Manual MARSSIM” developed in the United States. The scope and evaluation targets of deep soil may differ depending on the purpose, but it has been confirmed that the International Atomic Energy Agency and the U.S. Nuclear Regulatory Commission are targeting subsurface soil. MARSSIM outlines the need for an evaluation of this subsurface soil but does not suggest specific methods. In NUREG-1757, which complements MARSSIM, it is confirmed that subsurface soil specifically means a soil layer that is 15 to 30 cm deep in the surface layer. In the current study, using the previously verified computational code RESidual RADioactivity (RESRAD)-ONSITE, a methodology for summation is proposed to evaluate the impact of subsurface soil more flexibly and realistically while minimizing the exposure dose evaluation procedure. When using RESRAD-ONSITE according to this evaluation methodology, it was confirmed that it is possible to respond to changes in the depths of various soil layers. In addition, it was also confirmed that this methodology is adaptable to the contamination of nuclides, such as 60Co, 137Cs, 152Eu, and 154Eu, which are expected to be major nuclides when decommissioning nuclear facilities.

Comparison of occupational exposure according to dismantling strategy of Kori nuclear power plant unit #1 bio-shield

In this study, to evaluate the occupational exposure during dismantling of the bio-shield of the Kori nuclear power plant unit #1, the strategy and process of dismantling the concrete of the bio-shield were analyzed. Two strategies (cutting activated concrete prior to non-activated concrete and vice-versa) were analyzed with a focus on the cutting process, which differs significantly depending on the dismantling strategy, and additional costs related to the cutting equipment and disposal process were calculated accordingly. VISIPLAN was used to evaluate occupational exposure during cutting, and the exposure dose was assessed using the ALARA principle. The strategy where activated concrete was cut prior to non-activated concrete was evaluated to be 47% more inexpensive than the other strategy while satisfying the annual dose limit for workers in economic terms.

Occupational internal exposure calculation during dismantling of bio-shield of Kori nuclear power plant unit#1

In this study, a dismantling scenario was developed, and the internal dose of a worker was estimated during bio-shield dismantling of Kori nuclear power plant unit#1 for radiological safety assessment internal exposure of workers to bio-shield dismantling. The internal dose was estimated from the analysis of radioactivity concentration of five main radionuclides (including 3H, 55Fe, 60Co, 152Eu, and 154Eu) of radioactive concrete using MCNP and the cutting scenario of a dismantling worker was designed. The internal dose of a worker during Diamond Wire Saw(DWS) cutting was calculated as 1.44 mSv for dry cutting and 5.03 μSv for wet cutting. The cutting time for dry cutting was 1.43 times higher and the air release fraction was approximately 200 times larger compared to wet cutting. Most of the calculated exposures were based on work conducted in a highly contaminated area; this work is expected to be completed within 2000 h. Therefore, the estimated results could be directly compared to the annual dose limits. The results of this study would be useful for the preparation of dismantling work procedures of the bio-shield of Kori nuclear power plant unit#1 in accordance with radiation safety measures.

PRINCIPAL COMPONENTS BASED SUPPORT VECTOR REGRESSION MODEL FOR ON-LINE INSTRUMENT CALIBRATION MONITORING IN NPPS

In nuclear power plants (NPPs), periodic sensor calibrations are required to assure that sensors are operating correctly. By checking the sensor's operating status at every fuel outage, faulty sensors may remain undetected for periods of up to 24 months. Moreover, typically, only a few faulty sensors are found to be calibrated. For the safe operation of NPP and the reduction of unnecessary calibration, on-line instrument calibration monitoring is needed. In this study, principal component-based auto-associative support vector regression (PCSVR) using response surface methodology (RSM) is proposed for the sensor signal validation of NPPs. This paper describes the design of a PCSVR-based sensor validation system for a power generation system. RSM is employed to determine the optimal values of SVR hyperparameters and is compared to the genetic algorithm (GA). The proposed PCSVR model is confirmed with the actual plant data of Kori Nuclear Power Plant Unit 3 and is compared with the Auto-Associative support vector regression (AASVR) and the auto-associative neural network (AANN) model. The auto-sensitivity of AASVR is improved by around six times by using a PCA, resulting in good detection of sensor drift. Compared to AANN, accuracy and cross-sensitivity are better while the auto-sensitivity is almost the same. Meanwhile, the proposed RSM for the optimization of the PCSVR algorithm performs even better in terms of accuracy, auto-sensitivity, and averaged maximum error, except in averaged RMS error, and this method is much more time efficient compared to the conventional GA method.

Design of a Software Sensor for Feedwater Flow Measurement Using a Fuzzy Inference System

Venturi meters are used to measure the feedwater flow rate in most current pressurized water reactors. These meters can decrease the thermal performance of nuclear power plants because the feedwater flow rate can be overmeasured due to their fouling phenomena that make corrosion products caused by long-term operation accumulate in the feedwater flow meters. Therefore, in this paper, a software sensor using a fuzzy inference system is developed in order to increase the thermal efficiency by accurately estimating online the feedwater flow rate. The fuzzy inference system to be used for black-box modeling of the feedwater system is equipped with an automatic design algorithm that automates the selection of the input signals to the fuzzy inference system and its fuzzy rule generation including parameter optimization. The proposed algorithm was verified by using the numerical simulation data of the MARS code for Kori Nuclear Power Plant Unit 1 and also the real plant data of Yonggwang Nuclear Power Plant Unit 3. In the simulations using numerical simulation data and real plant data, the relative 2σ errors and the relative maximum error are small enough. The proposed method can be applied successfully to validate and monitor the existing feedwater flow meters.

High-Level Radiation Field Analysis Using Monte Carlo Simulation in Kori Nuclear Power Plant Unit 1

Radiation field analysis in nuclear power plant (NPP) was performed by using Monte Carlo simulation against whole containment building. In this work, the neutron energy spectra were calculated in the high-level radiation field of Kori Unit 1 in Korea, which is to be a basis of dose assessment for NPP workers. The Monte Carlo calculation was divided into two steps, neutron source term generation, and neutron energy spectrum calculation, to improve the calculation efficiency. The neutron energy spectra were, in this work, obtained at six points where the NPP workers should access. Even though the results are not compared with measurement data, it is noted that this approach can be employed for radiation field analysis. In the near future, a comparison between calculation results and measurement data will be carried out.

Development of integrated automatic diagnosis method for loose parts monitoring system

Generally, it is known that loose parts in the reactor coolant systems (RCS) bring serious damage into the system components and impede the normal function of the system. So, it is necessary to rapidly respond when the impact event has occurred. This paper presents a realization of automatic diagnosis algorithm for LPMS (Loose Parts Monitoring System) and application results to the impact test data at YongGwang Nuclear Power Plant Unit 3 (YGN3), Kori Nuclear Power Plant Unit 4 (KNU4) and the real data at YongGwang Nuclear Power Plant Unit 1 (YGN1). The integrated diagnosis algorithm is composed of three parts; prefiltering, impact location and mass estimation. The prefiltering is needed to reject low frequency background noises. To estimate the impact location, the starting points of impact are detected from the filtered signals and compared to produce the time differences, and then the modified triangular method is applied. To estimate the mass and energy of a loose part, we automatically compute the maximum amplitude and the initial half period. Additionally, a modified impact theory considering amplitude and energy attenuation effects is applied.To show the effectiveness of the proposed diagnostic method, the real impact test data at YGN3, KNU4 and the real impact data at YGN 1 is used. The analysis results show that the location estimation error is on average below 7.5%, and the average mass estimation is within 40%.

Development of Software for the Microsimulator for the KO-RI Nuclear Power Plant Unit 2

A workstation-based real-time simulator for two-loop pressurized water reactor plants is developed for classroom training in support of a full-scale simulator, on-site transient analysis, and engineering studies. The present simulator consists of three functional modules: plant module, graphic module, and man-machine interaction module. The plant module includes models for the core kinetics, reactor coolant system, steam generator, main steam line, balance of plant, and control and protection system. Each of the models is optimized to obtain the capability of real-time simulation. The graphic module is designed to provide the user with more information at a glance by dynamically displaying schematic diagrams of the systems, symbols indicating the operating status of each component, trend curves, and the main control room. As tools for the man-machine interface, the man-machine interaction model uses a color cathode ray tube monitor, a standard keyboard, and the mouse. The interactive communication module receives parameters from the user via the keyboard and mouse, and transfers them to the plant module so as to enable the user to perform his specific actions. This module provides the user with various initiating events (malfunctions and manual controls) through SYSTEM, CONTROL ROOM, and ACCIDENTS menus, and thus a wide range ofmore » nuclear steam supply system transients can be easily simulated. The FISA-2/WS is verified through comparisons with analytical solutions, separated tests and integral tests, and predictions by RETRAN-2 and RELAP5/MOD3.« less