The Application of FPGA for Anticipated Transients Without Scram Mitigation System

2010 ◽  
Author(s):  
Mao-Sheng Tseng ◽  
Hui-Wen Huang ◽  
Ming-Huei Chen ◽  
Tsung-Chieh Cheng ◽  
Hsiang-Han Chung ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Method ◽  
Computer Code ◽  
Test Facility ◽  
Software Faults ◽  
Control Rod ◽  
Field Programmable ◽  
Important Means ◽  
And Control

The digitalized Instrumentation and Control (I&C) system of nuclear power plants (NPP) could provide operator easily Human-Machine Interface (HMI) and more powerful overall operation capability. However, some software errors may cause a kind of Common Cause Failure (CCF). As a consequence, the event of Anticipated Transients Without Scram (ATWS) will occur. In order to assure that the plant can be shutdown safely and to follow the requirements of 10CFR50.62, the utility builds up various ATWS mitigation features in NPP. The features include Fine Motion Control Rod Drive Run In, Alternate Rod Insertion, Standby Liquid Control System, Reactor Internal Pump Trip or Runback, Feedwater Flow Runback and Inhibition of Automatic Depressurization System. This research developed an evaluation method of diverse back-up means for computerized I&C system. A diverse backup of digital I&C system is the most important means to defend against CCF and un-detectable software faults. Institute of Nuclear Energy Research (INER) is developing a computerized I&C test facility, which is incorporated a commercial grade I&C systems with Personal Computer Transient Analyzer (PCTran)/Advanced Boiling Water Reactor (ABWR), a NPP simulation computer code. By taking the technology of Field Programmable Gate Array (FPGA) to implement the methods of ATWS mitigation, the research built up a diverse back-up of digital I&C system to expect to defend against CCF and undetectable software faults. According to the testing and evaluation, the work can be achieved the analysis of Diversity and Defense-in-Depth (D3).

FPGA Technology and Platforms for NPP I&C systems

2020 ◽  
pp. 419-457
Author(s):  
Andriy Kovalenko ◽  
Ievgen Babeshko ◽  
Viktor Tokarev ◽  
Kostiantyn Leontiiev
Keyword(s):  
Field Programmable Gate Array ◽  
Nuclear Power ◽  
Power Plants ◽  
Critical Systems ◽  
Field Programmable ◽  
Assessment Techniques ◽  
Safety Critical Systems ◽  
New Generation ◽  
And Control ◽  
Element Base

This chapter describes an element base of new generation for NPP I&C, namely field programmable gate array (FPGA), and peculiarities of the FPGA application for designing safety critical systems. FPGA chips are modern complex electronic components that have been applied in nuclear power plants (NPPs) instrumentation and control systems (I&CSs) during the last 15-17 years. The advantages and some risks caused by application of the FPGA technology are analyzed. Safety assessment techniques of FPGA-based I&CSs and experience of their creation are described. The FPGA-based platform RadICS and its application for development of NPP I&CS is described.

A Review of the Current State of FPGA Systems in Nuclear Instrumentation and Control

2013 ◽  
Author(s):  
Phillip McNelles ◽  
Lixuan Lu
Keyword(s):  
Analog Circuits ◽  
Integrated Circuit ◽  
Nuclear Power ◽  
Power Plants ◽  
Future Research ◽  
Extensive Literature ◽  
Pressurized Water ◽  
Field Programmable ◽  
Water Reactors ◽  
And Control

A Field Programmable Gate Array, or FPGA, is a form of integrated circuit that is programmed (configured) after it has been built. These devices have recently become a topic of interest for various applications in the nuclear field. Most of the recent work put into these FPGA systems is for the purpose of Instrumentation and Control (I&C) systems, but other applications include health physics, particle detectors, and pulse measurement systems. These new FPGA based systems are thought of as possible replacements for older, analog systems that are commonly used in Nuclear Power Plants (NPPs). Many of these systems are becoming obsolete, and it can be difficult to repair and maintain them. FPGAs possess certain advantages over traditional analog circuits, as well as microprocessors, for nuclear I&C applications. This paper provides an extensive literature survey on the current research into FPGA-based systems in NPP applications, such as shutdown systems, neutron monitors, and feedwater controls. Current plans and plans for future FPGA implementations are also discussed. Research from different countries in North America, Europe and Asia is discussed, covering a variety of NPP types (CANDU, Pressurized Water Reactors, Boiling Water Reactors, etc.). The main companies and organizations involved in the FPGA research and development are examined, and a direction for future research is presented.

Application of Uncertainty Analysis in the Comparison of Void Fraction Calculations With Experiment

2010 ◽  
Author(s):  
Jason Pascoe ◽  
Yuksel Parlatan ◽  
B. McLaughlin ◽  
Sophia Fung
Keyword(s):  
Void Fraction ◽  
Nuclear Power ◽  
Power Plants ◽  
Computer Code ◽  
Scientific Basis ◽  
Output Parameter ◽  
Test Facility ◽  
Model Parameters ◽  
Uncertainty Range ◽  
Test Loop

Safety analysis computer codes are designed to simulate phenomena relevant to the assessment of normal and transient behaviour in nuclear power plants. In order to do so, models of relevant phenomena are developed and a set of such models constitutes a computer code. In accident or transient analysis the values of certain output parameters (margin parameters) are used to characterize the severity of the event. The accuracy of the computer code in calculating these margin parameters is usually obtained through validation and variation in the margin parameter is estimated through the propagation of variation in the code input. A method for estimating code uncertainty respect to a specific output parameter has been developed. The methodology has the following basic elements: (1) specification and ranking of phenomena that govern the behaviour of the output parameter for which an uncertainty range is required; (2) identification of models within the code that represent the relevant phenomena; (3) determination of the governing parameters for the phenomenological models and Identification of uncertainty ranges for the governing model parameters from validation or scientific basis, if available; (4) decomposition of the governing model parameters into related parameters; (5) identification of uncertainty ranges for the modelling parameters for use in Best Estimate Analysis; (6) design and execution of a case matrix; and (7) estimation of the code uncertainty through quantification of the variability in output parameters arising from uncertainty in modelling parameters. This methodology has been employed using simulations of Large Break Loss of Coolant Accident (LOCA) tests in the RD-14M test facility to calculate the uncertainty in the TUF thermal hydraulics code calculation of the coolant void fraction. The uncertainty has been estimated with and without plant parameters (parameters specific to the RD-14M test loop). The TUF coolant void fraction uncertainty without plant parameters was determined to be 0.08 while the uncertainty with plant parameters included was determined to be 0.11. The uncertainty value without plant parameters included is comparable to the uncertainty in the measurements (0.09). The uncertainty value with plant parameters included is larger than the variation in the bias (0.10) of the TUF calculation of void fraction. From these findings, it can be concluded that the estimated accuracy of the TUF code calculation of void fraction is consistent with the available experimental data.

Dynamic Flowgraph Methodology Assessment of an FPGA-Based Postaccident Monitoring System for Westinghouse AP1000 Nuclear Power Plants

2015 ◽  
Vol 1 (3) ◽  
Author(s):  
Phillip McNelles ◽  
Lixuan Lu ◽  
Marc-James Abi-Jaoude
Keyword(s):  
Monitoring System ◽  
Integrated Circuit ◽  
Nuclear Power ◽  
Power Plants ◽  
Design Parameters ◽  
Reactor Accident ◽  
Nuclear Plants ◽  
Field Programmable ◽  
And Control ◽  
Fine Tune

A field-programmable gate array (FPGA) is a type of integrated circuit that is programmed after being manufactured. These FPGA-based systems are considered to be viable alternatives to replace many obsolete instrumentation and control (I&C) systems that are used in nuclear plants. This paper describes an FPGA-based lab-scale implementation of a postaccident monitoring system (PAMS), for a Westinghouse AP1000 reactor. This system will monitor vital parameters in the event of a serious reactor accident. The system reliability was analyzed using the dynamic flowgraph methodology (DFM). DFM was applied to fine-tune the design parameters by determining the potential causes of faults in the design.

Cybersecurity Case for FPGA-Based NPP Instrumentation and Control Systems

2016 ◽  
Author(s):  
Oleg A. Illiashenko ◽  
Yevheniia V. Broshevan ◽  
Vyacheslav S. Kharchenko
Keyword(s):  
Control Systems ◽  
Nuclear Power ◽  
Power Plants ◽  
Subjective Assessment ◽  
Security Requirements ◽  
Critical Systems ◽  
International Regulations ◽  
Field Programmable ◽  
High Level ◽  
And Control

Modern industrial instrumentation and control systems (I&Cs) used in nuclear power plants (NPP) are facing more with cybersecurity threats and vulnerabilities, which were neglected before. Cybersecurity incidents are a subject to grow into more complex attacks with worse consequences than before. The use of field programmable gate arrays (FPGA) in such critical systems causes specific risks for ensuring of safety, as the master-property of such kind of systems, and security as a subordinate property primarily to the NPP reactor trip systems (RTS). Cybersecurity assessment results of industrial I&Cs are mainly based on subjective assessment of the expert judgment and they do not take into account all features of propagating FPGA technology. Nowadays there is a big gap in understanding how to assess and assure the security of FPGA-based NPP I&Cs (FNI&Cs). Conformance of FNI&Cs to security requirements, their verification to high-level standards often is subjective and depends on particular expert. Regulatory and certification bodies, developers and end-users of FNI&Cs are missing the understandable methodology for security assurance of such kind of systems taking into account specific context of the operating environment which allows decreasing time-to-market and thus providing benefits for all interested parties. The paper describes cybersecurity assurance technique of multi-version FNI&Cs. Requirements profile is formulated using the best practices from the following international regulations. The goal of the paper is presentation of the case-based methodology and tool of FNI&Cs cybersecurity assurance based on international regulations. Proposed methodology provides comparable and repeatable process of assurance.

The reactor research and test facility

2020 ◽  
pp. 69-79
Author(s):  
Andrey S. KIRILLOV ◽  
Aleksandr P. PYSHKO ◽  
Andrey A. ROMANENKO ◽  
Valery I. YARYGIN
Keyword(s):  
Measurement System ◽  
Nuclear Power ◽  
Power Plants ◽  
Test Facility ◽  
Physical Parameters ◽  
Pumping System ◽  
Current State ◽  
History Of ◽  
Special Measurement ◽  
Space Nuclear Power

The paper describes an overview of the history of development and the current state of JSC “SSC RF-IPPE” reactor research and test facility designed for assembly, research and full-scale life energy tests of space nuclear power plants with a thermionic reactor. The leading specialists involved in development and operation of this facility are represented. The most significant technological interfaces and upgrade operations carried out in the recent years are discussed. The authors consider the use of an oil-free pumping system as part of this facility during degassing and life testing. Proposed are up-to-date engineering solutions for development of the automated special measurement system designed to record NPP performance, including volt-ampere characteristics together with thermophysical and nuclear physical parameters of a ground prototype of the space nuclear power plant. Key words: reactor research and test facility, thermionic reactor, life energy tests, oil-free pumping system, automated special measurement system, volt-ampere characteristics.

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