Reactor Protection System Research Articles

Development of dynamic fault tree model for reactor protection system

AbstractDue to the importance of the reactor protection system (RPS) safety function in nuclear power plants (NPPs), its reliability analysis is one of the essential parts of the probabilistic safety assessment (PSA) of any NPP. The dynamic fault tree (DFT) technique extends the conventional static fault tree (SFT) by considering the time requirements to model and evaluate the NPP safety. This study develops the DFT model for RPS to achieve a more realistic and accurate representation of the RPS safety analysis taking into consideration the change in k‐out‐of‐n logic during operation due to test and maintenance (T&M) of any channel. Also, it shows the efficiency of the DFT model over the SFT model with respect to the accuracy, complexity, and calculation time which makes it a useful tool to assess the unavailability of RPS and design a proper maintenance program.

Bayesian belief-based model for reliability improvement of the digital reactor protection system

The digital reactor protection system (RPS) is one of the most important digital instrumentation and control (I&C) systems utilized in nuclear power plants (NPPs). It ensures a safe reactor trip when the safety-related parameters violate the operational limits and conditions of the reactor. Achieving high reliability and availability of digital RPS is essential to maintaining a high degree of reactor safety and cost savings. The main objective of this study is to develop a general methodology for improving the reliability of the RPS in NPP, based on a Bayesian Belief Network (BBN) model. The structure of BBN models is based on the incorporation of failure probability and downtime of the RPS I&C components. Various architectures with dual-state nodes for the I&C components were developed for reliability-sensitive analysis and availability optimization of the RPS and to demonstrate the effect of I&C components on the failure of the entire system. A reliability framework clarified as a reliability block diagram transformed into a BBN representation was constructed for each architecture to identify which one will fit the required reliability. The results showed that the highest availability obtained using the proposed method was 0.9999998. There are 120 experiments using two common component importance measures that are applied to define the impact of I&C modules, which revealed that some modules are more risky than others and have a larger effect on the failure of the digital RPS.

INFORMATION PROCESSING IN THE REACTOR PROTECTION SYSTEMS OF HIGH TEMPERATURE GAS-COOLED REACTORS

Reactor protection systems (RPS) transform process variable signals from the sensors into initiation and actuation signals to trip the reactor if the signal's value exceeds the predefined trip setpoints of the RPS. Information on the current value of the process variables signals and the trip setpoint should be displayed properly on the visual display unit (VDU) in order to maintain the situation awareness of the operators in main control rooms (MCR). In addition, it is also helpful for them to investigate the cause of an accident after the reactor trip and to mitigate the accident based on the appropriate emergency operating procedures. This paper investigates how the information is processed in the RPS of Experimental Power Reactor (EPR) based on high temperature reactor (HTR) technology, and how the information is displayed on the human machine interface (HMI) of the MCR of the EPR. It is conducted by classifying the RPS into three layers based on its components and their functions, followed by the investigation of the type and the information processing in each layer. The results show that the form of the information has been changed throughout the RPS, started from the sensors and until it is displayed on the VDU. The results of the investigation are necessary to understand the concept of RPS, especially for new operators, and to prepare the mitigation actions based on the process variable that cause the reactor trip.Keywords: Experimental power reactor, Reactor protection system, Human machine interface, Information processing, Situation awareness

Dynamic Modeling of Reactor Protection System in Nuclear Power Plant for Reliability Evaluation Based on State Transition Diagram

Reliability assessment of a digital dynamic system using traditional Fault Tree Analysis (FTA) is difficult. This paper addresses the dynamic modeling of safety-critical complex systems such as the digital Reactor Protection System (RPS) in Nuclear Power Plants (NPPs). The digital RPS is a safety system utilized in the NPPs for safe operation and shut-down of the reactor in emergency events. A quantitative evaluation reliability analysis for the digital RPS with 2-out-of-4 architecture using the state transition diagram is presented in this paper. The study assesses the effects of independent hardware failures, Common Cause Failures (CCFs), and software failures on the failure of the RPS through calculating Probability of Failure on Demand (PFD). The results prove the validity of the proposed method in analyzing and evaluating reliability of the digital RPS and also show that the CCFs and longer detection time are the main contributions to the PFD of digital RPS.

Stability and steady state analysis of control and safety systems of Nuclear Power Plants

Control Systems are the first layer of safety in Nuclear Power Plants, the failure of which invokes Safety Systems. The failure of safety systems can lead to radiation exposure to the public. Therefore, it is essential to ensure the stability of such systems. This paper proposes an effective methodology for stability and steady state analysis of control and safety systems of Nuclear Power Plants. The methodology includes Petri net modeling and analysis of its dynamic behavior. Despite of its simplicity, it is very accurate. The application of the proposed techniques is shown and validated on a reactor protection system, which is under operation in six Nuclear Power Plants.

Overcoming the uncertainty in a research reactor LOCA in level-1 PSA; Fuzzy based fault-tree/event-tree analysis

Probabilistic safety assessment (PSA) which plays a crucial role in risk evaluation is a quantitative approach intended to demonstrate how a nuclear reactor meets the safety margins as part of the licensing process. Despite PSA merits, some shortcomings associated with the final results exist. Conventional PSA uses crisp values to represent the failure probabilities of basic events. This causes a high level of uncertainty due to the inherent imprecision and vagueness of failure input data. In this paper, to tackle this imperfection, a fuzzy approach is employed with fault tree analysis and event tree analysis. Thus, instead of using the crisp values, a set of fuzzy numbers is applied as failure probabilities of basic events. Hence, in the fault tree and event tree analysis, the top events and the end-states frequencies are treated as fuzzy numbers. By introducing some fuzzy importance measures the critical components which contribute maximum to the system failure and total uncertainty are identified. As a practical example, under redesign Iranian heavy water research reactor loss of coolant accident is studied. The results show that the reactor protection system has the largest index in sequences lead to a core meltdown. In addition, the emergency core cooling system has a main role in preventing abnormal conditions.

A Software Fault Tree Analysis Technique for Formal Requirement Specifications of Nuclear Reactor Protection Systems

Rigorous safety demonstration of software, as well as systems, is required when developing digital reactor protection systems in nuclear power plants. Various safety analysis techniques try to identify, analyze, and find remedies for potential hazards at each stage of software development life-cycle. This paper proposes a software fault tree analysis technique for software requirements written in the NuSCR formal specification language. The proposed method can mechanically construct software fault trees and calculate minimal cut-sets, encompassing timing constraints of multi-cycles, from NuSCR formal specifications. We also improved the fault tree construction and analysis tool into “NuFTA 2.0” to cope with multi-cycled executions. The case study with a preliminary version of requirements specification for a Korean nuclear power plant in operation shows the proposed technique’s effectiveness and applicability to other V&Vs such as simulation.

Research on Improvement of FQNPP Reactor Protection System Hardware Logic Design

Fuqing Nuclear Power Plant Reactor Protection System has designed emergency control panel ECP, which is used to deal with the risk of loss of digital automatic protection functions due to common software failure. ECP is designed and implemented by analogy technology, which can bypass the computerized DCS system and use buttons, indicator lights, relays and other devices to achieve system-level protection functions. In practical engineering applications, it is found that the current ECP can not meet the requirements of post-accident control in some specific circumstances. Through the analysis of system design and accident management procedures, two design improvements are listed, compared and analysed, and the optimal improvement is given in this paper.

ANALYSIS ON THE PERFORMANCE OF THE BANDUNG CONVERSION FUEL-PLATE TRIGA REACTOR IN STEADY STATE WITH CONSTANT COOLANT FLOW RATE

Bandung TRIGA2000 Reactor, a General Atomic (GA)-made research reactor used for training, research andiIsotope production, has been upgraded to operate at power of 2000 kW using TRIGA fuel rod type. Recently, the TRIGA reactor fuel element producers are going to discontinue the production of TRIGA fuel element. To overcome the unavailability of TRIGA fuel element, BATAN planned to modify TRIGA2000 fuel type from rod-type to U3Si2-Al plate-type fuel with 19.75% enrichment, similar to the domestically fabricated one used in RSG-GAS. The carried out design emphasized on the determination of operation condition limits for setting the reactor protection system in accordance to the reactor safety calculation results. The conceptual design of the innovative fuel plate TRIGA reactor cooling system is expected to remove heat generated by fuels with nominal power of 1 MW up to 2 MW. The design is developed through modelling and safety analysis using COOLOD-N2 validated code. The safety margin is set to its flow instability at transient condition of the fuel plate, which is ≥ 2.38; departure from nucleate boiling ratio ≥1.50; and no onset of nucleate boiling, ΔTONB ≥ 0oC. The primary coolant flow rate accommodating the existing Bandung TRIGA reactor capability is as high as 50 kg/s. The analysis results show that at power of 1 MW, the reactor can safely operate, while at power of 2 MW the safety margin is exceeded. In other words, the plate TRIGA reactor that employs forced convection mode operates safely at 1 MW with excess power 120% of its nominal power.Keywords: 1 MW, Thermalhydraulic design, Steady state condition, TRIGA plate, Constant flowrate

Confluent Modeling of Heterogeneous Safety and Operational Instrumentation and Control Systems

Abstract Individual systems or groups of systems related to the safety and operational instrumentation & control (I&C), and electrical systems are gradually replaced in existing nuclear power plants (NPPs) as part of modernization projects. Modernizations are usually scheduled over multiple years. Different automation platform generations and safety-related product families are deployed to progressively replace legacy systems. Typically, each of the new I&C platforms and products have their own set of engineering tools. One challenge for the modernization of installed systems is the safety and security approval of these heterogeneous systems by regulatory bodies. An NPP consists of systems manufactured by different vendors, which target different application domains, like category A according to IEC 61226 for a reactor protection system, category C for some control room data processing systems or nonclassified for auxiliary systems. Thus, replacing a legacy system by introducing a new one might bring in unknown risks, especially when the new system has more complex interfaces, e.g., when replacing an analog system by a digital system. Moreover, considering the time and budget limitations, commercial-off-the-shelf (COTS) hardware and software are also involved in modernization projects. In contrast to specialized equipment, vulnerabilities for COTS are widespread. On the other hand, existing security measures/mitigations are also required to reflect the system's changes, e.g., mitigations for known vulnerabilities of COTS systems. In order to make an overall and integrated safety analysis after a system change in the frame of a modernization project, it is necessary to jointly consider these I&C systems targeting different application domains, in addition to the physical aggregates, like sensors, pumps, and valves that interact with the physical processes. The restrictive deployment of wireless technologies may also be modeled and analyzed. While wireless is not deployed by legacy systems, it is being covered by new nuclear IEC standards, as some utilities intend to simplify selected I&C maintenance procedures that involve temporary data collection. The key modeling concepts consider new developments in the critical infrastructure and industrial automation domain. With the integrated modeling approach, different disciplines can be addressed, like probabilistic and deterministic safety analyses, security assessments, need for testing and specialized trainings.

ReDICS: 5 years’ operation experiences and its technology

PUSPATI TRIGA Reactor (RTP) is the one and only 1MW research reactor in Malaysia that has been operating since 1982. RTP is used for radioisotope production, various R&D activities and operator training. After more than 3 decades of operations with an analog system, the control console is upgraded into a digital system. Reactor Digital Instrumentation and Control System (ReDICS) is designed, fabricated, manufactured, installed, tested and eventually commissioned in 2014. ReDICS mainly consists of Reactor Protection System (RPS), Operator Work Station (OWS), Data Acquisition and Control System (DACS) and Control Rod Drive Mechanism (CRDM). This paper shares the experiences and performance status throughout the 5 years of ReDICS operation after its commissioning. Besides, technical and operational data of ReDICS are presented and then compared with the old analog I&C system. This paper also describes the instrumentation and measurement technologies in ReDICS as well as the work performed to ensure ReDICS is always functioning at its best and safely.

Real time nuclear power plant operating state cognitive algorithm development using dynamic Bayesian network

In a reactor, various reactor instruments inform the operator of changes in the operating conditions. However, until now, the identification of the reactor state from interpretations of the instrumentation reading has been mostly left to the human operators, which sometimes can be flawed depending on circumstances. Therefore, an artificial cognitive system that can recognize the nuclear reactor state is suggested in this paper to aid the operator under severe environmental stress. Two reactor operating states are preliminarily considered: Normal state and loss of coolant accident state. The artificial cognitive system predicts the state of the reactor in real time, and it determines the type of LOCA if the reactor state is determined to be in an accident state. The proposed system uses only reactor protection system monitoring parameters among various available measurements. The proposed system is composed of two dynamic Bayesian models: Reactor State Determination (RSD) Model, and Accident Type Categorization (ATC) Model. When an accident occurs, the RSD model took 0.3 s to recognize it with 100% accuracy. The average accuracy of the ATC model is about 88%. The average accuracy increased more when the model was refined which suggests the model can be further improved in the future.

Experimental approach to evaluate software reliability in hardware-software integrated environment

Reliability in safety-critical systems and equipment is of vital importance, so the probabilistic safety assessment (PSA) has been widely used for many years in the nuclear industry to address reliability in a quantitative manner. As many nuclear power plants (NPPs) become digitalized, evaluating the reliability of safety-critical software has become an emerging issue. Due to a lack of available methods, in many conventional PSA models only hardware reliability is addressed with the assumption that software reliability is perfect or very high compared to hardware reliability. This study focused on developing a new method of safety-critical software reliability quantification, derived from hardware-software integrated environment testing. Since the complexity of hardware and software interaction makes the possible number of test cases for exhaustive testing well beyond a practically achievable range, an importance-oriented testing method that assures the most efficient test coverage was developed. Application to the test of an actual NPP reactor protection system demonstrated the applicability of the developed method and provided insight into complex software-based system reliability.

A Novel Monitoring System for the Data Integrity of Reactor Protection System Using Blockchain Technology

Nuclear Power Plants (NPPs) are physically isolated from external networks and have different operational environments than conventional information technology (IT) systems. Accordingly, NPPs were regarded as safe from external cyber-attacks. However, it was later determined that isolated networks are not safe from cyber-attacks. Malicious data injection attacks on Programmable Logic Controllers (PLCs) deployed in the safety system of NPPs are critical to nuclear facilities, as they were in the Stuxnet attack. It is necessary to monitor the integrity of PLC data and protect the PLCs from cyber threats such as modification of deployed logic or setpoints. To address this problem, this paper proposes a novel system for monitoring data integrity of PLCs using blockchain technologies. Considering the NPP environment, we developed a private blockchain system to monitor the data integrity of PLCs. The new concept that is Proof of Monitoring (PoM) for data integrity of PLCs was proposed to overcome the limitation for applying the private blockchain to the cybersecurity of NPPs. Additionally, we developed an integrity monitoring system for the Reactor Protection System (RPS)—a safety system in NPPs—using the developed blockchain. It can detect cyber-attacks (such as false code injection attacks on PLCs) and monitor which PLC integrity has been compromised in real-time. A validation experiment using a false data injection attack on PLCs was performed on the developed system, and the results confirmed that the developed system successfully monitored the modification of data in the PLCs.

МАРКОВСЬКІ МОДЕЛІ ОЦІНЮВАННЯ ФУНКЦІОНАЛЬНОЇ БЕЗПЕКИ ПРОГРАМНО-ТЕХНІЧНИХ КОМПЛЕКСІВ НА САМОДІАГНОСТОВНИХ ПРОГРАМОВНИХ ПЛАТФОРМАХ З УРАХУВАННЯМ ПОМИЛОК ЗАСОБІВ КОНТРОЛЮ

Ensuring the safe operation of nuclear power plants remains one of the most important tasks. An important role in ensuring the safety of nuclear power plants is played by instrumentation and control systems (ICS). Extremely high demands are made on such systems, first of all, on their reliability and functional safety. The object of research and analysis in this work is the Nuclear Island I&C Instrumentation System and Reactor Protection System, which are developed based on programmable RadICS Platform with self-diagnostic. The failure trees of such platforms and ICS based on them were developed. In the next stage, several Markov models of redundant ICS are developed. The article aims to develop and analyze the results of research on Markov models for reliability and safety assessment of ICS based on self-checking programmable platforms. The models take into account errors of checking and diagnostic tools, namely errors associated with identifying failures and failures of the corresponding components of the hardware and software/FPGA and channels of redundant structures. Models have been developed for various options of redundant structures and taking into account the diversity principle for ISC structure and failure rate caused by design defects. Multiple-fragment Markov Models of ICS were also investigated. The scientific novelty lies in the fact that the proposed models take into account an expanded set of parameters of self-diagnosing programmable platforms, ICS, the processes of their use and maintenance. Recommendations on the selection of parameters and structural ICS are formulated.

A Novel Weakest t-Norm based Fuzzy Importance Measure for Fuzzy Fault Tree Analysis of Combustion Engineering Reactor Protection System

Recently, a new fuzzy fault tree analysis (FFTA) has been developed to propagate and quantify the epistemic uncertainties occurring in qualitative data such as expert opinions or judgments. It is well known that the weakest triangular norm (Tw) based fuzzy arithmetic operations preserve the shape of the fuzzy numbers, provide more exact fuzzy results and effectively reduce uncertainty range. The objective of this paper is to develop a novel Tw-based fuzzy importance measure to identify the critical basic events in FFTA. The proposed approach has been demonstrated by applying it to a case study to identify the critical components of the Group 1 of the U.S. Combustion Engineering Reactor Protection System (CERPS). The obtained results are then compared to the results computed by the existing well-known importance measures of conventional as well as FFTA. The computed results confirm that the proposed Tw -based importance measure is feasible to identify the critical basic events in FFTA in more exact way.

Study of the PWR ex-core detector response simulation based on the 3D SN method

Simulation of the ex-core detector response function (DRF) is of great importance for reactor power monitoring, control and protection systems. However, the 3D DRF calculation requires vast majorities of computational resources due to its characteristics such as deep penetration, strong anisotropic scattering, large computational geometry, etc. In this paper, the step characteristics (SC) spatial differencing scheme and GMRES algorithm have been implemented in the 3D SN code to ensure the simulation efficient. The adjoint first-collision (FC) source method is developed to mitigate the ray effect in the DRF simulation. The Kobayashi benchmarks are calculated to verify the correctness of the FC module. Numerical results based on the Kori-1 reactor ex-core detector indicate that the SN-FC method has the comparable accuracy with multi-group Monte-Carlo method but achieves much higher efficiency.

Instrumentation & control system for “SAMOP” reactor experimental facility

The instrumentation and control system (ICS) for a subcritical assembly for molly-99 production (SAMOP) reactor experimental facility is presented in this paper. The system is designed based on the defense-in-depth concept. SAMOP reactor is fuelled by uranyl nitrate and driven by an external neutron source from Kartini reactor beam-port. The ICS was designed using 2 neutron detectors of fission chamber and compensated ionization chamber types and provided with a safety rod made from boral used for shutting-down the reactor in emergency condition. The method used for reactor protection system is based on hardwire with a fail-safe principle. The safety rod drive mechanism is designed by using rack and pinion, and provided by the safety button of key release and electromagnetic holder status.

Uncertainties in the Power Alignment at the Jordan Research and Training Reactor

Nuclear research reactors are required to be equipped with an active Reactor Protection System (RPS) that intervenes and shuts down the reactor at pre-set trip set-points, assuring that the reactor operating conditions do not exceed safety limits stated in the reactor Safety Analysis Report (SAR). However, for the RPS to truly prevent the reactor power from exceeding the reactor power safety limits, robust and accurate calibration of the RPS power signals is required.The accuracy of the power alignment at the Jordan Research and Training Reactor (JRTR) was evaluated by estimating sources of uncertainty in the measured thermal power, most significantly induced by assumptions in the thermal power evaluation method, and by the uncertainties in temperature and flow rate measurements.Based on the results of the analyses, a set of recommended modifications to the power alignment procedure and to the Operator Work Stations (OWS) alarm set points are provided to ensure that the RPS satisfies its function of shutting down the reactor without exceeding the safety limits.

Risk-informed approach to the safety improvement of the reactor protection system of the AGN-201K research reactor

Abstract Periodic safety reviews (PSRs) are conducted on operating nuclear power plants (NPPs) and have been mandated also for research reactors in Korea, in response to the Fukushima accident. One safety review tool, the probabilistic safety assessment (PSA), aims to identify weaknesses in the design and operation of the research reactor, and to evaluate and compare possible safety improvements. However, the PSA for research reactors is difficult due to scarce data availability. An important element in the analysis of research reactors is the reactor protection system (RPS), with its functionality and importance. In this view, we consider that of the AGN-201K, a zero-power reactor without forced decay heat removal systems, to demonstrate a risk-informed safety improvement study. By incorporating risk- and safety-significance importance measures, and sensitivity and uncertainty analyses, the proposed method identifies critical components in the RPS reliability model, systematically proposes potential safety improvements and ranks them to assist in the decision-making process.