Reactor Trip System Research Articles

Multi-fragmental and multi-phase availability models of the safety-critical I&C systems with two-cascade redundancy

Traditional availability, reliability, and safety models face the dimension problem due to a huge number of components in modern systems, motivating further research in this field. This paper focuses on multi-fragmental and multiphase models for availability and functional safety assessment of the information and control (I&C) systems with two-cascade redundancy considering design faults manifestation during operation. The methodology of the research is based on Markov and semi-Markov chains with the utilization of multi-phase modeling. Several multi-phase models are developed and investigated considering different conditions of operation and failures caused by version faults. The case study of the research is based on the analysis of safety-critical nuclear power plant I&C systems such as the reactor trip systems developed using the programmable platform RadICS.

Verification and validation optimization method for signal quality bits in digital control system application software of nuclear power plant

Abstract The digitalization enhances nuclear power plant (NPP) instrumentation and control (I&C) system performance, while simultaneously introducing safety and reliability challenges. To enhance the stability of distributed control system (DCS), and minimize the dangerous failure rate and safe failure rate of safety system, signal quality bits (SQBs) are widely used in NPP DCS to identify the validity of important signals. However, in the past decade or so, numerous unscheduled turbine shutdowns and/or reactor trips have resulted from incorrect SQB design or settings. Therefore, it is necessary to conduct optimization research on the design methods and setting principles of SQBs, establish an optimization method for DCS application software SQB verification and validation (V&V), then minimize DCS faults attributed to SQB. With the aim, this work conducts an in-depth investigation and analysis of the definition, design, setting and transmission of SQB in NPP. A specific optimization method for software V&V to conduct SQB on NPP safety DCS application software is proposed. Twelve concerns in the SQB V&V work are detailed. The reactor trip system ΔT protection logic is utilized as a case to illustrate the detail implementation process of this optimization method. This work provides an engineering technical reference and operational guidance for SQB software V&V, conducive to ensuring the safe and economical operation of NPPs.

A study on the uncertainty of setpoint for reactor trip system of NPPs considering rectangular distributions

The setpoint of the reactor trip system shall be set to consider the measurement uncertainty of the instrument channel and provide a reasonable and sufficient margin between the analytical limit and the trip setpoint.A comparative analysis was conducted to find out an appropriate uncertainty combination method through an example problem. The four methods were evaluated; 1) ISA-67.04.01 method, 2) the GUM95 method, 3) the modified GUM method developed by Fotowicz, and 4) the modified IEC61888 method proposed by authors for the pressure instrument channel presented in ISA-RP67.04.02 example. The appropriateness of each method was validated by comparing it with the result of Monte Carlo simulation.As a result of the evaluation, all methods are appropriate when all measurement uncertainty elements are normally distributed as expected. But ISA-67.04 method and GUM95 method overestimated the channel uncertainty if there is a dominant input element with rectangular distribution among the uncertainty input elements.Modified GUM95 methods developed by Fotowicz and modified IEC61888 method by authors are able to produce almost the same level of channel uncertainty as the Monte Carlo method, even when there is a dominant rectangular distribution among the uncertainty components, without computer-assisted simulations.

Design of novel fractional order FPGA based reactor protection and safety controllers for ACP1000 nuclear power plant in LabVIEW

In this research work, an advanced most modern ACP1000 Nuclear Power Plant is addressed. An enhanced fractional order model of ACP1000 nuclear power plant is adopted with an addition of protection and safety systems. The whole plant model is developed by using innovative hybrid technology of Visual Basic, LabVIEW, Fractional Order and Field Programmable Gate Array (FPGA). A reactor trip system is designed and modeled using FPGA technology in LabVIEW. Plant parameters are systematically modeled and panels are designed in LabVIEW for reactor protection controllers. Twenty one reactor trip controllers are designed and modeled based on complex digital logics using FPGA programming in LabVIEW. Two fractional order trip controllers are designed for over temperature protection and over power protection in LabVIEW. FPGA based safety controllers are designed for Engineered Safety Features (ESF) in LabVIEW. For enhanced model of ACP1000 nuclear power plant, 374 systems are modeled in modular form in Visual Basic Environment. Nine process controllers are configured in ANFIS framework in LabVIEW. In this research work, process controllers are used in conjunction with protection and safety controllers using FPGA. The parametric display of simulations is carried out in Visual Basic. The closed loop performance of proposed protection controllers is evaluated under reactor trip and turbine trip while that of safety controllers are evaluated under inadvertent opening of safety valves of pressurizer. Various parameters are simulated for severe transient conditions and the results are evaluated and validated against reference design data and Final Safety Analysis Report (FSAR) of ACP1000 nuclear power plant. All the results are well within the trip and safety systems design bounds under abnormal and severe operating conditions.

A Redundancy-Guided Approach for the Hazard Analysis of Digital Instrumentation and Control Systems in Advanced Nuclear Power Plants

Digital instrumentation and control (I&C) upgrades are a vital research area for the nuclear industry. Despite their performance benefits, deployment of digital I&C in nuclear power plants (NPPs) has been limited. Digital I&C systems exhibit complex failure modes including common cause failures (CCFs), which can be difficult to identify. This paper describes the development of a redundancy-guided application of the Systems-Theoretic Process Analysis and fault tree analysis for the hazard analysis of digital I&C in advanced NPPs. The resulting Redundancy-Guided Systems-Theoretic Hazard Analysis (RESHA) is applied for the case study of a representative state-of-the-art digital reactor trip system. The analysis qualitatively and systematically identifies the most critical CCFs and other hazards of digital I&C systems. Ultimately, the RESHA can help researchers make informed decisions for how, and to what degree, defensive measures such as redundancy, diversity, and defense in depth can be used to mitigate or eliminate the potential hazards of digital I&C systems.

A novel approach for software reliability analysis of digital instrumentation and control systems in nuclear power plants

This work presents a novel, integrated approach to the reliability analysis of digital instrumentation and control systems by incorporating Bayesian belief network (BBN), human reliability analysis (HRA), and common cause failure (CCF) modeling techniques. The Bayesian and HRA-Aided Method for the Reliability Analysis of Software (BAHAMAS) provides consideration of software development life cycle (SDLC) processes and their influence on software reliability. It is assumed that software failures can be traced to human errors in the SDLC, which can be modeled with HRA methods. Additionally, a system’s reliability can be predicted based on how its SDLC quality compares with existing similar systems. A case study demonstrates the quantification of results from a hazard analysis of a digital reactor trip system. The case study shows agreement with values reported in the literature. BAHAMAS is shown to be a flexible tool whose application is designed to conveniently incorporate with conventional probability risk assessments.

Level 1 probabilistic safety assessment of supercritical–CO2–cooled micro modular reactor in conceptual design phase

Micro reactors are increasingly being considered for utilization as distributed power sources. Hence, the probabilistic safety assessment (PSA) of a direct supercritical–CO2–cooled fast reactor, called micro modular reactor (MMR), was performed in this study; this reactor was developed using innovative design concepts. It adopted a modular design and passive safety systems to minimize site constraints. As the MMR is in its conceptual design phase, design weaknesses and valuable safety insights could be identified during PSA. Level 1 internal event PSA was carried out involving literature survey, system characterization, identification of initiating events, transient analyses, development of event trees and fault trees, and quantification. The initiating events and scenarios significantly contributing to core damage frequency (CDF) were determined to identify design weaknesses in MMR. The most significant initiating event category contributing to CDF was the transients with the power conversion system initially available category, owing to its relatively high occurrence frequency. Further, an importance analysis revealed that the safety of MMR can be significantly improved by improving the reliability of reactor trip and passive decay heat removal system operation. The findings presented in this paper are expected to contribute toward future applications of PSA for assessing unconventional nuclear reactors in their conceptual design phases.

Transient Reactor Test (TREAT) Facility Restart Heat Balance Calibration and Rod Worth Measurements

The Transient Reactor Test Facility has been restarted after more than 20 years in a safe standby condition. The plan to bring the reactor back into operation included a typical core characterization that was historically performed every time the core was reconfigured for a new experiment campaign. The core characterization included determining initial critical position of the control rods, a heat balance run for calibration of the nuclear instruments to enable the indication of reactor power, control rod worth measurements, and a series of three temperature-limited transients increasing in the amount of reactivity inserted as a step for the interpolation of set points for the reactor trip system and reactivity insertion limits. The heat balance and control rod worth measurements are discussed in this paper. After critical control rod position was determined, a heat balance operation was used to position the nuclear instruments for correct power indication. This was followed by control rod differential worth measurements to generate the control rod worth curves used by the automatic reactor control system for control of the reactor during transient operations. These restart evolutions are summarized here, and the results are compared to the historic measurements.

Multifragmentation Markov Modeling of a Reactor Trip System

Markov models (MM) are widely used in dependability assessment of complex safety-critical systems. The main computational difficulties in using MMs are model size and stiffness. Selection of the solution approach (SA) and tools based on analysis of MM stiffness and complexity increases the assessment accuracy. This paper presents the safety assessment of nuclear power plan instrumentation and control system (NPP I&Cs): a two-channel FPGA-based reactor trip system with three parallel tracks on “2-out-of-3” logic. The MM was built using a multifragmentation approach and solved with several SAs and tools. The analysis of results shows few application problems: the importance of usability-oriented tool selection, achieving an accurate result, and supporting the results verification.

マルコフ状態遷移モデルの2-out-of-4構成ディジタル原子炉停止系信頼性評価への応用

マルコフ状態遷移モデルの2-out-of-4構成ディジタル原子炉停止系信頼性評価への応用

총채널 불확실도를 적용한 원전 노심출구온도의 운전가능 판정기준

Nuclear power plants are equipped with the reactor trip system (RTS) and the engineered safety features actuation system (ESFAS) to improve safety on the normal operation. In the event of the design basis accident (DBA), a various of post accident monitor(PAM)systems support to provide important details (e.g. Containment pressure, temperature and pressure of reactor cooling system and core exit temperature) to determine action of main control room (MCR). Operator should be immediately activated for the accident mitigation with the information. Especially, core exit temperature is a critical parameter because the operating mode converts from normal mode to emergency mode when the temperature of core exit reaches 649 ℃. In this study, uncertainty which was caused by exterior environment, characteristic of thermocouple/connector and accuracy of calibrator/indicator was evaluated in accordance with ANSI-ISA 67.04. The square root of the sum of square (SRSS) methodology for combining uncertainty terms that are random and independent was used in the synthesis. Every uncertainty that may exist in the hardware which is used to measure the core exit temperature was conservatively applied and the associative relation between the elements of uncertainty was considered simultaneously. As a result of uncertainty evaluation, the channel statistical allowance (CSA) of single channel of core exit temperature was +1.042%Span. The range of uncertainty,╶0.35%Span (╶4.05℃) ~ +2.08%Span(24.25℃), was obtained as the operating criteria of

ATWS analysis for Maanshan PWR using TRACE/SNAP code

The objective of this study was to use TRACE code to analyze the reactor coolant system pressure transients under LONF ATWS for Maanshan PWR. According to Westinghouse Anticipated Transients Without Trip report, LONF ATWS was considered as the most severe plant condition. The ASME Code Level C service limit criteria of 22.06MPa (3200psig) was assumed to be an unacceptable plant condition in SECY-83-293. In accordance with 10 CFR 50.62, Maanshan NPP has established AMSAC. Unlike the reactor trip system, it automatically initiates the auxiliary feedwater system and turbine trip under conditions indicating an ATWS. Since the ATWS analysis is not specified in the FSAR (Taiwan Power Company, 1982), TRACE code was used to assess the RCS pressure for Maanshan NPP. Analysis results indicate that RCS pressure can be maintained within 22.06MPa with sufficient negative moderator temperature coefficient and normal work of AMSAC and valves.

Transient Analysis of Depressurized Loss-of-Forced-Circulation Accident Without Scram in High-Temperature Gas-Cooled Reactor

The transient response of the high-temperature gas-cooled reactor (HTGR) to depressurized loss of forced circulation combined with failure of all reactor trip systems, a beyond-design-basis accident, is analyzed for an extended period of time during which no active core cooling is resumed. The characteristic behavior of the reactor during the long-term conduction cooldown event is found to be shaped by several parameters that are usually not considered in the safety design of the HTGR. For example, while the Doppler effect is usually relied upon to provide inherent shutdown of the reactor, the reactivity coefficient of temperature of the graphite moderator is found to be a critical parameter for determining the final settling temperature of the fuel following the recriticality. Furthermore, this study finds that the peak fuel temperature reached during this event is correlated strongly even to the initial core operating temperature prior to the initiation of the transient event. These and other results of this study are expected to provide useful input to the development of enhanced safety design guidelines for commercial HTGRs in the aftermath of the Fukushima accident.

Sensitivity analysis on the effect of software-induced common cause failure probability in the computer-based reactor trip system unavailability

The reactor trip system has been digitized in advanced nuclear power plants, since the programmable nature of computer based systems has a number of advantages over non-programmable systems. However, software is still vulnerable to common cause failure (CCF). Residual software faults represent a CCF concern, which threat the implemented achievements.This study attempts to assess the effectiveness of so-called defensive strategies against software CCF with respect to reliability. Sensitivity analysis has been performed by re-quantifying the models upon changing the software failure probability. Importance measures then have been estimated in order to reveal the specific contribution of software CCF in the trip failure probability.The results reveal the importance and effectiveness of signal and software diversity as applicable strategies to ameliorate inefficiencies due to software CCF in the reactor trip system (RTS). No significant change has been observed in the rate of RTS failure probability for the basic software CCF greater than 1×10−4. However, the related Fussell–Vesley has been greater than 0.005, for the lower values.The study concludes that consideration of risk associated with the software based systems is a multi-variant function which requires compromising among them in more precise and comprehensive studies.

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.

China: CSRC — carbon black

The objective of this study was to use TRACE code to analyze the reactor coolant system pressure transients under LONF ATWS for Maanshan PWR. According to Westinghouse Anticipated Transients Without Trip report, LONF ATWS was considered as the most severe plant condition. The ASME Code Level C service limit criteria of 22.06 MPa (3200 psig) was assumed to be an unacceptable plant condition in SECY-83-293. In accordance with 10 CFR 50.62, Maanshan NPP has established AMSAC. Unlike the reactor trip system, it automatically initiates the auxiliary feedwater system and turbine trip under conditions indicating an ATWS. Since the ATWS analysis is not specified in the FSAR (Taiwan Power Company, 1982), TRACE code was used to assess the RCS pressure for Maanshan NPP. Analysis results indicate that RCS pressure can be maintained within 22.06 MPa with sufficient negative moderator temperature coefficient and normal work of AMSAC and valves.

Microprocessor-Based Testing Increases Power Plant Availability

Commercial nuclear power plants need equipment that increases plant availability, reduces test time and is both self-diagnostic and user-friendly. Westinghouse's Process Protection Upgrade System, called Eagle 21?, is a functional replacement for existing analog process protection equipment used to monitor nuclear power stations and to initiate the actuation of the reactor trip and engineering safeguards systems. Periodic testing of this protection equipment is required by industry regulations, and the advent of microprocessor systems has made it possible for plant test personnel to do this required testing in a short time, with great precision, and with few or no errors. The architecture of the testing system is described and the major benefits are highlighted.

A Microprocessor Tester for the Treat Upgrade Reactor Trip System

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.

Microprocessor-Based Architecture for a Modular Process Protection Upgrade System

The Westinghouse Process Protection Upgrade System (Eagle 21TM) is a functional replacement for the existing analog process protection equipment used to monitor nuclear generating stations and actuate the reactor trip and engineering safeguards systems as required. However, the Eagle 21TM system does more than merely replace the existing analog hardware; it exploits the power of a functionally decentralized architecture with multiple independent microprocessorbased subsystems to increase plant availablity, reduce test time, reduce reconfiguration time, and increase the amount of information available to the operator and to new diagnostic systems. The architecture of the system is described and the major benefits are highlighted.