Digital Plant Protection System Research Articles

The Development Process of Human Machine Interface of Plant Protection System of a Small Modular Reactor

The Plant Protection System (PPS), which consists of Reactor Protection Systems (RPS) and Engineered Safety Actuated Systems (ESFAS), is one of the most important safety systems in nuclear reactors, including Small Modular Reactors (SMRs). The RPS generates a signal to trip the reactor if the measured reactor parameters exceed the trip setpoint, and then the ESFAS is actuated to mitigate the consequences of the accident by minimizing fuel damage and radioactivity release into the environment. Therefore, a comprehensive Human-Machine Interface (HMI) is essential for monitoring and controlling the PPS to ensure its reliability and enhance the operators' situational awareness. This study discusses the development process of the HMI for the digital PPS of an SMR. In this study, various standards, guidance, and design criteria for PPS and HMI are incorporated and applied to ensure that the proposed design meets the required level of reliability. In the first stage, the proposed design is intended for assessing the functionality and reliability of the PPS. Moreover, in the future, it will play an essential role in the design phase of the HMI for the PPS of an SMR in Indonesia.

Plant Risk Effect Analysis Focusing on Digital I&C Equipment Failures

Safety-critical digital systems have been installed in nuclear power plants and thus their safety effect evaluation has become an emerging issue. The multi-tasking feature of digital instrumentation and control (I&C) equipment could increase the risk factor because the I&C equipment affects the actuation of the safety functions in several mechanisms. In this study, we quantify the safety of the digital plant protection system in Korean nuclear power plants based on probabilistic safety assessment (PSA) technology. Fifteen fault-tree models for the digital reactor-trip system and seven for the safety-feature actuation system are constructed and integrated into the plant safety assessment model. The result of the sensitivity study shows the boundaries of a plant risk and the effect of the digital equipment failures on the total plant risk.

Plant Risk Effect Analysis Focusing on Digital I&C Equipment Failures

Safety-critical digital systems have been installed in nuclear power plants and thus their safety effect evaluation has become an emerging issue. The multi-tasking feature of digital instrumentation and control (I&C) equipment could increase the risk factor because the I&C equipment affects the actuation of the safety functions in several mechanisms. In this study, we quantify the safety of the digital plant protection system in Korean nuclear power plants based on probabilistic safety assessment (PSA) technology. Fifteen fault-tree models for the digital reactor-trip system and seven for the safety-feature actuation system are constructed and integrated into the plant safety assessment model. The result of the sensitivity study shows the boundaries of a plant risk and the effect of the digital equipment failures on the total plant risk.

Evaluation of error detection coverage and fault-tolerance of digital plant protection system in nuclear power plants

Recently, traditional analog-based safety-related instrumentation and control (I&C) systems in nuclear power plants (NPPs) have been replaced with modern digital-based systems. Due to the digitalization of nuclear I&C systems, the safety assessment has become a major issue, as it is crucial to the system’s reliability. In the safety assessment of the digitalized system, evaluation of error detection coverage and fault-tolerance are critical factors. For the evaluation, we use C++ based hardware description instead of a board with integrated circuit components. We select the digital plant protection system (DPPS) in NPPs as a target system. Permanent fault is used as a possible fault in the system and some error detection methods are used to detect errors. From the experiment, we confirmed that the proposed approach can evaluate the error detection coverage and the fault-tolerance of DPPS in NPPs.

A method for evaluating fault coverage using simulated fault injection for digitalized systems in nuclear power plants

The fault coverage for digital system in nuclear power plants is evaluated using a simulated fault injection method. Digital systems have numerous advantages, such as hardware elements share and hardware replication of the needed number of independent channels. However, the application of digital systems to safety-critical systems in nuclear power plants has been limited due to reliability concerns. In the reliability issues, fault coverage is one of the most important factors. In this study, we propose an evaluation method of the fault coverage for safety-critical digital systems in nuclear power plants. The system under assessment is a local coincidence logic processor for a digital plant protection system at Ulchin nuclear power plant units 5 and 6. The assessed system is simplified and then a simulated fault injection method is applied to evaluate the fault coverage of two fault detection mechanisms. From the simulated fault injection experiment, the fault detection coverage of the watchdog timer is 44.2% and that of the read only memory (ROM) checksum is 50.5%. Our experiments show that the fault coverage of a safety-critical digital system is effectively quantified using the simulated fault injection method.

A formal software requirements specification method for digital nuclear plant protection systems

This article describes NuSCR, a formal software requirements specification method for digital plant protection system in nuclear power plants. NuSCR improves the readability and specifiability by providing graphical or tabular notations depending on the type of operations. NuSCR specifications can be formally analyzed for completeness, consistency, and against the properties specified in temporal logic. We introduce the syntax and semantics of NuSCR and demonstrate the effectiveness of the approach using reactor protection system, digital protection system being developed in Korea, as a case study.

Korean Standard Nuclear Power Plant Response to Common Mode Failure of the Digital Plant Protection System

The Korean Standard Nuclear Power Plant (KSNP) has adopted the digital plant protection system (PPS) to enhance the reliability and safety of the plant. Although the digital PPS can be designed with high reliability, it is considered to be vulnerable to common mode failure (CMF) in the system software, resulting in a total loss of the built-in hardware redundancy. Therefore, a comprehensive evaluation has been performed to demonstrate the intrinsic capability of the KSNP design in coping with the design-basis events concurrent with CMF in the digital PPS. Instead of the conservative bounding analysis methodology, a best-estimate analysis methodology has been developed and utilized since the design-basis events accompanied by CMF in the digital PPS are categorized as beyond-design-basis events. An additional reactor trip function on high containment pressure in the diverse protection system (DPS), which is totally diverse from the PPS and is not affected by the CMF in the digital PPS, has been proposed to meet the acceptance criteria of the evaluation results. A variety of diverse means such as the DPS, process control systems, and operator actions including design modification have been verified to be effective in mitigating the design-basis events with CMF in the digital PPS.

Defense-in-depth and diversity evaluation to cope with design bases events concurrent with common mode failure in digital plant protection system for KNGR

The Korean Next Generation Reactor (KNGR) has been evolved to adopt an advanced design feature, a digital Plant Protection System (PPS) as an effort of enhancing reliability and safety of the plant. Although the digital PPS can be designed with high reliability, it is considered to be vulnerable to the Common Mode Failure (CMF) in the system software resulting in a total loss of the built-in hardware redundancy. Therefore, a comprehensive evaluation has been performed to demonstrate the intrinsic capability of the KNGR design in coping with the design basis events concurrent with CMF in the digital PPS. Instead of the conservative bounding analysis methodology, a best-estimate analysis methodology has been developed and utilized since the design basis events accompanied by CMF in the digital PPS are categorized as beyond design bases events. A variety of diverse means such as Alternate Protection System (APS), process control systems, and timely operator actions have been verified to be effective in mitigating the design basis events with CMF in the digital PPS.