Main Feedwater System Research Articles

Gear detection and research for hydraulic couplers based on AC electromagnetic field techniques

Abstract To address the issue of fatigue cracks of hydraulic coupler gears in nuclear power plants, we construct an alternating current (AC) electromagnetic field detection system based on stress analysis of gear models. In this model, we design a probe suitable for gear detection in the main feedwater system of nuclear power plants. Moreover, artificial defects are machined into the gears to verify the feasibility of the AC electromagnetic techniques. Results indicate that the AC electromagnetic field techniques can detect the near-surface defects in the test piece with a greater detection depth and better detection effectiveness.

Mechanical and 1st Chemical Cleaning of NEK Steam Generators

Sludge removal is performed on two steam generators (SG’s) at the Krško Nuclear Power Plant (NEK) during every outage. SG’s are a meeting point of five major plant systems: Reactor Coolant System (RC) on the primary side and four systems on the secondary side – Auxiliary Feedwater System (AF), Blowdown System (BD), Main Feedwater System (FW), and Main Steam System (MS). Sludge removal activities take place on the secondary side of the SG’s on the top of the tube sheet. It always consists of classical Sludge Lancing (SL) which is done by spraying water at different angles (30°, 90°, 150°) between the tube gaps in the steam generator tube bundle with a pressure of around 220 bars. Another method is Inner Bundle Lancing (IBL) which means spraying water directly inside the tube bundle with a traveling lance tape with a spraying nozzle at the end. Water is sprayed at an angle or directly on the top of the tube sheet with a robot-guided manipulator which is placed inside a steam generator. The manipulator and therefore the spraying action is controlled by an operator and at times it is fully autonomous to provide the highest protection measures possible. Another method of sludge removal which was for the first time utilized in 2019 at the Krško site was Chemical Cleaning (CC) of both SG’s. During this process, a chemical was injected into the SG’s through the BD system and periodically pumped between the two SG’s to create a dynamic flow and maximize the cleaning effect. To achieve the best results, a constant temperature of the chemicals had to be maintained at all times. Upon completion of chemical cleaning, a rinsing phase was followed to remove any post-treatment chemicals. After all sludge removal activities, a televisual inspection (TVI) of the top of the tube sheet was performed to access the hard sludge area and to search for potential foreign objects in the SG’s. If for instance an object of importance during TV inspection is found, an attempt to retrieve it would usually take place. Other methods of sludge removal such as upper bundle flushing or ultrasonic cleaning have not been implemented in NEK thus far.
 Since the power plant uprate in May 2000, NEK conducted SL on both SG’s every outage also starting with IBL in 2013 and 2015, and the same method was used in the 2018 outage. During the outage in 2019, all three methods (SL, IBL, and CC) have been utilized with the main purpose to extend the full load operation of the plant, preventing and/or stopping denting processes in the SG’s from occurring, reducing and stopping the build-up of hard sludge area to increase/sustain efficiency and remove foreign objects found in the SG’s. SG’s U-tubes are a barrier between the primary side coolant and the secondary side of NEK and the environment. Therefore, it is crucial to keep the highest level of integrity of the U-tubes because any leak could potentially mean a release of radioactive material into the atmosphere. This paper describes the purpose and workflow of sludge removal activities in the outage of 2019 in NEK.

Sludge Removal at Krško NPP Steam Generators during Outage

Sludge removal is performed on two steam generators (SG’s) at the Krško Nuclear Power Plant (NEK) during every outage. SG’s are a meeting point of four major plant systems: Reactor Coolant System (RC) on the primary side and three systems on the secondary side – Auxiliary Feedwater System (AF), Main Feedwater System (FW) and Main Steam System (MS). Sludge removal activities take place on the secondary side of the SG’s on the top of the tube sheet. It consists of classical Sludge Lancing (SL) which is done by spraying water at different angles (30°, 90°, 150°) between the tube gaps in the steam generator tube bundle with a pressure around 220 bars. Another method is Inner Bundle Lancing (IBL) which means spraying water at a much higher pressure (NEK’s contractor reached a pressure of approximately 590 bars). Such water is sprayed directly on the top of the tube sheet with a robot guided lance which is placed inside a steam generator. The robot is controlled by an operator and at times fully autonomous to provide the highest protection measures possible. After these activities, a televisual inspection (TVI) of the top of the tube sheet is performed to access the hard sludge area and to search for potential foreign objects in the SG’s. If an object is found, an attempt to retrieve it would usually take place. Other methods of sludge removal as upper bundle flushing or chemical cleaning have not been implemented in NEK thus far. Since the power plant uprate in May 2000, NEK conducted SL on both SG’s every outage with IBL in 2013 and 2015 and the same method was used in the 2018 outage. The purpose of these activities is mainly to extend the full load operation of the plant, prevent denting processes in the SG’s from occurring, stop the buildup of hard sludge area to increase/sustain efficiency and remove foreign objects found in the SG’s. SG’s U-tubes are a barrier between the primary side coolant and the secondary side of NEK and the environment. Therefore, it is crucial to keep the highest level of integrity of the U-tubes because any leak could potentially mean a release of radioactive material to the atmosphere. This paper describes the purpose and workflow of sludge removal in NEK.

Investigation of Loss of Feedwater (LOFW) Accident in the APR-1400 Using Fault Tree Analysis

Nuclear power plants play a significant role in the contribution of electricity generation on a global scale. Various reactor designs have advantages over others in different aspects. APR-1400 is a pressurized water reactor that is deemed safe due to the redundancy and independence of the multiple safety systems. Probabilistic safety assessment (PSA) is well known for its effectiveness in the representation of risk and safety analysis of the systems in a nuclear power plant. It provides different scenarios of system failure and accident progression via fault tree analysis. A loss of feedwater (LOFW) accident may occur due to numerous reasons such as spurious closure of valves, component failure of heaters, pumps, tanks, or a loss of offsite power (LOOP) event. In the present research, a methodology has been developed that aims to investigate different factors contributing to the loss of feedwater. This research also aims to analyze LOFW accidents by developing fault tree models for the main feedwater system of the APR-1400 to identify the basic events, which may lead to a loss of feedwater accidents. The results of the top event probabilities, risk decrease factor (RDF), risk increase factor (RIF), minimal cut sets (MCS), basic event probabilities, and sensitivity analysis were compared with the WASH-1400 database. It has been found that the control valve (V04) and main feedwater isolation valve (V05) have more contribution to the LOFW accident. The common cause failure (CCF) analysis has been carried out, and it was found that the flow toward the check valve and steam generator are most critical for CCF.

1D-3D coupling investigation of hydraulic transient for power-supply failure of centrifugal pump-pipe system

Abstract In the pumping station, the main feedwater system and the reactor system of nuclear power plant, power-supply failure causes strong hydraulic transients. One-dimensional method of characteristics (1D-MOC) is used to calculate the transient process in the pipeline system while three-dimensional (3D) computational fluid dynamics is employed to analyze the turbulent flows inside the pump and to obtain the performance parameters of the pump, and the data exchanges on the boundary conditions of the shared interface between 1D and 3D domains are updated based on the MpCCI platform. Based on the equation of motion of the pump motion parts, the relationship between the external characteristics and the internal flow field in the pump is further investigated because the dynamic behavior of the pump and the detailed fluid field evolutions inside the pump are captured during the transition process, and the transient flow rate, rotating speed, and pressure inside the impeller are comprehensively investigated. Meanwhile, compared with the data gained by experiment and traditional 1D-MOC, the relative errors of rotating speed and the flow rate obtained by 1D-3D coupling method are smaller than those by 1D-MOC. Furthermore, the influences of the main coupling parameters and coupling modes on the calculation results are analyzed, and the cause of the deviation is further explained.

Fault Diagnosis for AP1000 Plant Feed Water System Based on the Fault Petri Net

Abstract. The feed water system is one of the most important system for nuclear power plant. Based on an analysis for the working principles of the AP1000 plant main feed water system, a novel fault Petri net model and fault diagnosis method are proposed. The main feed water flow low alarm is a serious fault for the feed water system, Petri net theory is adopted to establish the fault model for this fault signal. Through the consequence and account for this model, fault reasons and reasons probability for the main feed water flow low have been obtained. Application shows that, the fault Petri net has more advantages than the fault tree in qualitative analysis and quantitative calculation. And this method is a new route for the nuclear plant systems’ fault diagnosis.

발전소 주급수 계통 감발 과정에서의 스윙체크밸브 닫힘 시점의 실험적 결정

The reliable operation of a swing check valve in the main feed water system of a power plant is most essential for successful shout-down process. A failure to close the valve at proper time often leads to the instability of the main feed water system, or even to an emergency stop of the power plant. In reality it is a very difficult task to monitor the behavior of a swing check valve. Furthermore it is impossible to see the motion of the valve. In this work two measurements were carried out simultaneously to determine the precise valve closure time. The dynamic pressure measurements were made at the inlet and outlet regions of the swing check valve. The transient vibration of the valve housing in the direction of water flow was also measured, which enabled the measurement of the transient vibration of the valve housing near valve closure. By comparing the results produced from these measurements the precise valve closure time could be determined. By carrying out order tracking technique using the dynamic pressure signals and pump rpm signal, the complicated dynamic problems inside the main feed water system can be more easily dealt with. This measurement scheme might be implemented in a power plant on a real-time basis without much difficulty. If this could be implemented, valuable information essential for shut-down operations can readily be passed on to the main control room. The feasibility of this implementation was demonstrated by this experimental work.

A model-based approach to on-line process disturbance management: The models

Abstract A methodology is proposed which can be used to design real-time expert systems for on-line process disturbance management. This methodology encompasses diverse functional aspects that are required for an effective process disturbance management: (1) intelligent process monitoring and alarming, (2) on-line sensor data validation and sensor failure diagnosis, (3) on-line hardware (besides sensors) failure diagnosis, and (4) real-time corrective measure synthesis. Accomplishment of these functions is made possible through the integrated application of the various models: goal-tree success-tree, process monitor tree, sensor failure diagnosis, and hardware failure diagnosis models. This paper presents and discusses the various models along with the overall algorithm of the methodology. The application of the methodology to a target process, a typical main feedwater system of a nuclear power plant which employs a complex control mechanism, will be presented in a companion paper.

A model-based approach to on-line process disturbance management: The application

Abstract A model-based methodology for developing a real-time expert system for on-line process disturbance management has been presented in the companion paper ( Reliab. Engng Sys. Safety , Vol. 28, pp. 265–305). The methodology includes diverse functional aspects required for effective process disturbance management: intelligent process monitoring and alarming, sensor failure diagnosis, hardware (except sensors) failure diagnosis, and corrective measure synthesis. The application of the methodology to a target process—the main feedwater system of a pressurized water reactor (PWR) nuclear plant employing a complex control scheme—is presented in this paper. The performance tests of the real-time expert system, MOAS II, developed by the application of the methodology demonstrate that the expert system successfully carries out its intended functions: early detection of occurring disturbance, correct diagnosis of the disturbance cause, and presentation of optimal control advice to the operator. Therefore, the model-based technique lends itself to the development of a valuable operator aid for on-line process disturbance management.