• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3275-3285
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• 2021

A Nuclear Power Plant (NPP) is a complex dynamic system-of-systems with highly nonlinear behaviors. In order to control the plant operation under both normal and abnormal conditions, the different systems in NPPs (e.g., the reactor core components, primary and secondary coolant systems) are usually monitored continuously, resulting in very large amounts of data. This situation makes it possible to integrate relevant qualitative and quantitative knowledge with artificial intelligence techniques to provide faster and more accurate behavior predictions, leading to more rapid decisions, based on actual NPP operation data. Data-driven models (DDM) rely on artificial intelligence to learn autonomously based on patterns in data, and they represent alternatives to physics-based models that typically require significant computational resources and might not fully represent the actual operation conditions of an NPP. In this study, a feed-forward backpropagation artificial neural network (ANN) model was trained to simulate the interaction between the reactor core and the primary and secondary coolant systems in a pressurized water reactor. The transients used for model training included perturbations in reactivity, steam valve coefficient, reactor core inlet temperature, and steam generator inlet temperature. Uncertainties of the plant physical parameters and operating conditions were also incorporated in these transients. Eight training functions were adopted during the training stage to develop the most efficient network. The developed ANN model predictions were subsequently tested successfully considering different new transients. Overall, through prompt prediction of NPP behavior under different transients, the study aims at demonstrating the potential of artificial intelligence to empower rapid emergency response planning and risk mitigation strategies.

• Nuclear Engineering and Technology
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• v.16 no.1
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• pp.11-17
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• 1984

Sample calculations were done for KORI-1 to develop a better understanding of what happens after very small LOCA ($\leq$0.05 ft$^2$). For a water-side break with the break size larger than 0.006 ft$^2$, fluid-loss through break exceeds the makeup. If the break size is larger than 0.008ft$^2$, decay heat can be completely removed through break. Based on these results, it was concluded that KORI-1 is fairly safe for the whole spectrum of sizes in very small LOCA. However, for the reactor with 900 MWe or 1200 MWe, a certain spectrum of sizes in very small LOCA should be carefully considered. In the accident sequence the transition from natural circulation to pool boiling or from pool boiling to natural circulation may be troublesome to the operator or in the safety analysis. Operator's intervention was discussed; primary pump shutoff, HPI pump shutoff, break isolation, and opening relief valve. It was proved that continuous operation of HPI pumps after shutdown will not threaten the integrity of the primary system.

• Journal of Energy Engineering
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• v.6 no.2
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• pp.170-175
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• 1997

An experimental study was carried out with the Mock-up made for the improved water level control of the steam generator in nuclear power plant and for the confirmation of swell/shrink status of the water level by opening and closing the steam dump valve. We can confirmed the possibility of using the Mock-up by introducing the PI controller and the FUZZY controller. Accordingly, we can confirmed that the practical usability of advanced controllers, which will be developed for the improved water level control of the steam generator in nuclear power plant by using the Mock-up.

• Proceedings of the KSME Conference
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• 2001.11b
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• pp.735-740
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• 2001

In the nuclear power plant, emergency core coolant system(ECCS) is furnished at reactor coolant system(RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, it occurs thermal stratification phenomena in case that there is the mixing of cooling water and high temperature water due to valve leakage in ECCS. This thermal stratification phenomena raises excessive thermal stresses at pipe wall. Therefore, this phenomena causes the accident that reactor coolant flows in reactor containment in the nuclear power plant due to the deformation of pipe and thermal fatigue crack(TFC) at the pipe wall around the place that it exists. Hence, in order to fundamental identification of this phenomena, it requires the experimental research of modeling test in the pipe flow that occurs thermal stratification phenomena. So, this paper models RCS and ECCS pipe arrangement and analyzes the mechanism of thermal stratification phenomena by measuring of temperature in variance with leakage flow rate in ECCS modeled pipe and Reynold number in RCS modeled pipe. Besides, results of this experiment is compared with computational analysis which is done in advance.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.311-318
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• 1992

During mid-loop operation of Nuclear Power Plant, to prevent the Decay Heat Removal System (DHRS) from failure due to air entrainment of free surface vortex in the piping system, a set of simulating experiments was performed. Through these experiments, a relation between the non-dimensionalized numbers, such as H/d, Froude number, Reynolds number, was found. It was also found that the perturbation of the system by the disturbance such as pump start, valve operation, etc., has a strong effect on the free surface vortex. Furthermore, from viewpoint of reactor safety, a modified inlet device which is reducer type is strongly recommended for the prevention of air entrainment into DHRS.

• Analytical Science and Technology
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• v.37 no.2
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• pp.123-129
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• 2024

We present a rapid method for the determination of Cs, Sr, and Ba, heat generators found in highly active liquid wastes, by gas-pressurized extraction chromatography (GPEC) using a column containing a cation-exchange resin. GPEC is a microscale column chromatographic technique that uses a constant flow rate of solvent (0.07 mL/min) with pressurized nitrogen gas supplied through a valve. In particular, because this method uses a small sample volume (a few hundred microliters), it produces less chemical waste and allows for faster separation compared to traditional column chromatography. In this study, we evaluated the separation of Cs, Sr, and Ba using GPEC. The eluate from the column (GPEC or conventional column chromatography) was quantitatively analyzed using inductively coupled plasma-mass spectrometry to measure the column recovery and precision. The column reproducibility of the proposed GPEC system (RSDs of recoveries) ranged from 2.7 to 4.1 %, and the column recoveries for the three elements ranged from 72 to 98% when aqueous HCl was used as the eluent. The GPEC results are slightly different in efficiency and separation resolution compared to those of conventional column chromatography because of the differences in the eluent flow rate as well as the internal diameter and length of the column. However, the two methods had similar recoveries for Cs and Sr, and the precision of GPEC was improved by two-fold. Remarkably, the solvent volume required for GPEC analysis was five times lower than that of the conventional method, and the total analysis time was 11 times shorter.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.880-885
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• 2024

For Safety Assisted Engineering works, real-time simulators have emerged as a mandatory tool among all the key actors involved in the nuclear industry (utilities, designers and safety authorities). EDF, Electricité de France, as the leading worldwide nuclear power plant operator, has a crucial need for efficient and updated simulation tools for training, operating and safety analysis support. This paper will present the work performed at EDF/DT to develop a new generation of engineering simulator to fulfil these tasks. The project is called SiRENE, which is the acronym of Re-hosted Engineering Simulator in French. The project has been economically challenging. Therefore, to benefit from existing tools and experience, the SiRENE project combines: - A part of the process issued from the operating fleet training full-scope simulator. - An improvement of the simulator prediction reliability with the integration of High-Fidelity models, used in Safety Analysis. These High-Fidelity models address Nuclear Steam Supply System code, with CATHARE thermal-hydraulics system code and neutronics, with COCCINELLE code. - And taking advantage of the last generation and improvements of instructor station. The intensive and challenging uses of the new SiRENE engineering simulator are also discussed. The SiRENE simulator has to address different topics such as verification and validation of operating procedures, identification of safety paths, tests of I&C developments or modifications, tests on hydraulics system components (pump, valve etc.), support studies for Probabilistic Safety Analysis (PSA). etc. It also emerges that SiRENE simulator is a valuable tool for self-training of the newcomers in EDF nuclear engineering centers. As a modifiable tool and thanks to a skillful team managing the SiRENE project, specific and adapted modifications can be taken into account very quickly, in order to provide the best answers for our users' specific issues. Finally, the SiRENE simulator, and the associated configurations, has been distributed among the different engineering centers at EDF (DT in Lyon, DIPDE in Marseille and CNEPE in Tours). This distribution highlights a strong synergy and complementarity of the different engineering institutes at EDF, working together for a safer and a more profitable operating fleet.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.7 s.250
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• pp.707-714
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• 2006

A numerical analysis study has performed in terms of fluid dynamics to identify the wall thinning generated in the main feedwater isolation valve body of a nuclear power plant. To review the relations between flow characteristics and the wall thinning induced by flow accelerated corrosion (FAC), numerical analysis using FLUENT code and ultrasonic tests (UT) were performed. The local velocities according to the analysis results were compared with the distribution of the measured wall thickness by ultrasonic tests. The comparison results show that the local velocity in the x-direction had no correlation with the wall thinning but the local velocity in the y-direction and turbulence intensity had a great influence on that. These results provide a good match to those of the previous studies - locations colliding vertically against components undergo severe wall thinning. These results may be utilized to the design modification and the wall thinning management for main feedwater isolation valves for preventing the wall thinning degradation.

• Journal of Advanced Marine Engineering and Technology
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• v.36 no.5
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• pp.662-673
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• 2012

This study introduces a condition diagnosis technique for a turbine governor system. The governor system is an important control system to handle turbine speed in a nuclear power plant. The turbine governor system includes turbine valves and stop valves which have their own functions in the system. Because a turbine governor system is operated by high oil pressure, it is very difficult to maintain under stable operating conditions. Turbine valves supply oil pressure to the governor system for proper operation. Using the pressure variation of turbine and governor valves, operating conditions of the turbine governor control system are detected and identified. To achieve automatic detection of valve status, time-based and frequency-based analysis is employed. In this study, a new approach, wavelet decomposition, was used to extract specific features from the pressure signals of the governor and stop valves. The extracted features, which represent the operating conditions of the turbine governor system, include important information to control and diagnose the valves. After extracting the specific features, decision rules were used to classify the valve conditions. The rules were generated by a decision tree algorithm (a typical simple method for data-based rule generation). The results given by the wavelet-based analysis were compared to detection results using time- and frequency-based approaches. Compared with the several related studies, the wavelet transform-based analysis, the proposed in this study has the advantage of easier application without auxiliary features.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.54-60
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• 2010

Piping failures due to thermal fatigue have been widely reported in normally stagnant non-isolable reactor coolant branch lines. Since the thermal fatigue due to thermal stratification was not considered in the piping fatigue design in old NPPs, it is important to evaluate the effect of thermal stratification on the integrity of branch lines. In this study, geometrical screening criteria for Up-horizontal branch lines in MRP-132 were applied to SI(Safety Injection) lines of KSNP 2-loop and WH 3-loop. Some computational fluid dynamic(CFD) analyses on the Reactor Coolant System(RCS) branch lines were also performed to develop the regulatory guidelines for screening criteria. As a result of applying MRP-132 screening criteria, KSNP 2-loop and WH 3-loop SI lines are determined to need further detailed evaluation. Results of CFD analyses show that both valve isolation and amount of leakage through valve can be used as technical bases for the screening criteria on the thermal fatigue analysis.