safety control systems: Topics by WorldWideScience.org

Sample records for safety control systems

FOOD SAFETY CONTROL SYSTEM IN CHINA

Institute of Scientific and Technical Information of China (English)

Liu Wei-jun; Wei Yi-min; Han Jun; Luo Dan; Pan Jia-rong

2007-01-01

Most countries have expended much effort to develop food safety control systems to ensure safe food supplies within their borders. China, as one of the world's largest food producers and consumers,pays a lot of attention to food safety issues. In recent years, China has taken actions and implemented a series of plans in respect to food safety. Food safety control systems including regulatory, supervisory,and science and technology systems, have begun to be established in China. Using, as a base, an analysis of the current Chinese food safety control system as measured against international standards, this paper discusses the need for China to standardize its food safety control system. We then suggest some policies and measures to improve the Chinese food safety control system.
Safety-related control air systems

International Nuclear Information System (INIS)

Anon.

1977-01-01

This Standard applies to those portions of the control air system that furnish air required to support, control, or operate systems or portions of systems that are safety related in nuclear power plants. This Standard relates only to the air supply system(s) for safety-related air operated devices and does not apply to the safety-related air operated device or to air operated actuators for such devices. The objectives of this Standard are to provide (1) minimum system design requirements for equipment, piping, instruments, controls, and wiring that constitute the air supply system; and (2) the system and component testing and maintenance requirements
Quantitative safety assessment of air traffic control systems through system control capacity

Science.gov (United States)

Guo, Jingjing

Quantitative Safety Assessments (QSA) are essential to safety benefit verification and regulations of developmental changes in safety critical systems like the Air Traffic Control (ATC) systems. Effectiveness of the assessments is particularly desirable today in the safe implementations of revolutionary ATC overhauls like NextGen and SESAR. QSA of ATC systems are however challenged by system complexity and lack of accident data. Extending from the idea "safety is a control problem" in the literature, this research proposes to assess system safety from the control perspective, through quantifying a system's "control capacity". A system's safety performance correlates to this "control capacity" in the control of "safety critical processes". To examine this idea in QSA of the ATC systems, a Control-capacity Based Safety Assessment Framework (CBSAF) is developed which includes two control capacity metrics and a procedural method. The two metrics are Probabilistic System Control-capacity (PSC) and Temporal System Control-capacity (TSC); each addresses an aspect of a system's control capacity. And the procedural method consists three general stages: I) identification of safety critical processes, II) development of system control models and III) evaluation of system control capacity. The CBSAF was tested in two case studies. The first one assesses an en-route collision avoidance scenario and compares three hypothetical configurations. The CBSAF was able to capture the uncoordinated behavior between two means of control, as was observed in a historic midair collision accident. The second case study compares CBSAF with an existing risk based QSA method in assessing the safety benefits of introducing a runway incursion alert system. Similar conclusions are reached between the two methods, while the CBSAF has the advantage of simplicity and provides a new control-based perspective and interpretation to the assessments. The case studies are intended to investigate the
Safety-related control air systems - approved 1977

International Nuclear Information System (INIS)

Anon.

1978-01-01

This standard applies to those portions of the control air system that furnish air required to support, control, or operate systems or portions of systems that are safety related in nuclear power plants. This standard relates only to the air supply system(s) for safety-related air operated devices and does not apply to the safety-related air operated device or to air operated actuators for such devices. The objectives of this standard are to provide (1) minimum system design requirements for equipment, piping, instruments, controls, and wiring that constitute the air supply system; and (2) the system and component testing and maintenance requirements
Development of digital safety system logic and control

International Nuclear Information System (INIS)

Nishikawa, H.; Sakamoto, H.

1995-01-01

Advanced-BWR (ABWR) uses total digital control and instrumentation (C and I) system. In particular, ABWR adopts a newly developed safety system using advanced digital technology. In the presentation the digital safety system design, manufacturing and factory validation test method are shortly overviewed. The digital safety system consists of micro-processor based digital controllers, data and information transmission by optical fibers and human-machine interface using color flat displays. This new developed safety system meet the nuclear safety requirements such as high reliability, independence of divisions, operability and maintainability. (2 refs., 4 figs., 1 tab.)
Safety implications of control systems

International Nuclear Information System (INIS)

Smith, O.L.

1983-01-01

The Safety Implications of Control Systems Program has three major activities in support of USI-A47. The first task is a failure mode and effects analysis of all plant systems which may potentially induce control system disturbance that have safety implications. This task has made a preliminary study of overfill events and recommended cases for further analysis on the hybrid simulator. Work continues on overcooling and undercooling. A detailed investigation of electric power network is in progress. LERs are providing guidance on important failure modes that will provide initial conditions for further simulator studies. The simulator taks is generating a detailed model of the control system supported by appropriate neutronics, hydraulics, and thermodynamics submodels of all other principal plant components. The simulator is in the last stages of development. Checkout calculations are in progress to establish model stability, robustness, and qualitative credibility. Verification against benchmark codes and plant data will follow
Instrumentation and control systems important to safety in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2005-01-01

This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. It supplements Safety Standards Series No. NS-R-1: Safety of Nuclear Power Plants: Design (the Requirements for Design), which establishes the design requirements for ensuring the safety of nuclear power plants. This Safety Guide describes how the requirements should be met for instrumentation and control (I and C) systems important to safety. This publication is a revision and combination of two previous Safety Guides: Safety Series Nos 50-SG-D3 and 50-SG-D8, which are superseded by this new Safety Guide. The revision takes account of developments in I and C systems important to safety since the earlier Safety Guides were published in 1980 and 1984, respectively. The objective of this Safety Guide is to provide guidance on the design of I and C systems important to safety in nuclear power plants, including all I and C components, from the sensors allocated to the mechanical systems to the actuated equipment, operator interfaces and auxiliary equipment. This Safety Guide deals mainly with design requirements for those I and C systems that are important to safety. It expands on paragraphs of Ref in the area of I and C systems important to safety. This publication is intended for use primarily by designers of nuclear power plants and also by owners and/or operators and regulators of nuclear power plants. This Safety Guide provides general guidance on I and C systems important to safety which is broadly applicable to many nuclear power plants. More detailed requirements and limitations for safe operation specific to a particular plant type should be established as part of the design process. The present guidance is focused on the design principles for systems important to safety that warrant particular attention, and should be applied to both the design of new I and C systems and the modernization of existing systems. Guidance is provided on how design
Safety Management System in Croatia Control Ltd.

OpenAIRE

Pavlin, Stanislav; Sorić, Vedran; Bilać, Dragan; Dimnik, Igor; Galić, Daniel

2009-01-01

International Civil Aviation Organization and other international aviation organizations regulate the safety in civil aviation. In the recent years the International Civil Aviation Organization has introduced the concept of the safety management system through several documents among which the most important is the 2006 Safety Management Manual. It treats the safety management system in all the segments of civil aviation, from carriers, aerodromes and air traffic control to design, constructi...
Spallation Neutron Source Accelerator Facility Target Safety and Non-safety Control Systems

International Nuclear Information System (INIS)

Battle, Ronald E.; DeVan, B.; Munro, John K. Jr.

2006-01-01

The Spallation Neutron Source (SNS) is a proton accelerator facility that generates neutrons for scientific researchers by spallation of neutrons from a mercury target. The SNS became operational on April 28, 2006, with first beam on target at approximately 200 W. The SNS accelerator, target, and conventional facilities controls are integrated by standardized hardware and software throughout the facility and were designed and fabricated to SNS conventions to ensure compatibility of systems with Experimental Physics Integrated Control System (EPICS). ControlLogix Programmable Logic Controllers (PLCs) interface to instruments and actuators, and EPICS performs the high-level integration of the PLCs such that all operator control can be accomplished from the Central Control room using EPICS graphical screens that pass process variables to and from the PLCs. Three active safety systems were designed to industry standards ISA S84.01 and IEEE 603 to meet the desired reliability for these safety systems. The safety systems protect facility workers and the environment from mercury vapor, mercury radiation, and proton beam radiation. The facility operators operated many of the systems prior to beam on target and developed the operating procedures. The safety and non-safety control systems were tested extensively prior to beam on target. This testing was crucial to identify wiring and software errors and failed components, the result of which was few problems during operation with beam on target. The SNS has continued beam on target since April to increase beam power, check out the scientific instruments, and continue testing the operation of facility subsystems
Safety Metrics for Human-Computer Controlled Systems

Science.gov (United States)

Leveson, Nancy G; Hatanaka, Iwao

2000-01-01

The rapid growth of computer technology and innovation has played a significant role in the rise of computer automation of human tasks in modem production systems across all industries. Although the rationale for automation has been to eliminate "human error" or to relieve humans from manual repetitive tasks, various computer-related hazards and accidents have emerged as a direct result of increased system complexity attributed to computer automation. The risk assessment techniques utilized for electromechanical systems are not suitable for today's software-intensive systems or complex human-computer controlled systems.This thesis will propose a new systemic model-based framework for analyzing risk in safety-critical systems where both computers and humans are controlling safety-critical functions. A new systems accident model will be developed based upon modem systems theory and human cognitive processes to better characterize system accidents, the role of human operators, and the influence of software in its direct control of significant system functions Better risk assessments will then be achievable through the application of this new framework to complex human-computer controlled systems.
Operation safety of control systems. Principles and methods

International Nuclear Information System (INIS)

Aubry, J.F.; Chatelet, E.

2008-01-01

This article presents the main operation safety methods that can be implemented to design safe control systems taking into account the behaviour of the different components with each other (binary 'operation/failure' behaviours, non-consistent behaviours and 'hidden' failures, dynamical behaviours and temporal aspects etc). To take into account these different behaviours, advanced qualitative and quantitative methods have to be used which are described in this article: 1 - qualitative methods of analysis: functional analysis, preliminary risk analysis, failure mode and failure effects analyses; 2 - quantitative study of systems operation safety: binary representation models, state space-based methods, event space-based methods; 3 - application to the design of control systems: safe specifications of a control system, qualitative analysis of operation safety, quantitative analysis, example of application; 4 - conclusion. (J.S.)
A new radiation safety control system for Ganil

International Nuclear Information System (INIS)

Saint Jores, P. De; Luong, T.T.; Martina, L.; Vega, G.

1991-01-01

A second generation radiation safety control system has been installed to upgrade the initial system which was not flexible enough to support new ion beams and new experimental conditions required by the accelerator operation. The main reasons which necessitated the improvement of the safety control system are presented. The new system which controls the Ganil accelerator from the first quarter of 1990 is described. It uses a star structured architecture, VME standard processors and front-end modules activated by pDOS operating system and high level language (C and Fortran) tasks, associated with enhanced resolution color displays for real time synoptics. (R.P.) 4 refs., 4 figs
Safety Evaluation Approach with Security Controls for Safety I and C Systems on Nuclear Power Plants

International Nuclear Information System (INIS)

Kim, D. H.; Jeong, S. Y.; Kim, Y. M.; Park, H. S.; Lee, M. S.; Kim, T. H.

2016-01-01

This paper addresses concepts of safety and security and relations between them for assessing effects of security features in safety systems. Also, evaluation approach for avoiding confliction with safety requirements and cyber security features which may be adopted in safety-related digital I and C system will be described. In this paper, safety-security life cycle model based confliction avoidance method was proposed to evaluate the effects when the cyber security control features are implemented in the safety I and C system. Also, safety effect evaluation results using the proposed evaluation method were described. In case of technical security controls, many of them are expected to conflict with safety requirements, otherwise operational and managerial controls are not relatively. Safety measures and cyber security measures for nuclear power plants should be implemented not to conflict with one another. Where safety function and security features are both required within the systems, and also where security features are implemented within safety systems, they should be justified
Safety Evaluation Approach with Security Controls for Safety I and C Systems on Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Kim, D. H.; Jeong, S. Y.; Kim, Y. M.; Park, H. S. [KINS, Daejeon (Korea, Republic of); Lee, M. S.; Kim, T. H. [Formal Works Inc., Seoul (Korea, Republic of)

2016-05-15

This paper addresses concepts of safety and security and relations between them for assessing effects of security features in safety systems. Also, evaluation approach for avoiding confliction with safety requirements and cyber security features which may be adopted in safety-related digital I and C system will be described. In this paper, safety-security life cycle model based confliction avoidance method was proposed to evaluate the effects when the cyber security control features are implemented in the safety I and C system. Also, safety effect evaluation results using the proposed evaluation method were described. In case of technical security controls, many of them are expected to conflict with safety requirements, otherwise operational and managerial controls are not relatively. Safety measures and cyber security measures for nuclear power plants should be implemented not to conflict with one another. Where safety function and security features are both required within the systems, and also where security features are implemented within safety systems, they should be justified.
Safety regulations concerning instrumentation and control systems for research reactors

International Nuclear Information System (INIS)

El-Shanshoury, A.I.

2009-01-01

A brief study on the safety and reliability issues related to instrumentation and control systems in nuclear reactor plants is performed. In response, technical and strategic issues are used to accomplish instrumentation and control systems safety. For technical issues there are ; systems aspects of digital I and C technology, software quality assurance, common-mode software, failure potential, safety and reliability assessment methods, and human factors and human machine interfaces. The strategic issues are the case-by-case licensing process and the adequacy of the technical infrastructure. The purpose of this work was to review the reliability of the safety systems related to these technical issues for research reactors
Segmentation Scheme for Safety Enhancement of Engineered Safety Features Component Control System

International Nuclear Information System (INIS)

Lee, Sangseok; Sohn, Kwangyoung; Lee, Junku; Park, Geunok

2013-01-01

Common Caused Failure (CCF) or undetectable failure would adversely impact safety functions of ESF-CCS in the existing nuclear power plants. We propose the segmentation scheme to solve these problems. Main function assignment to segments in the proposed segmentation scheme is based on functional dependency and critical function success path by using the dependency depth matrix. The segment has functional independence and physical isolation. The segmentation structure is that prohibit failure propagation to others from undetectable failures. Therefore, the segmentation system structure has robustness to undetectable failures. The segmentation system structure has functional diversity. The specific function in the segment defected by CCF, the specific function could be maintained by diverse control function that assigned to other segments. Device level control signals and system level control signals are separated and also control signal and status signals are separated due to signal transmission paths are allocated independently based on signal type. In this kind of design, single device failure or failures on signal path in the channel couldn't result in the loss of all segmented functions simultaneously. Thus the proposed segmentation function is the design scheme that improves availability of safety functions. In conventional ESF-CCS, the single controller generates the signal to control the multiple safety functions, and the reliability is achieved by multiplication within the channel. This design has a drawback causing the loss of multiple functions due to the CCF (Common Cause Failure) and single failure Heterogeneous controller guarantees the diversity ensuring the execution of safety functions against the CCF and single failure, but requiring a lot of resources like manpower and cost. The segmentation technology based on the compartmentalization and functional diversification decreases the CCF and single failure nonetheless the identical types of controllers
Segmentation Scheme for Safety Enhancement of Engineered Safety Features Component Control System

Energy Technology Data Exchange (ETDEWEB)

Lee, Sangseok; Sohn, Kwangyoung [Korea Reliability Technology and System, Daejeon (Korea, Republic of); Lee, Junku; Park, Geunok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

Common Caused Failure (CCF) or undetectable failure would adversely impact safety functions of ESF-CCS in the existing nuclear power plants. We propose the segmentation scheme to solve these problems. Main function assignment to segments in the proposed segmentation scheme is based on functional dependency and critical function success path by using the dependency depth matrix. The segment has functional independence and physical isolation. The segmentation structure is that prohibit failure propagation to others from undetectable failures. Therefore, the segmentation system structure has robustness to undetectable failures. The segmentation system structure has functional diversity. The specific function in the segment defected by CCF, the specific function could be maintained by diverse control function that assigned to other segments. Device level control signals and system level control signals are separated and also control signal and status signals are separated due to signal transmission paths are allocated independently based on signal type. In this kind of design, single device failure or failures on signal path in the channel couldn't result in the loss of all segmented functions simultaneously. Thus the proposed segmentation function is the design scheme that improves availability of safety functions. In conventional ESF-CCS, the single controller generates the signal to control the multiple safety functions, and the reliability is achieved by multiplication within the channel. This design has a drawback causing the loss of multiple functions due to the CCF (Common Cause Failure) and single failure Heterogeneous controller guarantees the diversity ensuring the execution of safety functions against the CCF and single failure, but requiring a lot of resources like manpower and cost. The segmentation technology based on the compartmentalization and functional diversification decreases the CCF and single failure nonetheless the identical types of
Risk assessment of safety data link and network communication in digital safety feature control system of nuclear power plant

International Nuclear Information System (INIS)

Lee, Sang Hun; Son, Kwang Seop; Jung, Wondea; Kang, Hyun Gook

2017-01-01

Highlights: • Safety data communication risk assessment framework and quantitative scheme were proposed. • Fault-tree model of ESFAS unavailability due to safety data communication failure was developed. • Safety data link and network risk were assessed based on various ESF-CCS design specifications. • The effect of fault-tolerant algorithm reliability of safety data network on ESFAS unavailability was assessed. - Abstract: As one of the safety-critical systems in nuclear power plants (NPPs), the Engineered Safety Feature-Component Control System (ESF-CCS) employs safety data link and network communication for the transmission of safety component actuation signals from the group controllers to loop controllers to effectively accommodate various safety-critical field controllers. Since data communication failure risk in the ESF-CCS has yet to be fully quantified, the ESF-CCS employing data communication systems have not been applied in NPPs. This study therefore developed a fault tree model to assess the data link and data network failure-induced unavailability of a system function used to generate an automated control signal for accident mitigation equipment. The current aim is to provide risk information regarding data communication failure in a digital safety feature control system in consideration of interconnection between controllers and the fault-tolerant algorithm implemented in the target system. Based on the developed fault tree model, case studies were performed to quantitatively assess the unavailability of ESF-CCS signal generation due to data link and network failure and its risk effect on safety signal generation failure. This study is expected to provide insight into the risk assessment of safety-critical data communication in a digitalized NPP instrumentation and control system.
Assessment of Safety Standards for Automotive Electronic Control Systems

Science.gov (United States)

2016-06-01

This report summarizes the results of a study that assessed and compared six industry and government safety standards relevant to the safety and reliability of automotive electronic control systems. These standards include ISO 26262 (Road Vehicles - ...
Process Control Systems in the Chemical Industry: Safety vs. Security

Energy Technology Data Exchange (ETDEWEB)

Jeffrey Hahn; Thomas Anderson

2005-04-01

Traditionally, the primary focus of the chemical industry has been safety and productivity. However, recent threats to our nation’s critical infrastructure have prompted a tightening of security measures across many different industry sectors. Reducing vulnerabilities of control systems against physical and cyber attack is necessary to ensure the safety, security and effective functioning of these systems. The U.S. Department of Homeland Security has developed a strategy to secure these vulnerabilities. Crucial to this strategy is the Control Systems Security and Test Center (CSSTC) established to test and analyze control systems equipment. In addition, the CSSTC promotes a proactive, collaborative approach to increase industry's awareness of standards, products and processes that can enhance the security of control systems. This paper outlines measures that can be taken to enhance the cybersecurity of process control systems in the chemical sector.

The micro-processor controlled process radiation monitoring system for reactor safety systems

International Nuclear Information System (INIS)

Mizuno, K.; Noguchi, A.; Kumagami, S.; Gotoh, Y.; Kumahara, T.; Arita, S.

1986-01-01

Digital computers are soon expected to be applied to various real-time safety and safety-related systems in nuclear power plants. Hitachi is now engaged in the development of a micro-processor controlled process radiation monitoring system, which operates on digital processing methods employed with a log ratemeter. A newly defined methodology of design and test procedures is being applied as a means of software program verification for these safety systems. Recently implemented micro-processor technology will help to achieve an advanced man-machine interface and highly reliable performance. (author)
A survey of approaches combining safety and security for industrial control systems

International Nuclear Information System (INIS)

Kriaa, Siwar; Pietre-Cambacedes, Ludovic; Bouissou, Marc; Halgand, Yoran

2015-01-01

The migration towards digital control systems creates new security threats that can endanger the safety of industrial infrastructures. Addressing the convergence of safety and security concerns in this context, we provide a comprehensive survey of existing approaches to industrial facility design and risk assessment that consider both safety and security. We also provide a comparative analysis of the different approaches identified in the literature. - Highlights: • We raise awareness of safety and security convergence in numerical control systems. • We highlight safety and security interdependencies for modern industrial systems. • We give a survey of approaches combining safety and security engineering. • We discuss the potential of the approaches to model safety and security interactions
Development of embedded Control System for Control and Safety Rod Drive Mechanisms (CSRDMs) of PFBR

International Nuclear Information System (INIS)

Kameswari, K.; Palanisami, K.; Thirugnana Murthy, D.; Murali, N.; Satyamurty, S.A.V.

2013-01-01

Prototype Fast Breeder Reactor (PFBR), a 500 MWe, Sodium cooled, fast breeder reactor is nearing completion at Kalpakkam, Tamil Nadu. PFBR has two independent, fast acting and diverse shutdown systems, one with nine Control and Safety Rods (CSRs) and another with three Diverse Safety Rods (DSRs), with independent driving mechanisms called CSRDMs and DSRDMs respectively. This paper deals with the development of Real Time Computer based Control system for controlling nine CSRDMs with model based software development environment - SCADE (Safety Critical Application Development Environment). (author)
Analysis approach for common cause failure on non-safety digital control system

Energy Technology Data Exchange (ETDEWEB)

Kim, Yun Goo; Oh, Eungse [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

2014-05-15

The effects of common cause failure (CCF) on safety digital instrumentation and control (I and C) system had been considered in defense in depth and diversity coping analysis with safety analysis method. For the non-safety system, single failure had been considered for safety analysis. IEEE Std. 603-1991, Clause 5.6.3.1(2), 'Isolation' states that no credible failure on the non-safety side of an isolation device shall prevent any portion of a safety system from meeting its minimum performance requirements during and following any design basis event requiring that safety function. The software CCF is one of the credible failure on the non-safety side. In advanced digital I and C system, same hardware component is used for different control system and the defect in manufacture or common external event can generate CCF. Moreover, the non-safety I and C system uses complex software for its various function and software quality assurance for the development process is less severe than safety software for the cost effective design. Therefore the potential defects in software cannot be ignored and the effect of software CCF on non-safety I and C system is needed to be evaluated. This paper proposes the general process and considerations for the analysis of CCF on non-safety I and C system.
Analysis Method of Common Cause Failure on Non-safety Digital Control System

Energy Technology Data Exchange (ETDEWEB)

Kim, Yun Goo; Oh, Eun Gse [KHNP, Daejeon (Korea, Republic of)

2014-08-15

The effects of common cause failure on safety digital instrumentation and control system had been considered in defense in depth analysis with safety analysis method. However, the effects of common cause failure on non-safety digital instrumentation and control system also should be evaluated. The common cause failure can be included in credible failure on the non-safety system. In the I and C architecture of nuclear power plant, many design feature has been applied for the functional integrity of control system. One of that is segmentation. Segmentation defenses the propagation of faults in the I and C architecture. Some of effects from common cause failure also can be limited by segmentation. Therefore, in this paper there are two type of failure mode, one is failures in one control group which is segmented, and the other is failures in multiple control group because that the segmentation cannot defense all effects from common cause failure. For each type, the worst failure scenario is needed to be determined, so the analysis method has been proposed in this paper. The evaluation can be qualitative when there is sufficient justification that the effects are bounded in previous safety analysis. When it is not bounded in previous safety analysis, additional analysis should be done with conservative assumptions method of previous safety analysis or best estimation method with realistic assumptions.
Control system of labour safety measures in the higher educational institution

Directory of Open Access Journals (Sweden)

O. G. Feoktistova

2015-01-01

Full Text Available The article examines a system of labour safety measures control. With the introduction of the integrated system of management the competitive ability of production and organization, the effectiveness of its activity rise, and sinnergicheskiy effect is also reached and the savings of all forms of resources are ensured. Objectives and methods of control system of labour safety measures in enterprises are developed, including in the educational institutions.
Impacts of safety on the design of light remotely-piloted helicopter flight control systems

International Nuclear Information System (INIS)

Di Rito, G.; Schettini, F.

2016-01-01

This paper deals with the architecture definition and the safety assessment of flight control systems for light remotely-piloted helicopters for civil applications. The methods and tools to be used for these activities are standardised for conventional piloted aircraft, while they are currently a matter of discussion in case of light remotely-piloted systems flying into unsegregated airspaces. Certification concerns are particularly problematic for aerial systems weighing from 20 to 150 kgf, since the airworthiness permission is granted by national authorities. The lack of specific requirements actually requires to analyse both the existing standards for military applications and the certification guidelines for civil systems, up to derive the adequate safety objectives. In this work, after a survey on applicable certification documents for the safety objectives definition, the most relevant functional failures of a light remotely-piloted helicopter are identified and analysed via Functional Hazard Assessment. Different architectures are then compared by means of Fault-Tree Analysis, highlighting the contributions to the safety level of the main elements of the flight control system (control computers, servoactuators, antenna) and providing basic guidelines on the required redundancy level. - Highlights: • A method for architecture definition and safety assessment of light RW‐UAS flight control systems is proposed. • Relevant UAS failures are identified and analysed via Functional Hazard Assessment and Fault‐Tree Analysis. • The key safety elements are control computers, servoactuators and TX/RX system. • Single‐simplex flight control systems have inadequate safety levels. • Dual‐duplex flight control systems demonstrate to be safety compliant, with safety budgets dominated by servoactuators.
Design of the Control System for Engineered Safety Features of KIJANG Research Reactor

International Nuclear Information System (INIS)

Kim, Hagtae; Kim, Jun-Yeon; Chae, Hee-Taek

2015-01-01

The purpose of this paper is to design an effective control system for the Engineered Safety Features (ESF) of KJRR such as the Safety Residual Heat Removal System (SRHRS) pumps and Siphon Break Valve (SBV) without an Engineered Safety Features-Component Control System (ESF-CCS). This control system is called a 'local motor starter', because this system controls motors in the SRHRS pumps and SBVs by receiving the signal from Reactor Protection System (RPS) and Alternate Protection System (APS) when the differential pressure or pool level reach the set points. In this paper, the design concepts and requirements of the local motor starter based on the design features of KJRR is proposed. An ESF is a safety system that mitigates consequences of the Anticipated Operational Occurrence (AOO) and Design Basis Accident (DBA). The results of this paper are able to be used for the development of control systems for research reactors similar to KJRR. The precondition for such application is to have a few ESFs and conduct simple logic. The proposed control system called a local motor starter is being designed, and a manufacture of the actual systems is expected in the foreseeable future
77 FR 22637 - Federal Motor Vehicle Safety Standards; Accelerator Control Systems

Science.gov (United States)

2012-04-16

... revise the Federal Motor Vehicle Safety Standard for accelerator control systems (ACS) in two ways. First... Standard (FMVSS) No. 124, Accelerator Control Systems,\\2\\ in two ways. First, we are proposing to update... February 2011 final report ``Technical Assessment of Toyota Electronic Throttle Control Systems,'' the...
Continuous restraint control systems: safety improvement for various occupants

NARCIS (Netherlands)

Laan, E. van der; Jager, B. de; Veldpaus, F.; Steinbuch, M.; Nunen, E. van; Willemsen, D.

2009-01-01

Occupant safety can be significantly improved by continuous restraint control systems. These restraint systems adjust their configuration during the impact according to the actual operating conditions, such as occupant size, weight, occupant position, belt usage and crash severity. In this study,
New Methodology for a Comprehensive Modular Safety Control System in a Cyclotron Site

International Nuclear Information System (INIS)

Kaufman, Y.; Kravitz, M.; Arad, M.; Osovizky, A.; Paran, J.; Sarussi, B.; Ellenbogen, M.; Tal, N.

2004-01-01

This Paper describes a new methodology for a comprehensive modular Safety Control System (SCS), for a cyclotron site. The developed SCS is a modular approach for controlling the production procedures, safety conditions and documentation aspects in the Cyclotron site. Usually, the safety conditions in cyclotron sites are maintained by a variety of sensors. The cyclotron is supplied from the manufacturer with a self-integrated control system for its operation, yet the comprehensive SCS has to be defined and setup by the customer. Therefore, customers face a lot of integration problems in trying to combine all the signals from the different safety systems such as radiation monitoring, environmental and access control, in order to maintain proper safety working conditions. The presented SCS design provides main user interface and the complete safety solution required by including preset control logic definitions and open logic for specific user applications. The knowledge for the preset control logic definitions was gathered in previous projects. Failure Mode and Effects Analysis (FMEA) method has been implemented on the SCS to analyze the potential failure modes and their impact on the product reliability
Automated Systems for Road Safety control in a Developing World ...

African Journals Online (AJOL)

An Automated system was finally designed and developed for road safety control. This Automated system is believed to have the capacity to minimize or eliminate the problems identified in this study on traffic control in a developing world. Key words: drivers, traffic situation information, accident causation, FRSC ...
Safety implications of using programmable digital computers in nuclear safety and control systems

International Nuclear Information System (INIS)

Adams, D.M.; Rohrdanz, R.R.

1982-01-01

This papers describes the activities being conducted at the Idaho National Engineering Laboratory associated with the use of stored-program computers for protection and control systems. This project has recently been initiated and a preliminary report will be available. The use of computers in plant control and protection (and more generally in system important to safety) represents a major departure from the systems which have been used in the past. The design, development, and audit methods used for these systems are significantly different, thus requiring different skills and different perspectives
Cold Vacuum Drying Safety Class Instrumentation and Control System Design Description

International Nuclear Information System (INIS)

WHITEHURST, R.

1999-01-01

This document describes the Cold Vacuum Drying Facility (CVDF) Safety Class Instrumentation and Control system (SCIC). The SCIC provides safety functions and features to protect the environment, off-site and on-site personnel and equipment. The function of the SCIC is to provide automatic trip features, valve interlocks, alarms, indication and control for the cold vacuum drying process
Research on conceptual design of simplified nuclear safety instrument and control system

International Nuclear Information System (INIS)

Huang Jie

2015-01-01

The Nuclear safety instrument and control system is directly related to the safety of the reactor. So redundant and diversity design is used to ensure the system's security and reliability. This make the traditional safety system large, more cabinets and wiring complexity. To solve these problem, we can adopt new technology to make the design more simple. The simplify conceptual design can make the system less cabinets, less wiring, but high security, strong reliability. (author)
New design of engineered safety features-component control system to improve performance and reliability

International Nuclear Information System (INIS)

Kim, S.T.; Jung, H.W.; Lee, S.J.; Cho, C.H.; Kim, D.H.; Kim, H.

2006-01-01

Full text: Full text: The Engineered Safety Features-Component Control System (ESF-CCS) controls the engineered safety features of a Nuclear Power Plant such as Solenoid Operated Valves (SOV), Motor Operated Valves (MOV), pumps, dampers, etc. to mitigate the effects of a Design Basis Accident (DBA) or an abnormal operation. ESF-CCS serves as an interface system between the Plant Protection System (PPS) and remote actuation devices. ESF-CCS is composed of fault tolerant Group Controllers GC, Loop Controllers (LC), ESF-CCS Test and Interface Processor (ETIP) and Cabinet Operator Module (COM) and Control Channel Gateway (CCG) etc. GCs in each division are designed to be fully independent triple configuration, which perform system level NSSS and BOP ESFAS logic (2-out-of-4 logic and l-out-of-2 logic, respectively) making it possible to test each GC individually during normal operation. In the existing configuration, the safety-related plant component control is part of the Plant Control System (PCS) non-safety system. For increased safety and reliability, this design change incorporates this part into the LCs, and is therefore designed according to the safety-critical system procedures. The test and diagnosis capabilities of ETIP and COM are reinforced. By means of an automatic periodic test for all main functions of the system, it is possible to quickly determine an abnormal status of the system, and to decrease the elapsed time for tests, thus effectively increasing availability. ESF-CCS consists of four independent divisions (A, B, C, and D) in the Advanced Power Reactor 1400 (APR1400). One prototype division is being manufactured and will be tested
The safety feature of hydraulic driving system of control rod for 200 MW nuclear heating reactor

International Nuclear Information System (INIS)

Chi Zongbo; Wu Yuanqiang

1997-01-01

The hydraulic driving system of control rod is used as control rod drive mechanism in 200 MW nuclear heating reactor. Design of this system is based on passive system, integrating drive and guide of control rod. The author analyzes the inherent safety and the design safety of this system, with mechanism of control rod not ejecting when the pressure of pressure vessel is lost, and calculating result of core not exposing when the amount of coolant is drained by broken pipe. The results indicate that this system has good safety feature, and assures reactor safety under any accident conditions, providing important technology support for 200 MW nuclear heating reactor with inherent safety feature
Description of the control and safety systems of the RA reactor

International Nuclear Information System (INIS)

Popovic, B.; Pesic, M.

1962-01-01

This report contains detailed description and scheme of the control and safety system of the RA reactor. It consists of interconnected five systems: for automated regulation; compensation rods; safety rods; power density measurement device; period meter; automated D 2 O level meter in the core. Automated regulation system is divided into two parts: basic system for reactor operation regime at power from 10kW - 10 MW and precise regulation system for operation at set-up power level up to 10 kW which is used occasionally
Safety critical FPGA-based NPP instrumentation and control systems: assessment, development and implementation

International Nuclear Information System (INIS)

Bakhmach, E. S.; Siora, A. A.; Tokarev, V. I.; Kharchenko, V. S.; Sklyar, V. V.; Andrashov, A. A.

2010-10-01

The stages of development, production, verification, licensing and implementation methods and technologies of safety critical instrumentation and control systems for nuclear power plants (NPP) based on FPGA (Field Programmable Gates Arrays) technologies are described. A life cycle model and multi-version technologies of dependability and safety assurance of FPGA-based instrumentation and control systems are discussed. An analysis of NPP instrumentation and control systems construction principles developed by Research and Production Corporation Radiy using FPGA-technologies and results of these systems implementation and operation at Ukrainian and Bulgarian NPP are presented. The RADIY TM platform has been designed and developed by Research and Production Corporation Radiy, Ukraine. The main peculiarity of the RADIY TM platform is the use of FPGA as programmable components for logic control operation. The FPGA-based RADIY TM platform used for NPP instrumentation and control systems development ensures sca lability of system functions types, volume and peculiarities (by changing quantity and quality of sensors, actuators, input/output signals and control algorithms); sca lability of dependability (safety integrity) (by changing a number of redundant channel, tiers, diagnostic and reconfiguration procedures); sca lability of diversity (by changing types, depth and method of diversity selection). (Author)
JACoW Safety instrumented systems and the AWAKE plasma control as a use case

CERN Document Server

Blanco Viñuela, Enrique; Fernández Adiego, Borja; Speroni, Roberto

2018-01-01

Safety is likely the most critical concern in many process industries, yet there is a general uncertainty on the proper engineering to reduce the risks and ensure the safety of persons or material at the same time as providing the process control system. Some of the reasons for this misperception are unclear requirements, lack of functional safety engineering knowledge or incorrect protection functionalities attributed to the BPCS (Basic Process Control System). Occasionally the control engineers are not aware of the hazards inherent to an industrial process and this causes an incorrect design of the overall controls. This paper illustrates the engineering of the SIS (Safety Instrumented System) and the BPCS of the plasma vapour controls of the AWAKE R&D; project, the first proton-driven plasma wakefield acceleration experiment in the world. The controls design and implementation refers to the IEC61511/ISA84 standard, including technological choices, design, operation and maintenance. Finally, the publica...

Safety control and risk management

International Nuclear Information System (INIS)

Rasmussen, J.

1987-01-01

The acceptable probability of major accidents in nuclear power is very small, and can not be determined from direct empirical evidence. Therefore, control of the level of safety is a complex problem. The difficulty is related to the fact that a variable, 'safety', which is not accessible to direct measurement, is to be tightly controlled. Control, therefore, depends on a systematic, analytical prediction of the target state, i.e., the level of safety, from indirect evidence. From a control theoretic point of view this means that safety is controlled by a system which includes openloop as well as closed loop control paths. The aim of the paper is to take a general systems view on the complex mechanisms involved in the control of safety of industrial installations like nuclear power. From this, the role of probabilistic risk analysis is evaluated and needs for further development discussed. (author)
Technical considerations for the development of an engineering safety features control system with PLC

International Nuclear Information System (INIS)

Lee, C. K.; Kim, C. H.; Han, J. B.; Kim, H.; Lee, S. S.

2002-01-01

Technical considerations are summarized for the development of an ESFCS(Engineered Safety Features Control System) with PLC (Programmable Logic Controller). The ESFCS is required for the mitigation of plant accident conditions and therefore developed in conformance with the design requirements applied to the safety critical system. The design of ESFCS primarily considered its safety, and the system has an architecture that will be able to minimize spurious actuation. The PLC based functional distribution and redundant design features are adopted, and the fieldbus is applied in the communication of information and control signals between PLC processors. It is expected that the ESFCS will have several advanced design features compared with the conventional systems supplied by foreign vendors
Development of FPGA-based safety-related instrumentation and control systems

Energy Technology Data Exchange (ETDEWEB)

Oda, N.; Tanaka, A.; Izumi, M.; Tarumi, T.; Sato, T. [Toshiba Corporation, Isogo Nuclear Engineering Center, Yokohama (Japan)

2004-07-01

Toshiba has developed systems which perform signal processing by field programmable gate arrays (FPGA) for safety-related instrumentation and control systems. FPGA is a device which consists only of defined digital circuit: hardware, which performs defined processing. FPGA-based system solves issues existing both in the conventional systems operated by analog circuits (analog-based system) and the systems operated by central processing units (CPU-based system). The advantages of applying FPGA are to keep the long-life supply of products, improving testability (verification), and to reduce the drift which may occur in analog-based system. Considering application to safety-related systems, nonvolatile and non rewritable FPGA which is impossible to be changed after once manufactured has been adopted in Toshiba FPGA-based system. The systems which Toshiba developed this time are Power range Monitor (PRM) and Trip Module (TM). These systems are compatible with the conventional analog-based systems and the CPU-based systems. Therefore, requested cost for upgrading will be minimized. Toshiba is planning to expand application of FPGA-based technology by adopting this development method to the other safety-related systems from now on. (authors)
Risk assessment of computer-controlled safety systems for fusion reactors

International Nuclear Information System (INIS)

Fryer, M.O.; Bruske, S.Z.

1983-01-01

The complexity of fusion reactor systems and the need to display, analyze, and react promptly to large amounts of information during reactor operation will require a number of safety systems in the fusion facilities to be computer controlled. Computer software, therefore, must be included in the reactor safety analyses. Unfortunately, the science of integrating computer software into safety analyses is in its infancy. Combined plant hardware and computer software systems are often treated by making simple assumptions about software performance. This method is not acceptable for assessing risks in the complex fusion systems, and a new technique for risk assessment of combined plant hardware and computer software systems has been developed. This technique is an extension of the traditional fault tree analysis and uses structured flow charts of the software in a manner analogous to wiring or piping diagrams of hardware. The software logic determines the form of much of the fault trees
Safety critical FPGA-based NPP instrumentation and control systems: assessment, development and implementation

Energy Technology Data Exchange (ETDEWEB)

Bakhmach, E. S.; Siora, A. A.; Tokarev, V. I. [Research and Production Corporation Radiy, 29 Geroev Stalingrada Str., Kirovograd 25006 (Ukraine); Kharchenko, V. S.; Sklyar, V. V.; Andrashov, A. A., E-mail: marketing@radiy.co [Center for Safety Infrastructure-Oriented Research and Analysis, 37 Astronomicheskaya Str., Kharkiv 61085 (Ukraine)

2010-10-15

The stages of development, production, verification, licensing and implementation methods and technologies of safety critical instrumentation and control systems for nuclear power plants (NPP) based on FPGA (Field Programmable Gates Arrays) technologies are described. A life cycle model and multi-version technologies of dependability and safety assurance of FPGA-based instrumentation and control systems are discussed. An analysis of NPP instrumentation and control systems construction principles developed by Research and Production Corporation Radiy using FPGA-technologies and results of these systems implementation and operation at Ukrainian and Bulgarian NPP are presented. The RADIY{sup TM} platform has been designed and developed by Research and Production Corporation Radiy, Ukraine. The main peculiarity of the RADIY{sup TM} platform is the use of FPGA as programmable components for logic control operation. The FPGA-based RADIY{sup TM} platform used for NPP instrumentation and control systems development ensures sca lability of system functions types, volume and peculiarities (by changing quantity and quality of sensors, actuators, input/output signals and control algorithms); sca lability of dependability (safety integrity) (by changing a number of redundant channel, tiers, diagnostic and reconfiguration procedures); sca lability of diversity (by changing types, depth and method of diversity selection). (Author)
On safety classification of instrumentation and control systems and their components

International Nuclear Information System (INIS)

Yastrebenetskij, M.A.; Rozen, Yu.V.

2004-01-01

Safety classification of instrumentation and control systems (I and C) and their components (hardware, software, software-hardware complexes) is described: - evaluation of classification principles and criteria in Ukrainian standards and rules; comparison between Ukrainian and international principles and criteria; possibility and ways of coordination of Ukrainian and international standards related to (I and C) safety classification
Operation safety of control systems. Principles and methods; Surete de fonctionnement des systemes de commande. Principes et methodes

Energy Technology Data Exchange (ETDEWEB)

Aubry, J.F. [Institut National Polytechnique, 54 - Nancy (France); Chatelet, E. [Universite de Technologie de Troyes, 10 (France)

2008-09-15

This article presents the main operation safety methods that can be implemented to design safe control systems taking into account the behaviour of the different components with each other (binary 'operation/failure' behaviours, non-consistent behaviours and 'hidden' failures, dynamical behaviours and temporal aspects etc). To take into account these different behaviours, advanced qualitative and quantitative methods have to be used which are described in this article: 1 - qualitative methods of analysis: functional analysis, preliminary risk analysis, failure mode and failure effects analyses; 2 - quantitative study of systems operation safety: binary representation models, state space-based methods, event space-based methods; 3 - application to the design of control systems: safe specifications of a control system, qualitative analysis of operation safety, quantitative analysis, example of application; 4 - conclusion. (J.S.)
Study concerning the power plant control and safety equipment by integrated distributed systems

International Nuclear Information System (INIS)

Optea, I.; Oprea, M.; Stanescu, P.

1995-01-01

The paper deals with the trends existing in the field of nuclear control and safety equipment and systems, proposing a high-efficiency integrated system. In order to enhance the safety of the plant and reliability of the structure system and components, we present a concept based on the latest computer technology with an open, distributed system, connected by a local area network with high redundancy. A modern conception for the control and safety system is to integrate all the information related to the reactor protection, active engineered safeguard and auxiliary systems parameters, offering a fast flow of information between all the agencies concerned so that situations can be quickly assessed. The integrated distributed control is based on a high performance operating system for realtime applications, flexible enough for transparent networking and modular for demanding configurations. The general design considerations for nuclear reactors instrumentation reliability and testing methods for real-time functions under dynamic regime are presented. Taking into account the fast progress in information technology, we consider the replacement of the old instrumentation of Cernavoda-1 NPP by a modern integrated system as an economical and efficient solution for the next units. (Author) 20 Refs
Division of Cyber Safety and Security Responsibilities Between Control System Owners and Suppliers

OpenAIRE

Skotnes , Ruth

2016-01-01

Part 2: CONTROL SYSTEMS SECURITY; International audience; The chapter discusses the important issue of responsibility for information and communications technology (ICT) – or cyber – safety and security for industrial control systems and the challenges involved in dividing the responsibility between industrial control system owners and suppliers in the Norwegian electric power supply industry. Industrial control system owners are increasingly adopting information and communications technologi...
Impact Safety Control Strategy for the Battery System of an Example Electric Bus

Directory of Open Access Journals (Sweden)

Zhen-po Wang

2015-01-01

Full Text Available This paper proposes a side impact safety control strategy for the battery system, aiming at defusing the hazards of unacceptable behaviors of the battery system such as high-voltage hazards. Based on some collision identification metrics, a side impact discrimination algorithm and a side impact severity algorithm are developed for electric buses. Based on the study on the time to break for power battery, the side impact discrimination algorithm response time is about 20 ms posing a great challenge to the side impact discrimination algorithm. At the same time, the reliability of the impact safety control strategy developed in this paper is evaluated for other plausible side impact signals generated by finite element analysis. The results verify that the impact safety control strategy exhibits robust performance and is able to trigger a breaking signal for power battery system promptly and accurately.
The engineering project and reliability research of the safety interlock slow control system in BESIII

International Nuclear Information System (INIS)

Zhang Yinhong; Zhao Jingwei; Li Xiaonan; Xie Xiaoxi; Gao Cuishan; Bai Jingzhi; Chen Xihui; Min Jian; Nie Zhendong

2008-01-01

The new safety interlock slow control system of BESIII is designed to ensure that the BESIII interior equipments and the accelerator control center to work in coordination, and to guarantee the safety of the operating staff and all the important equipments at the same time. This paper introduces the hardware and software design of safety interlock system from the engineering requirements angle, including a detailed research on the software implementation technique of the state machine on PLC and the reliability of the system. (authors)
78 FR 979 - Petition for Positive Train Control Safety Plan Approval and System Certification of the...

Science.gov (United States)

2013-01-07

...] Petition for Positive Train Control Safety Plan Approval and System Certification of the Electronic Train... the Federal Railroad Administration (FRA) for Positive Train Control (PTC) Safety Plan (PTCSP) approval and system certification of the Electronic Train Management System (ETMS) as required by 49 U.S.C...
Technical self reliance of digital safety systems

Energy Technology Data Exchange (ETDEWEB)

Kwon, Kee Choon; Lee, Dong Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Kook Hun [Doosan Heavy Industries and Construction, Changwon (Korea, Republic of); Choi, Seung Gap [POSCON, Pohang (Korea, Republic of)

2009-04-15

This paper summarizes the development results of the Korea Nuclear Instrumentation and Control System (KNICS) project sponsored by the Korean government. In this project, Man Machine Interface System (MMIS) architecture, two digital platforms, and several control systems are developed. One platform is a programmable Logic Controller (PLC) for a safety system and another platform is a Distributed Control System (DCS) for a non safety system. With the POSAFE Q PLC, a Reactor Protection System (RPS) and an Engineered Safety Feature Component Control System (ESF CCS) are developed. A Power Control System (PCS) is developed based on the DCS. The safety grade platform and the digital safety systems obtained approval for the Topical Report from the Korean regulatory body in February of 2009. Also a Korean utility and a vendor company determined KNICS results to apply them to the planned Nuclear Power Plant (NPP) in March 2009. This paper introduces the technical self reliance experiences of the safety grade platform and the digital safety systems developed in the KNICS R and D project.
Safety problems in vehicles with adaptive cruise control system

Directory of Open Access Journals (Sweden)

Yadav Arun K.

2017-06-01

Full Text Available In today’s world automotive industries are still putting efforts towards more autonomous vehicles (AVs. The main concern of introducing the autonomous technology is safety of driver. According to a survey 90% of accidents happen due to mistake of driver. The adaptive cruise control system (ACC is a system which combines cruise control with a collision avoidance system. The ACC system is based on laser and radar technologies. This system is capable of controlling the velocity of vehicle automatically to match the velocity of car, bus or truck in front of vehicle. If the lead vehicle gets slow down or accelerate, than ACC system automatically matches that velocity. The proposed paper is focusing on more accurate methods of detecting the preceding vehicle by using a radar and lidar sensors by considering the vehicle side slip and by controlling the distance between two vehicles. By using this approach i.e. logic for calculation of former vehicle distance and controlling the throttle valve of ACC equipped vehicle, an improvement in driving stability was achieved. The own contribution results with fuel efficient driving and with more safer and reliable driving system, but still some improvements are going on to make it more safe and reliable.
Developing and maintaining national food safety control systems ...

African Journals Online (AJOL)

The establishment of effective food safety systems is pivotal to ensuring the safety of the national food supply as well as food products for regional and international trade. The development, structure and implementation of modern food safety systems have been driven over the years by a number of developments.
Verification of the safety communication protocol in train control system using colored Petri net

International Nuclear Information System (INIS)

Chen Lijie; Tang Tao; Zhao Xianqiong; Schnieder, Eckehard

2012-01-01

This paper deals with formal and simulation-based verification of the safety communication protocol in ETCS (European Train Control System). The safety communication protocol controls the establishment of safety connection between train and trackside. Because of its graphical user interface and modeling flexibility upon the changes in the system conditions, this paper proposes a composition Colored Petri Net (CPN) representation for both the logic and the timed model. The logic of the protocol is proved to be safe by means of state space analysis: the dead markings are correct; there are no dead transitions; being fair. Further analysis results have been obtained using formal and simulation-based verification approach. The timed models for the open transmit system and the application process are created for the purpose of performance analysis of the safety communication protocol. The models describe the procedure of data transmission and processing, and also provide relevant timed and stochastic factors, as well as time delay and lost packet, which may influence the time for establishment of safety connection of the protocol. Time for establishment of safety connection of the protocol in normal state is verified by formal verification, and then time for establishment of safety connection with different probability of lost packet is simulated. After verification it is found that the time for establishment of safety connection of the safety communication protocol satisfies the safety requirements.
Design a Smart Control Strategy to Implement an Intelligent Energy Safety and Management System

OpenAIRE

Jing-Min Wang; Ming-Ta Yang

2014-01-01

The energy saving and electricity safety are today a cause for increasing concern for homes and buildings. Integrating the radio frequency identification (RFID) and ZigBee wireless sensor network (WSN) mature technologies, the paper designs a smart control strategy to implement an intelligent energy safety and management system (IESMS) which performs energy measuring, controlling, monitoring, and saving of the power outlet system. The presented RFID and billing module is used to identify user...
Safety and security analysis for distributed control system in nuclear power plants

International Nuclear Information System (INIS)

Lu Zhigang; Liu Baoxu

2011-01-01

The Digital Distributed Control System (DCS) is the core that manages all monitoring and operation tasks in a Nuclear Power Plant (NPP). So, Digital Distributed Control System in Nuclear Power Plant has strict requirements for control and automation device safety and security due to many factors. In this article, factors of safety are analyzed firstly, while placing top priority on reliability, quality of supply and stability have also been carefully considered. In particular, advanced digital and electronic technologies are adopted to maintain sufficient reliability and supervisory capabilities in nuclear power plants. Then, security of networking and information technology have been remarked, several design methodologies considering the security characteristics are suggested. Methods and technologies of this article are being used in testing and evaluation for a real implement of a nuclear power plant in China. (author)
A comparison of the difference of requirements between functional safety and nuclear safety controllers

Energy Technology Data Exchange (ETDEWEB)

Chen, C.K.; Lee, C.L.; Shyu, S.S. [Inst. of Nuclear Energy Research, Taoyuan, Taiwan (China)

2014-07-01

In order to establish self-reliant capabilities of nuclear I&C systems in Taiwan, Taiwan's Nuclear I&C System (TNICS) project had been established by Institute of Nuclear Energy Research (INER). A Triple Modular Redundant (TMR) safety controller (SCS-2000) has been completed and gone through the IEC 61508 Safety Integrity Level 3 (SIL3) certification of Functional Safety for industries. Based on the certification processes, the difference of requirements between Functional Safety and Nuclear Safety controllers in term of hardware and software are addressed in this study. Besides, the measures used to determine and verify the reliability of the safety control system design are presented. (author)
A Practical Risk Assessment Methodology for Safety-Critical Train Control Systems

Science.gov (United States)

2009-07-01

This project proposes a Practical Risk Assessment Methodology (PRAM) for analyzing railroad accident data and assessing the risk and benefit of safety-critical train control systems. This report documents in simple steps the algorithms and data input...

Safety logic systems of PFBR

International Nuclear Information System (INIS)

Sambasivan, S. Ilango

2004-01-01

Full text : PFBR is provided with two independent, fast acting and diverse shutdown systems to detect any abnormalities and to initiate safety action. Each system consists of sensors, signal processing systems, logics, drive mechanisms and absorber rods. The absorber rods of the first system are Control and Safety Rods (CSR) and that of the second are called as Diverse Safety Rods (DSR). There are nine CSR and three DSR. While CSR are used for startup, control of reactor power, controlled shutdown and SCRAM, the DSR are used only for SCRAM. The respective drive mechanisms are called as CSRDM and DSRDM. Each of these two systems is capable of executing the shutdown satisfactorily with single failure criteria. Two independent safety logic systems based on diverse principles have been designed for the two shut down systems. The analog outputs of the sensors of Core Monitoring Systems comprising of reactor flux monitoring, core temperature monitoring, failed fuel detection and core flow monitoring systems are processed and converted into binary signals depending on their instantaneous values. Safety logic systems receive the binary signals from these core-monitoring systems and process them logically to protect the reactor against postulated initiating events. Neutronic and power to flow (P/Q) signals form the inputs to safety logic system-I and temperature signals are inputs to the safety logic system II. Failed fuel detection signals are processed by both the shut down systems. The two logic systems to actuate the safety rods are also based on two diverse designs and implemented with solid-state devices to meet all the requirements of safety systems. Safety logic system I that caters to neutronic and P/Q signals is designed around combinational logic and has an on-line test facility to detect struck at faults. The second logic system is based on dynamic logic and hence is inherently safe. This paper gives an overview of the two logic systems that have been
Safety of mechanical devices. Safety of automation systems

International Nuclear Information System (INIS)

Pahl, G.; Schweizer, G.; Kapp, K.

1985-01-01

The paper deals with the classic procedures of safety engineering in the sectors mechanical engineering, electrical and energy engineering, construction and transport, medicine technology and process technology. Particular stress is laid on the safety of automation systems, control technology, protection of mechanical devices, reactor safety, mechanical constructions, transport systems, railway signalling devices, road traffic and protection at work in chemical plans. (DG) [de
Assessment of shaft safety and management system of controlling engineering information

Energy Technology Data Exchange (ETDEWEB)

Liu Rui-xin; Xu Yan-chun [Yanzhou Mining Group Ltd., Zoucheng (China)

2008-02-15

Evaluating shaft safety and establishing a system for controlling engineering information is very important because more than 90 shafts in thick alluvial areas suddenly have shaft wall fracturing or breaking problems and there are more than a few hundred shafts of similar geologic conditions. Taking shaft control in the Yangzhou Coal Mining Group as an example, an assessment and management system and related software were established. This system includes basic information of the mine, measurement results and analysis, and functions of empirical and theoretical forecasting and finite element analysis, which are confirmed to be very effective for guiding shaft well control engineering in practice. 8 refs., 3 figs., 2 tabs.
Cold Vacuum Drying Safety Class Instrumentation and Control System Design Description SYS 93-2

International Nuclear Information System (INIS)

WHITEHURST, R.

1999-01-01

This document describes the Cold Vacuum Drying Facility (CVDF) Safety Class Instrumentation and Control system (SCIC). The SCIC provides safety functions and features to protect the environment, off-site and on-site personnel and equipment. The function of the SCIC is to provide automatic trip features, valve interlocks, alarms, indication and control for the cold vacuum drying process
Implementation of amplifiers, control and safety subsystems of radiofrequency system of VINCY Cyclotron

International Nuclear Information System (INIS)

Drndarevic, V.; Obradovic, M.; Samardic, B.; Djuric, B.; Bojovic, B.; Trajic, M.I.; Golubicic, Z.; Smiljakovic, V.

1996-01-01

Concept and design of power amplifiers, control subsystem and safety subsystems for the RF system of the VINCY cyclotron are described. The power amplifiers subsystem consists of two amplifiers of 30 kW nominal power that operate in class B or class C. High stability of voltage amplitude of 5x10 -4 and phase stability between two resonators better than ± 0.5 0 in the range from 16.5 to 31 MHz is being providing by RF control subsystem. Autonomous safety system serves to protect staff from high voltage and to protect equipment from damage. (author)
Probabilistic safety assessment for instrumentation and control systems in nuclear power plants: an overview

International Nuclear Information System (INIS)

Lu, Lixuan; Jiang, Jin

2004-01-01

Deregulation in the electricity market has resulted in a number of challenges in the nuclear power industry. Nuclear power plants must find innovative ways to remain competitive by reducing operating costs without jeopardizing safety. Instrumentation and Control (I and C) systems not only play important roles in plant operation, but also in reducing the cost of power generation while maintaining and/or enhancing safety. Therefore, it is extremely important that I and C systems are managed efficiently and economically. With the increasing use of digital technologies, new methods are needed to solve problems associated with various aspects of digital I and C systems. Probabilistic Safety Assessment (PSA) has proved to be an effective method for safety analysis and risk-based decisions, even though challenges are still present. This paper provides an overview of PSA applications in three areas of digital I and C systems in nuclear power plants. These areas are Graded Quality Assurance, Surveillance Testing, and Instrumentation and Control System Design. In addition, PSA application in the regulation of nuclear power plants that adopt digital I and C systems is also investigated. (author)
Design measures to increase safety and reliability of power station control and protection systems

International Nuclear Information System (INIS)

Edelmann, J.; Spieth, W.

1977-06-01

The paper reviews a few criteria which exert a considerable influence on the safety and reliability of monitoring and control systems. When judging the safety and reliability of a system, it is of importance not only to look at the failures of just one part of a system but also to take into account the effect these failures have on the overall process. In this respect there is a marked difference between a centralized and a decentralized system. With the technical equipment nowadays at our disposal a high safety standard has been reached. Redundant and dynamic protection systems make the occurrence of a dangerous failure hypothetic. (Author)
Safety-related instrumentation and control systems for nuclear power plants

International Nuclear Information System (INIS)

1984-01-01

This Safety Guide deals mainly with design requirements for those I and C systems that are important to safety but are not safety systems. The Guide is intended to expand paragraphs 3.1, 3.2 and 3.3 of the Code of Practice on Design for Safety of Nuclear Power Plants (IAEA Safety Series No.50-C-D) in the area of I and C systems important to safety and refers to them as safety-related I and C systems. It also gives guidance and enumerates requirements for multiplexing and the use of the digital computers employed in this area
ACCIDENT ANALYSES & CONTROL OPTIONS IN SUPPORT OF THE SLUDGE WATER SYSTEM SAFETY ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

WILLIAMS, J.C.

2003-11-15

This report documents the accident analyses and nuclear safety control options for use in Revision 7 of HNF-SD-WM-SAR-062, ''K Basins Safety Analysis Report'' and Revision 4 of HNF-SD-SNF-TSR-001, ''Technical Safety Requirements - 100 KE and 100 KW Fuel Storage Basins''. These documents will define the authorization basis for Sludge Water System (SWS) operations. This report follows the guidance of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', for calculating onsite and offsite consequences. The accident analysis summary is shown in Table ES-1 below. While this document describes and discusses potential control options to either mitigate or prevent the accidents discussed herein, it should be made clear that the final control selection for any accident is determined and presented in HNF-SD-WM-SAR-062.
Software quality assurance and software safety in the Biomed Control System

International Nuclear Information System (INIS)

Singh, R.P.; Chu, W.T.; Ludewigt, B.A.; Marks, K.M.; Nyman, M.A.; Renner, T.R.; Stradtner, R.

1989-01-01

The Biomed Control System is a hardware/software system used for the delivery, measurement and monitoring of heavy-ion beams in the patient treatment and biology experiment rooms in the Bevalac at the Lawrence Berkeley Laboratory (LBL). This paper describes some aspects of this system including historical background philosophy, configuration management, hardware features that facilitate software testing, software testing procedures, the release of new software quality assurance, safety and operator monitoring. 3 refs
Who is in control of road safety? A STAMP control structure analysis of the road transport system in Queensland, Australia.

Science.gov (United States)

Salmon, Paul M; Read, Gemma J M; Stevens, Nicholas J

2016-11-01

Despite significant progress, road trauma continues to represent a global safety issue. In Queensland (Qld), Australia, there is currently a focus on preventing the 'fatal five' behaviours underpinning road trauma (drug and drink driving, distraction, seat belt wearing, speeding, and fatigue), along with an emphasis on a shared responsibility for road safety that spans road users, vehicle manufacturers, designers, policy makers etc. The aim of this article is to clarify who shares the responsibility for road safety in Qld and to determine what control measures are enacted to prevent the fatal five behaviours. This is achieved through the presentation of a control structure model that depicts the actors and organisations within the Qld road transport system along with the control and feedback relationships that exist between them. Validated through a Delphi study, the model shows a diverse set of actors and organisations who share the responsibility for road safety that goes beyond those discussed in road safety policies and strategies. The analysis also shows that, compared to other safety critical domains, there are less formal control structures in road transport and that opportunities exist to add new controls and strengthen existing ones. Relationships that influence rather than control are also prominent. Finally, when compared to other safety critical domains, the strength of road safety controls is brought into question. Copyright © 2016 Elsevier Ltd. All rights reserved.
Strategy to safety grade systems replacements

International Nuclear Information System (INIS)

Stimler, M.; Sullivan, K.E.; Trebincevic, I.

1993-01-01

The introduction of digital instrumentation and control systems in nuclear power plants is characterized by the need to satisfy the requirements of safety, reliability and man-machine ergonomics. Today digital instrumentation and control systems meet these requirements and the trend in Europe is towards full digital based nuclear power plant control systems. This paper describes Siemens (KWU) experience in nuclear power plants and development in trends within Europe. Topics which are the subject of major concern to NPP operators addressed in this paper are: human performance factors - man-machine interface; operating philosophy; safety, availability and reliability. Other aspects addressed are: Siemens open-quotes defense in depthclose quotes concept, description of Siemens digital I ampersand C systems, safety requirements and systems, I ampersand C qualification, control room ergonomics, information systems and retrofitting experience
Safety and interlock system for Tristan

International Nuclear Information System (INIS)

Takeda, S.; Kudo, K.; Katoh, T.; Akiyama, A.

1987-01-01

This report describes alarm and interlock system of TRISTAN, concentrating on personnel safety. The basis of TRISTAN machine-control system (TMS) is an N-to-N computer network and KEK NODAL which offers high software productivity. TMC achieves high flexibility of operation both for normal operation and for the fast commissioning. However, to assure the safety of personnel and the TRISTAN machine operation, the safety system has to continue functioning during TMC failure as well. A distributed safety and interlock system (DSIS) is used for diversification of risks in TRISTAN system. DSIS is functionally subdivided along local system lines and has a hierarchical structure of 12 programmable sequence controllers (PSCs). Optical fiber links connect the PSCs at subsystem level and a PSC at the supervisory level of TRISTAN central control room (TCCR). The subsystem PSCs provide the interlock functions between their local devices. The local PSCs interact with the central system through a limited number of summarized signals. The central PSC provides the interlock functions between the subsystems and interacts with an operator's panel. Personnel safety is based on a system of electrical interlock keys, emergency push-buttons around the tunnel, at the entrance gates or in the control room
Validation of Safety-Critical Systems for Aircraft Loss-of-Control Prevention and Recovery

Science.gov (United States)

Belcastro, Christine M.

2012-01-01

Validation of technologies developed for loss of control (LOC) prevention and recovery poses significant challenges. Aircraft LOC can result from a wide spectrum of hazards, often occurring in combination, which cannot be fully replicated during evaluation. Technologies developed for LOC prevention and recovery must therefore be effective under a wide variety of hazardous and uncertain conditions, and the validation framework must provide some measure of assurance that the new vehicle safety technologies do no harm (i.e., that they themselves do not introduce new safety risks). This paper summarizes a proposed validation framework for safety-critical systems, provides an overview of validation methods and tools developed by NASA to date within the Vehicle Systems Safety Project, and develops a preliminary set of test scenarios for the validation of technologies for LOC prevention and recovery
Requirements and analysis of electromagnetic compatibility of safety-related instrumentation and control system in nuclear power plants

International Nuclear Information System (INIS)

Liu Sujuan

2002-01-01

The state-of-the-art instrumentation and control system and the influence of their application to the electromagnetic compatibility is analyzed. Based on the present situation of nuclear safety in China and relevant experiences from other countries, the author tries to probe into the requirements and test methods about how safety-related instrument and control system to accommodate electromagnetic interference, radio-frequency interference and power surges in the environments of nuclear power plant so as to develop Chinese safety standards
Experience in the review of utility control room design review and safety parameter display system programs

International Nuclear Information System (INIS)

Moore, V.A.

1985-01-01

The Detailed Control Room Design Review (DCRDR) and the Safety Parameter Display System (SPDS) had their origins in the studies and investigations conducted as the result of the TMI-2 accident. The President's Commission (Kemeny Commission) critized NRC for not examining the man-machine interface, over-emphasizing equipment, ignoring human beings, and tolerating outdated technology in control rooms. The Commission's Special Inquiry Group (Rogovin Report) recommended greater application of human factors engineering including better instrumentation displays and improved control room design. The NRC Lessons Learned Task Force concluded that licensees should review and improve control rooms using NRC Human engineering guidelines, and install safety parameter display systems (then called the safety staff vector). The TMI Action Plan Item I.D.1 and I.D.2 were based on these recommendations
Architecture Level Safety Analyses for Safety-Critical Systems

Directory of Open Access Journals (Sweden)

K. S. Kushal

2017-01-01

Full Text Available The dependency of complex embedded Safety-Critical Systems across Avionics and Aerospace domains on their underlying software and hardware components has gradually increased with progression in time. Such application domain systems are developed based on a complex integrated architecture, which is modular in nature. Engineering practices assured with system safety standards to manage the failure, faulty, and unsafe operational conditions are very much necessary. System safety analyses involve the analysis of complex software architecture of the system, a major aspect in leading to fatal consequences in the behaviour of Safety-Critical Systems, and provide high reliability and dependability factors during their development. In this paper, we propose an architecture fault modeling and the safety analyses approach that will aid in identifying and eliminating the design flaws. The formal foundations of SAE Architecture Analysis & Design Language (AADL augmented with the Error Model Annex (EMV are discussed. The fault propagation, failure behaviour, and the composite behaviour of the design flaws/failures are considered for architecture safety analysis. The illustration of the proposed approach is validated by implementing the Speed Control Unit of Power-Boat Autopilot (PBA system. The Error Model Annex (EMV is guided with the pattern of consideration and inclusion of probable failure scenarios and propagation of fault conditions in the Speed Control Unit of Power-Boat Autopilot (PBA. This helps in validating the system architecture with the detection of the error event in the model and its impact in the operational environment. This also provides an insight of the certification impact that these exceptional conditions pose at various criticality levels and design assurance levels and its implications in verifying and validating the designs.
Design of Instrumentation and Control Systems for Nuclear Power Plants. Specific Safety Guide

International Nuclear Information System (INIS)

2016-01-01

This publication is a revision and combination of two Safety Guides, IAEA Safety Standards Series No. NS-G-1.1 and No. NS-G-1.3. The revision takes into account developments in instrumentation and control (I&C) systems since the publication of the earlier Safety Guides. The main changes relate to the continuing development of computer applications and the evolution of the methods necessary for their safe, secure and practical use. In addition, account is taken of developments in human factors engineering and the need for computer security. This Safety Guide references and takes into account other IAEA Safety Standards and Nuclear Security Series publications that provide guidance relating to I&C design
Safety system status monitoring

International Nuclear Information System (INIS)

Lewis, J.R.; Morgenstern, M.H.; Rideout, T.H.; Cowley, P.J.

1984-03-01

The Pacific Northwest Laboratory has studied the safety aspects of monitoring the preoperational status of safety systems in nuclear power plants. The goals of the study were to assess for the NRC the effectiveness of current monitoring systems and procedures, to develop near-term guidelines for reducing human errors associated with monitoring safety system status, and to recommend a regulatory position on this issue. A review of safety system status monitoring practices indicated that current systems and procedures do not adequately aid control room operators in monitoring safety system status. This is true even of some systems and procedures installed to meet existing regulatory guidelines (Regulatory Guide 1.47). In consequence, this report suggests acceptance criteria for meeting the functional requirements of an adequate system for monitoring safety system status. Also suggested are near-term guidelines that could reduce the likelihood of human errors in specific, high-priority status monitoring tasks. It is recommended that (1) Regulatory Guide 1.47 be revised to address these acceptance criteria, and (2) the revised Regulatory Guide 1.47 be applied to all plants, including those built since the issuance of the original Regulatory Guide
Safety system status monitoring

Energy Technology Data Exchange (ETDEWEB)

Lewis, J.R.; Morgenstern, M.H.; Rideout, T.H.; Cowley, P.J.

1984-03-01

The Pacific Northwest Laboratory has studied the safety aspects of monitoring the preoperational status of safety systems in nuclear power plants. The goals of the study were to assess for the NRC the effectiveness of current monitoring systems and procedures, to develop near-term guidelines for reducing human errors associated with monitoring safety system status, and to recommend a regulatory position on this issue. A review of safety system status monitoring practices indicated that current systems and procedures do not adequately aid control room operators in monitoring safety system status. This is true even of some systems and procedures installed to meet existing regulatory guidelines (Regulatory Guide 1.47). In consequence, this report suggests acceptance criteria for meeting the functional requirements of an adequate system for monitoring safety system status. Also suggested are near-term guidelines that could reduce the likelihood of human errors in specific, high-priority status monitoring tasks. It is recommended that (1) Regulatory Guide 1.47 be revised to address these acceptance criteria, and (2) the revised Regulatory Guide 1.47 be applied to all plants, including those built since the issuance of the original Regulatory Guide.

New technique for determining unavailability of computer controlled safety systems

International Nuclear Information System (INIS)

Fryer, M.O.; Bruske, S.Z.

1984-04-01

The availability of a safety system for a fusion reactor is determined. A fusion reactor processes tritium and requires an Emergency Tritium Cleanup (ETC) system for accidental tritium releases. The ETC is computer controlled and because of its complexity, is an excellent candidate for this analysis. The ETC system unavailability, for preliminary untested software, is calculated based on different assumptions about operator response. These assumptions are: (a) the operator shuts down the system after the first indication of plant failure; (b) the operator shuts down the system after following optimized failure verification procedures; or (c) the operator is taken out of the decision process, and the computer uses the optimized failure verification procedures
Availability analysis of safety grade multiple redundant controller used in advanced nuclear safety systems

International Nuclear Information System (INIS)

Son, Kwang Seop; Kim, Dong Hoon; Park, Gee Yong; Kang, Hyun Gook

2018-01-01

Highlights: •The multiple redundant controller, SPLC is configured as the combination of DMR and TMR architecture. •We construct the Markov model of SPLC using the concept of the system unavailability rate. •To satisfy the availability requirement of safety grade controller, the fault coverage factor (FCF) should be ≥0.8 and the MTTR of each module should be ≤100 h when FCF is 0.9. •The availability of SPLC is better than that of PLC having iTMR architecture however it is poorer than iTMR considering the off-line test and inspection on the assumption that MTTR of each module is ≤200 h. -- Abstract: We analyze the availability of the Safety Programmable Logic Controller (SPLC) having multiple redundant architectures. In the SPLC, input/output and processor module are configured as triple modular redundancy (TMR), and backplane bus, power and communication modules are configured as dual modular redundancy (DMR). The voting logics for redundant architectures are based on the forwarding error detection. It means that the receivers perform the voting logics based on the status information of transmitters. To analyze the availability of SPLC, we construct the Markov model and simplify the model adopting the system unavailability rate. The results show that the fault coverage factor should be ≥0.8 and Mean Time To Repair (MTTR) should be ≤100 h in order to satisfy the requirement that the availability of the safety grade PLC should be ≥0.995. Also we evaluate the availability of SPLC comparing to other PLCs such as simplex, processor DMR (pDMR) and independent TMR (iTMR) PLCs used in the existing nuclear safety systems. The availability of SPLC is higher than those of the simplex, pDMR but is lower than that of iTMR for one month which is the periodic off-line test and inspection. That’s why the number of redundant modules used in PLC is more dominant to increasing the availability than the number of fault masking methods such as voting logics used
A safety assessment methodology applied to CNS/ATM-based air traffic control system

Energy Technology Data Exchange (ETDEWEB)

Vismari, Lucio Flavio, E-mail: lucio.vismari@usp.b [Safety Analysis Group (GAS), School of Engineering at University of Sao Paulo (Poli-USP), Av. Prof. Luciano Gualberto, Trav.3, n.158, Predio da Engenharia de Eletricidade, Sala C2-32, CEP 05508-900, Sao Paulo (Brazil); Batista Camargo Junior, Joao, E-mail: joaocamargo@usp.b [Safety Analysis Group (GAS), School of Engineering at University of Sao Paulo (Poli-USP), Av. Prof. Luciano Gualberto, Trav.3, n.158, Predio da Engenharia de Eletricidade, Sala C2-32, CEP 05508-900, Sao Paulo (Brazil)

2011-07-15

In the last decades, the air traffic system has been changing to adapt itself to new social demands, mainly the safe growth of worldwide traffic capacity. Those changes are ruled by the Communication, Navigation, Surveillance/Air Traffic Management (CNS/ATM) paradigm , based on digital communication technologies (mainly satellites) as a way of improving communication, surveillance, navigation and air traffic management services. However, CNS/ATM poses new challenges and needs, mainly related to the safety assessment process. In face of these new challenges, and considering the main characteristics of the CNS/ATM, a methodology is proposed at this work by combining 'absolute' and 'relative' safety assessment methods adopted by the International Civil Aviation Organization (ICAO) in ICAO Doc.9689 , using Fluid Stochastic Petri Nets (FSPN) as the modeling formalism, and compares the safety metrics estimated from the simulation of both the proposed (in analysis) and the legacy system models. To demonstrate its usefulness, the proposed methodology was applied to the 'Automatic Dependent Surveillance-Broadcasting' (ADS-B) based air traffic control system. As conclusions, the proposed methodology assured to assess CNS/ATM system safety properties, in which FSPN formalism provides important modeling capabilities, and discrete event simulation allowing the estimation of the desired safety metric.
A safety assessment methodology applied to CNS/ATM-based air traffic control system

International Nuclear Information System (INIS)

Vismari, Lucio Flavio; Batista Camargo Junior, Joao

2011-01-01

In the last decades, the air traffic system has been changing to adapt itself to new social demands, mainly the safe growth of worldwide traffic capacity. Those changes are ruled by the Communication, Navigation, Surveillance/Air Traffic Management (CNS/ATM) paradigm , based on digital communication technologies (mainly satellites) as a way of improving communication, surveillance, navigation and air traffic management services. However, CNS/ATM poses new challenges and needs, mainly related to the safety assessment process. In face of these new challenges, and considering the main characteristics of the CNS/ATM, a methodology is proposed at this work by combining 'absolute' and 'relative' safety assessment methods adopted by the International Civil Aviation Organization (ICAO) in ICAO Doc.9689 , using Fluid Stochastic Petri Nets (FSPN) as the modeling formalism, and compares the safety metrics estimated from the simulation of both the proposed (in analysis) and the legacy system models. To demonstrate its usefulness, the proposed methodology was applied to the 'Automatic Dependent Surveillance-Broadcasting' (ADS-B) based air traffic control system. As conclusions, the proposed methodology assured to assess CNS/ATM system safety properties, in which FSPN formalism provides important modeling capabilities, and discrete event simulation allowing the estimation of the desired safety metric.
Considerations on nuclear reactor passive safety systems

International Nuclear Information System (INIS)

2016-01-01

After having indicated some passive safety systems present in electronuclear reactors (control bars, safety injection system accumulators, reactor cooling after stoppage, hydrogen recombination systems), this report recalls the main characteristics of passive safety systems, and discusses the main issues associated with the assessment of new passive systems (notably to face a sustained loss of electric supply systems or of cold water source) and research axis to be developed in this respect. More precisely, the report comments the classification of safety passive systems as it is proposed by the IAEA, outlines and comments specific aspects of these systems regarding their operation and performance. The next part discusses the safety approach, the control of performance of safety passive systems, issues related to their reliability, and the expected contribution of R and D (for example: understanding of physical phenomena which have an influence of these systems, capacities of simulation of these phenomena, needs of experimentations to validate simulation codes)
Analysis and design on airport safety information management system

Directory of Open Access Journals (Sweden)

Yan Lin

2017-01-01

Full Text Available Airport safety information management system is the foundation of implementing safety operation, risk control, safety performance monitor, and safety management decision for the airport. The paper puts forward the architecture of airport safety information management system based on B/S model, focuses on safety information processing flow, designs the functional modules and proposes the supporting conditions for system operation. The system construction is helpful to perfecting the long effect mechanism driven by safety information, continually increasing airport safety management level and control proficiency．
The LHC personnel safety system

International Nuclear Information System (INIS)

Ninin, P.; Valentini, F.; Ladzinski, T.

2011-01-01

Large particle physics installations such as the CERN Large Hadron Collider require specific Personnel Safety Systems (PSS) to protect the personnel against the radiological and industrial hazards. In order to fulfill the French regulation in matter of nuclear installations, the principles of IEC 61508 and IEC 61513 standard are used as a methodology framework to evaluate the criticality of the installation, to design and to implement the PSS.The LHC PSS deals with the implementation of all physical barriers, access controls and interlock devices around the 27 km of underground tunnel, service zones and experimental caverns of the LHC. The system shall guarantee the absence of personnel in the LHC controlled areas during the machine operations and, on the other hand, ensure the automatic accelerator shutdown in case of any safety condition violation, such as an intrusion during beam circulation. The LHC PSS has been conceived as two separate and independent systems: the LHC Access Control System (LACS) and the LHC Access Safety System (LASS). The LACS, using off the shelf technologies, realizes all physical barriers and regulates all accesses to the underground areas by identifying users and checking their authorizations.The LASS has been designed according to the principles of the IEC 61508 and 61513 standards, starting from a risk analysis conducted on the LHC facility equipped with a standard access control system. It consists in a set of safety functions realized by a dedicated fail-safe and redundant hardware guaranteed to be of SIL3 class. The integration of various technologies combining electronics, sensors, video and operational procedures adopted to establish an efficient personnel safety system for the CERN LHC accelerator is presented in this paper. (authors)
Safety Verification for Probabilistic Hybrid Systems

DEFF Research Database (Denmark)

Zhang, Lijun; She, Zhikun; Ratschan, Stefan

2010-01-01

The interplay of random phenomena and continuous real-time control deserves increased attention for instance in wireless sensing and control applications. Safety verification for such systems thus needs to consider probabilistic variations of systems with hybrid dynamics. In safety verification o...... on a number of case studies, tackled using a prototypical implementation....
Outline of the requirements of application of computer based instrumentation and control systems in the systems important to safety on Bohunice NPPs

International Nuclear Information System (INIS)

Bacurik, J.

1997-01-01

The most important regulatory requirements and issues are described related to the review, evaluation and assessment of computer-based safety-related IandC systems, with emphasis on safety instrumentation and control. These aspects include safety classification and categorization of IandC, ranking of applicable codes and standards, design evaluation on the system level, and software assessment. (author)
OBTAINING FOOD SAFETY BY APPLYING HACCP SYSTEM

Directory of Open Access Journals (Sweden)

ION CRIVEANU

2012-01-01

Full Text Available In order to increase the confidence of the trading partners and consumers in the products which are sold on the market, enterprises producing food are required to implement the food safety system HACCP,a particularly useful system because the manufacturer is not able to fully control finished products . SR EN ISO 22000:2005 establishes requirements for a food safety management system where an organization in the food chain needs to proove its ability to control food safety hazards in order to ensure that food is safe at the time of human consumption. This paper presents the main steps which ensure food safety using the HACCP system, and SR EN ISO 20000:2005 requirements for food safety.
Control maintenance training program for special safety systems at Bruce B

International Nuclear Information System (INIS)

Reinwald, G.

1997-01-01

It was recognized from the early days of commissioning of Bruce B that Control Maintenance staff would require a level of expertise to be able to maintain Special Safety Systems in proper running order. In the early 80's this was achieved through hands on experience during the original commissioning, troubleshooting and placing of the various systems in service. Control maintenance procedures were developed and implemented as the new systems came available for commissioning, as were operating manuals,training manuals etc. Under the development of the Maintenance Manager, a Conduct of Maintenance section was organized. One of the responsibilities of this section was to develop a series of Maintenance Administrative Procedures (MAPs) that set the standards for maintenance activities including training
NASA System Safety Handbook. Volume 1; System Safety Framework and Concepts for Implementation

Science.gov (United States)

Dezfuli, Homayoon; Benjamin, Allan; Everett, Christopher; Smith, Curtis; Stamatelatos, Michael; Youngblood, Robert

2011-01-01

System safety assessment is defined in NPR 8715.3C, NASA General Safety Program Requirements as a disciplined, systematic approach to the analysis of risks resulting from hazards that can affect humans, the environment, and mission assets. Achievement of the highest practicable degree of system safety is one of NASA's highest priorities. Traditionally, system safety assessment at NASA and elsewhere has focused on the application of a set of safety analysis tools to identify safety risks and formulate effective controls.1 Familiar tools used for this purpose include various forms of hazard analyses, failure modes and effects analyses, and probabilistic safety assessment (commonly also referred to as probabilistic risk assessment (PRA)). In the past, it has been assumed that to show that a system is safe, it is sufficient to provide assurance that the process for identifying the hazards has been as comprehensive as possible and that each identified hazard has one or more associated controls. The NASA Aerospace Safety Advisory Panel (ASAP) has made several statements in its annual reports supporting a more holistic approach. In 2006, it recommended that "... a comprehensive risk assessment, communication and acceptance process be implemented to ensure that overall launch risk is considered in an integrated and consistent manner." In 2009, it advocated for "... a process for using a risk-informed design approach to produce a design that is optimally and sufficiently safe." As a rationale for the latter advocacy, it stated that "... the ASAP applauds switching to a performance-based approach because it emphasizes early risk identification to guide designs, thus enabling creative design approaches that might be more efficient, safer, or both." For purposes of this preface, it is worth mentioning three areas where the handbook emphasizes a more holistic type of thinking. First, the handbook takes the position that it is important to not just focus on risk on an individual
Replacement of the complete control system of the NPP Oskarshamn 1 by digital distributed control system

International Nuclear Information System (INIS)

Berger, E.

1998-01-01

As part of an ongoing modernization program, the I and C system and the control room of Oskarshamn 1 will be upgraded by ABB using its 'Advant Power' range of digital, programmable process control system. Besides ensuring the higher level of safety that is demanded today, the new equipment provides the plant with an integrated system which will improve operator interaction with the plant and reduce the risk of human error. The newly installed DCS system will serve also as a platform for further improvements of the control room. This paper discusses Oskarshamn 1 exchange of the complete control system of a nuclear power plant, the technical solution and the time schedule. Oskarshamn 1 is the first nuclear power plant in Sweden. It is a boiling water reactor built between 1966 and 1971 by ABB ATOM in Sweden. According to the plant age the control system is relay-based, while instrumentation and analogue control is semiconductor-based. This makes maintenance expensive and even worse, makes extensions nearly impossible. According to the safety standards of the 1960s, there is no separation between safety and non safety control and no seismic qualification. To extend the life of this plant the owner has decided to improve the safety system as well as to replace the reactor protection system, the safety related control and the non safety related control by a state-of-the-art digital distributed control system from ABB. In March 1997, ABB got the order to replace the reactor protection system, the safety control system and to start the replacement of all control systems. The old control room has to be replaced by a new ergonomically design. Together with the exchange of the control system the safety features of the plant and the emergency power supply has to be extended. (author)
Safety Information System Guide

International Nuclear Information System (INIS)

Bullock, M.G.

1977-03-01

This Guide provides guidelines for the design and evaluation of a working safety information system. For the relatively few safety professionals who have already adopted computer-based programs, this Guide may aid them in the evaluation of their present system. To those who intend to develop an information system, it will, hopefully, inspire new thinking and encourage steps towards systems safety management. For the line manager who is working where the action is, this Guide may provide insight on the importance of accident facts as a tool for moving ideas up the communication ladder where they will be heard and acted upon; where what he has to say will influence beneficial changes among those who plan and control his operations. In the design of a safety information system, it is suggested that the safety manager make friends with a computer expert or someone on the management team who has some feeling for, and understanding of, the art of information storage and retrieval as a new and better means for communication
Introduction of the system of hazard analysis critical control point to ensure the safety of irradiated food

International Nuclear Information System (INIS)

Sajet, A.S.

2014-01-01

Hazard Analysis Critical Control Point (HACCP) is a preventive system for food safety. It identifies safety risks faced by food. Identified points are controlled ensuring product safety. Because of presence of many of the pathogenic microorganisms and parasites in food which caused cases of food poisoning and many diseases transmitted through food, the current methods of food production could not prevent food contamination or prevent the growth of these pathogens completely because of being a part of the normal flora in the environment. Irradiation technology helped to control diseases transmitted through food, caused by pathological microorganisms and parasites present in food. The application of a system based on risk analysis as a means of risk management in food chain, demonstrated the importance of food irradiation. (author)
Reactivity control in HTR power plants with respect to passive safety system. Summary

Energy Technology Data Exchange (ETDEWEB)

Barnert, H; Kugeler, K [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Sicherheitsforschung und Reaktortechnik

1996-12-01

The R and D and Demonstration of the High Temperature Reactor (HTR) is described in overview. The HTR-MODULE power plant, as the most advanced concept, is taken for the description of the reactivity control in general. The idea of the ``modularization of the core`` of the HTR has been developed as the answer on the experiences of the core melt accident at Three Miles Island. The HTR module has two shutdown systems: The ``6 rods``-system for hot shutdown at the ``18 small absorber pebbles units`` - system for cold shutdown. With respect to the definition of ``Passive Systems`` of IAEA-TECDOC-626 the total reactivity control system of the HTR-MODULE is a passive system of category D, because it is an emergency reactor shutdown system based on gravity driven rods, and devices, activated by fail-safe trip logic. But reactivity control of the HTR does not only consist of these engineered safety system but does have a self-acting stabilization by the negative temperature coefficient of the reactivity, being rather effective in reactivity control. Examples from computer calculations are presented, and, in addition, experimental results from the ``Stuck Rod Experiment`` at the AVR reactor in Juelich. On the basis of this the proposal is made that ``self-acting stabilization as a quality of the function`` should be discussed as a new category in addition to the active and passive engineered safety systems, structures and components of IAEA-TECDOC-626. The requirements for a future ``catastrophe-free`` nuclear technology are presented. In the appendix the 7th amendment of the atomic energy act of the Federal Republic of Germany, effective 28 July 94, is given. (author).
Safety analysis of control rod drive computers

International Nuclear Information System (INIS)

Ehrenberger, W.; Rauch, G.; Schmeil, U.; Maertz, J.; Mainka, E.U.; Nordland, O.; Gloee, G.

1985-01-01

The analysis of the most significant user programmes revealed no errors in these programmes. The evaluation of approximately 82 cumulated years of operation demonstrated that the operating system of the control rod positioning processor has a reliability that is sufficiently good for the tasks this computer has to fulfil. Computers can be used for safety relevant tasks. The experience gained with the control rod positioning processor confirms that computers are not less reliable than conventional instrumentation and control system for comparable tasks. The examination and evaluation of computers for safety relevant tasks can be done with programme analysis or statistical evaluation of the operating experience. Programme analysis is recommended for seldom used and well structured programmes. For programmes with a long, cumulated operating time a statistical evaluation is more advisable. The effort for examination and evaluation is not greater than the corresponding effort for conventional instrumentation and control systems. This project has also revealed that, where it is technologically sensible, process controlling computers or microprocessors can be qualified for safety relevant tasks without undue effort. (orig./HP) [de
Integrated Chassis Control System with Fail Safety Using Optimum Yaw Moment Distribution

International Nuclear Information System (INIS)

Yim, Seongjin

2014-01-01

This paper presents an integrated chassis control system with fail safety using optimum yaw moment distribution for a vehicle with steer-by-wire and brake-by-wire devices. The proposed system has two-level structure: upper- and lower-level controllers. In the upper-level controller, the control yaw moment is computed with sliding mode control theory. In the lower-level controller, the control yaw moment is distributed into the tire forces of active front steering(AFS) and electronic stability control(ESC) with the weighted pseudo-inverse based control allocation(WPCA) method. By setting the variable weights in WPCA, it is possible to take the sensor/actuator failure into account. In this framework, it is necessary to optimize the variables weights in order to enhance the yaw moment distribution. For this purpose, simulation-based tuning is proposed. To show the effectiveness of the proposed method, simulations are conducted on a vehicle simulation package, CarSim
Integrated Chassis Control System with Fail Safety Using Optimum Yaw Moment Distribution

Energy Technology Data Exchange (ETDEWEB)

Yim, Seongjin [Seoul Nat' l Univ. of Sci. and Tech., Seoul (Korea, Republic of)

2014-03-15

This paper presents an integrated chassis control system with fail safety using optimum yaw moment distribution for a vehicle with steer-by-wire and brake-by-wire devices. The proposed system has two-level structure: upper- and lower-level controllers. In the upper-level controller, the control yaw moment is computed with sliding mode control theory. In the lower-level controller, the control yaw moment is distributed into the tire forces of active front steering(AFS) and electronic stability control(ESC) with the weighted pseudo-inverse based control allocation(WPCA) method. By setting the variable weights in WPCA, it is possible to take the sensor/actuator failure into account. In this framework, it is necessary to optimize the variables weights in order to enhance the yaw moment distribution. For this purpose, simulation-based tuning is proposed. To show the effectiveness of the proposed method, simulations are conducted on a vehicle simulation package, CarSim.
Development and application of digital safety system in NPPs

International Nuclear Information System (INIS)

Kwon, Keechoon; Kim, Changhwoi; Lee, Dongyoung

2012-01-01

This paper describes the development of digital safety system in NPPs based on safety- grade programmable logic controller (PLC) platform and its application to real NPP construction. The digital safety system consists of a reactor protection system and an engineered safety feature-component control system. The safety-grade PLC platform was developed so that it meets the requirements of the regulation. The PLC consists of various modules such as a power module, a processor module, communication modules, digital input/output modules, analog input/output modules, a LOCA bus extension module, and a high-speed pulse counter module. The reactor protection system is designed with a redundant 4-channel architecture, and every channel is implemented with the same architecture. A single channel consists of a redundant bi-stable processor, a redundant coincidence processor, an automatic test and interface processor, and a cabinet operator module. The engineered safety feature-component control system is designed with four redundant divisions, and implemented with the PLC platform. The principal components of an individual division are fault tolerant group controllers, loop controllers, a test and interface processor, a cabinet operator module and a control channel gateway. The topical report is submitted to the regulatory body, and got safety evaluation report from the regulatory body. Also, the developed system is tested in the integrated performance validation facility. It is decided that the digital safety system applied to Shin-Uljin unit 1 and 2 after a topical report approval and validation test. Design changes occur in the digital safety system that is applied to an actual nuclear power plant construction, and the PLC has also been upgraded

Development and application of digital safety system in NPPs

Energy Technology Data Exchange (ETDEWEB)

Kwon, Keechoon; Kim, Changhwoi; Lee, Dongyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-03-15

This paper describes the development of digital safety system in NPPs based on safety- grade programmable logic controller (PLC) platform and its application to real NPP construction. The digital safety system consists of a reactor protection system and an engineered safety feature-component control system. The safety-grade PLC platform was developed so that it meets the requirements of the regulation. The PLC consists of various modules such as a power module, a processor module, communication modules, digital input/output modules, analog input/output modules, a LOCA bus extension module, and a high-speed pulse counter module. The reactor protection system is designed with a redundant 4-channel architecture, and every channel is implemented with the same architecture. A single channel consists of a redundant bi-stable processor, a redundant coincidence processor, an automatic test and interface processor, and a cabinet operator module. The engineered safety feature-component control system is designed with four redundant divisions, and implemented with the PLC platform. The principal components of an individual division are fault tolerant group controllers, loop controllers, a test and interface processor, a cabinet operator module and a control channel gateway. The topical report is submitted to the regulatory body, and got safety evaluation report from the regulatory body. Also, the developed system is tested in the integrated performance validation facility. It is decided that the digital safety system applied to Shin-Uljin unit 1 and 2 after a topical report approval and validation test. Design changes occur in the digital safety system that is applied to an actual nuclear power plant construction, and the PLC has also been upgraded.
Towards a decision support system for control of multiple food safety hazards in raw milk production

NARCIS (Netherlands)

Spiegel, van der M.; Sterrenburg, P.; Haasnoot, W.; Fels-Klerx, van der H.J.

2013-01-01

Decision support systems (DSS) for controlling multiple food safety hazards in raw milk production have not yet been developed, but the underlying components are fragmentarily available. This article presents the state-of-the-art of essential DSS elements for judging food safety compliance of raw
The safety interlocking system at the NAC

International Nuclear Information System (INIS)

Visser, K.; Mostert, H.

1984-01-01

The central safety interlocking system (CSIS) controls the higher level of interlocking between the various cyclotron subsystems. It ensures the safe operation of the entire cyclotron facility as regards personnel safety and proper instrument operation. The system consists of a micro-processor with a ROM-based safety interlocking program, relay output modules providing ''safety OK'' instructions to all interlocked apparatus, alarm input modules connected to transducers providing binary alarm status signals and an interface to the central control computer. All solid state electronic components of the system are situated in a low level radiation area and are interfaced to cyclotron equipment by means of 24 V relays
Field Programmable Gate Array-based I and C Safety System

International Nuclear Information System (INIS)

Kim, Hyun Jeong; Kim, Koh Eun; Kim, Young Geul; Kwon, Jong Soo

2014-01-01

Programmable Logic Controller (PLC)-based I and C safety system used in the operating nuclear power plants has the disadvantages of the Common Cause Failure (CCF), high maintenance costs and quick obsolescence, and then it is necessary to develop the other platform to replace the PLC. The Field Programmable Gate Array (FPGA)-based Instrument and Control (I and C) safety system is safer and more economical than Programmable Logic Controller (PLC)-based I and C safety system. Therefore, in the future, FPGA-based I and C safety system will be able to replace the PLC-based I and C safety system in the operating and the new nuclear power plants to get benefited from its safety and economic advantage. FPGA-based I and C safety system shall be implemented and verified by applying the related requirements to perform the safety function
Field Programmable Gate Array-based I and C Safety System

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun Jeong; Kim, Koh Eun; Kim, Young Geul; Kwon, Jong Soo [KEPCO, Daejeon (Korea, Republic of)

2014-08-15

Programmable Logic Controller (PLC)-based I and C safety system used in the operating nuclear power plants has the disadvantages of the Common Cause Failure (CCF), high maintenance costs and quick obsolescence, and then it is necessary to develop the other platform to replace the PLC. The Field Programmable Gate Array (FPGA)-based Instrument and Control (I and C) safety system is safer and more economical than Programmable Logic Controller (PLC)-based I and C safety system. Therefore, in the future, FPGA-based I and C safety system will be able to replace the PLC-based I and C safety system in the operating and the new nuclear power plants to get benefited from its safety and economic advantage. FPGA-based I and C safety system shall be implemented and verified by applying the related requirements to perform the safety function.
Intermediate probabilistic safety assessment approach for safety critical digital systems

International Nuclear Information System (INIS)

Taeyong, Sung; Hyun Gook, Kang

2001-01-01

Even though the conventional probabilistic safety assessment methods are immature for applying to microprocessor-based digital systems, practical needs force to apply it. In the Korea, UCN 5 and 6 units are being constructed and Korean Next Generation Reactor is being designed using the digital instrumentation and control equipment for the safety related functions. Korean regulatory body requires probabilistic safety assessment. This paper analyzes the difficulties on the assessment of digital systems and suggests an intermediate framework for evaluating their safety using fault tree models. The framework deals with several important characteristics of digital systems including software modules and fault-tolerant features. We expect that the analysis result will provide valuable design feedback. (authors)
INTEGRATED SAFETY MANAGEMENT SYSTEM IN AIR TRAFFIC SERVICES

Directory of Open Access Journals (Sweden)

Volodymyr Kharchenko

2014-06-01

Full Text Available The article deals with the analysis of the researches conducted in the field of safety management systems.Safety management system framework, methods and tools for safety analysis in Air Traffic Control have been reviewed.Principles of development of Integrated safety management system in Air Traffic Services have been proposed.
PWR hybrid computer model for assessing the safety implications of control systems

International Nuclear Information System (INIS)

Smith, O.L.; Renier, J.P.; Difilippo, F.C.; Clapp, N.E.; Sozer, A.; Booth, R.S.; Craddick, W.G.; Morris, D.G.

1986-03-01

The ORNL study of safety-related aspects of nuclear power plant control systems consists of two interrelated tasks: (1) failure mode and effects analysis (FMEA) that identified single and multiple component failures that might lead to significant plant upsets and (2) computer models that used these failures as initial conditions and traced the dynamic impact on the control system and remainder of the plant. This report describes the simulation of Oconee Unit 1, the first plant analyzed. A first-principles, best-estimate model was developed and implemented on a hybrid computer consisting of AD-4 analog and PDP-10 digital machines. Controls were placed primarily on the analog to use its interactive capability to simulate operator action. 48 refs., 138 figs., 15 tabs
Tests on instrumentation and control systems important to safety in nuclear power stations. Systempruefung der leittechnischen Einrichtungen des Sicherheitssystems in Kernkraftwerken

Energy Technology Data Exchange (ETDEWEB)

1985-01-01

The rule applies to the reactor protection system, to the protection and state boundaries, to control devices important to safety, and to danger alarms of the classes S and I. The system inspection of the control devices of the safety system comprises in-service testing and recurrent testing.
Design of an Active Automotive Safety System

Directory of Open Access Journals (Sweden)

Y. Wang

2013-07-01

Full Text Available With the development of the national economy, the people's standard of living got corresponding improvement, cars has been one of the indispensable traffic tools in many families. An active safety system is proposed, which can real-time detect the vehicle's running status and judge the security status of the vehicle. The system, which takes single-chip microcomputer as the controlling core and combines with millimeter-wave and ultrasonic distance measurement technology, can detect the distance from vehicle to vehicle and judge the security status of the vehicle. The hardware composition of the system and the data acquiring circuit are proposed, the mathematic model for different situation is established, and the controlling algorithm is completed. This system can accurately measure speed and distance between vehicles; the active safety control system can meet the relevant data measurement and transmission requirement; and can meet the functional requirement of the active safety control system
Design lessons from using programmable controllers in the MFTF-B personnel safety and interlocks system

International Nuclear Information System (INIS)

Branum, J.D.

1983-01-01

Applying programmable controllers in critical applications such as personnel safety and interlocks systems requires special considerations in the design of both hardware and software. All modern programmable controller systems feature extensive internal diagnostic capabilities to protect against problems such as program memory errors; however most, if not all present designs lack an intrinsic capability for detecting and countering failures on the field-side of their I/O modules. Many of the most common styles of I/O modules can also introduce potentially dangerous sneak circuits, even without component failure. This paper presents the most significant lessons learned to date in the design of the MFTF-B Personnel Safety and Interlocks System, which utilizes two non-redundant programmable controllers with over 800 I/O points each. Specific problems recognized during the design process as well as those discovered during initial testing and operation are discussed along with their specific solutions in hardware and software
LOFT integral test system final safety analysis report

International Nuclear Information System (INIS)

1974-03-01

Safety analyses are presented for the following LOFT Reactor systems: engineering safety features; support buildings and facilities; instrumentation and controls; electrical systems; and auxiliary systems. (JWR)
Nuclear safety considerations with emphasis on instrumentation and control systems

International Nuclear Information System (INIS)

Beare, J.W.

1978-01-01

The conceptual model of a nuclear power plant in Canada is that it consists basically of two kinds of systems. The first kind is the process systems, that is, those structures and components associated with the production of nuclear energy and its conversion to other forms of energy. The second kind is the special safety systems, whose purpose it is to protect the public in the event of a serious failure in the process systems which might otherwise lead to unacceptable radiological consequences. Quantitative limits are set on the unavailability of the special safety systems. These limits are low enough to be consistent with low overall risk and yet can be demonstrated by test during operation of the plant. Low unavailability is an important but not the only condition required for low unrealiability for the special safety systems. The special safety systems minimize the chance of a cross-linked failure particularly under the conditions experienced as a result of the more severe types of postulated serious process failures. Nuclear power plants must also withstand, without a major hazard to the public, certain rare events associated with natural phenomena or man-made activities off-site and also certain in-plant events such as fire or break-up of a turbine-generator which might have a cross-linking effect on process and safety systems. In the latest designs, Canadian nuclear power plants have emergency systems to deal with such events. The emergency systems have an enhanced degree of physical and functional separation from other plant systems. (author)
Safety climate and culture: Integrating psychological and systems perspectives.

Science.gov (United States)

Casey, Tristan; Griffin, Mark A; Flatau Harrison, Huw; Neal, Andrew

2017-07-01

Safety climate research has reached a mature stage of development, with a number of meta-analyses demonstrating the link between safety climate and safety outcomes. More recently, there has been interest from systems theorists in integrating the concept of safety culture and to a lesser extent, safety climate into systems-based models of organizational safety. Such models represent a theoretical and practical development of the safety climate concept by positioning climate as part of a dynamic work system in which perceptions of safety act to constrain and shape employee behavior. We propose safety climate and safety culture constitute part of the enabling capitals through which organizations build safety capability. We discuss how organizations can deploy different configurations of enabling capital to exert control over work systems and maintain safe and productive performance. We outline 4 key strategies through which organizations to reconcile the system control problems of promotion versus prevention, and stability versus flexibility. (PsycINFO Database Record (c) 2017 APA, all rights reserved).
Safety control system and its interface to EPICS for the off-line front end of the SPES project

International Nuclear Information System (INIS)

Vasquez, J.; Andrighetto, A.; Bassato, G.; Costa, L.; Giacchini, M.; Bertocco, M.

2012-01-01

The SPES (Selective Production of Exotic Species) project is based on a facility for the production of neutron-rich radioactive ion beams using the isotope separation on-line technique. The SPES off-line front-end apparatus involves a number of subsystems and procedures that are potentially dangerous both for human operators and for the equipment. The high voltage power supply, the ion source complex power supplies, the target chamber handling systems and the laser source are some example of these subsystems. For that reason, a safety control system has been developed. It is based on Schneider Electrics Preventa family safety modules that control the power supply of critical subsystems in combination with safety detectors that monitor critical variables. A Programmable Logic Controller (PLC), model BMXP342020 from the Schneider Electrics Modicon M340 family, is used for monitoring the status of the system as well as controlling the sequence of some operations in automatic way. A touch screen, model XBTGT5330 from the Schneider Electrics Magelis family, is used as Human Machine Interface (HMI) and communicates with the PLC using MODBUS-TCP. Additionally, an interface to the EPICS control network was developed using a home-made MODBUS-TCP EPICS driver in order to integrate it to the control system of the Front End as well as present the status of the system to the users on the main control panel. (authors)
Safety Evaluation of Kartini Reactor Based on Instrumentation System Design

International Nuclear Information System (INIS)

Tjipta Suhaemi; Djen Djen Dj; Itjeu K; Johnny S; Setyono

2003-01-01

The safety of Kartini reactor has been evaluated based on instrumentation system aspect. The Kartini reactor is designed by BATAN. Design power of the reactor is 250 kW, but it is currently operated at 100 kW. Instrumentation and control system function is to monitor and control the reactor operation. Instrumentation and control system consists of safety system, start-up and automatic power control, and process information system. The linear power channel and logarithmic power channel are used for measuring power. There are 3 types of control rod for controlling the power, i.e. safety rod, shim rod, and regulating rod. The trip and interlock system are used for safety. There are instrumentation equipment used for measuring radiation exposure, flow rate, temperature and conductivity of fluid The system of Kartini reactor has been developed by introducing a process information system, start-up system, and automatic power control. It is concluded that the instrumentation of Kartini reactor has followed the requirement and standard of IAEA. (author)
Safety Review related to Commercial Grade Digital Equipment in Safety System

International Nuclear Information System (INIS)

Yu, Yeongjin; Park, Hyunshin; Yu, Yeongjin; Lee, Jaeheung

2013-01-01

The upgrades or replacement of I and C systems on safety system typically involve digital equipment developed in accordance with non-nuclear standards. However, the use of commercial grade digital equipment could include the vulnerability for software common-mode failure, electromagnetic interference and unanticipated problems. Although guidelines and standards for dedication methods of commercial grade digital equipment are provided, there are some difficulties to apply the methods to commercial grade digital equipment for safety system. This paper focuses on regulatory guidelines and relevant documents for commercial grade digital equipment and presents safety review experiences related to commercial grade digital equipment in safety system. This paper focuses on KINS regulatory guides and relevant documents for dedication of commercial grade digital equipment and presents safety review experiences related to commercial grade digital equipment in safety system. Dedication including critical characteristics is required to use the commercial grade digital equipment on safety system in accordance with KEPIC ENB 6370 and EPRI TR-106439. The dedication process should be controlled in a configuration management process. Appropriate methods, criteria and evaluation result should be provided to verify acceptability of the commercial digital equipment used for safety function
Replacement of the complete control system of the NPP Oskarshamn 1 by digital distributed control system

International Nuclear Information System (INIS)

Berger, E.

1998-01-01

As part of an ongoing modernization program, Oskarshamn 1's I and C system and control room will be upgraded by ABB using its Advant Power range of digital, programmable process control system. Besides ensuring the higher level of safety that is demanded today, the new equipment provides the plant with an integrated system which will improve operator overview of operation and reduce risk of human error and serve as a platform for further improvements of the control room. This paper discusses in the example of Oskarshamn 1 how the complete control system of a nuclear power plant may be exchanged, the technical solution and the time schedule. Oskarshamn 1 is the first nuclear power plant in Sweden. It is a boiling water reactor built by ABB ATOM in Sweden between 1966 and 1971. According to this age the control system is semiconductor based and the reactor protection system is relays based. This makes the maintenance expensive and extensions nearly impossible. To extend the life period of this plant the owner has decided to improve the safety system and to replace the reactor protection system and safety related control and the non safety related control by a state of the art digital distributed control system of ABB. In March 1997 ABB got the order to replace the reactor protection system, the safety control system, to start the replacement of all control systems and to replace the old control room by a new ergonomically designed control room. Together with the exchange of the control system an enhancement of the safety system and of the emergency power supply will be implemented
Controls in new construction reactors-factory testing of the non-safety portion of the Lungmen nuclear power plant distributed control system

International Nuclear Information System (INIS)

Wu, Y. S.; Dick, J. W.; Tetirick, C. W.

2006-01-01

The construction permit for Taipower's Lungmen Nuclear Units 1 and 2, two ABWR plants, was issued on March 17, 1999[1], The construction of these units is progressing actively at site. The digital I and C system supplied by GE, which is designated as the Distributed Control and Information System (DCIS) in this project, is being implemented primarily at one vendor facility. In order to ensure the reliability, safety and availability of the DCIS, it is required to comprehensively test the whole DCIS in factory. This article describes the test requirements and acceptance criteria for functional testing of the Non-Safety Distributed Control and Information system (DCIS) for Taiwan Power's Lungmen Units 1 and 2 GE selected Invensys as the equipment supplier for this Non-Safety portion of DCIS. The DCIS system of the Lungmen Units is a physically distributed control system. Field transmitters are connected to hard I/O terminal inputs on the Invensys I/A system. Once the signal is digitized on FBMs (Field Bus Modules) in Remote Multiplexing Units (RMUs), the signal is passed into an integrated control software environment. Control is based on the concept of compounds and blocks where each compound is a logical collection of blocks that performs a control function. Each point identified by control compound and block can be individually used throughout the DCIS system by referencing its unique name. In the Lungmen Project control logic and HSI (Human System Interface) requirements are divided into individual process systems called MPLs (Master Parts List). Higher-level Plant Computer System (PCS) algorithms access control compounds and blocks in these MPLs to develop functions. The test requirements and acceptance criteria for the DCIS system of the Lungmen Project are divided into three general categories (see 1,2,3 below) of verification, which in turn are divided into several specific tests: 1. DCIS System Physical Checks a) RMU Test - To confirm that the hard I
Controls in new construction reactors-factory testing of the non-safety portion of the Lungmen nuclear power plant distributed control system

Energy Technology Data Exchange (ETDEWEB)

Wu, Y. S. [Taiwan Power Company, 242, Roosevelt Road, Sec. 3, Taipei 100, Taiwan (China); Dick, J. W. [Invensys System Inc., 33 Commercial St., Foxboro, MA 02035 (United States); Tetirick, C. W. [GE Energy, 1989 Little Orchard Street, San Jose, CA 95125-1030 (United States)

2006-07-01

The construction permit for Taipower's Lungmen Nuclear Units 1 and 2, two ABWR plants, was issued on March 17, 1999[1], The construction of these units is progressing actively at site. The digital I and C system supplied by GE, which is designated as the Distributed Control and Information System (DCIS) in this project, is being implemented primarily at one vendor facility. In order to ensure the reliability, safety and availability of the DCIS, it is required to comprehensively test the whole DCIS in factory. This article describes the test requirements and acceptance criteria for functional testing of the Non-Safety Distributed Control and Information system (DCIS) for Taiwan Power's Lungmen Units 1 and 2 GE selected Invensys as the equipment supplier for this Non-Safety portion of DCIS. The DCIS system of the Lungmen Units is a physically distributed control system. Field transmitters are connected to hard I/O terminal inputs on the Invensys I/A system. Once the signal is digitized on FBMs (Field Bus Modules) in Remote Multiplexing Units (RMUs), the signal is passed into an integrated control software environment. Control is based on the concept of compounds and blocks where each compound is a logical collection of blocks that performs a control function. Each point identified by control compound and block can be individually used throughout the DCIS system by referencing its unique name. In the Lungmen Project control logic and HSI (Human System Interface) requirements are divided into individual process systems called MPLs (Master Parts List). Higher-level Plant Computer System (PCS) algorithms access control compounds and blocks in these MPLs to develop functions. The test requirements and acceptance criteria for the DCIS system of the Lungmen Project are divided into three general categories (see 1,2,3 below) of verification, which in turn are divided into several specific tests: 1. DCIS System Physical Checks a) RMU Test - To confirm that the hard

Design and research of safety monitor and control system based on CAN BUS

International Nuclear Information System (INIS)

Wen Xinling; Chen Yu; Zhang Zhen; Zhao Yubin

2007-01-01

In Order to protect machine operator under danger work area in producing-manufacturing industry, we present a distributed safety monitor and control system based on CAN BUS technology. The detection signal is collected based on the photo-voltage characteristics of the infrared sensor and it was processed with the core of AT89C51. The microprocessor controls the CAN BUS controller SJA1000/transceiver PCA82C250 to structure CAN BUS communication system to transmit the data. Through the serial interface MAX232 connected main controller with each control node, PC can monitor and control each machine in real time and renew control scheme. This paper introduces composition principle and the methods of hardware design in detail. Experiments shown that the system has yield control precision of 0.1 mm, defend distance more than 15 m and the measurement accuracy of 100%. Moreover, it can realize to reform FA431 and monitor cotton-breaking, yarn-breaking and product quality. Productivity is improved about 25%-35%. (authors)
Upgrading instrumentation and control systems for plant safety and operation

International Nuclear Information System (INIS)

Martin, M.; Prehler, H.J.; Schramm, W.

1997-01-01

Upgrading the electrical systems and instrumentation and control systems has become increasingly more important in the past few years for nuclear power plants currently in operation. As the requirements to be met in terms of plant safety and availability have become more stringent in the past few years, Western plants built in the sixties and seventies have been the subject of manifold backfitting and upgrading measures in the past. In the meantime, however, various nuclear power plants are facing much more thorough upgrading phases because of the difficulties in obtaining spare parts for older equipment systems. As digital technology has become widespread in many areas because of its advantages, and as applications are continuously expanding, conventional equipment and systems are losing more and more ground as a consequence of decreasing demand. Merely because of the pronounced decline in demand for conventional electronic components it is possible for equipment manufacturers to guarantee spare parts deliveries for older systems only for specific future periods of time. In addition, one-off manufacture entails high costs in purchases of spare parts. As a consequence of current thinking more and more focusing on availability and economy, upgrading of electrical systems and instrumentation and control systems is becoming a more and more topical question, for older plants even to ensure completion of full service life. (orig.) [de
Improving the safety and protective automatic actions of the CMS electromagnetic calorimeter detector control system

CERN Document Server

Jimenez Estupinan, Raul; Cirkovic, Predrag; Di Calafiori, Diogo Raphael; Dissertori, Guenther; Djambazov, Lubomir; Jovanovic, Dragoslav; Lustermann, Werner; Milenovic, Predrag; Zelepoukine, Serguei

2017-01-01

The CMS ECAL Detector Control System (DCS) features several monitoring mechanisms able to react and perform automatic actions based on pre-defined action matrices. The DCS is capable of early detection of anomalies inside the ECAL and on its off-detector support systems, triggering automatic actions to mitigate the impact of these events and preventing them from escalating to the safety system. The treatment of such events by the DCS allows for a faster recovery process, better understanding of the development of issues, and in most cases, actions with higher granularity than the safety system. This paper presents the details of the DCS automatic action mechanisms, as well as their evolution based on several years of CMS ECAL operations.
Safety and control of accelerator-driven subcritical systems

Energy Technology Data Exchange (ETDEWEB)

Rief, H. [Ispra Establishment (Italy); Takahashi, H. [Brookhaven National Laboratory, Long Island, NY (United States)

1995-10-01

To study control and safety of accelertor driven nuclear systems, a one point kinetic model was developed and programed. It deals with fast transients as a function of reactivity insertion. Doppler feedback, and the intensity of an external neutron source. The model allows for a simultaneous calculation of an equivalent critical reactor. It was validated by a comparison with a benchmark specified by the Nuclear Energy Agency Committee of Reactor Physics. Additional features are the possibility of inserting a linear or quadratic time dependent reactivity ramp which may account for gravity induced accidents like earthquakes, the possibility to shut down the external neutron source by an exponential decay law of the form exp({minus}t/{tau}), and a graphical display of the power and reactivity changes. The calculations revealed that such boosters behave quite benignly even if they are only slightly subcritical.
SCOPE safety-controls optimization by performance evaluation: A systematic approach for safety-related decisions at the Hanford Tank Remediation System. Phase 1, final report

International Nuclear Information System (INIS)

Bergeron, K.D.; Williams, D.C.; Slezak, S.E.; Young, M.L.

1996-12-01

The Department of Energy's Hanford Tank Waste Remediation system poses a significant challenge for hazard management because of the uncertainty that surrounds many of the variables that must be considered in decisions on safety and control strategies. As a result, site managers must often operate under excessively conservative and expensive assumptions. This report describes a systematic approach to quantifying the uncertainties surrounding the critical parameters in control decisions (e.g., condition of the tanks, kinds of wastes, types of possible accidents) through the use of expert elicitation methods. The results of the elicitations would then be used to build a decision support system and accident analysis model that would allow managers to see how different control strategies would affect the cost and safety of a facility configuration
Safety aspects of core power distribution surveillance and control

International Nuclear Information System (INIS)

Beraha, D.; Grumbach, R.; Hoeld, A.; Werner, W.

1978-01-01

The incentives for improved core surveillance and core control systems are outlined. An efficient code for evaluating the power distribution is indispensable for designing and testing such a system. The characteristics of the core simulator QUABOX/CUBBOX and the features required for off-line and on-line applications are described. The important role of the simulator for the safety assessment of a digital core control system is underlined. With regard to the safety aspects of core control, possible disturbances are classified. Simulation results are given concerning the failure of a control actuator. It is shown that means can be devised to prevent unstable behaviour of the control system and, furthermore, to contribute to a safe reactor operation by accounting for process disturbances. (author)
Automated safety control by video cameras

NARCIS (Netherlands)

Lefter, I.; Rothkrantz, L.; Somhorst, M.

2012-01-01

At this moment many surveillance systems are installed in public domains to control the safety of people and properties. They are constantly watched by human operators who are easily overloaded. To support the human operators, a surveillance system model is designed that detects suspicious behaviour
Towards the certification of non-deterministic control systems for safety-critical applications: analysing aviation analogies for possible certification strategies

CSIR Research Space (South Africa)

Burger, CR

2011-11-01

Full Text Available Current certification criteria for safety-critical systems exclude non-deterministic control systems. This paper investigates the feasibility of using human-like monitoring strategies to achieve safe non-deterministic control using multiple...
Radiation safety and control

International Nuclear Information System (INIS)

Kim, Jang Hee; Kim, Gi Sub.

1996-12-01

The principal objective of radiological safety control is intended for achievement and maintenance of appropriately safe condition in environmental control for activities involving exposure from the use of radiation. In order to establish these objective, we should be to prevent deterministic effects and to limit the occurrence stochastic effects to level deemed to be acceptable by the application of general principles of radiation protection and systems of dose limitation based on ICRP recommendations. (author). 22 tabs., 13 figs., 11 refs
Radiological safety and control

International Nuclear Information System (INIS)

Kim, Jang Hee; Kim, Ki Sub

1995-01-01

The practical objective of radiological safety control is intended for achievement and maintenance of appropreately safe condition in environmental control for activities involving exposure from the use of radiation. In order to establish these objectives, we should be to prevent deterministic effects and to limit the occurrence stochastic effects to level deemed to be acceptable by the application of general principles of radiation protection and systems of dose limitation based on ICRP recommendations. 34 tabs., 19 figs., 11 refs. (Author) .new
Evaluating Safety Culture Under the Socio-Technical Complex Systems Perspective

International Nuclear Information System (INIS)

Lemos, F. L. de

2016-01-01

procedures for equipment operation are ignored. However, when it comes to more subtle interactions between system components, it becomes harder to detect potentially hazardous situations that are hidden, and can lead the system to hazardous states. For example, leaders can take decisions that are in conflict with decisions taken by other colleagues at a very different department, and without knowing, be contributing to future unintended consequences to the system. Such a situation may not be easily detected by direct observation. This explains why having a good safety culture seems not to be enough to assure the safety of the system. According to STAMP principals, safety is a problem of flaws in the control of the interactions between components of the system, and not only a problem of failures of components of the system. Remember that safety culture defines a property of part of the system, which could be considered as a component of the system. We can find examples of companies that, even having well evaluated safety culture, or organizational culture, fail to keep their high safety standards. In this work we propose a methodology that integrates safety culture in the control structure of the system. It is based on STAMP: Systems Theoretic Accident Models and Processes, and the Three Lenses: Strategic, Political and Cultural Approaches. It can help evaluate either the existing safety culture of a Nuclear Power Plant or the implementation of new safety culture projects. STAMP is based on the assumption that accidents are a result of flawed control over the interactions between components of a system. Where, control structure is a model of the system in terms of control loops. To understand how the control structure of a system can be corrupted, and therefore, leading the system to hazardous conditions, the methodology of the Three Lenses is applied. By following this approach it becomes possible to keep all the safety culture traits but, instead of focus on safety culture
Technical features of ABWR safety systems

International Nuclear Information System (INIS)

Sugisaki, Toshihiko; Tominaga, Kenji; Horiuchi, Tetsuo

1986-01-01

The engineering safety facilities of ABWRs have been disigned so as to have many excellent characteristics such as safety, reliability and economy, reflecting the merit of adopting new technology such as internal pumps and new control rod driving mechanism, and coupled with the safety peculiar to BWRs. In this paper, about ECCS, containment vessels and others which compose the engineering safety facilities of ABWRs, the characteristics related to the safety owing to the adoption of internal pumps and others, and the evaluation of the performance at the time of various accidents are discussed. As the results of safety evaluation, it was clarified that due to the safety peculiar to ABWRs and the characteristics of the safety facilities, the large increases of safety, reliability and economy have been planned in the ABWRs, and for example, core flooding can be maintained even at the time of a hypothetical loss of coolant accident. BWRs have the simple system constitution, good self controllability, large natural circulation ability, simple operation control method and excellent ability of confining heat and radioactivity. BWRs have three safety functions to stop reactors, to remove heat from reactors, and to confine radioactive substances. These functions of ABWRs were evaluated, and very high safety was confirmed. (Kako, I.)
Adoption of digital safety protection system in Japan

International Nuclear Information System (INIS)

Ogiso, Z.

1998-01-01

The application of micro-processor-based digital controllers has been widely propagated among various industries in recent years. While in the nuclear power plant industry, the application of them has also been expanding gradually starting from non-safety related systems, taking advantage of their reliability and maintainability over the conventional analog devices. Based on the careful study of the feasibility of digital controllers to the safety protection system, the Tokyo Electric Power Company proposed on May 1989 the adoption of digital controllers to the safety protection system in the Application for Permission of Establishment of Kashiwazaki-Kariwa units 6 and 7 (ABWR-1350Mwe each). MITI, Ministry of International Trade and Industry, the Japanese regulatory body for electric power generating facilities, had approved this application after careful review. This paper describes a series of supporting activities leading to the MITI's approval of the digital safety protection system and the MITI's licensing activities. (author)
SCOPE safety-controls optimization by performance evaluation: A systematic approach for safety-related decisions at the Hanford Tank Remediation System. Phase 1, final report

Energy Technology Data Exchange (ETDEWEB)

Bergeron, K.D.; Williams, D.C.; Slezak, S.E.; Young, M.L. [and others

1996-12-01

The Department of Energy`s Hanford Tank Waste Remediation system poses a significant challenge for hazard management because of the uncertainty that surrounds many of the variables that must be considered in decisions on safety and control strategies. As a result, site managers must often operate under excessively conservative and expensive assumptions. This report describes a systematic approach to quantifying the uncertainties surrounding the critical parameters in control decisions (e.g., condition of the tanks, kinds of wastes, types of possible accidents) through the use of expert elicitation methods. The results of the elicitations would then be used to build a decision support system and accident analysis model that would allow managers to see how different control strategies would affect the cost and safety of a facility configuration.
30 CFR 7.103 - Safety system control test.

Science.gov (United States)

2010-07-01

... Areas of Underground Coal Mines Where Permissible Electric Equipment is Required § 7.103 Safety system... operate immediately when activated and stop the engine within 15 seconds. (6) The total intake air inlet...
A holistic strategy for quality and safety control of traditional Chinese medicines by the "iVarious" standard system.

Science.gov (United States)

Chen, Anzhen; Sun, Lei; Yuan, Hang; Wu, Aiying; Lu, Jingguang; Ma, Shuangcheng

2017-10-01

An effective quality control system is the key to ensuring the quality, safety and efficacy of traditional Chinese medicines (TCMs). However, the current quality standard research lacks the top-design and systematic design, mostly based on specific technologies or evaluation methods. To resolve the challenges and questions of quality control of TCMs, a brand-new quality standard system, named "iVarious", was proposed. The system comprises eight elements in a modular format. Meaning of every element was specifically illustrated via corresponding research instances. Furthermore, frankincense study was taken as an example for demonstrating standards and research process, based on the "iVarious" system. This system highlighted a holistic strategy for effectiveness, security, integrity and systematization of quality and safety control standards of TCMs. The establishment of "iVarious" integrates multi-disciplinary technologies and progressive methods, basis elements and key points of standard construction. The system provides a novel idea and technological demonstration for regulation establishment of TCMs quality standards.
Logical safety system for triggering off the protection action of a safety actuator

International Nuclear Information System (INIS)

Plaige, Yves.

1982-01-01

This invention applies in particular to the emergency triggering of safety actuators controlling the shutdown of a nuclear reactor. This logical safety system includes four redundant lines each composed, inter alia, of a logical circuit for controlling the triggering of a protection action, a logical alarm circuit connected to the control circuit and a logical inhibiting circuit making it impossible to inhibit several alarm circuits simultaneously [fr
PWR hybrid computer model for assessing the safety implications of control systems

International Nuclear Information System (INIS)

Smith, O.L.; Booth, R.S.; Clapp, N.E.; DiFilippo, F.C.; Renier, J.P.; Sozer, A.

1985-01-01

The ORNL study of safety-related aspects of control systems consists of two interrelated tasks, (1) a failure mode and effects analysis that, in part, identifies single and multiple component failures that may lead to significant plant upsets, and (2) a hybrid computer model that uses these failures as initial conditions and traces the dynamic impact on the control system and remainder of the plant. The second task is reported here. The initial step in model development was to define a suitable interface between the FMEA and computer simulation tasks. This involved identifying primary plant components that must be simulated in dynamic detail and secondary components that can be treated adequately by the FMEA alone. The FMEA in general explores broader spectra of initiating events that may collapse into a reduced number of computer runs. A portion of the FMEA includes consideration of power supply failures. Consequences of the transients may feedback on the initiating causes, and there may be an interactive relationship between the FMEA and the computer simulation. Since the thrust of this program is to investigate control system behavior, the controls are modeled in detail to accurately reproduce characteristic response under normal and off-normal transients. The balance of the model, including neutronics, thermohydraulics and component submodels, is developed in sufficient detail to provide a suitable support for the control system
Control and safety systems for TRIGA irradiation facilities C5 and C9

International Nuclear Information System (INIS)

Talpalariu Cornel Talpalariu Jeni Crucean Mircea Matei Corina

2008-01-01

Full text: The Institute for Nuclear Research conducted research for designing and manufacturing of microprocessor equipment for some irradiation facilities operating by the TRIGA reactor. This equipment has accumulated a wide operating time allowing the conclusions referring to reliability, ergonomics, and design of the operating facilities. Based upon these studies a new program was initiated for the design and manufacturing of a modern equipment with improved reliability and flexibility performances. The system provides the user with a multitude of options, numerical and analog interfaces, keyboard and high reliability local display. The main functional components of the system are: - 8 PID full options regulating loops; - 8 safety analog channels having 4 preset trips; - watch dog restart and fault tolerant facilities; - 8 high precision analog with an input of 0 - 15 mV from thermocouple; - 8 computer controlled power supplies of 220 V, 1 kWA; - alphanumeric display and keyboard; - fault tolerant analog scanner. A real improvement of the system is the future remote control computer, a PC AT Pentium working like a system controller, real time data acquisition, and operator's adviser. This new facility allows the operator to set the trips or to control remotely all the power supply and step-by-step positioner of irradiation device. Software design for acquisition and data processing provides modern techniques for operator interfacing, representation recording and protection of test results. Software implementation keeps a special organization supported by a real time executive that is the best method to achieve the performance required. Following this objective, the software structure consists of: 1. Tasks as follows: - testing parameters setup; - data processing routines; - engineering and electrical conversion; - numerical / graphical data representation; - test results recording routines and data base management. 2. Drivers as follows: - A/I and D
Control of Industrial Safety Based on Dynamic Characteristics of a Safety Budget-Industrial Accident Rate Model in Republic of Korea.

Science.gov (United States)

Choi, Gi Heung; Loh, Byoung Gook

2017-06-01

Despite the recent efforts to prevent industrial accidents in the Republic of Korea, the industrial accident rate has not improved much. Industrial safety policies and safety management are also known to be inefficient. This study focused on dynamic characteristics of industrial safety systems and their effects on safety performance in the Republic of Korea. Such dynamic characteristics are particularly important for restructuring of the industrial safety system. The effects of damping and elastic characteristics of the industrial safety system model on safety performance were examined and feedback control performance was explained in view of cost and benefit. The implications on safety policies of restructuring the industrial safety system were also explored. A strong correlation between the safety budget and the industrial accident rate enabled modeling of an industrial safety system with these variables as the input and the output, respectively. A more effective and efficient industrial safety system could be realized by having weaker elastic characteristics and stronger damping characteristics in it. A substantial decrease in total social cost is expected as the industrial safety system is restructured accordingly. A simple feedback control with proportional-integral action is effective in prevention of industrial accidents. Securing a lower level of elastic industrial accident-driving energy appears to have dominant effects on the control performance compared with the damping effort to dissipate such energy. More attention needs to be directed towards physical and social feedbacks that have prolonged cumulative effects. Suggestions for further improvement of the safety system including physical and social feedbacks are also made.

Density Control of Multi-Agent Systems with Safety Constraints: A Markov Chain Approach

Science.gov (United States)

Demirer, Nazli

The control of systems with autonomous mobile agents has been a point of interest recently, with many applications like surveillance, coverage, searching over an area with probabilistic target locations or exploring an area. In all of these applications, the main goal of the swarm is to distribute itself over an operational space to achieve mission objectives specified by the density of swarm. This research focuses on the problem of controlling the distribution of multi-agent systems considering a hierarchical control structure where the whole swarm coordination is achieved at the high-level and individual vehicle/agent control is managed at the low-level. High-level coordination algorithms uses macroscopic models that describes the collective behavior of the whole swarm and specify the agent motion commands, whose execution will lead to the desired swarm behavior. The low-level control laws execute the motion to follow these commands at the agent level. The main objective of this research is to develop high-level decision control policies and algorithms to achieve physically realizable commanding of the agents by imposing mission constraints on the distribution. We also make some connections with decentralized low-level motion control. This dissertation proposes a Markov chain based method to control the density distribution of the whole system where the implementation can be achieved in a decentralized manner with no communication between agents since establishing communication with large number of agents is highly challenging. The ultimate goal is to guide the overall density distribution of the system to a prescribed steady-state desired distribution while satisfying desired transition and safety constraints. Here, the desired distribution is determined based on the mission requirements, for example in the application of area search, the desired distribution should match closely with the probabilistic target locations. The proposed method is applicable for both
Programmable logic controller (PLC) for safety systems of nuclear plants

International Nuclear Information System (INIS)

Sen, S.K.; Karmakar, G.; Joseph, Jose; Patil, R.K.

2002-01-01

Full text: A programmable logic controller (PLC) has been developed by RCnD, BARC for use in the safety critical systems in nuclear power plants. This PLC uses qualified hardware developed in RCnD for use in NPP. The programming software conforms to IEC-61131 part 3. The application programming is done on function block diagram (FBD) editor and the FBD is automatically converted into code in high level language (C / C++). This feature makes the application easily decipherable and therefore easily subjected to reviews and other validation techniques. The key to make quality software for use in nuclear systems is to enforce various standards in the design and development of the software, something, which is not possible to do with a commercially available PLC. This PLC with its software completely transparent lends itself to rigorous verification and validation easily
System and software safety analysis for the ERA control computer

International Nuclear Information System (INIS)

Beerthuizen, P.G.; Kruidhof, W.

2001-01-01

The European Robotic Arm (ERA) is a seven degrees of freedom relocatable anthropomorphic robotic manipulator system, to be used in manned space operation on the International Space Station, supporting the assembly and external servicing of the Russian segment. The safety design concept and implementation of the ERA is described, in particular with respect to the central computer's software design. A top-down analysis and specification process is used to down flow the safety aspects of the ERA system towards the subsystems, which are produced by a consortium of companies in many countries. The user requirements documents and the critical function list are the key documents in this process. Bottom-up analysis (FMECA) and test, on both subsystem and system level, are the basis for safety verification. A number of examples show the use of the approach and methods used
New Paradigm in Nuclear Safety from Quality Assurance to Safety Management System

International Nuclear Information System (INIS)

Lim, Nam-Jin; Park, Chan-Gook; Nam, Ji-Hee; Kim, Kwan-Hyun; Kwon, Hyuk-il; Lee, Young-Gun Lee

2006-01-01

The initial concept of Quality Control (QC) controlling the quality of products is now evolving toward the Management System (MS) achieving safety, through Quality Assurance (QA) ensuring the quality of products and Quality Management (QM) managing the quality by a systematic approach. Nuclear safety can be achieved through an integrated MS that ensures the health, environmental, security, quality and economic requirements being considered together with nuclear safety requirements. MS approach is developed through realizing that most of nuclear accidents had occurred not by the malfunction of hardware or equipment, but by the human error. The MS is a set of inter-related or interacting elements (system) that establishes policies and objectives and which enables those objectives to be achieved in an efficient and effective way
Software for computer based systems important to safety in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based
Software for computer based systems important to safety in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2005-01-01

Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based
Software for computer based systems important to safety in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2000-01-01

Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based
ACP Facility Safety Surveillance System Installation

International Nuclear Information System (INIS)

You, Gil Sung; Kook, D. H.; Choung, W. M.; Ku, J. H.; Cho, I. J.; You, G. S.; Kwon, K. C.; Lee, W. K.; Lee, E. P.

2006-10-01

The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO 2 into U-metal. For demonstration of this process, α-γ type new hotcell was built in the IMEF basement. All facilities which treat radioactive materials must manage CCTV system which is under control of Health Physics department. Three main points (including hotcell rear door area) have each camera, but operators who are in charge of facility management need to check the safety of the facility immediately through the network in his office. This needs introduce additional network cameras installation and this new surveillance system is expected to update the whole safety control ability with existing system
A study of software safety analysis system for safety-critical software

International Nuclear Information System (INIS)

Chang, H. S.; Shin, H. K.; Chang, Y. W.; Jung, J. C.; Kim, J. H.; Han, H. H.; Son, H. S.

2004-01-01

The core factors and requirements for the safety-critical software traced and the methodology adopted in each stage of software life cycle are presented. In concept phase, Failure Modes and Effects Analysis (FMEA) for the system has been performed. The feasibility evaluation of selected safety parameter was performed and Preliminary Hazards Analysis list was prepared using HAZOP(Hazard and Operability) technique. And the check list for management control has been produced via walk-through technique. Based on the evaluation of the check list, activities to be performed in requirement phase have been determined. In the design phase, hazard analysis has been performed to check the safety capability of the system with regard to safety software algorithm using Fault Tree Analysis (FTA). In the test phase, the test items based on FMEA have been checked for fitness guided by an accident scenario. The pressurizer low pressure trip algorithm has been selected to apply FTA method to software safety analysis as a sample. By applying CASE tool, the requirements traceability of safety critical system has been enhanced during all of software life cycle phases
Access safety systems - New concepts from the LHC experience

International Nuclear Information System (INIS)

Ladzinski, T.; Delamare, C.; Luca, S. di; Hakulinen, T.; Hammouti, L.; Havart, F.; Juget, J.F.; Ninin, P.; Nunes, R.; Riesco, T.; Sanchez-Corral Mena, E.; Valentini, F.

2012-01-01

The LHC Access Safety System has introduced a number of new concepts into the domain of personnel protection at CERN. These can be grouped into several categories: organisational, architectural and concerning the end-user experience. By anchoring the project on the solid foundations of the IEC 61508/61511 methodology, the CERN team and its contractors managed to design, develop, test and commission on time a SIL3 safety system. The system uses a successful combination of the latest Siemens redundant safety programmable logic controllers with a traditional relay logic hard wired loop. The external envelope barriers used in the LHC include personnel and material access devices, which are interlocked door-booths introducing increased automation of individual access control, thus removing the strain from the operators. These devices ensure the inviolability of the controlled zones by users not holding the required credentials. To this end they are equipped with personnel presence detectors and the access control includes a state of the art bio-metry check. Building on the LHC experience, new projects targeting the refurbishment of the existing access safety infrastructure in the injector chain have started. This paper summarises the new concepts introduced in the LHC access control and safety systems, discusses the return of experience and outlines the main guiding principles for the renewal stage of the personnel protection systems in the LHC injector chain in a homogeneous manner. (authors)
Regulatory Control of Radiation Sources. Safety Guide

International Nuclear Information System (INIS)

2009-01-01

This Safety Guide is intended to assist States in implementing the requirements established in Safety Standards Series No. GS-R-1, Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety, for a national regulatory infrastructure to regulate any practice involving radiation sources in medicine, industry, research, agriculture and education. The Safety Guide provides advice on the legislative basis for establishing regulatory bodies, including the effective independence of the regulatory body. It also provides guidance on implementing the functions and activities of regulatory bodies: the development of regulations and guides on radiation safety; implementation of a system for notification and authorization; carrying out regulatory inspections; taking necessary enforcement actions; and investigating accidents and circumstances potentially giving rise to accidents. The various aspects relating to the regulatory control of consumer products are explained, including justification, optimization of exposure, safety assessment and authorization. Guidance is also provided on the organization and staffing of regulatory bodies. Contents: 1. Introduction; 2. Legal framework for a regulatory infrastructure; 3. Principal functions and activities of the regulatory body; 4. Regulatory control of the supply of consumer products; 5. Functions of the regulatory body shared with other governmental agencies; 6. Organization and staffing of the regulatory body; 7. Documentation of the functions and activities of the regulatory body; 8. Support services; 9. Quality management for the regulatory system.
Manufacture of Platform Prototype for Digital Safety System

International Nuclear Information System (INIS)

Lee, S. Y.; Kim, J. S.; Kim, J. M.

2010-01-01

Unit controller is a basic unit of digital safety system platform prototype. The typical unit controller is comprised of CPB(CPU board), CMB(communication board), AIB(Analog input board), AOB(Analog output board), CIB(contact input board), COB(contact output board), and a subrack. It is developed according to H/W development procedure and S/W development life cycle. A digital safety system(for example, plant protection system) is the assemblies of unit controllers. CPB performs the function of each system. DSP(digital signal processor) is built in CPB. CMB is responsible for communication between unit controllers. NSD(Network Switching Device) exchanges data between the unit controllers. Each unit controller of the platform are connected to NSD through CMB. Reliability analyses on unit controller and NSD are performed. These reliability data are used as input of technical validation
Cost benefit analysis of reactor safety systems

International Nuclear Information System (INIS)

Maurer, H.A.

1984-01-01

Cost/benefit analysis of reactor safety systems is a possibility appropriate to deal with reactor safety. The Commission of the European Communities supported a study on the cost-benefit or cost effectiveness of safety systems installed in modern PWR nuclear power plants. The following systems and their cooperation in emergency cases were in particular investigated in this study: the containment system (double containment), the leakage exhaust and control system, the annulus release exhaust system and the containment spray system. The benefit of a safety system is defined according to its contribution to the reduction of the radiological consequences for the environment after a LOCA. The analysis is so far performed in two different steps: the emergency core cooling system is considered to function properly, failure of the emergency core cooling system is assumed (with the possible consequence of core melt-down) and the results may demonstrate the evidence that striving for cost-effectiveness can produce a safer end result than the philosophy of safety at any cost. (orig.)
Study of fieldbus technology confiability when applied in a Sterilization plant control and safety systems

International Nuclear Information System (INIS)

Karma, D.; Sampa, M.H.O.; Rela, P.R.

2001-01-01

Several sterilization processes have been used in these years for treatment of countless products. Some processes use high temperatures, thermal shocks and chemical agents. With the discovery of the ionizing radiation and its posterior technological developments turned possible the application of that process, in 1960, also in the sterilization, denominated radiation sterilization. This process became also applied in another areas of health and industrial as food conservation, gemstones enhancement and others. The radiation sterilization requests an effective control and it needs a high level of safety. The commercial use of the computers applied in industrial automation provides and the domain of new technologies in this field provides news applications then new designs now is possible. The Fieldbus technology, a new digital communication protocol, like a Local Area Network, can be an alternative in the cobalt-60 irradiation plant. This paper show preliminary study about confiability in systems using Fieldbus technology. This technology was simulated in sterilization plant control and safety systems and the fail probability was quantified using Fail Tree Analysis Method. Fieldbus technology can be used in sterilization plants because the confiability in this systems is like PLCs and relays systems, was the conclusion
Software V and V methods for a safety - grade programmable logic controller

International Nuclear Information System (INIS)

Jang Yeol Kim; Young Jun Lee; Kyung Ho Cha; Se Woo Cheon; Jang Soo Lee; Kee Choon Kwon

2006-01-01

This paper addresses the Verification and Validation(V and V) process and the methodology for an embedded real time software of a safety-grade Programmable Logic Controller(PLC). This safety- grade PLC is being developed as one of the Korean Nuclear Instrumentation and Control System (KNICS) projects. KNICS projects are developing a Reactor Protection System(RPS) and an Engineered Safety Feature-Component Control System(ESF-CCS) as well as a safety-grade PLC. The safety-grade PLC will be a major component that encomposes the RPS systems and the ESF-CCS systems as nuclear instruments and control equipment. This paper describes the V and V guidelines and procedures, V and V environment, V and V process and methodology, and the V and V tools in the KNICS projects. Specifically, it describes the real-time operating system V and V experience which corresponds to the requirement analysis phase, design phase and the implementation and testing phase of the software development life cycle. Main activities of the V and V for the PLC system software are a technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, software safety analysis, and a software configuration management. The proposed V and V methodology satisfies the Standard Review Plan(SRP)/Branch Technical Position(BTP)-14 criteria for the safety software in nuclear power plants. The proposed V and V methodology is going to be used to verify the upcoming software life cycle in the KNICS projects. (author)
Performance Testing Methodology for Safety-Critical Programmable Logic Controller

International Nuclear Information System (INIS)

Kim, Chang Ho; Oh, Do Young; Kim, Ji Hyeon; Kim, Sung Ho; Sohn, Se Do

2009-01-01

The Programmable Logic Controller (PLC) for use in Nuclear Power Plant safety-related applications is being developed and tested first time in Korea. This safety-related PLC is being developed with requirements of regulatory guideline and industry standards for safety system. To test that the quality of the developed PLC is sufficient to be used in safety critical system, document review and various product testings were performed over the development documents for S/W, H/W, and V/V. This paper provides the performance testing methodology and its effectiveness for PLC platform conducted by KOPEC
Nuclear power plants - Instrumentation and control systems important for safety - Classification (International Electrotechnical Commission Standard Publication 1226:1993)

International Nuclear Information System (INIS)

Stefanik, J.

1996-01-01

This international standard established a method of classification of the information and command functions for nuclear power plants, and the I and C and equipment that provide those functions, into categories that designate the importance for safety of the functions, and the associated systems and equipment. The resulting classification then determines relevant design criteria. The design criteria are the measures of quality by which the adequacy of each functions, and the associated systems and equipment in relation to its importance to plant safety is ensured. In this standard, the criteria are those of functionality, reliability, performance, environmental durability and quality assurance. This standard is applicable to all the information and command functions, and the instrumentation and control systems and equipment that provide those functions. The functions, systems and equipment under consideration provide automated protection, closed or open loop control, and information to the operating staff. They keep the NPP conditions inside the safe operating envelope and provide automatic actions, or enable manual actions, that mitigate accidents or prevent or minimize radioactive releases to the site or wider environment. The functions, and the associated systems and equipment that fulfill these roles safeguard the health and safety of the NPP operators and the public. This standard complements, and does not replace or supersede, the Safety Guides and Codes of Practice published by the International Atomic Energy Agency
Documents pertaining to safety control of nuclear facilities

International Nuclear Information System (INIS)

1998-01-01

The Finnish Radiation and Nuclear Safety Authority (STUK) controls the safety of nuclear facilities in Finland. This control encompasses on one hand the evaluation of plant safety on the basis of plans and analyses pertaining to the plant and on the other hand the inspection of plant structures, systems and components as well as of operational activity. STUK also monitors plants operational experience feedback and technical developments in the field, as well as the development of safety research and takes the necessary measures on their basis. Guide YVL 1.1 describes how STUK controls the design, construction and operation of nuclear power plants. The documents to be submitted to STUK are described in the nuclear energy legislation and YVL guides. This guide presents the mode of delivery, quality, contents and number of documents to be submitted to STUK
A Nuclear Safety System based on Industrial Computer

International Nuclear Information System (INIS)

Kim, Ji Hyeon; Oh, Do Young; Lee, Nam Hoon; Kim, Chang Ho; Kim, Jae Hack

2011-01-01

The Plant Protection System(PPS), a nuclear safety Instrumentation and Control (I and C) system for Nuclear Power Plants(NPPs), generates reactor trip on abnormal reactor condition. The Core Protection Calculator System (CPCS) is a safety system that generates and transmits the channel trip signal to the PPS on an abnormal condition. Currently, these systems are designed on the Programmable Logic Controller(PLC) based system and it is necessary to consider a new system platform to adapt simpler system configuration and improved software development process. The CPCS was the first implementation using a micro computer in a nuclear power plant safety protection system in 1980 which have been deployed in Ulchin units 3,4,5,6 and Younggwang units 3,4,5,6. The CPCS software was developed in the Concurrent Micro5 minicomputer using assembly language and embedded into the Concurrent 3205 computer. Following the micro computer based CPCS, PLC based Common-Q platform has been used for the ShinKori/ShinWolsong units 1,2 PPS and CPCS, and the POSAFE-Q PLC platform is used for the ShinUlchin units 1,2 PPS and CPCS. In developing the next generation safety system platform, several factors (e.g., hardware/software reliability, flexibility, licensibility and industrial support) can be considered. This paper suggests an Industrial Computer(IC) based protection system that can be developed with improved flexibility without losing system reliability. The IC based system has the advantage of a simple system configuration with optimized processor boards because of improved processor performance and unlimited interoperability between the target system and development system that use commercial CASE tools. This paper presents the background to selecting the IC based system with a case study design of the CPCS. Eventually, this kind of platform can be used for nuclear power plant safety systems like the PPS, CPCS, Qualified Indication and Alarm . Pami(QIAS-P), and Engineering Safety
A Nuclear Safety System based on Industrial Computer

Energy Technology Data Exchange (ETDEWEB)

Kim, Ji Hyeon; Oh, Do Young; Lee, Nam Hoon; Kim, Chang Ho; Kim, Jae Hack [Korea Electric Power Corporation Engineering and Construction, Daejeon (Korea, Republic of)

2011-05-15

The Plant Protection System(PPS), a nuclear safety Instrumentation and Control (I and C) system for Nuclear Power Plants(NPPs), generates reactor trip on abnormal reactor condition. The Core Protection Calculator System (CPCS) is a safety system that generates and transmits the channel trip signal to the PPS on an abnormal condition. Currently, these systems are designed on the Programmable Logic Controller(PLC) based system and it is necessary to consider a new system platform to adapt simpler system configuration and improved software development process. The CPCS was the first implementation using a micro computer in a nuclear power plant safety protection system in 1980 which have been deployed in Ulchin units 3,4,5,6 and Younggwang units 3,4,5,6. The CPCS software was developed in the Concurrent Micro5 minicomputer using assembly language and embedded into the Concurrent 3205 computer. Following the micro computer based CPCS, PLC based Common-Q platform has been used for the ShinKori/ShinWolsong units 1,2 PPS and CPCS, and the POSAFE-Q PLC platform is used for the ShinUlchin units 1,2 PPS and CPCS. In developing the next generation safety system platform, several factors (e.g., hardware/software reliability, flexibility, licensibility and industrial support) can be considered. This paper suggests an Industrial Computer(IC) based protection system that can be developed with improved flexibility without losing system reliability. The IC based system has the advantage of a simple system configuration with optimized processor boards because of improved processor performance and unlimited interoperability between the target system and development system that use commercial CASE tools. This paper presents the background to selecting the IC based system with a case study design of the CPCS. Eventually, this kind of platform can be used for nuclear power plant safety systems like the PPS, CPCS, Qualified Indication and Alarm . Pami(QIAS-P), and Engineering Safety

Multi-core System Architecture for Safety-critical Control Applications

DEFF Research Database (Denmark)

Li, Gang

and size, and high power consumption. Increasing the frequency of a processor is becoming painful now due to the explosive power consumption. Furthermore, components integrated into a single-core processor have to be certified to the highest SIL, due to that no isolation is provided in a traditional single...... certification cost. Meanwhile, hardware platforms with improved processing power are required to execute the applications of larger size. To tackle the two issues mentioned above, the state of the art approaches are using more Electronic Control Units (ECU) in a federated architecture or increasing......-core processor. A promising alternative to improve processing power and provide isolation is to adopt a multi-core architecture with on-chip isolation. In general, a specific multi-core architecture can facilitate the development and certification of safety-related systems, due to its physical isolation between...
A holistic strategy for quality and safety control of traditional Chinese medicines by the “iVarious” standard system

Directory of Open Access Journals (Sweden)

Anzhen Chen

2017-10-01

Full Text Available An effective quality control system is the key to ensuring the quality, safety and efficacy of traditional Chinese medicines (TCMs. However, the current quality standard research lacks the top-design and systematic design, mostly based on specific technologies or evaluation methods. To resolve the challenges and questions of quality control of TCMs, a brand-new quality standard system, named “iVarious”, was proposed. The system comprises eight elements in a modular format. Meaning of every element was specifically illustrated via corresponding research instances. Furthermore, frankincense study was taken as an example for demonstrating standards and research process, based on the “iVarious” system. This system highlighted a holistic strategy for effectiveness, security, integrity and systematization of quality and safety control standards of TCMs. The establishment of “iVarious” integrates multi-disciplinary technologies and progressive methods, basis elements and key points of standard construction. The system provides a novel idea and technological demonstration for regulation establishment of TCMs quality standards.
Means to improve underground coal mine safety by automated control of methane drainage systems

Directory of Open Access Journals (Sweden)

Babut Gabriel Bujor

2017-01-01

Full Text Available Based on the critical analysis of the presently employed management of methane drainage systems operation in Jiu Valley collieries, the paper aims to assess the basic elements required to develop an automated monitoring and control system of these. The results obtained after studies and researches carried out also allowed formulating certain proposals regarding the modification of manual control procedures of methane drainage systems operation, in order to correlate them with the prescriptions of legislation requirements from countries having a well-developed mining industry. Putting in practice the mentioned proposals could have immediate and beneficial effects on increasing the methane drainage process efficiency, leading meanwhile to an improved working environment and, implicitly, to a higher level of occupational safety and health in Jiu Valley collieries.
System safety education focused on flight safety

Science.gov (United States)

Holt, E.

1971-01-01

The measures necessary for achieving higher levels of system safety are analyzed with an eye toward maintaining the combat capability of the Air Force. Several education courses were provided for personnel involved in safety management. Data include: (1) Flight Safety Officer Course, (2) Advanced Safety Program Management, (3) Fundamentals of System Safety, and (4) Quantitative Methods of Safety Analysis.
Programmable controllers replace relays in MFTF-B personnel-safety interlocks

International Nuclear Information System (INIS)

Branum, J.D.

1981-01-01

This paper describes a new approach for implementing personnel safety interlocks logic using industrial-type programmable controllers. The logic for all personnel safety interlocks except those totally internal to a subsystem is implemented in two non-redundant controllers. A high degree of fail-safe reliability is achieved by augmenting the protective features intrinsic to each controller with those provided by a small amount of external support hardware. The controllers are interfaced to the host computer system via fiber optic data links to enable display of interlock and overall system status on the control room graphic displays. When fully implemented, the controllers will perform the equivalent of over 2000 discreet relay functions
Antisideslip and Antirollover Safety Speed Controller Design for Vehicle on Curved Road

Directory of Open Access Journals (Sweden)

Guo Lie

2014-01-01

Full Text Available When the drivers cannot be aware of the existing of forthcoming curved roads and fail to regulate their safety speeds accordingly, sideslip or rollover may occur with high probability. The antisideslip and antirollover control of vehicle on curved road in automatic highway systems is studied. The safety speed warning system is set before entering the curved road firstly. The speed adhesion control is adopted to shorten the braking distance while decelerating and to guarantee the safety speed. The velocity controller when decelerating on the straight path and the posture controller when driving on curved road are designed, respectively, utilizing integral backstepping technology. Simulation results demonstrate that this control system is characterized by quick and precise tracking and global stability. Consequently, it is able to avoid the dangerous operating conditions, such as sideslip and rollover, and guarantee the safety and directional stability when driving on curved road.
Microbiological performance of Hazard Analysis Critical Control Point (HACCP)-based food safety management systems: A case of Nile perch processing company

NARCIS (Netherlands)

Kussaga, J.B.; Luning, P.A.; Tiisekwa, B.P.M.; Jacxsens, L.

2017-01-01

This study aimed at giving insight into microbiological safety output of a Hazard Analysis Critical Control Point (HACCP)-based Food Safety Management System (FSMS) of a Nile perch exporting company by using a combined assessment, This study aimed at giving insight into microbiological safety output
The NASA Aviation Safety Reporting System

Science.gov (United States)

1983-01-01

This is the fourteenth in a series of reports based on safety-related incidents submitted to the NASA Aviation Safety Reporting System by pilots, controllers, and, occasionally, other participants in the National Aviation System (refs. 1-13). ASRS operates under a memorandum of agreement between the National Aviation and Space Administration and the Federal Aviation Administration. The report contains, first, a special study prepared by the ASRS Office Staff, of pilot- and controller-submitted reports related to the perceived operation of the ATC system since the 1981 walkout of the controllers' labor organization. Next is a research paper analyzing incidents occurring while single-pilot crews were conducting IFR flights. A third section presents a selection of Alert Bulletins issued by ASRS, with the responses they have elicited from FAA and others concerned. Finally, the report contains a list of publications produced by ASRS with instructions for obtaining them.
Description of the control and safety systems of the RA reactor; Opis sistema za upravljanje i sigurnosnu zastitu RA

Energy Technology Data Exchange (ETDEWEB)

Popovic, B; Pesic, M [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Yugoslavia)

1962-07-01

This report contains detailed description and scheme of the control and safety system of the RA reactor. It consists of interconnected five systems: for automated regulation; compensation rods; safety rods; power density measurement device; period meter; automated D{sub 2}O level meter in the core. Automated regulation system is divided into two parts: basic system for reactor operation regime at power from 10kW - 10 MW and precise regulation system for operation at set-up power level up to 10 kW which is used occasionally.
Neutron flux control systems validation

International Nuclear Information System (INIS)

Hascik, R.

2003-01-01

In nuclear installations main requirement is to obtain corresponding nuclear safety in all operation conditions. From the nuclear safety point of view is commissioning and start-up after reactor refuelling appropriate period for safety systems verification. In this paper, methodology, performance and results of neutron flux measurements systems validation is presented. Standard neutron flux measuring chains incorporated into the reactor protection and control system are used. Standard neutron flux measuring chain contains detector, preamplifier, wiring to data acquisition unit, data acquisition unit, wiring to display at control room and display at control room. During reactor outage only data acquisition unit and wiring and displaying at reactor control room is verified. It is impossible to verify detector, preamplifier and wiring to data acquisition recording unit during reactor refuelling according to low power. Adjustment and accurate functionality of these chains is confirmed by start-up rate (SUR) measurement during start-up tests after refuelling of the reactors. This measurement has direct impact to nuclear safety and increase operational nuclear safety level. Briefly description of each measuring system is given. Results are illustrated on measurements performed at Bohunice NPP during reactor start-up tests. Main failures and their elimination are described (Authors)
49 CFR 659.19 - System safety program plan: contents.

Science.gov (United States)

2010-10-01

... implementation of the system safety program. (j) A description of the process used by the rail transit agency to... the rail transit agency to manage safety issues. (d) The process used to control changes to the system... hazard management program. (n) A description of the process used for facilities and equipment safety...
Probabilistic safety assessment for instrumentation and control systems in nuclear power plants. A literature survey

International Nuclear Information System (INIS)

Lu, Lixuan; Jiang, Jin

2003-01-01

Deregulation in electricity market will create a great deal of challenges for Nuclear Power Plants (NPP). To stay competitive, NPP will need to find new ways to reduce their operation costs. In NPP, Instrumentation and Control (I and C) systems play an important role in reducing the cost of producing electricity while maintaining and/or enhancing safety. Therefore, it is extremely important that one should manage the I and C systems more efficiently and economically. Meanwhile, obsolescence problem associated with I and C systems encouraged the usage of advanced digital techniques in I and C systems. Thus, new methodologies are needed to analyze the reliability and determine the maintenance strategy for the digital I and C systems. Probabilistic Safety Assessment (PSA) has been probed to be a promising method to deal with this issue. This paper provides a literature survey on the development of digital I and C systems in NPP, followed by a detailed review of PSA including its benefits, limitations and the future direction of its development. Most importantly, potential applications of PSA in various aspects of I and C systems are brought into perspective throughout the paper. Furthermore, the applicability of PSA in the regulation of safety-related I and C systems is demonstrated. Detailed information on PSA applications in 1) the resource allocation for I and C systems: 2) the determination of surveillance testing strategies; and 3) I and C system designs, is provided. (author)
CERN safety system monitoring - SSM

International Nuclear Information System (INIS)

Hakulinen, T.; Ninin, P.; Valentini, F.; Gonzalez, J.; Salatko-Petryszcze, C.

2012-01-01

CERN SSM (Safety System Monitoring) is a system for monitoring state-of-health of the various access and safety systems of the CERN site and accelerator infrastructure. The emphasis of SSM is on the needs of maintenance and system operation with the aim of providing an independent and reliable verification path of the basic operational parameters of each system. Included are all network-connected devices, such as PLCs (local purpose control unit), servers, panel displays, operator posts, etc. The basic monitoring engine of SSM is a freely available system-monitoring framework Zabbix, on top of which a simplified traffic-light-type web-interface has been built. The web-interface of SSM is designed to be ultra-light to facilitate access from hand-held devices over slow connections. The underlying Zabbix system offers history and notification mechanisms typical of advanced monitoring systems. (authors)
NASA aviation safety reporting system

Science.gov (United States)

1981-01-01

Aviation safety reports that relate to loss of control in flight, problems that occur as a result of similar sounding alphanumerics, and pilot incapacitation are presented. Problems related to the go around maneuver in air carrier operations, and bulletins (and FAA responses to them) that pertain to air traffic control systems and procedures are included.
System analysis of vehicle active safety problem

Science.gov (United States)

Buznikov, S. E.

2018-02-01

The problem of the road transport safety affects the vital interests of the most of the population and is characterized by a global level of significance. The system analysis of problem of creation of competitive active vehicle safety systems is presented as an interrelated complex of tasks of multi-criterion optimization and dynamic stabilization of the state variables of a controlled object. Solving them requires generation of all possible variants of technical solutions within the software and hardware domains and synthesis of the control, which is close to optimum. For implementing the task of the system analysis the Zwicky “morphological box” method is used. Creation of comprehensive active safety systems involves solution of the problem of preventing typical collisions. For solving it, a structured set of collisions is introduced with its elements being generated also using the Zwicky “morphological box” method. The obstacle speed, the longitudinal acceleration of the controlled object and the unpredictable changes in its movement direction due to certain faults, the road surface condition and the control errors are taken as structure variables that characterize the conditions of collisions. The conditions for preventing typical collisions are presented as inequalities for physical variables that define the state vector of the object and its dynamic limits.
Compositional Synthesis of Safety Controllers

NARCIS (Netherlands)

Kuijper, W.

2012-01-01

In my thesis I investigate compositional techniques for synthesis of safety controllers. A safety controller, in this context, is a state machine that gives the set of safe control outputs for every possible sequence of observations from the plant under control. Compositionality, in this context,
A formal safety analysis for PLC software-based safety critical system using Z

International Nuclear Information System (INIS)

Koh, Jung Soo; Seong, Poong Hyun

1997-01-01

This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC (Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formed safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system
Analysis of Aviation Safety Reporting System Incident Data Associated With the Technical Challenges of the Vehicle Systems Safety Technology Project

Science.gov (United States)

Withrow, Colleen A.; Reveley, Mary S.

2014-01-01

This analysis was conducted to support the Vehicle Systems Safety Technology (VSST) Project of the Aviation Safety Program (AVsP) milestone VSST4.2.1.01, "Identification of VSST-Related Trends." In particular, this is a review of incident data from the NASA Aviation Safety Reporting System (ASRS). The following three VSST-related technical challenges (TCs) were the focus of the incidents searched in the ASRS database: (1) Vechicle health assurance, (2) Effective crew-system interactions and decisions in all conditions; and (3) Aircraft loss of control prevention, mitigation, and recovery.
Emerging standards with application to accelerator safety systems

International Nuclear Information System (INIS)

Mahoney, K.L.; Robertson, H.P.

1997-01-01

This paper addresses international standards which can be applied to the requirements for accelerator personnel safety systems. Particular emphasis is given to standards which specify requirements for safety interlock systems which employ programmable electronic subsystems. The work draws on methodologies currently under development for the medical, process control, and nuclear industries
Reliability Analysis Multiple Redundancy Controller for Nuclear Safety Systems

International Nuclear Information System (INIS)

Son, Gwangseop; Kim, Donghoon; Son, Choulwoong

2013-01-01

This controller is configured for multiple modular redundancy (MMR) composed of dual modular redundancy (DMR) and triple modular redundancy (TMR). The architecture of MRC is briefly described, and the Markov model is developed. Based on the model, the reliability and Mean Time To Failure (MTTF) are analyzed. In this paper, the architecture of MRC for nuclear safety systems is described. The MRC is configured for multiple modular redundancy (MMR) composed of dual modular redundancy (DMR) and triple modular redundancy (TMR). Markov models for MRC architecture was developed, and then the reliability was analyzed by using the model. From the reliability analyses for the MRC, it is obtained that the failure rate of each module in the MRC should be less than 2 Χ 10 -4 /hour and the MTTF average increase rate depending on FCF increment, i. e. ΔMTTF/ΔFCF, is 4 months/0.1

Rapid prototyping of the Central Safety System for Nuclear Risk in ITER

Energy Technology Data Exchange (ETDEWEB)

Scibile, L. [ITER Organization, CS 90 046, St. Paul-lez-Durance, Cedex (France); Ambrosino, G. [Consorzio CREATE, Universita degli Studi di Napoli Federico II, via Claudio 21, 80125, Napoli (Italy); De Tommasi, G., E-mail: detommas@unina.i [Consorzio CREATE, Universita degli Studi di Napoli Federico II, via Claudio 21, 80125, Napoli (Italy); Pironti, A. [Consorzio CREATE, Universita degli Studi di Napoli Federico II, via Claudio 21, 80125, Napoli (Italy)

2010-07-15

The Central Safety System for Nuclear Risk (CSS-N) coordinates the safety control systems to ensure nuclear safety for the ITER complex. Since the CSS-N is a safety critical system, its validation and commissioning play a very important role; in particular the required level of reliability must be demonstrated. In such a scenario, it is strongly recommended to use modeling and simulation tools since the early design phase. Indeed, the modeling tools will help in the definition of the control system requirements. Furthermore the models can than be used for the rapid prototyping of the safety system. Hardware-in-the-loop simulations can also be performed in order to assess the performance of the control hardware against a plant simulator. The proposed approach relies on the availability of a plant simulator to develop the prototype of the control system. This paper introduces the methodology used to design and develop both the CSS-N Oriented Plant Simulator and the CSS-N Prototype.
Rapid prototyping of the Central Safety System for Nuclear Risk in ITER

International Nuclear Information System (INIS)

Scibile, L.; Ambrosino, G.; De Tommasi, G.; Pironti, A.

2010-01-01

The Central Safety System for Nuclear Risk (CSS-N) coordinates the safety control systems to ensure nuclear safety for the ITER complex. Since the CSS-N is a safety critical system, its validation and commissioning play a very important role; in particular the required level of reliability must be demonstrated. In such a scenario, it is strongly recommended to use modeling and simulation tools since the early design phase. Indeed, the modeling tools will help in the definition of the control system requirements. Furthermore the models can than be used for the rapid prototyping of the safety system. Hardware-in-the-loop simulations can also be performed in order to assess the performance of the control hardware against a plant simulator. The proposed approach relies on the availability of a plant simulator to develop the prototype of the control system. This paper introduces the methodology used to design and develop both the CSS-N Oriented Plant Simulator and the CSS-N Prototype.
The use of microprocessors at TRIUMF in the control of radiation safety interlock systems

International Nuclear Information System (INIS)

King, L.

1988-01-01

At TRIUMF the cyclotron vault, all primary beam lines, and each experimental area has a dedicated control unit to manage the safety interlock control of the area lockup sequence, beam blocker drive and area access. Typically each area has 24 devices which are monitored to control 16 outputs. These control units (Area Safety Units) were first implemented through the use of relay logic. The relay logic was reliable but difficult to modify to incorporate changes to the areas. In 1979 it was decided to use microprocessors in the form of single board computers to control the Area Safety Units. The details of the hardware and software is discussed as well as the advantages of microprocessor control
The passive safety systems of the Swr 1000

International Nuclear Information System (INIS)

Neumann, D.

2001-01-01

In recent years, a new boiling water reactor (BWR) plant called the SWR 1000 has been developed by Siemens on behalf of Germany's electric utilities. This new plant design concept incorporates the wide range of operating experience gained with German BWRs. The main objective behind developing the SWR 1000 was to design a plant with a rated electric output of approximately 1000 MW which would not only have a lower capital cost and lower power generating costs but would also provide a much higher level of nuclear safety compared to plants currently in operation. This safety-related goal has been met through, for example, the use of passive safety equipment. Passive systems make a significant contribution towards increasing the over-all level of plant safety due to the way in which they operate. They function solely accord-ing to basic laws of nature, such as gravity, and perform their designated functions with-out any need for electric power or other sources of external energy, or signals from instrumentation and control (I and C) equipment. The passive safety systems have been designed such that design basis accidents can be controlled using just these systems alone. However, the design concept of the SWR 1000 is nevertheless still based on the provision of active safety systems in addition to passive systems. (author)
Probabilistic safety assessment for digital instrumentation and control systems in nuclear power plants - a review

International Nuclear Information System (INIS)

Lu, L.; Jiang, J.

2003-01-01

Deregulation in electricity market has created a great deal of challenges for nuclear power industries [1]. To stay competitive, Nuclear Power Plants (NPPs) will have to find ways to reduce their operational costs and to improve the plant safety. Instrumentation and Control (I and C) systems play an important role in this regard. Thus, new methodologies need to be developed to manage the operation of I and C systems more economically without jeopardizing the overall plant safety. Probabilistic Safety Assessment (PSA) technique is one of the promising methods to deal with such an issue, because PSA analyzes various system operational issues from a probabilistic sense, rather than a worst-case approach. However, there are several limitations when PSA is applied to I and C systems directly. A possible solution to this problem can be found by incorporating PSA with several other approaches. To better understand the issues involved, an attempt has been made in this paper to carry out a literature survey on this and related subject, particularly the effort will be made on: 1) the development of digital I and C systems in NPP, 2) PSA and its potential benefits and limitations, and 3) applications of PSA in various aspects of I and C systems including the resource allocation, the determination of surveillance testing strategies and the design of I and C systems. Finally, some solutions to overcome the aforementioned obstacles when applying PSA in I and C systems are also examined critically. (author)
Design for safety: theoretical framework of the safety aspect of BIM system to determine the safety index

Directory of Open Access Journals (Sweden)

Ai Lin Evelyn Teo

2016-12-01

Full Text Available Despite the safety improvement drive that has been implemented in the construction industry in Singapore for many years, the industry continues to report the highest number of workplace fatalities, compared to other industries. The purpose of this paper is to discuss the theoretical framework of the safety aspect of a proposed BIM System to determine a Safety Index. An online questionnaire survey was conducted to ascertain the current workplace safety and health situation in the construction industry and explore how BIM can be used to improve safety performance in the industry. A safety hazard library was developed based on the main contributors to fatal accidents in the construction industry, determined from the formal records and existing literature, and a series of discussions with representatives from the Workplace Safety and Health Institute (WSH Institute in Singapore. The results from the survey suggested that the majority of the firms have implemented the necessary policies, programmes and procedures on Workplace Safety and Health (WSH practices. However, BIM is still not widely applied or explored beyond the mandatory requirement that building plans should be submitted to the authorities for approval in BIM format. This paper presents a discussion of the safety aspect of the Intelligent Productivity and Safety System (IPASS developed in the study. IPASS is an intelligent system incorporating the buildable design concept, theory on the detection, prevention and control of hazards, and the Construction Safety Audit Scoring System (ConSASS. The system is based on the premise that safety should be considered at the design stage, and BIM can be an effective tool to facilitate the efforts to enhance safety performance. IPASS allows users to analyse and monitor key aspects of the safety performance of the project before the project starts and as the project progresses.
A study on LAN applications in nuclear safety systems

International Nuclear Information System (INIS)

Kim, Sung; Lee, Young Ryul; Koo, Jun Mo; Han, Jai Bok

1995-01-01

It is a general tendency to digitalize the conventional relay based I and C systems in nuclear power plant. But, the digitalisation of nuclear safety systems has many a difficulty to surmount. The typical one thing of many difficulties is the data communication problem between local controllers and systems. The network architecture built with LAN (Local Area Network) in digital systems of the other industries are general. But in case of nuclear safety systems many considerations in point of safety and license are required to implement it in the field. In this parer, some considerations for applying LAN in nuclear safety systems were reviewed
Efficiency of the functioning of the state control system for the safety and quality of animal products in Ukraine

Directory of Open Access Journals (Sweden)

I. Kyryliuk

2017-12-01

Full Text Available The study reveals the results of evaluating the effectiveness of the state control system (supervision on the safety and individual indicators of the quality of livestock products in Ukraine. The necessity of application of such components of efficiency as legislation, management and its organizational structure, inspection and laboratory service, information, training and communications is substantiated. It has been determined that during a sufficiently long period of time (until 2015, the system of state control (supervision was archaic and actually focused on the principles of command and administrative economy. The modern tendencies and specifics of the improvement of the Ukrainian control system in the direction of its harmonization with the European one are shown. The emphasis was on significant volumes of work that needed to be done in a very short time, as well as in the absence of adequate funding and appropriate skilled specialists. The emergence of clarity and unambiguousness in determining the responsibility of market operators for violating the legislation requirements in the field of production and circulation of animal origin food products was emphasized. Along with the achievements, there were identified systemic problems related to the technical regulation of safety assurance processes and individual quality indicators in Ukraine. Also it was noted and revealed that legislation in the area of guaranteeing the quality and safety of livestock products in Ukraine remains incomplete and not fully developed. The necessity of development of a number of by-laws and allocation of necessary financing for effective functioning of the state control system over product safety is substantiated. Article specified on the presence of insufficient number of professional inspection and laboratory services is underlined. The mechanisms of avoiding corruption risks and excessive pressure on the subjects of the livestock production market are
Performance Test Results of Safety I and C Systems of SMART MMIS

International Nuclear Information System (INIS)

Suh, Yong Suk; Keum, Jong Yong; Jeong, Kwang Il; Lee, Joon Ku; Lee, Sang Seok; Kim, Kwan Woong

2011-01-01

KAERI has developed SMART (System-integrated Modular Advanced ReacTor), a 330MWt integral pressurized light water reactor that integrates four reactor coolant pumps, one pressurizer, eight steam generators, and one reactor core into a reactor vessel, since 1997 and submitted a SSAR (Standard design Safety Analysis Report) to Korea institute of nuclear safety (KINS) at the end of 2010 for the purpose of achieving the standard design approval (SDA) by the end of 2011. SMART MMIS has been designed with fully digitalized systems. Non-safety instrumentation and control (I and C) systems are designed based on the commercial distributed control systems. The safety I and C systems are designed using a new platform that was developed and validated by KAERI. Safety I and C systems are modularized using the platform. In the protection systems (PSs), datalinks are used to transmit data in a one-way direction in order to meet the independency requirement. In the engineered safety features-component control system (ESF-CCS), network switch devices (NSDs) are used to connect the group and loop controllers. The NSD was also newly developed and validated by KAERI. After validating the platform and NSD, a test facility was developed using the platform and NSDs to validate the performance of safety I and C systems. This paper presents the development and test results from the test facility
Physics related to control and safety of hybrid systems; Physique associee au controle et a la surete des systemes hybrides

Energy Technology Data Exchange (ETDEWEB)

Gueton, O

2001-12-01

Regarding nuclear waste management, ADS can be considered as large minor actinides burners. In a first part, a critical analysis of different reactor types shows that fast spectrum, helium coolant and nitride fuel, containing 100% minor actinides, agree perfectly with the high transmutation requirements of ADS. The control and safety demonstration of this system represents the main purpose of this study. Understanding spatial and dynamic behaviour of ADS flux is absolutely necessary. For this purpose, we have defined an indicator to quantify spatial decoupling. It shows, on the one hand, point kinetic deficiency to study local transients, and on the other hand, perturbations propagation differences between ADS and critical cores. Then, in a more concrete approach, accidental sequences (source transient, beam de-focalization, reactivity insertions, loss of flow, depressurization) are evaluated for this core, strongly loaded with minor actinides. It is shown that the automatic beam shutdown leads to preserve large safety margins for all studied transients. The accelerator emergency stop is induced by an unexpected evolution of the core control parameters. These parameters, except reactivity, can be directly measured in subcritical systems like in critical ones. Concerning reactivity, we suggest a new method for its absolute determination in ADS: at the time of reactor start-up, the reactivity must be calibrated by coupling two methods of relative reactivity measurements (pulsed source and Approached Source Multiplication) for successive subcritical levels. After that, the on-line follow-up of reactivity is obtained from this calibration like in a critical core. (authors)
The ATLAS Detector Safety System

CERN Multimedia

Helfried Burckhart; Kathy Pommes; Heidi Sandaker

The ATLAS Detector Safety System (DSS) has the mandate to put the detector in a safe state in case an abnormal situation arises which could be potentially dangerous for the detector. It covers the CERN alarm severity levels 1 and 2, which address serious risks for the equipment. The highest level 3, which also includes danger for persons, is the responsibility of the CERN-wide system CSAM, which always triggers an intervention by the CERN fire brigade. DSS works independently from and hence complements the Detector Control System, which is the tool to operate the experiment. The DSS is organized in a Front- End (FE), which fulfills autonomously the safety functions and a Back-End (BE) for interaction and configuration. The overall layout is shown in the picture below. ATLAS DSS configuration The FE implementation is based on a redundant Programmable Logical Crate (PLC) system which is used also in industry for such safety applications. Each of the two PLCs alone, one located underground and one at the s...
Development of Non-safety System Architecture and Evaluation of Components/Systems

International Nuclear Information System (INIS)

Oh, I. S.; Lee, C. K.; Kim, D. H.; Lee, J. W.; Lee, D. Y.; Park, W. M.; Hwang, I. K.; Hur, S.; Kim, J. T.; Park, J. C.; Lee, J. W.

2007-10-01

We describe in this report the works performed for a technical evaluation of the non-safety digital control system of the KNICS, the non-safety process control system of the KNICS, a communication load analysis for the MMIS (including both the non-safety and the safety systems) of the KNICS, the development of MMI and an implementation of the logic for the CVCS, and the works performed to support writing a proposal needed for bidding an I and C system based on the KNICS. The technical evaluation results were aimed to be used by the designers to detect parts needed to be corrected or to be newly inserted, and also by the developers during the development phase. The requirement specifications and the data requirement characteristics have been identified for each subsystem of the determined KNICS structure. For each communication node, the specifications related to the data transfer including the data capacity for interfaces, delay time for the data transfer, and the marginal availability of its performance capabilities have been analyzed to identify the amount of data transfer and hence to verify that both of the designed structures for the safety related communications network and for the digital communications network are appropriate. The results of the supporting work performed for writing the technical specifications related to each subsystem of the KNICS structure, are expected to be useful in writing a proposal for the expected Uljin new units 1 and 2, and in the I and C upgrade for any of the existing nuclear power plants under operation. Also included in this report are the descriptions on a design of the chemical volume control system (CVCS), on the supporting work performed to draw the logic diagrams for CVCS using the tool ISaGRAF, and on the generation of a set of system displays to be used as references
Development of Non-safety System Architecture and Evaluation of Components/Systems

Energy Technology Data Exchange (ETDEWEB)

Oh, I. S.; Lee, C. K.; Kim, D. H.; Lee, J. W.; Lee, D. Y.; Park, W. M.; Hwang, I. K.; Hur, S.; Kim, J. T.; Park, J. C.; Lee, J. W

2007-10-15

We describe in this report the works performed for a technical evaluation of the non-safety digital control system of the KNICS, the non-safety process control system of the KNICS, a communication load analysis for the MMIS (including both the non-safety and the safety systems) of the KNICS, the development of MMI and an implementation of the logic for the CVCS, and the works performed to support writing a proposal needed for bidding an I and C system based on the KNICS. The technical evaluation results were aimed to be used by the designers to detect parts needed to be corrected or to be newly inserted, and also by the developers during the development phase. The requirement specifications and the data requirement characteristics have been identified for each subsystem of the determined KNICS structure. For each communication node, the specifications related to the data transfer including the data capacity for interfaces, delay time for the data transfer, and the marginal availability of its performance capabilities have been analyzed to identify the amount of data transfer and hence to verify that both of the designed structures for the safety related communications network and for the digital communications network are appropriate. The results of the supporting work performed for writing the technical specifications related to each subsystem of the KNICS structure, are expected to be useful in writing a proposal for the expected Uljin new units 1 and 2, and in the I and C upgrade for any of the existing nuclear power plants under operation. Also included in this report are the descriptions on a design of the chemical volume control system (CVCS), on the supporting work performed to draw the logic diagrams for CVCS using the tool ISaGRAF, and on the generation of a set of system displays to be used as references.
Software Safety Risk in Legacy Safety-Critical Computer Systems

Science.gov (United States)

Hill, Janice L.; Baggs, Rhoda

2007-01-01

Safety Standards contain technical and process-oriented safety requirements. Technical requirements are those such as "must work" and "must not work" functions in the system. Process-Oriented requirements are software engineering and safety management process requirements. Address the system perspective and some cover just software in the system > NASA-STD-8719.13B Software Safety Standard is the current standard of interest. NASA programs/projects will have their own set of safety requirements derived from the standard. Safety Cases: a) Documented demonstration that a system complies with the specified safety requirements. b) Evidence is gathered on the integrity of the system and put forward as an argued case. [Gardener (ed.)] c) Problems occur when trying to meet safety standards, and thus make retrospective safety cases, in legacy safety-critical computer systems.
A formal safety analysis for PLC software-based safety critical system using Z

International Nuclear Information System (INIS)

Koh, Jung Soo

1997-02-01

This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC (Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formal safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system. And also, we have found that some errors or mismatches in user requirement and final implemented PLC ladder logic while analyzing the process of the consistency and completeness of Z translated formal specifications. In the case of relatively small systems like Beamline hutch door interlock system, a formal safety analysis including explicit proof is highly recommended so that the safety of PLC-based critical system may be enhanced and guaranteed. It also provides a helpful benefits enough to comprehend user requirement expressed by ambiguous natural language
Reliability analysis of diverse safety logic systems of fast breeder reactor

International Nuclear Information System (INIS)

Ravi Kumar, Bh.; Apte, P.R.; Srivani, L.; Ilango Sambasivan, S.; Swaminathan, P.

2006-01-01

Safety Logic for Fast Breeder Reactor (FBR) is designed to initiate safety action against Design Basis Events. Based on the outputs of various processing circuits, Safety logic system drives the control rods of the shutdown system. So, Safety Logic system is classified as safety critical system. Therefore, reliability analysis has to be performed. This paper discusses the Reliability analysis of Diverse Safety logic systems of FBRs. For this literature survey on safety critical systems, system reliability approach and standards to be followed like IEC-61508 are discussed in detail. For Programmable Logic device based systems, Hardware Description Languages (HDL) are used. So this paper also discusses the Verification and Validation for HDLs. Finally a case study for the Reliability analysis of Safety logic is discussed. (author)
Software qualification for digital safety system in KNICS project

International Nuclear Information System (INIS)

Kwon, Kee-Choon; Lee, Dong-Young; Choi, Jong-Gyun

2012-01-01

In order to achieve technical self-reliance in the area of nuclear instrumentation and control, the Korea Nuclear Instrumentation and Control System (KNICS) project had been running for seven years from 2001. The safety-grade Programmable Logic Controller (PLC) and the digital safety system were developed by KNICS project. All the software of the PLC and digital safety system were developed and verified following the software development life cycle Verification and Validation (V and V) procedure. The main activities of the V and V process are preparation of software planning documentations, verification of the Software Requirement Specification (SRS), Software Design Specification (SDS) and codes, and a testing of the software components, the integrated software, and the integrated system. In addition, a software safety analysis and a software configuration management are included in the activities. For the software safety analysis at the SRS and SDS phases, the software Hazard Operability (HAZOP) was performed and then the software fault tree analysis was applied. The software fault tree analysis was applied to a part of software module with some critical defects identified by the software HAZOP in SDS phase. The software configuration management was performed using the in-house tool developed in the KNICS project. (author)
The safety implications of control systems program at ORNL

International Nuclear Information System (INIS)

Smith, O.L.

1987-01-01

Simulations of two pressurized water reactors (PWRs) point to several conclusions that bear on the principle interests of Unresolved Safety Issue A-47: (1) the simulated control systems of both plants exhibit considerable ability to respond to the investigated classes of off-normal disturbances; (2) overfill of the steam generators usually produced only minor cooling of the primary side; (3) despite protective features, substantial amounts of water could be injected into the steam lines because of low steam quality or high water level, but further analysis is needed to determine whether this creates the potential for water-hammer damage or other mass or momentum effects; and (4) potential core-uncovery scenarios explored steam generator tube rupture and other small breaks that might lead to loss of primary inventory without actuation of high pressure injection. The results indicated situations in which automatic actuation of high pressure injection would terminate the leak and others in which operator intervention appeared necessary
The safety implications of control systems program at ORNL

International Nuclear Information System (INIS)

Smith, O.L.

1987-01-01

Simulations of two pressurized water reactors (PWRs) point to several conclusions that bear on the principle interests of Unresolved Safety Issue A-47: the simulated control systems of both plants exhibit considerable ability to respond to the investigated classes of off-normal disturbances; overfill of the steam generators usually produced only minor cooling of the primary side; despite protective features, substantial amounts of water could be injected into the steam lines because of low steam quality or high water level, but further analysis is needed to determine whether this creates the potential for water-hammer damage or other mass or momentum effects; and potential core-uncovery scenarios explored steam generator tube rupture and other small breaks that might lead to loss of primary inventory without actuation of high pressure injection. The results indicated situations in which automatic actuation of high pressure injection would terminate the leak and others in which operator intervention appeared necessary
Control system for NPP powerfull turbines

International Nuclear Information System (INIS)

Osipenko, V.D.; Rozhanskij, V.E.; Rokhlenko, V.Yu.

1985-01-01

A control system for NPP 1000 MW turbines safety is described. The turbine safety system has a hydraulic drive to actuate in case of increasipg of rotational speed of a turbine rotor and an electrohydraulic drce to operate in case of pressure reduction in the lubrication system, axial displacement deviation, etc. The system is highly reliable due to application of a safety system without slide valves and long-term operation of hydraulic controls in guarding conditions; the system epsures multifunctional control with high accuracy and speed due to application of the intricate electronic part, high speed of response with a limited use of high pressure oil due to application of two-pressure pumps, pneumohydraulic accumulators and oil discharge valves. Steady-state serviceability of the system is maintained by devices for valve cooling dawn. A shockless change from electrohydraulic to hydraulic control channels is provided

Stabilization with guaranteed safety using Control Lyapunov–Barrier Function

NARCIS (Netherlands)

Romdlony, Muhammad Zakiyullah; Jayawardhana, Bayu

2016-01-01

We propose a novel nonlinear control method for solving the problem of stabilization with guaranteed safety for nonlinear systems. The design is based on the merging of the well-known Control Lyapunov Function (CLF) and the recent concept of Control Barrier Function (CBF). The proposed control
Survey of electronic safety systems in accelerator applications

International Nuclear Information System (INIS)

Mahoney, K.

1997-01-01

This paper presents the preliminary results and analysis of a comprehensive survey of the implementation of accelerator safety interlock systems from over 30 international labs. At the present time there is not a self consistent means to evaluate both the experiences and level of protection provided by electronic safety interlock systems. This research is intended to analyze the strength and weaknesses of several different types of interlock system implementation methodologies. Research, medical, and industrial accelerators are compared. Thomas Jefferson National Accelerator Facility (TJNAF) was one of the first large particle accelerators to implement a safety interlock system using programmable logic controllers. Since that time all of the major new U.S. accelerator construction projects plan to use some form of programmable electronics as part of a safety interlock system in some capacity
Safety evaluation for instrumentation and control system upgrading project of Malaysian TRIGA MARK II PUSPATI Research reactor

International Nuclear Information System (INIS)

Ridha Roslan; Nik Mohd Faiz Khairuddin

2013-01-01

Full-text: Malaysian TRIGA MARK II research reactor has been in safe operation since its first criticality in 1982. The reactor is licensed to be operated by Malaysian Nuclear Agency to perform training and research development related activities. Due to its extensive operation since last three decades, the option of modifications for safety and safety-related item and component become a necessary to replace the outdated equipment to a stat-of-art, reliable technologies. This paper will present the current regulatory activities performed by Atomic Energy Licensing Board (AELB) to ensure the upgrading of analogue to digital instrumentation and control system is implemented in safe manner. The review activity includes documentation review, manufacturer quality audit and on-site inspection for commissioning. The review performed by AELB is based on The International Atomic Energy Agency (IAEA) Safety Requirements NS-R-4, entitled Safety of Research Reactors. During this endeavour, AELB seeks technical cooperation from Korea Institute of Nuclear Safety (KINS), the nuclear experts organization of the country of origin of the instrumentation and control technology. The regulatory activity is still on-going and is expected to be completed by issuance of Authorization for Restart on December 2013. (author)
Technical safety requirements control level verification

International Nuclear Information System (INIS)

STEWART, J.L.

1999-01-01

A Technical Safety Requirement (TSR) control level verification process was developed for the Tank Waste Remediation System (TWRS) TSRs at the Hanford Site in Richland, WA, at the direction of the US. Department of Energy, Richland Operations Office (RL). The objective of the effort was to develop a process to ensure that the TWRS TSR controls are designated and managed at the appropriate levels as Safety Limits (SLs), Limiting Control Settings (LCSs), Limiting Conditions for Operation (LCOs), Administrative Controls (ACs), or Design Features. The TSR control level verification process was developed and implemented by a team of contractor personnel with the participation of Fluor Daniel Hanford, Inc. (FDH), the Project Hanford Management Contract (PHMC) integrating contractor, and RL representatives. The team was composed of individuals with the following experience base: nuclear safety analysis; licensing; nuclear industry and DOE-complex TSR preparation/review experience; tank farm operations; FDH policy and compliance; and RL-TWRS oversight. Each TSR control level designation was completed utilizing TSR control logic diagrams and TSR criteria checklists based on DOE Orders, Standards, Contractor TSR policy, and other guidance. The control logic diagrams and criteria checklists were reviewed and modified by team members during team meetings. The TSR control level verification process was used to systematically evaluate 12 LCOs, 22 AC programs, and approximately 100 program key elements identified in the TWRS TSR document. The verification of each TSR control required a team consensus. Based on the results of the process, refinements were identified and the TWRS TSRs were modified as appropriate. A final report documenting key assumptions and the control level designation for each TSR control was prepared and is maintained on file for future reference. The results of the process were used as a reference in the RL review of the final TWRS TSRs and control suite. RL
Safety implications of computerized process control in nuclear power plants

International Nuclear Information System (INIS)

1991-02-01

Modern nuclear power plants are making increasing use of computerized process control because of the number of potential benefits that accrue. This practice not only applies to new plants but also to those in operation. Here, the replacement of both conventional process control systems and outdated computerized systems is seen to be of benefit. Whilst this contribution is obviously of great importance to the viability of nuclear electricity generation, it must be recognized that there are major safety concerns in taking this route. However, there is the potential for enhancing the safety of nuclear power plants if the full power of microcomputers and the associated electronics is applied correctly through well designed, engineered, installed and maintained systems. It is essential that areas where safety can be improved be identified and that the pitfalls are clearly marked so that they can be avoided. The deliberations of this Technical Committee Meeting are a step on the road to this goal of improved safety through computerized process control. This report also contains the papers presented at the technical committee meeting by participants. A separate abstract was prepared for each of these 15 presentations. Refs, figs and tabs
NASA System Safety Handbook. Volume 2: System Safety Concepts, Guidelines, and Implementation Examples

Science.gov (United States)

Dezfuli, Homayoon; Benjamin, Allan; Everett, Christopher; Feather, Martin; Rutledge, Peter; Sen, Dev; Youngblood, Robert

2015-01-01

This is the second of two volumes that collectively comprise the NASA System Safety Handbook. Volume 1 (NASASP-210-580) was prepared for the purpose of presenting the overall framework for System Safety and for providing the general concepts needed to implement the framework. Volume 2 provides guidance for implementing these concepts as an integral part of systems engineering and risk management. This guidance addresses the following functional areas: 1.The development of objectives that collectively define adequate safety for a system, and the safety requirements derived from these objectives that are levied on the system. 2.The conduct of system safety activities, performed to meet the safety requirements, with specific emphasis on the conduct of integrated safety analysis (ISA) as a fundamental means by which systems engineering and risk management decisions are risk-informed. 3.The development of a risk-informed safety case (RISC) at major milestone reviews to argue that the systems safety objectives are satisfied (and therefore that the system is adequately safe). 4.The evaluation of the RISC (including supporting evidence) using a defined set of evaluation criteria, to assess the veracity of the claims made therein in order to support risk acceptance decisions.
A study on enforcement effects of radiation safety control regulations for diagnostic X-ray equipment

International Nuclear Information System (INIS)

Sung, Mo IL; Park, Myeong Hwan; Kwon, Duk Moon; Lee, Joon IL

1999-01-01

The purposes of this study are to analyze the realities after enforcements of safety control regulations for diagnostic X-ray equipment and to suggest means for an improvement of low radiation safety control. A questionnaire survey for medical radiologic technologists was carried out to determine enforcement effects of the safety control regulations. The results of analysis from the survey are as follows. That is, most of he respondents realized the importance of the radiation safety control system, but about a half of them revealed that regulations were not well observed in accordance with their purposes. Only 43.9 percent of the respondents took an active part in quality control of radiation. And responsibility, sex, age, and knowledge for safety control were important indicators for observations of the regulations. Training for the safety control regulations are needed to ensure safety control and proper usage of diagnostic X-ray equipment. And management of organizations using diagnostic X-ray equipment have to understand and stress the importance of radiation safety control system. (author)
Safety parameter display system for Kalinin NPP

International Nuclear Information System (INIS)

Andreev, V.I.; Videneev, E.N.; Tissot, J.C.; Joonekindt, D.; Davidenko, N.N.; Shaftan, G.I.; Dounaev, V.G.; Neboyan, V.T.

1995-01-01

The paper discusses the safety parameter display system (SPDS), which is being designed for Kalinin NPP. The assessment of the safety status of the plant is done by the continuous monitoring of six critical safety functions and the corresponding status trees. Besides, a number of additional functions are realized within the scope of KlnNPP, aimed at providing the operator and the safety engineer in the main control room with more detailed information in accidental situation as well as during the normal operation. In particular, these functions are: archiving, data logs and alarm handling, safety actions monitoring, mnemonic diagrams indicating the state of main technological equipment and basic plant parameters, reference data, etc. As compared with the traditional scope of functions of this kind of systems, the functionality of KlnNPP SPDS is significantly expanded due to the inclusion in it the operator support function ''computerized procedures''. The basic SPDS implementation platform is ADACS of SEMA GROUP design. The system architecture includes two workstations in the main control room: one is for reactor operator and the other one for safety engineer. Every station has two CRT screens which ensures computerized procedures implementation and provides for extra services for the operator. Also, the information from the SPDS is transmitted to the local crisis center and to the crisis center of the State utility organization concern ''Rosenergoatom''. (author). 3 refs, 6 figs, 1 tab
Luxury cruise? The safety potential of advanced cruise control.

NARCIS (Netherlands)

Oei, H.L.

2003-01-01

The principles of advanced cruise control (ACC) are outlined and the requirements for an ACC system are described. An intelligent cruise control system fitted in a Nissan Primera was tested on the road over a 2-week period by 10 drivers, eight of which were experts in road safety. Most test-drives
Qualification of FPGA-Based Safety-Related PRM System

International Nuclear Information System (INIS)

Miyazaki, Tadashi; Oda, Naotaka; Goto, Yasushi; Hayashi, Toshifumi

2011-01-01

Toshiba has developed Non-rewritable (NRW) Field Programmable Gate Array (FPGA)-based safety-related Instrumentation and Control (I and C) system. Considering application to safety-related systems, nonvolatile and non-rewritable FPGA which is impossible to be changed after once manufactured has been adopted in Toshiba FPGA-based system. FPGA is a device which consists only of basic logic circuits, and FPGA performs defined processing which is configured by connecting the basic logic circuit inside the FPGA. FPGA-based system solves issues existing both in the conventional systems operated by analog circuits (analog-based system) and the systems operated by central processing unit (CPU-based system). The advantages of applying FPGA are to keep the long-life supply of products, improving testability (verification), and to reduce the drift which may occur in analog-based system. The system which Toshiba developed this time is Power Range Neutron Monitor (PRM). Toshiba is planning to expand application of FPGA-based technology by adopting this development process to the other safety-related systems such as RPS from now on. Toshiba developed a special design process for NRW-FPGA-based safety-related I and C systems. The design process resolves issues for many years regarding testability of the digital system for nuclear safety application. Thus, Toshiba NRW-FPGA-based safety-related I and C systems has much advantage to be a would standard of the digital systems for nuclear safety application. (author)
Software Quality Assurance for Nuclear Safety Systems

International Nuclear Information System (INIS)

Sparkman, D R; Lagdon, R

2004-01-01

The US Department of Energy has undertaken an initiative to improve the quality of software used to design and operate their nuclear facilities across the United States. One aspect of this initiative is to revise or create new directives and guides associated with quality practices for the safety software in its nuclear facilities. Safety software includes the safety structures, systems, and components software and firmware, support software and design and analysis software used to ensure the safety of the facility. DOE nuclear facilities are unique when compared to commercial nuclear or other industrial activities in terms of the types and quantities of hazards that must be controlled to protect workers, public and the environment. Because of these differences, DOE must develop an approach to software quality assurance that ensures appropriate risk mitigation by developing a framework of requirements that accomplishes the following goals: (sm b ullet) Ensures the software processes developed to address nuclear safety in design, operation, construction and maintenance of its facilities are safe (sm b ullet) Considers the larger system that uses the software and its impacts (sm b ullet) Ensures that the software failures do not create unsafe conditions Software designers for nuclear systems and processes must reduce risks in software applications by incorporating processes that recognize, detect, and mitigate software failure in safety related systems. It must also ensure that fail safe modes and component testing are incorporated into software design. For nuclear facilities, the consideration of risk is not necessarily sufficient to ensure safety. Systematic evaluation, independent verification and system safety analysis must be considered for software design, implementation, and operation. The software industry primarily uses risk analysis to determine the appropriate level of rigor applied to software practices. This risk-based approach distinguishes safety
Research on Integration of NPP Operational Safety Management Performance Systems

International Nuclear Information System (INIS)

Chi, Miao; Shi, Liping

2014-01-01

The operational safety management of Nuclear Power Plants demands systematic planning and integrated control. NPPs are following the well-developed safety indicator systems proposed by IAEA Operational Safety Performance Indicator Programme, NRC Reactor Oversight Process or the other institutions. Integration of the systems is proposed to benefiting from the advantages of both systems and avoiding improper application into the real world. The authors analyzed the possibility and necessity for system integration, and propose an indicator system integrating method
Reactor safety: the Nova computer system

International Nuclear Information System (INIS)

Eisgruber, H.; Stadelmann, W.

1991-01-01

After instances of maloperation, the causes of defects, the effectiveness of the measures taken to control the situation, and possibilities to avoid future recurrences need to be investigated above all before the plant is restarted. The most important aspect in all these efforts is to check the sequence in time, and the completeness, of the control measures initiated automatically. For this verification, a computer system is used instead of time-consuming manual analytical techniques, which produces the necessary information almost in real time. The results are available within minutes after completion of the measures initiated automatically. As all short-term safety functions are initiated by automatic systems, their consistent and comprehensive verification results in a clearly higher level of safety. The report covers the development of the computer system, and its implementation, in the Gundremmingen nuclear power station. Similar plans are being pursued in Biblis and Muelheim-Kaerlich. (orig.) [de
Modeling for safety in a synthesis-centric systems engineering framework

NARCIS (Netherlands)

Markovski, J.; Mortel - Fronczak, van de J.M.; Ortmeier, F.; Daniel, P.

2012-01-01

The ever-increasing complexity of safety-critical systems puts high demands on safety assurance and certification. We focus on the development of control software, where safety) requirements engineering plays a crucial and delicate role. Nowadays, most of the safety features are ensured by the
Maintaining and improving the control and safety systems for the Electromagnetic Calorimeter of the CMS experiment

CERN Document Server

Di Calafiori, Diogo Raphael; Dissertori, Günther; Holme, Oliver; Jovanovic, Dragoslav; Lustermann, Werner; Zelepoukine, Serguei

2012-01-01

This paper presents the current architecture of the control and safety systems designed and implemented for the Electromagnetic Calorimeter (ECAL) of the Compact Muon Solenoid (CMS) experiment at the Large Hadron Collider (LHC). An evaluation of system performance during all CMS physics data taking periods is reported, with emphasis on how software and hardware solutions are used to overcome limitations, whilst maintaining and improving reliability and robustness. The outcomes of the CMS ECAL Detector Control System (DCS) Software Analysis Project were a fundamental step towards the integration of all control system applications and the consequent piece-by-piece software improvements allowed a smooth transition to the latest revision of the system. The ongoing task of keeping the system in-line with new hardware technologies and software platforms specified by the CMS DCS Group is discussed. The structure of the comprehensive support service with detailed incident logging is presented in addition to a complet...
Noise problems in PFR safety system power supplies

Energy Technology Data Exchange (ETDEWEB)

Hodgson, D

1978-08-15

Reactor power on PFR is controlled by eleven neutron absorbing rods. They are held vertically above the core by electromagnets which, in turn, are mounted on motorised lead screws. Five of the rods are used as power control elements; five are used as shut off rods and the eleventh is a safety rod, positioned vertically above the centre of the reactor core. All rods fall under gravity into the reactor when the electro magnets are de-energised. In addition the safety rod can be driven into the reactor by a pneumatic piston and cylinder mechanism. This provides some degree of diversity from the common design of mechanism used by the control and shut off rods. Also being in the center it can be armed and ready for use when the above core shield is rotated. The automatic protective system for the plant provides the electro magnet currents and hence the reactor trip signals via two separate and diverse safety systems, namely a ''Relay System'' and a ''Solid State System.'' The two systems are completely separate and independent. The overall probability of failure of the complete system to release at least 3 rods when called upon to do so is less than 10/sup -6/.
Development of a safety parameter supervision system for Angra-1

International Nuclear Information System (INIS)

Silva, R.A. da; Thome Filho, Z.D.; Schirru, R.; Martinez, A.S.; Oliveira, L.F.S. de

1986-01-01

The Safety Parameter Supervision System (SSPS) which is a computerized system for monitoring essential parameters in real time, determining the safety status and emergency procedures for returning normal reactor operation, in case of an anomaly occurrence, is presented. The SSPS consists of three sub-systems: Integrated parameter monitoring system which gives to operators an integrated vision of values of a parameter set, able to detect any deviation of normal reactor operation; safety critical function system which evaluates safety status in terms of a safety critical function set appointed in advance, and in case of violation of any critical function, it initiates the adequate emergency procedure to return normal operation; and safety parameter computer system which carries out the arquirement of analogic and digital control signals of nuclear power plant. (M.C.K.) [pt
KAERI software safety guideline for developing safety-critical software in digital instrumentation and control system of nuclear power plant

International Nuclear Information System (INIS)

Lee, Jang Soo; Kim, Jang Yeol; Eum, Heung Seop.

1997-07-01

Recently, the safety planning for safety-critical software systems is being recognized as the most important phase in the software life cycle, and being developed new regulatory positions and standards by the regulatory and the standardization organization. The requirements for software important to safety of nuclear reactor are described in such positions and standards. Most of them are describing mandatory requirements, what shall be done, for the safety-critical software. The developers of such a software. However, there have been a lot of controversial factors on whether the work practices satisfy the regulatory requirements, and to justify the safety of such a system developed by the work practices, between the licenser and the licensee. We believe it is caused by the reason that there is a gap between the mandatory requirements (What) and the work practices (How). We have developed a guidance to fill such gap, which can be useful for both licenser and licensee to conduct a justification of the safety in the planning phase of developing the software for nuclear reactor protection systems. (author). 67 refs., 13 tabs., 2 figs
KAERI software safety guideline for developing safety-critical software in digital instrumentation and control system of nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Lee, Jang Soo; Kim, Jang Yeol; Eum, Heung Seop

1997-07-01

Recently, the safety planning for safety-critical software systems is being recognized as the most important phase in the software life cycle, and being developed new regulatory positions and standards by the regulatory and the standardization organization. The requirements for software important to safety of nuclear reactor are described in such positions and standards. Most of them are describing mandatory requirements, what shall be done, for the safety-critical software. The developers of such a software. However, there have been a lot of controversial factors on whether the work practices satisfy the regulatory requirements, and to justify the safety of such a system developed by the work practices, between the licenser and the licensee. We believe it is caused by the reason that there is a gap between the mandatory requirements (What) and the work practices (How). We have developed a guidance to fill such gap, which can be useful for both licenser and licensee to conduct a justification of the safety in the planning phase of developing the software for nuclear reactor protection systems. (author). 67 refs., 13 tabs., 2 figs.
V and V methods of a safety-critical software for a programmable logic controller

Energy Technology Data Exchange (ETDEWEB)

Kim, Jang Yeol; Lee, Young Jun; Cha, Kyung Ho; Cheon, Se Woo; Lee, Jang Soo; Kwon, Kee Choon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kong, Seung Ju [Korea Hydro and Nuclear Power Co., Ltd, Daejeon (Korea, Republic of)

2005-11-15

This paper addresses the Verification an Validation(V and V) process and the methodology for an embedded real time software of a safety-grade Programmable Logic Controller(PLC). This safety-grade PLC is being developed as one of the Korean Nuclear Instrumentation and Control System(KNICS) project KNICS projects are developing a Reactor Protection System(RPS) and an Engineered Safety Feature-Component Control System(ESF-CCS) as well as a safety-grade PLC. The safety-grade PLC will be a major component that encomposes the RPS systems and the ESF-CCS systems as nuclear instruments and control equipment. This paper describes the V and V guidelines an procedures, V and V environment, V and V process and methodology, and the V and V tools in the KNICS projects. Specifically, it describes the real-time operating system V and V experience which corresponds to the requirement analysis phase, design phase and the implementation and testing phase of the software development life cycle. Main activities of the V and V for the PLC system software are a technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, software safety analysis, and a software configuration management. The proposed V and V methodology satisfies the Standard Review Plan(SRP)/Branch Technical Position(BTP)-14 criteria for the safety software in nuclear power plants. The proposed V and V methodology is going to be used to verify the upcoming software life cycle in the KNICS projects.

V and V methods of a safety-critical software for a programmable logic controller

International Nuclear Information System (INIS)

Kim, Jang Yeol; Lee, Young Jun; Cha, Kyung Ho; Cheon, Se Woo; Lee, Jang Soo; Kwon, Kee Choon; Kong, Seung Ju

2005-01-01

This paper addresses the Verification an Validation(V and V) process and the methodology for an embedded real time software of a safety-grade Programmable Logic Controller(PLC). This safety-grade PLC is being developed as one of the Korean Nuclear Instrumentation and Control System(KNICS) project KNICS projects are developing a Reactor Protection System(RPS) and an Engineered Safety Feature-Component Control System(ESF-CCS) as well as a safety-grade PLC. The safety-grade PLC will be a major component that encomposes the RPS systems and the ESF-CCS systems as nuclear instruments and control equipment. This paper describes the V and V guidelines an procedures, V and V environment, V and V process and methodology, and the V and V tools in the KNICS projects. Specifically, it describes the real-time operating system V and V experience which corresponds to the requirement analysis phase, design phase and the implementation and testing phase of the software development life cycle. Main activities of the V and V for the PLC system software are a technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, software safety analysis, and a software configuration management. The proposed V and V methodology satisfies the Standard Review Plan(SRP)/Branch Technical Position(BTP)-14 criteria for the safety software in nuclear power plants. The proposed V and V methodology is going to be used to verify the upcoming software life cycle in the KNICS projects
Contributions to the research programs in nuclear and industrial electronics, domestic production of instrumentation, safety and control systems and equipment for nuclear reactors and auxiliary installations

International Nuclear Information System (INIS)

Talpariu, C; Talpariu, J.; Matei, C.

2001-01-01

Domestic production of component system and equipment for the control and safety of nuclear facilities was one of the priority objective of the Nuclear Research Institute Pitesti. The problems addressed were particularly related to design and production of analog and digital equipment for measurements, triggering and display of the values of process parameters as well as to regulating complex functions of this equipment. Associated to this effort were the research works concerning: - reliability and in-service life-time of the electronic components and equipment in the safety and control systems for nuclear processes; - radiation endurance of industrial electronic components; utilization of whirling currents in calandria tube testing; - expert systems and applications in nuclear reactor control and safety; design and testing methods of process real time software packages for safety in control critical systems for nuclear domain. There are presented characteristics of the following equipment: 1. amplifier for ionization chambers with triggering comparator circuits for the CANDU 600 reactor shut down system; 2. amplifier for ionization chambers without triggering comparator circuits for power regulating system; 3. safety and regulating computerized system for C9 and C5 cans; 4. acquisition system for dosimetric data in nuclear facilities; 5. program able digital comparator for the reactor shut down system; 6. stationary gamma areal monitors for CANDU 600 reactors and other nuclear facilities
Some safety considerations in laser-controlled thermonuclear reactors. Final report

International Nuclear Information System (INIS)

Botts, T.E.; Breton, D.; Chan, C.K.; Levy, S.I.; Sehnert, M.; Ullman, A.Z.

1978-07-01

A major objective of this study was to identify potential safety questions for laser controlled thermonuclear reactors. From the safety viewpoint, it does not appear that the actual laser controlled thermonuclear reactor conceptual designs present hazards very different than those of magnetically confined fusion reactors. Some aspects seem beneficial, such as small lithium inventories, and the absence of cryogenic devices, while other aspects are new, for example the explosion of pressure vessels and laser hazards themselves. Major aspects considered in this report include: (a) general safety considerations, (b) tritium inventories, (c) system behavior during loss of flow accidents, and (d) safety considerations of laser related penetrations
Reactor safety systems

International Nuclear Information System (INIS)

Kafka, P.

1975-01-01

The spectrum of possible accidents may become characterized by the 'maximum credible accident', which will/will not happen. Similary, the performance of safety systems in a multitude of situations is sometimes simplified to 'the emergency system will/will not work' or even 'reactors are/ are not safe'. In assessing safety, one must avoid this fallacy of reducing a complicated situation to the simple black-and-white picture of yes/no. Similarly, there is a natural tendency continually to improve the safety of a system to assure that it is 'safe enough'. Any system can be made safer and there is usually some additional cost. It is important to balance the increased safety against the increased costs. (orig.) [de
Road identification for its-integrated systems of automotive active safety

Directory of Open Access Journals (Sweden)

V. Ivanov

2005-04-01

Full Text Available The paper discusses several aspects of active safety control for automotive application. Particular emphasis is placed on the fuzzy logic determination of friction properties of a tyre-road contact. An example of vehicle control systems equipped with off-board sensors of road roughness, temperature, moisture and rain intensity demonstrates the implementation of this approach. The paper proposes conceptual solutions for preventive active safety control applied to vehicles which are integrated in an intelligent transportation system.
Reactor system safety assurance

International Nuclear Information System (INIS)

Mattson, R.J.

1984-01-01

The philosophy of reactor safety is that design should follow established and conservative engineering practices, there should be safety margins in all modes of plant operation, special systems should be provided for accidents, and safety systems should have redundant components. This philosophy provides ''defense in depth.'' Additionally, the safety of nuclear power plants relies on ''safety systems'' to assure acceptable response to design basis events. Operating experience has shown the need to study plant response to more frequent upset conditions and to account for the influence of operators and non-safety systems on overall performance. Defense in depth is being supplemented by risk and reliability assessment
76 FR 49532 - Federal Motor Vehicle Safety Standards; Electronic Stability Control; Technical Report on the...

Science.gov (United States)

2011-08-10

...-0112] Federal Motor Vehicle Safety Standards; Electronic Stability Control; Technical Report on the Effectiveness of Electronic Stability Control Systems for Cars and LTVs AGENCY: National Highway Traffic Safety..., Electronic Stability Control Systems. The report's title is: Crash Prevention Effectiveness in Light-Vehicle...
Safety control and minimization of radioactive wastes

International Nuclear Information System (INIS)

Wang Jinming; Rong Feng; Li Jinyan; Wang Xin

2010-01-01

Compared with the developed countries, the safety control and minimization of the radwastes in China are under-developed. The research of measures for the safety control and minimization of the radwastes is very important for the safety control of the radwastes, and the reduction of the treatment and disposal cost and environment radiation hazards. This paper has systematically discussed the safety control and the minimization of the radwastes produced in the nuclear fuel circulation, nuclear technology applications and the process of decommission of nuclear facilities, and has provided some measures and methods for the safety control and minimization of the radwastes. (authors)
Promises in intelligent plant control systems

International Nuclear Information System (INIS)

Otaduy, P.J.

1987-01-01

The control system is the brain of a power plant. The traditional goal of control systems has been productivity. However, in nuclear power plants the potential for disaster requires safety to be the dominant concern, and the worldwide political climate demands trustworthiness for nuclear power plants. To keep nuclear generation as a viable option for power in the future, trust is the essential critical goal which encompasses all others. In most of today's nuclear plants the control system is a hybrid of analog, digital, and human components that focuses on productivity and operates under the protective umbrella of an independent engineered safety system. Operation of the plant is complex, and frequent challenges to the safety system occur which impact on their trustworthiness. Advances in nuclear reactor design, computer sciences, and control theory, and in related technological areas such as electronics and communications as well as in data storage, retrieval, display, and analysis have opened a promise for control systems with more acceptable human brain-like capabilities to pursue the required goals. This paper elaborates on the promise of futuristic nuclear power plants with intelligent control systems and addresses design requirements and implementation approaches
Upgrading safety systems of industrial irradiation facilities

International Nuclear Information System (INIS)

Gomes, R.S.; Gomes, J.D.R.L.; Costa, E.L.C.; Costa, M.L.L.; Thomé, Z.D.

2017-01-01

The first industrial irradiation facility in operation in Brazil was designed in the 70s. Nowadays, twelve commercial and research facilities are in operation and two already decommissioned. Minor modifications and upgrades, as sensors replacement, have been introduced in these facilities, in order to reduce the technological gap in the control and safety systems. The safety systems are designed in agreement with the codes and standards at the time. Since then, new standards, codes and recommendations, as well as lessons learned from accidents, have been issued by various international committees or regulatory bodies. The rapid advance of the industry makes the safety equipment used in the original construction become obsolete. The decreasing demand for these older products means that they are no longer produced, which can make it impossible or costly to obtain spare parts and the expansion of legacy systems to include new features. This work aims to evaluate existing safety systems at Brazilian irradiation facilities, mainly the oldest facilities, taking into account the recommended IAEA's design requirements. Irrespective of the fact that during its operational period no event with victims have been recorded in Brazilian facilities, and that the regulatory inspections do not present any serious deviations regarding the safety procedures, it is necessary an assessment of safety system with the purpose of bringing their systems to 'the state of the art', avoiding their rapid obsolescence. This study has also taken into account the knowledge, concepts and solutions developed to upgrading safety system in irradiation facilities throughout the world. (author)
Upgrading safety systems of industrial irradiation facilities

Energy Technology Data Exchange (ETDEWEB)

Gomes, R.S.; Gomes, J.D.R.L.; Costa, E.L.C.; Costa, M.L.L., E-mail: rogeriog@cnen.gov.br, E-mail: jlopes@cnen.gov.br, E-mail: evaldo@cnen.gov.br, E-mail: mara@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil). Diretoria de Radioproteção e Segurança Nuclear; Thomé, Z.D., E-mail: zielithome@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

2017-07-01

The first industrial irradiation facility in operation in Brazil was designed in the 70s. Nowadays, twelve commercial and research facilities are in operation and two already decommissioned. Minor modifications and upgrades, as sensors replacement, have been introduced in these facilities, in order to reduce the technological gap in the control and safety systems. The safety systems are designed in agreement with the codes and standards at the time. Since then, new standards, codes and recommendations, as well as lessons learned from accidents, have been issued by various international committees or regulatory bodies. The rapid advance of the industry makes the safety equipment used in the original construction become obsolete. The decreasing demand for these older products means that they are no longer produced, which can make it impossible or costly to obtain spare parts and the expansion of legacy systems to include new features. This work aims to evaluate existing safety systems at Brazilian irradiation facilities, mainly the oldest facilities, taking into account the recommended IAEA's design requirements. Irrespective of the fact that during its operational period no event with victims have been recorded in Brazilian facilities, and that the regulatory inspections do not present any serious deviations regarding the safety procedures, it is necessary an assessment of safety system with the purpose of bringing their systems to 'the state of the art', avoiding their rapid obsolescence. This study has also taken into account the knowledge, concepts and solutions developed to upgrading safety system in irradiation facilities throughout the world. (author)
Procedures for controlling the risks of reliability, safety, and availability of technical systems

International Nuclear Information System (INIS)

1987-01-01

The reference book covers four sections. Apart from the fundamental aspects of the reliability problem, of risk and safety and the relevant criteria with regard to reliability, the material presented explains reliability in terms of maintenance, logistics and availability, and presents procedures for reliability assessment and determination of factors influencing the reliability, together with suggestions for systems technical integration. The reliability assessment consists of diagnostic and prognostic analyses. The section on factors influencing reliability discusses aspects of organisational structures, programme planning and control, and critical activities. (DG) [de
Model-based safety analysis of a control system using Simulink and Simscape extended models

Directory of Open Access Journals (Sweden)

Shao Nian

2017-01-01

Full Text Available The aircraft or system safety assessment process is an integral part of the overall aircraft development cycle. It is usually characterized by a very high timely and financial effort and can become a critical design driver in certain cases. Therefore, an increasing demand of effective methods to assist the safety assessment process arises within the aerospace community. One approach is the utilization of model-based technology, which is already well-established in the system development, for safety assessment purposes. This paper mainly describes a new tool for Model-Based Safety Analysis. A formal model for an example system is generated and enriched with extended models. Then, system safety analyses are performed on the model with the assistance of automation tools and compared to the results of a manual analysis. The objective of this paper is to improve the increasingly complex aircraft systems development process. This paper develops a new model-based analysis tool in Simulink/Simscape environment.
Railing for safety: job demands, job control, and safety citizenship role definition.

Science.gov (United States)

Turner, Nick; Chmiel, Nik; Walls, Melanie

2005-10-01

This study investigated job demands and job control as predictors of safety citizenship role definition, that is, employees' role orientation toward improving workplace safety. Data from a survey of 334 trackside workers were framed in the context of R. A. Karasek's (1979) job demands-control model. High job demands were negatively related to safety citizenship role definition, whereas high job control was positively related to this construct. Safety citizenship role definition of employees with high job control was buffered from the influence of high job demands, unlike that of employees with low job control, for whom high job demands were related to lower levels of the construct. Employees facing both high job demands and low job control were less likely than other employees to view improving safety as part of their role orientation. Copyright (c) 2005 APA, all rights reserved.
Regulatory Control of Radiation Sources. Safety Guide (Arabic Edition)

International Nuclear Information System (INIS)

2012-01-01

This Safety Guide is intended to assist States in implementing the requirements established in Safety Standards Series No. GS-R-1, Legal and Governmental Infrastructure for Nuclear, Radiation, Radioactive Waste and Transport Safety, for a national regulatory infrastructure to regulate any practice involving radiation sources in medicine, industry, research, agriculture and education. The Safety Guide provides advice on the legislative basis for establishing regulatory bodies, including the effective independence of the regulatory body. It also provides guidance on implementing the functions and activities of regulatory bodies: the development of regulations and guides on radiation safety; implementation of a system for notification and authorization; carrying out regulatory inspections; taking necessary enforcement actions; and investigating accidents and circumstances potentially giving rise to accidents. The various aspects relating to the regulatory control of consumer products are explained, including justification, optimization of exposure, safety assessment and authorization. Guidance is also provided on the organization and staffing of regulatory bodies. Contents: 1. Introduction; 2. Legal framework for a regulatory infrastructure; 3. Principal functions and activities of the regulatory body; 4. Regulatory control of the supply of consumer products; 5. Functions of the regulatory body shared with other governmental agencies; 6. Organization and staffing of the regulatory body; 7. Documentation of the functions and activities of the regulatory body; 8. Support services; 9. Quality management for the regulatory system.
Rapid Prototyping of the Central Safety System for Nuclear Risk in ITER

Energy Technology Data Exchange (ETDEWEB)

Scibile, L. [ITER Organization, 13 - St. Paul lez Durance (France); Ambrosino, G.; De Tommasi, G.; Pironti, A. [Euratom-ENEA-CREATE, Universita di Napoli Federico II, Napoli (Italy)

2009-07-01

Full text of publication follows: In the current ITER Baseline design, the Central Safety System for Nuclear Risk (CSS-N) is the safety control system in charge to assure nuclear safety for the plant, personnel and environment. In particular it is envisaged that the CSS shall interface to the plant safety systems for nuclear risk and shall coordinate the individual protection provided by the intervention of these systems by the activation, where required, of additional protections. The design of such a system, together with its implementation, strongly depends on the requirements, particularly in terms of reliability. The CSS-N is a safety critical system, thus its validation and commissioning play a very important role, since the required level of reliability must be demonstrated. In such a scenario, where a new and non-conventional system has to be deployed, it is strongly recommended to use modeling and simulation tools since the early design phase. Indeed, the modeling tools will help in the definition of the system requirements, and they will be used to test and validate the control logic. Furthermore these tools can be used to rapid design the safety system and to carry out hardware-in-the-loop (HIL) simulations, which permit to assess the performance of the control hardware against a plant simulator. Both a control system prototype and a safety system oriented plant simulator have been developed to assess first the requirements and then the performance of the CSS-N. In particular the presented SW/HW framework permits to design and verify the CSS protection logics and to test and validate these logics by means of HIL simulations. This work introduces both the prototype and plant simulator architectures, together with the methodology adopted to design and implement these validation tools. (authors)
Aviation Safety Reporting System: Process and Procedures

Science.gov (United States)

Connell, Linda J.

1997-01-01

The Aviation Safety Reporting System (ASRS) was established in 1976 under an agreement between the Federal Aviation Administration (FAA) and the National Aeronautics and Space Administration (NASA). This cooperative safety program invites pilots, air traffic controllers, flight attendants, maintenance personnel, and others to voluntarily report to NASA any aviation incident or safety hazard. The FAA provides most of the program funding. NASA administers the program, sets its policies in consultation with the FAA and aviation community, and receives the reports submitted to the program. The FAA offers those who use the ASRS program two important reporting guarantees: confidentiality and limited immunity. Reports sent to ASRS are held in strict confidence. More than 350,000 reports have been submitted since the program's beginning without a single reporter's identity being revealed. ASRS removes all personal names and other potentially identifying information before entering reports into its database. This system is a very successful, proof-of-concept for gathering safety data in order to provide timely information about safety issues. The ASRS information is crucial to aviation safety efforts both nationally and internationally. It can be utilized as the first step in safety by providing the direction and content to informed policies, procedures, and research, especially human factors. The ASRS process and procedures will be presented as one model of safety reporting feedback systems.
Nuclear reactor safety system

International Nuclear Information System (INIS)

Ball, R.M.; Roberts, R.C.

1983-01-01

The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions
Nuclear reactor safety systems

International Nuclear Information System (INIS)

Ball, R.M.; Roberts, R.C.

1980-01-01

A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)
Safety management systems and their role in achieving high standards of operational safety

International Nuclear Information System (INIS)

Coulston, D.J.; Baylis, C.C.

2000-01-01

Achieving high standards of operational safety requires a robust management framework that is visible to all personnel with responsibility for its implementation. The structure of the management framework must ensure that all processes used to manage safety interlink in a logical and coherent manner, that is, they form a management system that leads to continuous improvement in safety performance. This Paper describes BNFL's safety management system (SMS). The SMS has management processes grouped within 5 main elements: 1. Policy, 2. Organisation, 3. Planning and Implementation, 4. Measuring and Reviewing Performance, 5. Audit. These elements reflect the overall process of setting safety objective (from Policy), measuring success and reviewing the performance. Effective implementation of the SMS requires senior managers to demonstrate leadership through their commitment and accountability. However, the SMS as a whole reflects that every employee at every level within BNFL is responsible for safety of operations under their control. The SMS therefore promotes a proactive safety culture and safe operations. The system is formally documented in the Company's Environmental, Health and Safety (EHS) Manual. Within in BNFL Group, the Company structures enables the Manual to provide overall SMS guidance and co-ordination to its range of nuclear businesses. Each business develops the SMS to be appropriate at all levels of its organisation, but ensuring that each level is consistent with the higher level. The Paper concludes with a summary of BNFL's safety performance. (author)

Safety evaluation of BWR off-gas treatment systems

International Nuclear Information System (INIS)

Schultz, R.J.; Schmitt, R.C.

1975-01-01

Some of the results of a safety evaluation performed on current generic types of BWR off-gas treatment systems including cooled and ambient temperature adsorber beds and cryogenics are presented. The evaluation covered the four generic types of off-gas systems and the systems of five major vendors. This study was part of original work performed under AEC contract for the Directorate of Regulatory Standards. The analysis techniques employed for the safety evaluation of these systems include: Fault Tree Analysis; FMECA (Failure Mode Effects and Criticality Analysis); general system comparisons, contaminant, system control, and design adequacy evaluations; and resultant Off-Site Dose Calculations. The salient areas presented are some of the potential problem areas, the approach that industry has taken to mitigate or design against potential upset conditions, and areas where possible deficiencies still exist. Potential problem areas discussed include hydrogen detonation, hydrogen release to equipment areas, operator/automatic control interface, and needed engineering evaluation to insure safe system operation. Of the systems reviewed, most were in the category of advanced or improved over that commonly in use today, and a conclusion from the study was that these systems offer excellent potential for noble gas control for BWR power plants where more stringent controls may be specified -- now or in the future. (U.S.)
Quality and safety control in the production of recreational-purpose cottage cheese

OpenAIRE

Мостова, Людмила Миколаївна; Клусович, Тетяна Валеріївна

2015-01-01

The paper deals with the relevant problem of implementing safety control systems according to HACCP principles for Ukrainian catering establishments in the integration of trade space. Taking into account high competition and consumer audience gaining priority, the work emphasizes the need to develop and implement safety control system according to HACCP principles as a precondition for leadership of catering establishments in the national market, and get significant competitive advantages in...
Quantitative risk assessment of digitalized safety systems

Energy Technology Data Exchange (ETDEWEB)

Shin, Sung Min; Lee, Sang Hun; Kang, Hym Gook [KAIST, Daejeon (Korea, Republic of); Lee, Seung Jun [UNIST, Ulasn (Korea, Republic of)

2016-05-15

A report published by the U.S. National Research Council indicates that appropriate methods for assessing reliability are key to establishing the acceptability of digital instrumentation and control (I and C) systems in safety-critical plants such as NPPs. Since the release of this issue, the methodology for the probabilistic safety assessment (PSA) of digital I and C systems has been studied. However, there is still no widely accepted method. Kang and Sung found three critical factors for safety assessment of digital systems: detection coverage of fault-tolerant techniques, software reliability quantification, and network communication risk. In reality the various factors composing digitalized I and C systems are not independent of each other but rather closely connected. Thus, from a macro point of view, a method that can integrate risk factors with different characteristics needs to be considered together with the micro approaches to address the challenges facing each factor.
Cyber Security Risk Assessment for the KNICS Safety Systems

International Nuclear Information System (INIS)

Lee, C. K.; Park, G. Y.; Lee, Y. J.; Choi, J. G.; Kim, D. H.; Lee, D. Y.; Kwon, K. C.

2008-01-01

In the Korea Nuclear I and C Systems Development (KNICS) project the platforms for plant protection systems are developed, which function as a reactor shutdown, actuation of engineered safety features and a control of the related equipment. Those are fully digitalized through the use of safety-grade programmable logic controllers (PLCs) and communication networks. In 2006 the Regulatory Guide 1.152 (Rev. 02) was published by the U.S. NRC and it describes the application of a cyber security to the safety systems in the Nuclear Power Plant (NPP). Therefore it is required that the new requirements are incorporated into the developed platforms to apply to NPP, and a cyber security risk assessment is performed. The results of the assessment were input for establishing the cyber security policies and planning the work breakdown to incorporate them
Lightweight monitoring and control system for coal mine safety using REST style.

Science.gov (United States)

Cheng, Bo; Cheng, Xin; Chen, Junliang

2015-01-01

The complex environment of a coal mine requires the underground environment, devices and miners to be constantly monitored to ensure safe coal production. However, existing coal mines do not meet these coverage requirements because blind spots occur when using a wired network. In this paper, we develop a Web-based, lightweight remote monitoring and control platform using a wireless sensor network (WSN) with the REST style to collect temperature, humidity and methane concentration data in a coal mine using sensor nodes. This platform also collects information on personnel positions inside the mine. We implement a RESTful application programming interface (API) that provides access to underground sensors and instruments through the Web such that underground coal mine physical devices can be easily interfaced to remote monitoring and control applications. We also implement three different scenarios for Web-based, lightweight remote monitoring and control of coal mine safety and measure and analyze the system performance. Finally, we present the conclusions from this study and discuss future work. Copyright © 2014 ISA. Published by Elsevier Ltd. All rights reserved.
Design requirements of communication architecture of SMART safety system

International Nuclear Information System (INIS)

Park, H. Y.; Kim, D. H.; Sin, Y. C.; Lee, J. Y.

2001-01-01

To develop the communication network architecture of safety system of SMART, the evaluation elements for reliability and performance factors are extracted from commercial networks and classified the required-level by importance. A predictable determinacy, status and fixed based architecture, separation and isolation from other systems, high reliability, verification and validation are introduced as the essential requirements of safety system communication network. Based on the suggested requirements, optical cable, star topology, synchronous transmission, point-to-point physical link, connection-oriented logical link, MAC (medium access control) with fixed allocation are selected as the design elements. The proposed architecture will be applied as basic communication network architecture of SMART safety system
DESIGN PACKAGE 1E SYSTEM SAFETY ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

M. Salem

1995-06-23

The purpose of this analysis is to systematically identify and evaluate hazards related to the Yucca Mountain Project Exploratory Studies Facility (ESF) Design Package 1E, Surface Facilities, (for a list of design items included in the package 1E system safety analysis see section 3). This process is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach was used since a radiological System Safety Analysis is not required. The risk assessment in this analysis characterizes the accident scenarios associated with the Design Package 1E structures/systems/components(S/S/Cs) in terms of relative risk and includes recommendations for mitigating all identified risks. The priority for recommending and implementing mitigation control features is: (1) Incorporate measures to reduce risks and hazards into the structure/system/component design, (2) add safety devices and capabilities to the designs that reduce risk, (3) provide devices that detect and warn personnel of hazardous conditions, and (4) develop procedures and conduct training to increase worker awareness of potential hazards, on methods to reduce exposure to hazards, and on the actions required to avoid accidents or correct hazardous conditions.
Traceability of Software Safety Requirements in Legacy Safety Critical Systems

Science.gov (United States)

Hill, Janice L.

2007-01-01

How can traceability of software safety requirements be created for legacy safety critical systems? Requirements in safety standards are imposed most times during contract negotiations. On the other hand, there are instances where safety standards are levied on legacy safety critical systems, some of which may be considered for reuse for new applications. Safety standards often specify that software development documentation include process-oriented and technical safety requirements, and also require that system and software safety analyses are performed supporting technical safety requirements implementation. So what can be done if the requisite documents for establishing and maintaining safety requirements traceability are not available?
Security for safety critical space borne systems

Science.gov (United States)

Legrand, Sue

1987-01-01

The Space Station contains safety critical computer software components in systems that can affect life and vital property. These components require a multilevel secure system that provides dynamic access control of the data and processes involved. A study is under way to define requirements for a security model providing access control through level B3 of the Orange Book. The model will be prototyped at NASA-Johnson Space Center.
Technical safety requirements control level verification; TOPICAL

International Nuclear Information System (INIS)

STEWART, J.L.

1999-01-01

A Technical Safety Requirement (TSR) control level verification process was developed for the Tank Waste Remediation System (TWRS) TSRs at the Hanford Site in Richland, WA, at the direction of the US. Department of Energy, Richland Operations Office (RL). The objective of the effort was to develop a process to ensure that the TWRS TSR controls are designated and managed at the appropriate levels as Safety Limits (SLs), Limiting Control Settings (LCSs), Limiting Conditions for Operation (LCOs), Administrative Controls (ACs), or Design Features. The TSR control level verification process was developed and implemented by a team of contractor personnel with the participation of Fluor Daniel Hanford, Inc. (FDH), the Project Hanford Management Contract (PHMC) integrating contractor, and RL representatives. The team was composed of individuals with the following experience base: nuclear safety analysis; licensing; nuclear industry and DOE-complex TSR preparation/review experience; tank farm operations; FDH policy and compliance; and RL-TWRS oversight. Each TSR control level designation was completed utilizing TSR control logic diagrams and TSR criteria checklists based on DOE Orders, Standards, Contractor TSR policy, and other guidance. The control logic diagrams and criteria checklists were reviewed and modified by team members during team meetings. The TSR control level verification process was used to systematically evaluate 12 LCOs, 22 AC programs, and approximately 100 program key elements identified in the TWRS TSR document. The verification of each TSR control required a team consensus. Based on the results of the process, refinements were identified and the TWRS TSRs were modified as appropriate. A final report documenting key assumptions and the control level designation for each TSR control was prepared and is maintained on file for future reference. The results of the process were used as a reference in the RL review of the final TWRS TSRs and control suite. RL
Enhancing Safety at Airline Operations Control Centre

Directory of Open Access Journals (Sweden)

Lukáš Řasa

2015-04-01

Full Text Available In recent years a new term of Safety Management System (SMS has been introduced into aviation legislation. This system is being adopted by airline operators. One of the groundbased actors of everyday operations is Operations Control Centre (OCC. The goal of this article has been to identify and assess risks and dangers which occur at OCC and create a template for OCC implementation into SMS.
75 FR 15620 - Federal Motor Vehicle Safety Standards; Air Brake Systems

Science.gov (United States)

2010-03-30

... fully develop improved brake systems and also to ensure vehicle control and stability while braking... [Docket No. NHTSA 2009-0175] RIN 2127-AK62 Federal Motor Vehicle Safety Standards; Air Brake Systems... Federal motor vehicle safety standard for air brake systems by requiring substantial improvements in...
Test Results of a Platform for Safety I and C Systems of SMART MMIS

International Nuclear Information System (INIS)

Suh, Yong Suk; Keum, Jong Yong; Jeong, Kwang Il; Lee, Joon Ku; Lee, Sang Seok; Kim, Kwan Woong

2011-01-01

SMART (System-integrated Modular Advanced ReacTor), a 330MWt integral pressurized light water reactor that integrates four reactor coolant pumps, one pressurizer, eight steam generators, and one reactor core into a reactor vessel, has been under development at KAERI since 1997. A standard design safety analysis report of the SMART prepared by KAERI was submitted to Korea institute of nuclear safety (KINS) at the end of 2010. KAERI aims to achieve standard design approval (SDA) from KINS by the end of 2011. SMART MMIS has been designed using digital systems. It has digital-based compact control rooms. Its instrumentation and control (I and C) systems are designed using modular equipment connected through datalinks. Non-safety I and C systems are designed based on the commercial distributed control systems. Safety I and C systems are based on a new platform developed by KAERI. The platform is a high-speed digital signal processor (DSP)-based control unit. It plays the role of a module that provides control functions of the safety I and C systems. The test facilities have been developed at KAERI since 2009. This paper presents the development and test results of the platform
Mathematical modelling of active safety system functions as tools for development of driverless vehicles

Science.gov (United States)

Ryazantsev, V.; Mezentsev, N.; Zakharov, A.

2018-02-01

This paper is dedicated to a solution of the issue of synthesis of the vehicle longitudinal dynamics control functions (acceleration and deceleration control) based on the element base of the vehicle active safety system (ESP) - driverless vehicle development tool. This strategy helps to reduce time and complexity of integration of autonomous motion control systems (AMCS) into the vehicle architecture and allows direct control of actuators ensuring the longitudinal dynamics control, as well as reduction of time for calibration works. The “vehicle+wheel+road” longitudinal dynamics control is complicated due to the absence of the required prior information about the control object. Therefore, the control loop becomes an adaptive system, i.e. a self-adjusting monitoring system. Another difficulty is the driver’s perception of the longitudinal dynamics control process in terms of comfort. Traditionally, one doesn’t pay a lot of attention to this issue within active safety systems, and retention of vehicle steerability, controllability and stability in emergency situations are considered to be the quality criteria. This is mainly connected to its operational limits, since it is activated only in critical situations. However, implementation of the longitudinal dynamics control in the AMCS poses another challenge for the developers - providing the driver with comfortable vehicle movement during acceleration and deceleration - while the possible highest safety level in terms of the road grip is provided by the active safety system (ESP). The results of this research are: universal active safety system - AMCS interaction interface; block diagram for the vehicle longitudinal acceleration and deceleration control as one of the active safety system’s integrated functions; ideology of adaptive longitudinal dynamics control, which enables to realize the deceleration and acceleration requested by the AMCS; algorithms synthesised; analytical experiments proving the
Importance of Decision Support Systems About Food Safety in Raw Milk Production

Directory of Open Access Journals (Sweden)

Ecem Akan

2015-12-01

Full Text Available In raw milk production decision support systems for control of food safety hazards has not been developed but main points of this system are available. The decision support systems’ elements include data identification at critical points in the milk supply chain, an information management system and data exchange. Decision supports systems has been developed on the basis of these elements. In dairy sector decision support systems are significant for controlling of food safety hazards and preferred by producers. When these systems are implemented in the milk supply chain, it can be prevented unnecessary sampling and analysis. In this article it will be underlined effects of decision support system elements on food safety of raw milk.
Impact of Passive Safety on FHR Instrumentation Systems Design and Classification

International Nuclear Information System (INIS)

Holcomb, David Eugene

2015-01-01

Fluoride salt-cooled high-temperature reactors (FHRs) will rely more extensively on passive safety than earlier reactor classes. 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants, establishes minimum design requirements to provide reasonable assurance of adequate safety. 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, provides guidance on how the safety significance of systems, structures, and components (SSCs) should be reflected in their regulatory treatment. The Nuclear Energy Institute (NEI) has provided 10 CFR 50.69 SSC Categorization Guideline (NEI-00-04) that factors in probabilistic risk assessment (PRA) model insights, as well as deterministic insights, through an integrated decision-making panel. Employing the PRA to inform deterministic requirements enables an appropriately balanced, technically sound categorization to be established. No FHR currently has an adequate PRA or set of design basis accidents to enable establishing the safety classification of its SSCs. While all SSCs used to comply with the general design criteria (GDCs) will be safety related, the intent is to limit the instrumentation risk significance through effective design and reliance on inherent passive safety characteristics. For example, FHRs have no safety-significant temperature threshold phenomena, thus enabling the primary and reserve reactivity control systems required by GDC 26 to be passively, thermally triggered at temperatures well below those for which core or primary coolant boundary damage would occur. Moreover, the passive thermal triggering of the primary and reserve shutdown systems may relegate the control rod drive motors to the control system, substantially decreasing the amount of safety-significant wiring needed. Similarly, FHR decay heat removal systems are intended to be running continuously to minimize the amount of safety-significant instrumentation needed to initiate
Safety design guide for safety related systems for CANDU 9

International Nuclear Information System (INIS)

Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

1996-03-01

In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new
Safety design guide for safety related systems for CANDU 9

Energy Technology Data Exchange (ETDEWEB)

Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Wright, A.C.D. [Atomic Energy of Canada Ltd., Toronto (Canada)

1996-03-01

In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new.
75 FR 67450 - Pipeline Safety: Control Room Management Implementation Workshop

Science.gov (United States)

2010-11-02

... regulations to address human factors and other aspects of control room management for certain pipelines where controllers use supervisory control and data acquisition (SCADA) systems. Under the final rule, pipeline... Washington, DC on October 22, 2010. Jeffrey D. Wiese, Associate Administrator for Pipeline Safety. [FR Doc...
Safety assessment of computerized control and protection systems. Report of a technical committee meeting held in Vienna, 12-16 October 1992

International Nuclear Information System (INIS)

1994-12-01

In developing the views expressed in this document, papers were presented by delegates from Member States. A total of 6 papers were presented in all on topics ranging from applications of computerized control and protection systems in older plants and in new advanced reactors to methods for improving software reliability. In addition two informal presentations were provided by a vendor and a licensing authority. These presentations provided valuable insights into the application of computerized control and protection systems and into the concern of software reliability with proposals for diverse 'backup' systems of different types. This was supplemented by utility and vendor presentations on system designs. Following the presentations, three working groups were formed to produce their views on the licensing of software based safety systems on reliability models and techniques for assessment of computerized safety systems, and on systems considered for computerized upgrading (need, criteria, approach, pitfalls and benefits). This document represents these collected views with the papers presented attached as an annex. Refs, figs and tabs

Safety system function trends

International Nuclear Information System (INIS)

Johnson, C.

1989-01-01

This paper describes research to develop risk-based indicators of plant safety performance. One measure of the safety-performance of operating nuclear power plants is the unavailability of important safety systems. Brookhaven National Laboratory and Science Applications International Corporation are evaluating ways to aggregate train-level or component-level data to provide such an indicator. This type of indicator would respond to changes in plant safety margins faster than the currently used indicator of safety system unavailability (i.e., safety system failures reported in licensee event reports). Trends in the proposed indicator would be one indication of trends in plant safety performance and maintenance effectiveness. This paper summarizes the basis for such an indicator, identifies technical issues to be resolved, and illustrates the potential usefullness of such indicators by means of computer simulations and case studies
The Daresbury personnel safety system

International Nuclear Information System (INIS)

Poole, D.E.; Ring, T.

1989-01-01

The personnel safety system designed for the SRS at Daresbury is a unified system covering the three accelerators of the source itself, the beamlines and the experimental stations. The system has also been applied to the experimental areas of the Nuclear Structure Facility, and is therefore established as a site standard. A dual guardline interlock module forms a building block for a relay based interlock system completely independent of the machine control system, although comprehensive monitoring of the system status via the control system computer is a feature. An outline of the design criteria adopted for the system is presented together with a more detailed description of the philosophy of the guardline logic and the way this is implemented in a standard modular form. The emphasis is on the design features of a modern microprocessor based variant of the original SRS system. Experience with the original system during build-up and operation of the SRS facility is described. 2 refs., 4 figs
Probabilistic safety analysis for control rod drive system of ET-RR-1

International Nuclear Information System (INIS)

Nasr, M.; Nasser, O.

1988-01-01

The International Atomic Energy Agency (IAEA) co-ordinated a Research programme on Probabilistic Safety Analysis (PSA) for research reactors; with the participation of several countries. In the framework of this project (Project Int. 9/063) the Egyptian Atomic Energy Authority decided to perform a PSA study on the ET-RR-1 (Egypt Thermal Research Reactor). The study is conducted in collaboration between the nuclear regulatory and safety centre (NRSC) and the reactor department of the nuclear research centre at Inchass. The present work is a part of the PSA study on ET-RR- it is concerning a probabilistic safety analysis of the control rod drive mechanism
A study on the revision of nuclear safety act to build the foundation of nuclear export and import control system in Korea

International Nuclear Information System (INIS)

Yang, Seung Hyo; Choi, Sun Do

2012-01-01

Nuclear related items require export and import control beyond the multilateral export control system according to Safeguard Agreement, Additional Protocol and bilateral agreements. Besides Korea as a nuclear supplier is needed to actively cope with its export control system, which is being reinforced internationally. In regard to this trend, this study drew the revision plan of present Nuclear Safety Act to found the nuclear export and import control system in Korea by examining the related legislations and analyzing the implementation status of nuclear export and import control
A study on the revision of nuclear safety act to build the foundation of nuclear export and import control system in Korea

Energy Technology Data Exchange (ETDEWEB)

Yang, Seung Hyo; Choi, Sun Do [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

2012-10-15

Nuclear related items require export and import control beyond the multilateral export control system according to Safeguard Agreement, Additional Protocol and bilateral agreements. Besides Korea as a nuclear supplier is needed to actively cope with its export control system, which is being reinforced internationally. In regard to this trend, this study drew the revision plan of present Nuclear Safety Act to found the nuclear export and import control system in Korea by examining the related legislations and analyzing the implementation status of nuclear export and import control.
Safety classification of nuclear power plant systems, structures and components

International Nuclear Information System (INIS)

1992-01-01

The Safety Classification principles used for the systems, structures and components of a nuclear power plant are detailed in the guide. For classification, the nuclear power plant is divided into structural and operational units called systems. Every structure and component under control is included into some system. The Safety Classes are 1, 2 and 3 and the Class EYT (non-nuclear). Instructions how to assign each system, structure and component to an appropriate safety class are given in the guide. The guide applies to new nuclear power plants and to the safety classification of systems, structures and components designed for the refitting of old nuclear power plants. The classification principles and procedures applying to the classification document are also given
Integrated therapy safety management system.

Science.gov (United States)

Podtschaske, Beatrice; Fuchs, Daniela; Friesdorf, Wolfgang

2013-09-01

The aim is to demonstrate the benefit of the medico-ergonomic approach for the redesign of clinical work systems. Based on the six layer model, a concept for an 'integrated therapy safety management' is drafted. This concept could serve as a basis to improve resilience. The concept is developed through a concept-based approach. The state of the art of safety and complexity research in human factors and ergonomics forms the basis. The findings are synthesized to a concept for 'integrated therapy safety management'. The concept is applied by way of example for the 'medication process' to demonstrate its practical implementation. The 'integrated therapy safety management' is drafted in accordance with the six layer model. This model supports a detailed description of specific work tasks, the corresponding responsibilities and related workflows at different layers by using the concept of 'bridge managers'. 'Bridge managers' anticipate potential errors and monitor the controlled system continuously. If disruptions or disturbances occur, they respond with corrective actions which ensure that no harm results and they initiate preventive measures for future procedures. The concept demonstrates that in a complex work system, the human factor is the key element and final authority to cope with the residual complexity. The expertise of the 'bridge managers' and the recursive hierarchical structure results in highly adaptive clinical work systems and increases their resilience. The medico-ergonomic approach is a highly promising way of coping with two complexities. It offers a systematic framework for comprehensive analyses of clinical work systems and promotes interdisciplinary collaboration. © 2013 The Authors. British Journal of Clinical Pharmacology © 2013 The British Pharmacological Society.
Integrated therapy safety management system

Science.gov (United States)

Podtschaske, Beatrice; Fuchs, Daniela; Friesdorf, Wolfgang

2013-01-01

Aims The aim is to demonstrate the benefit of the medico-ergonomic approach for the redesign of clinical work systems. Based on the six layer model, a concept for an ‘integrated therapy safety management’ is drafted. This concept could serve as a basis to improve resilience. Methods The concept is developed through a concept-based approach. The state of the art of safety and complexity research in human factors and ergonomics forms the basis. The findings are synthesized to a concept for ‘integrated therapy safety management’. The concept is applied by way of example for the ‘medication process’ to demonstrate its practical implementation. Results The ‘integrated therapy safety management’ is drafted in accordance with the six layer model. This model supports a detailed description of specific work tasks, the corresponding responsibilities and related workflows at different layers by using the concept of ‘bridge managers’. ‘Bridge managers’ anticipate potential errors and monitor the controlled system continuously. If disruptions or disturbances occur, they respond with corrective actions which ensure that no harm results and they initiate preventive measures for future procedures. The concept demonstrates that in a complex work system, the human factor is the key element and final authority to cope with the residual complexity. The expertise of the ‘bridge managers’ and the recursive hierarchical structure results in highly adaptive clinical work systems and increases their resilience. Conclusions The medico-ergonomic approach is a highly promising way of coping with two complexities. It offers a systematic framework for comprehensive analyses of clinical work systems and promotes interdisciplinary collaboration. PMID:24007448
[Infection control and safety culture in German hospitals].

Science.gov (United States)

Hansen, Sonja; Schwab, Frank; Gropmann, Alexander; Behnke, Michael; Gastmeier, Petra

2016-07-01

Healthcare-associated infections (HAI) are the most frequent adverse events in the healthcare setting and their prevention is an important contribution to patient safety in hospitals. To analyse to what extent safety cultural aspects with relevance to infection control are implemented in German hospitals. Safety cultural aspects of infection control were surveyed with an online questionnaire; data were analysed descriptively. Data from 543 hospitals with a median of [IQR] 275 [157; 453] beds were analysed. Almost all hospitals (96.6 %) had internal guidelines for infection control (IC) in place; 82 % defined IC objectives, most often regarding hand hygiene (HH) (93 %) and multidrug resistant organisms (72 %) and less frequently for antibiotic stewardship (48 %) or prevention of specific HAI. In 94 % of hospitals, a reporting system for adverse events was in place, which was also used to report low compliance with HH, outbreaks and Clostridium difficile-associated infections. Members of the IC team were most often seen to hold daily responsibility for IC in the hospital, but rarely other hospital staff (94 versus 19 %). Safety cultural aspects are not fully implemented in German hospitals. IC should be more strongly implemented in healthcare workers' daily routine and more visibly supported by hospital management.
Safety design integrated in the building delivery system

DEFF Research Database (Denmark)

Jørgensen, Kirsten

2013-01-01

. The purpose of this article is to demonstrate how safety and health can be integrated in the design phases integrated in the management delivery systems within construction, The method for the research was to go through the building delivery system step by step and create a normative description of what, when......In construction, it is important to view safety and health as an integrated part of the way that “designers” are working. The designers cowers architects, constructors, engineers and others who carry out their consulting services in the design phase of a construction project. The philosophy...... and how to fully integrate safety in each part of the process. The result is a concept and guideline including control forms for how to integrate safety design in the Building Delivery System plus what to do and when. The concept has been tested in an educational context. The practical value...
IAEA Safety Standards on Management Systems and Safety Culture

International Nuclear Information System (INIS)

Persson, Kerstin Dahlgren

2007-01-01

The IAEA has developed a new set of Safety Standard for applying an integrated Management System for facilities and activities. The objective of the new Safety Standards is to define requirements and provide guidance for establishing, implementing, assessing and continually improving a Management System that integrates safety, health, environmental, security, quality and economic related elements to ensure that safety is properly taken into account in all the activities of an organization. With an integrated approach to management system it is also necessary to include the aspect of culture, where the organizational culture and safety culture is seen as crucial elements of the successful implementation of this management system and the attainment of all the goals and particularly the safety goals of the organization. The IAEA has developed a set of service aimed at assisting it's Member States in establishing. Implementing, assessing and continually improving an integrated management system. (author)
The Optimization of power reactor control system

International Nuclear Information System (INIS)

Danupoyo, S.D.

1997-01-01

A power reactor is an important part in nuclear powered electrical plant systems. Success in controlling the power reactor will establish safety of the whole power plant systems. Until now, the power reactor has been controlled by a classical control system that was designed based on output feedback method. To meet the safety requirements that are now more restricted, the recently used power reactor control system should be modified. this paper describes a power reactor control system that is designed based on a state feedback method optimized with LQG (Linear-quadrature-gaussian) method and equipped with a state estimator. A pressurized-water type reactor has been used as the model. by using a point kinetics method with one group delayed neutrons. the result of simulation testing shows that the optimized control system can control the power reactor more effective and efficient than the classical control system
Reactor control system. PWR

International Nuclear Information System (INIS)

2009-01-01

At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)
Access Safety Systems – New Concepts from the LHC Experience

CERN Document Server

Ladzinski, T; di Luca, S; Hakulinen, T; Hammouti, L; Riesco, T; Nunes, R; Ninin, P; Juget, J-F; Havart, F; Valentini, F; Sanchez-Corral Mena, E

2011-01-01

The LHC Access Safety System has introduced a number of new concepts into the domain of personnel protection at CERN. These can be grouped into several categories: organisational, architectural and concerning the end-user experience. By anchoring the project on the solid foundations of the IEC 61508/61511 methodology, the CERN team and its contractors managed to design, develop, test and commission on time a SIL3 safety system. The system uses a successful combination of the latest Siemens redundant safety programmable logic controllers with a traditional relay logic hardwired loop. The external envelope barriers used in the LHC include personnel and material access devices, which are interlocked door-booths introducing increased automation of individual access control, thus removing the strain from the operators. These devices ensure the inviolability of the controlled zones by users not holding the required credentials. To this end they are equipped with personnel presence detectors and th...
A Microbial Assessment Scheme to measure microbial performance of Food Safety Management Systems

NARCIS (Netherlands)

Jacxsens, L.; Kussaga, J.; Luning, P.A.; Spiegel, van der M.; Devlieghere, F.; Uyttendaele, M.

2009-01-01

A Food Safety Management System (FSMS) implemented in a food processing industry is based on Good Hygienic Practices (GHP), Hazard Analysis Critical Control Point (HACCP) principles and should address both food safety control and assurance activities in order to guarantee food safety. One of the
Safety management of a complex R and D ground operating system

Science.gov (United States)

Connors, J. F.; Maurer, R. A.

1975-01-01

A perspective on safety program management was developed for a complex R&D operating system, such as the NASA-Lewis Research Center. Using a systems approach, hazardous operations are subjected to third-party reviews by designated-area safety committees and are maintained under safety permit controls. To insure personnel alertness, emergency containment forces and employees are trained in dry-run emergency simulation exercises. The keys to real safety effectiveness are top management support and visibility of residual risks.
Reactivity requirements and safety systems for heavy water reactors

International Nuclear Information System (INIS)

Kati, S.L.; Rustagi, R.S.

1977-01-01

The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined
A Study of Cyber Security Activities for Development of Safety-related Controller

Energy Technology Data Exchange (ETDEWEB)

Lee, Myeongkyun; Song, Seunghwan; Yoo, Kwanwoo; Yun, Donghwa [Korea Univ., Seoul (Korea, Republic of)

2014-05-15

Nuclear Power Plant Regulatory guide describes the regulatory requirements to implement cyber security activities to ensure that design and operate to respond to cyber threats that exploited to vulnerability of digital-based technologies associated with safety-related digital instrumentation and control systems at nuclear power plants. Cyber security activities coverage is instrumentation and control systems to perform safety functions and digital-based equipment to use development, test, analysis and asset for instrumentation and control systems. Regulatory guidance is required to the cyber security activities that should be performed in each development phase of safety-related controller. Development organization should establish and implement to cyber security plans for responding to cyber threats throughout each lifecycle phase and the result of the cyber security activities should be generated to the documents. In addition, the independent verification and validation organization should perform simulated penetration test for enhancing response capabilities to cyber security threats and development organization should establish and implement response hardening solutions for the cyber security vulnerabilities identified in the simulated penetration test.
A Study of Cyber Security Activities for Development of Safety-related Controller

International Nuclear Information System (INIS)

Lee, Myeongkyun; Song, Seunghwan; Yoo, Kwanwoo; Yun, Donghwa

2014-01-01

Nuclear Power Plant Regulatory guide describes the regulatory requirements to implement cyber security activities to ensure that design and operate to respond to cyber threats that exploited to vulnerability of digital-based technologies associated with safety-related digital instrumentation and control systems at nuclear power plants. Cyber security activities coverage is instrumentation and control systems to perform safety functions and digital-based equipment to use development, test, analysis and asset for instrumentation and control systems. Regulatory guidance is required to the cyber security activities that should be performed in each development phase of safety-related controller. Development organization should establish and implement to cyber security plans for responding to cyber threats throughout each lifecycle phase and the result of the cyber security activities should be generated to the documents. In addition, the independent verification and validation organization should perform simulated penetration test for enhancing response capabilities to cyber security threats and development organization should establish and implement response hardening solutions for the cyber security vulnerabilities identified in the simulated penetration test
Risk-based rules for crane safety systems

Energy Technology Data Exchange (ETDEWEB)

Ruud, Stian [Section for Control Systems, DNV Maritime, 1322 Hovik (Norway)], E-mail: Stian.Ruud@dnv.com; Mikkelsen, Age [Section for Lifting Appliances, DNV Maritime, 1322 Hovik (Norway)], E-mail: Age.Mikkelsen@dnv.com

2008-09-15

The International Maritime Organisation (IMO) has recommended a method called formal safety assessment (FSA) for future development of rules and regulations. The FSA method has been applied in a pilot research project for development of risk-based rules and functional requirements for systems and components for offshore crane systems. This paper reports some developments in the project. A method for estimating target reliability for the risk-control options (safety functions) by means of the cost/benefit decision criterion has been developed in the project and is presented in this paper. Finally, a structure for risk-based rules is proposed and presented.

Risk-based rules for crane safety systems

International Nuclear Information System (INIS)

Ruud, Stian; Mikkelsen, Age

2008-01-01

The International Maritime Organisation (IMO) has recommended a method called formal safety assessment (FSA) for future development of rules and regulations. The FSA method has been applied in a pilot research project for development of risk-based rules and functional requirements for systems and components for offshore crane systems. This paper reports some developments in the project. A method for estimating target reliability for the risk-control options (safety functions) by means of the cost/benefit decision criterion has been developed in the project and is presented in this paper. Finally, a structure for risk-based rules is proposed and presented
Development of an FPGA-based controller for safety critical application

International Nuclear Information System (INIS)

Xing, A.; De Grosbois, J.; Sklyar, V.; Archer, P.; Awwal, A.

2011-01-01

In implementing safety functions, Field Programmable Gate Arrays (FPGA) technology offers a distinct combination of benefits and advantages over microprocessor-based systems. FPGAs can be designed such that the final product is purely hardware, without any overhead runtime software, bringing the design closer to a conventional hardware-based solution. On the other hand, FPGAs can implement more complex safety logic that would generally require microprocessor-based safety systems. There are now qualified FPGA-based platforms available on the market with a credible use history in safety applications in nuclear power plants. Atomic Energy of Canada (AECL), in collaboration with RPC Radiy, has initiated a development program to define a vigorous FPGA engineering process suitable for implementing safety critical functions at the application development level. This paper provides an update on the FPGA development program along with the proposed design model using function block diagrams for the development of safety controllers in CANDU applications. (author)
A Fiber Bragg Grating-Based Monitoring System for Roof Safety Control in Underground Coal Mining

Directory of Open Access Journals (Sweden)

Yiming Zhao

2016-10-01

Full Text Available Monitoring of roof activity is a primary measure adopted in the prevention of roof collapse accidents and functions to optimize and support the design of roadways in underground coalmines. However, traditional monitoring measures, such as using mechanical extensometers or electronic gauges, either require arduous underground labor or cannot function properly in the harsh underground environment. Therefore, in this paper, in order to break through this technological barrier, a novel monitoring system for roof safety control in underground coal mining, using fiber Bragg grating (FBG material as a perceived element and transmission medium, has been developed. Compared with traditional monitoring equipment, the developed, novel monitoring system has the advantages of providing accurate, reliable, and continuous online monitoring of roof activities in underground coal mining. This is expected to further enable the prevention of catastrophic roof collapse accidents. The system has been successfully implemented at a deep hazardous roadway in Zhuji Coal Mine, China. Monitoring results from the study site have demonstrated the advantages of FBG-based sensors over traditional monitoring approaches. The dynamic impacts of progressive face advance on roof displacement and stress have been accurately captured by the novel roadway roof activity and safety monitoring system, which provided essential references for roadway support and design of the mine.
SBO simulations for Integrated Passive Safety System (IPSS) using MARS

International Nuclear Information System (INIS)

Kim, Sang Ho; Jeong, Sung Yeop; Chang, Soon Heung

2012-01-01

The current nuclear power plants have lots of active safety systems with some passive safety systems. The safety of current and future nuclear power plants can be enhanced by the application of additional passive safety systems for the ultimate safety. It is helpful to install the passive safety systems on current nuclear power plants without the design change for the licensibility. For solving the problem about the system complexity shown in the Fukushima accidents, the current nuclear power plants are needed to be enhanced by an additional integrated and simplified system. As a previous research, the integrated passive safety system (IPSS) was proposed to solve the safety issues related with the decay heat removal, containment integrity and radiation release. It could be operated by natural phenomena like gravity, natural circulation and pressure difference without AC power. The five main functions of IPSS are: (a) Passive decay heat removal, (b) Passive emergency core cooling, (c) Passive containment cooling, (d) Passive in vessel retention and ex-vessel cooling, and (e) Filtered venting and pressure control. The purpose of this research is to analyze the performances of each function by using MARS code. The simulated accident scenarios were station black out (SBO) and the additional accidents accompanied by SBO
A systems engineering approach to implementation of safety management systems in the Norwegian fishing fleet

International Nuclear Information System (INIS)

McGuinness, Edgar; Utne, Ingrid B.

2014-01-01

The fishing industry is plagued by a long history of fatality and injury occurrence. Commercial fishing is hence recognized as the most dangerous and difficult of professional callings, in all jurisdictions. Fishing vessels have their own unique set of hazards, a myriad collection of complex occupational accident potentials, barely controlled, co-existing in a perilous work environment. The work in this article is directed by the Norwegian Systematic Health, Environmental and Safety Activities in Enterprises (1997) (Internal Control Regulations [1]), the ISM Code [2] for vessels and their recent applicability to the fishing fleet of Norway. Both safety management works place requirements on the vessel operators and crew to actively manage safety as an on-going concern. The application of these safety management system (SMS) control documents to fishing vessels is just the latest instalment in a continual drive to improve safety in this sector. The difficulty is that there has been no previous systematic approach to safety within the fishing fleet. This article uses the tenants of systems engineering to determine the requirements for such a SMS, detailing the limiting factors and restrictive issues of this complex operating environment. - Highlights: • Systems engineer is applied as a tool for determining requirements for design and construction of a safety management system (SMS). • Outlining a simplistic format, identifying, designingand facilitating improvement opportunities in the conduction and application of SMS’s on fishing vessels. • Knowledge provision is a key requirement of management systems, through provision of understanding, detail orientation and applicable skills for realization. • Outlining, what is to be done and how it is to be completed to accomplish compliance with pertinent legislative requirements. • Promoting a combination of documentation and communication arrangements by which the actionsnecessary for management can be
Reliability Improved Design for a Safety System Channel

Energy Technology Data Exchange (ETDEWEB)

Oh, Eung Se; Kim, Yun Goo [KHNP, Daejeon (Korea, Republic of)

2016-05-15

Nowadays, these systems are implemented with a same platform type, such as a qualified programmable logic controller (PLC). The platform intensively uses digital communication with fiber-optic links to reduce cabling costs and to achieve effective signal isolation. These communication interface and redundancies within a channel increase the complexness of an overall system design. This paper proposes a simpler channel architecture design to reduce the complexity and to enhance overall channel reliability. Simplified safety channel configuration is proposed and the failure probabilities are compared with baseline safety channel configuration using an estimated generic value. The simplified channel configuration achieves 40 percent failure reduction compare to baseline safety channel configuration. If this configuration can be implemented within a processor module, overall safety channel reliability is increase and costs of fabrication and maintenance will be greatly reduced.
Reliability Improved Design for a Safety System Channel

International Nuclear Information System (INIS)

Oh, Eung Se; Kim, Yun Goo

2016-01-01

Nowadays, these systems are implemented with a same platform type, such as a qualified programmable logic controller (PLC). The platform intensively uses digital communication with fiber-optic links to reduce cabling costs and to achieve effective signal isolation. These communication interface and redundancies within a channel increase the complexness of an overall system design. This paper proposes a simpler channel architecture design to reduce the complexity and to enhance overall channel reliability. Simplified safety channel configuration is proposed and the failure probabilities are compared with baseline safety channel configuration using an estimated generic value. The simplified channel configuration achieves 40 percent failure reduction compare to baseline safety channel configuration. If this configuration can be implemented within a processor module, overall safety channel reliability is increase and costs of fabrication and maintenance will be greatly reduced
The verification methodologies for a software modeling of Engineered Safety Features- Component Control System (ESF-CCS)

International Nuclear Information System (INIS)

Lee, Young-Jun; Cheon, Se-Woo; Cha, Kyung-Ho; Park, Gee-Yong; Kwon, Kee-Choon

2007-01-01

The safety of a software is not guaranteed through a simple testing of the software. The testing reviews only the static functions of a software. The behavior, dynamic state of a software is not reviewed by a software testing. The Ariane5 rocket accident and the failure of the Virtual Case File Project are determined by a software fault. Although this software was tested thoroughly, the potential errors existed internally. There are a lot of methods to solve these problems. One of the methods is a formal methodology. It describes the software requirements as a formal specification during a software life cycle and verifies a specified design. This paper suggests the methods which verify the design to be described as a formal specification. We adapt these methods to the software of a ESF-CCS (Engineered Safety Features-Component Control System) and use the SCADE (Safety Critical Application Development Environment) tool for adopting the suggested verification methods
Safety assessment and quality control of medical x-ray facilities in some hospitals in Ghana

International Nuclear Information System (INIS)

Darko, E.O.; Charles, D.F.

1998-01-01

Safety assessment and quality control measurements of diagnostic x-ray installations were carried out in five hospitals in Ghana. The study was focused on the siting, design and construction of the buildings housing the x-ray units, assessment of safety systems and devices and measurements of the technical performance, and film processing conditions. The location, inadequacies in the design/construction, unavailability of relevant safety systems and devices, violation of basic safety principles and poor performance of some of the x-ray facilities indicate the need to improve quality control programmes, safety culture and enforcement of regulatory standards in diagnostic x-ray examinations in Ghana. (author). 8 refs., 11 tabs., 8 figs
Research on the evaluation model of the software reliability in nuclear safety class digital instrumentation and control system

International Nuclear Information System (INIS)

Liu Ying; Yang Ming; Li Fengjun; Ma Zhanguo; Zeng Hai

2014-01-01

In order to analyze the software reliability (SR) in nuclear safety class digital instrumentation and control system (D-I and C), firstly, the international software design standards were analyzed, the standards' framework was built, and we found that the D-I and C software standards should follow the NUREG-0800 BTP7-14, according to the NRC NUREG-0800 review of requirements. Secondly, the quantitative evaluation model of SR using Bayesian Belief Network and thirteen sub-model frameworks were established. Thirdly, each sub-models and the weight of corresponding indexes in the evaluation model were analyzed. Finally, the safety case was introduced. The models lay a foundation for review and quantitative evaluation on the SR in nuclear safety class D-I and C. (authors)
Cyber Security Penetration Test for Digital Safety I and C Systems

International Nuclear Information System (INIS)

Lee, C. K.; Kim, D. H.; Kwon, K. C.; Joo, H. K.; Song, J. S.

2010-01-01

In the Korea Nuclear I and C Systems Development project the platforms for plant protection systems are developed, which function as a reactor shutdown, actuation of engineered safety features and a control of the related equipment. Those are fully digitalized through the use of safety-grade programmable logic controllers (PLCs) and few types of communication network. However the Regulatory Guide 1.152 (Rev. 02) was published by the U.S. NRC in 2006 and it recommended the application of a cyber security to the safety systems in the Nuclear Power Plant (NPP). Therefore to incorporate the new licensing requirement, a cyber security risk assessment is performed for the platforms. Then the vulnerabilities identified by the risk assessment are validated by penetration test. This paper summarizes test scenario, test results and their incorporation into system design
Use of feedback control to address flight safety issues

Science.gov (United States)

Ganguli, Subhabrata

This thesis addresses three control problems related to flight safety. The first problem relates to the scope of improvement in performance of conventional flight control laws. In particular, aircraft longitudinal axis control based on the Total Energy Control System (TECS) is studied. The research draws attention to a potentially sluggish and undesirable aircraft response when the engine dynamics is slow (typically the case). The proposed design method uses a theoretically well-developed modern design method based on Hinfinity optimization to improve the aircraft dynamic behavior in spite of slow engine characteristics. At the same time, the proposed design method achieves other desirable performance goals such as insensitivity to sensor noise and wind gust rejection: all addressed in one unified framework. The second problem is based on a system level analysis of control structure hierarchy for aircraft flight control. The objective of the analysis problem is to translate outer-loop stability and performance specifications into a comprehensive inner-loop metric. The prime motivation is to make the flight control design process more systematic and the system-integration reliable and independent of design methodology. The analysis problem is posed within the robust control analysis framework. Structured singular value techniques and free controller parameterization ideas are used to impose a hierarchical structure for flight control architecture. The third problem involves development and demonstration of a new reconfiguration strategy in the flight control architecture that has the potential of improving flight safety while keeping cost and complexity low. This research proposes a fault tolerant feature based on active robust reconfiguration. The fault tolerant control problem is formulated in the Linear Parameter Varying (LPV) design framework. A prime advantage of this approach is that the synthesis results in a single nonlinear controller (as opposed to a bank
Control of Nuclear Materials and Special Equipment (Nuclear Safety Regulations)

International Nuclear Information System (INIS)

Cizmek, A.; Prah, M.; Medakovic, S.; Ilijas, B.

2008-01-01

Based on Nuclear Safety Act (OG 173/03) the State Office for Nuclear Safety (SONS) in 2008 adopted beside Ordinance on performing nuclear activities (OG 74/06) and Ordinance on special conditions for individual activities to be performed by expert organizations which perform activities in the area of nuclear safety (OG 74/06) the new Ordinance on the control of nuclear material and special equipment (OG 15/08). Ordinance on the control of nuclear material and special equipment lays down the list of nuclear materials and special equipment as well as of nuclear activities covered by the system of control of production of special equipment and non-nuclear material, the procedure for notifying the intention to and filing the application for a license to carry out nuclear activities, and the format and contents of the forms for doing so. This Ordinance also lays down the manner in which nuclear material records have to be kept, the procedure for notifying the State administration organization (regulatory body) responsible for nuclear safety by the nuclear material user, and the keeping of registers of nuclear activities, nuclear material and special equipment by the State administration organization (regulatory body) responsible for nuclear safety, as well as the form and content of official nuclear safety inspector identification card and badge.(author)
Evaluating safety management system implementation

International Nuclear Information System (INIS)

Preuss, M.

2009-01-01

Canada is committed to not only maintaining, but also improving upon our record of having one of the safest aviation systems in the world. The development, implementation and maintenance of safety management systems is a significant step towards improving safety performance. Canada is considered a world leader in this area and we are fully engaged in implementation. By integrating risk management systems and business practices, the aviation industry stands to gain better safety performance with less regulatory intervention. These are important steps towards improving safety and enhancing the public's confidence in the safety of Canada's aviation system. (author)
Progress report: 1996 Radiation Safety Systems Division

International Nuclear Information System (INIS)

Bhagwat, A.M.; Sharma, D.N.; Abani, M.C.; Mehta, S.K.

1997-01-01

The activities of Radiation Safety Systems Division include (i) development of specialised monitoring systems and radiation safety information network, (ii) radiation hazards control at the nuclear fuel cycle facilities, the radioisotope programmes at Bhabha Atomic Research Centre (BARC) and for the accelerators programme at BARC and Centre for Advanced Technology (CAT), Indore. The systems on which development and upgradation work was carried out during the year included aerial gamma spectrometer, automated environment monitor using railway network, radioisotope package monitor and air monitors for tritium and alpha active aerosols. Other R and D efforts at the division included assessment of risk for radiation exposures and evaluation of ICRP 60 recommendations in the Indian context, shielding evaluation and dosimetry for the new upcoming accelerator facilities and solid state nuclear track detector techniques for neutron measurements. The expertise of the divisional members was provided for 36 safety committees of BARC and Atomic Energy Regulatory Board (AERB). Twenty three publications were brought out during the year 1996. (author)
System Design and the Safety Basis

International Nuclear Information System (INIS)

Ellingson, Darrel

2008-01-01

The objective of this paper is to present the Bechtel Jacobs Company, LLC (BJC) Lessons Learned for system design as it relates to safety basis documentation. BJC has had to reconcile incomplete or outdated system description information with current facility safety basis for a number of situations in recent months. This paper has relevance in multiple topical areas including documented safety analysis, decontamination and decommissioning (D and D), safety basis (SB) implementation, safety and design integration, potential inadequacy of the safety analysis (PISA), technical safety requirements (TSR), and unreviewed safety questions. BJC learned that nuclear safety compliance relies on adequate and well documented system design information. A number of PIS As and TSR violations occurred due to inadequate or erroneous system design information. As a corrective action, BJC assessed the occurrences caused by systems design-safety basis interface problems. Safety systems reviewed included the Molten Salt Reactor Experiment (MSRE) Fluorination System, K-1065 fire alarm system, and the K-25 Radiation Criticality Accident Alarm System. The conclusion was that an inadequate knowledge of system design could result in continuous non-compliance issues relating to nuclear safety. This was especially true with older facilities that lacked current as-built drawings coupled with the loss of 'historical knowledge' as personnel retired or moved on in their careers. Walkdown of systems and the updating of drawings are imperative for nuclear safety compliance. System design integration with safety basis has relevance in the Department of Energy (DOE) complex. This paper presents the BJC Lessons Learned in this area. It will be of benefit to DOE contractors that manage and operate an aging population of nuclear facilities
Full scale impact testing for environmental and safety control of energy material shipping container systems

International Nuclear Information System (INIS)

Seagren, R.D.

1978-01-01

Heavily-shielded energy material shipping systems, similar in size and weight to those presently employed to transport irradiated reactor fuel elements, are being destructively tested under dynamic conditions. In these tests, the outer and inner steel shells interact in a complex manner with the massive biological shielding in the system. Results obtained from these tests provide needed information for new design concepts. Containment failure (and the resulting release of radioactive material to the environment which might occur in an extremely severe accident) is most likely through the seals and other ancillary features of the shipping systems. Analyses and experiments provide engineering data on the behavior of these shipping systems under severe accident conditions and information for predicting potential survivability and environmental control with a rational margin of safety
Safety review for human factors engineering and control rooms of nuclear power plants

International Nuclear Information System (INIS)

Yang Mengzhuo

1998-01-01

Safety review for human factors engineering and control rooms of nuclear power plants (NPP) is in a forward position of science and technology, which began at American TMI severe accident and had been implemented in China. The importance and the significance of the safety review are expounded, the requirements of its scope and profundity are explained in detail. In addition, the situation of the technical document system for nuclear safety regulation on human factors engineering and control rooms of NPP in China is introduced briefly, on which the safety review is based
Improving safety margin of LWRs by rethinking the emergency core cooling system criteria and safety system capacity

Energy Technology Data Exchange (ETDEWEB)

Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Bokyung, E-mail: bkkim2@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

2016-10-15

Highlights: • Zircaloy embrittlement criteria can increase to 1370 °C for CP-ECR lower than 13%. • The draft ECCS criteria of U.S. NRC allow less than 5% in power margin. • The Japanese fracture-based criteria allow around 5% in power margin. • Increasing SIT inventory is effective in assuring safety margin for power uprates. - Abstract: This study investigates the engineering compatibility between emergency core cooling system criteria and safety water injection systems, in the pursuit of safety margin increase of light water reactors. This study proposes an acceptable temperature increase to 1370 °C as long as equivalent cladding reacted calculated by the Cathcart–Pawel equation is below 13%, after an extensive literature review. The influence of different ECCS criteria on the safety margin during large break loss of coolant accident is investigated for OPR-1000 by the system code MARS-KS, implemented with the KINS-REM method. The fracture-based emergency core cooling system (ECCS) criteria proposed in this study are shown to enable power margins up to 10%. In the meantime, the draft U.S. NRC’s embrittlement criteria (burnup-sensitive) and Japanese fracture-based criteria are shown to allow less than 5%, and around 5% of power margins, respectively. Increasing safety injection tank (SIT) water inventory is the key, yet convenient, way of assuring safety margin for power increase. More than 20% increase in the SIT water inventory is required to allow 15% power margins, for the U.S. NRC’s burnup-dependent embrittlement criteria. Controlling SIT water inventory would be a useful option that could allow the industrial desire to pursue power margins even under the recent atmosphere of imposing stricter ECCS criteria for the considerable burnup effects.
Tasks and structure of the WENDELSTEIN 7-X control system

International Nuclear Information System (INIS)

Schacht, Joerg; Niedermeyer, Helmut; Laqua, Heike; Spring, Anett; Mueller, Ina; Pingel, Steffen; Woelk, Andreas

2006-01-01

The super conducting stellarator WENDELSTEIN 7-X will run pulses of up to 30 min duration with full heating power. Short pulses with arbitrary intervals, steady state long discharges and arbitrary sequences of short phases with different characteristics in one discharge will be supported by the control system. Each technical component and each diagnostic system including its data acquisition will have its own control system permitting autonomous operation for commissioning and testing. During the experimental sessions the activity of these devices will be coordinated by a central control system and the machine runs more or less automatically with predefined programs. A session leader program allows the leader of the experiment to choose and chain predefined segments, to start or stop a segment chain as a discharge. The progress of the discharge is shown by a sequence monitor attached to the central sequence controller and the session leader program. W7-X has to be prepared for the experiment and monitored by means of the PLC based operational management system. A safety system working independently of the operational management consists of local units responsible for the safety of each component and a central unit ensuring the safety of the whole W7-X system. This safety system provides interlocks and controls the human access to the device. A safety analysis is the basis for the development of the safety system

Self-powered remotely controlled machines and tools for safety improvement in mining

Energy Technology Data Exchange (ETDEWEB)

Mirzaeva, G. [University of Newcastle, Callaghan, NSW (Australia)

2005-07-01

This paper addresses the problem of meeting the safety requirements of mining industry for implementation of control and monitoring equipment without external wiring. Local power generation and accumulation combined with remote control and wireless data transmission are suggested as an appropriate way to make the implementation of such device safe and convenient, which in its turn would facilitate their wider application for automation and safety improvement. A rope shovel dipper trip system is discussed in detail as an example of a self-powered remotely-controlled system. Other possible applications of the concept are also identified, such as Armoured Face Conveyor (AFC) and water jet drilling operation monitoring. 5 refs., 6 figs.
Safety of High Speed Ground Transportation Systems : Analytical Methodology for Safety Validation of Computer Controlled Subsystems : Volume 2. Development of a Safety Validation Methodology

Science.gov (United States)

1995-01-01

This report describes the development of a methodology designed to assure that a sufficiently high level of safety is achieved and maintained in computer-based systems which perform safety cortical functions in high-speed rail or magnetic levitation ...
HETDEX tracker control system design and implementation

Science.gov (United States)

Beno, Joseph H.; Hayes, Richard; Leck, Ron; Penney, Charles; Soukup, Ian

2012-09-01

To enable the Hobby-Eberly Telescope Dark Energy Experiment, The University of Texas at Austin Center for Electromechanics and McDonald Observatory developed a precision tracker and control system - an 18,000 kg robot to position a 3,100 kg payload within 10 microns of a desired dynamic track. Performance requirements to meet science needs and safety requirements that emerged from detailed Failure Modes and Effects Analysis resulted in a system of 13 precision controlled actuators and 100 additional analog and digital devices (primarily sensors and safety limit switches). Due to this complexity, demanding accuracy requirements, and stringent safety requirements, two independent control systems were developed. First, a versatile and easily configurable centralized control system that links with modeling and simulation tools during the hardware and software design process was deemed essential for normal operation including motion control. A second, parallel, control system, the Hardware Fault Controller (HFC) provides independent monitoring and fault control through a dedicated microcontroller to force a safe, controlled shutdown of the entire system in the event a fault is detected. Motion controls were developed in a Matlab-Simulink simulation environment, and coupled with dSPACE controller hardware. The dSPACE real-time operating system collects sensor information; motor commands are transmitted over a PROFIBUS network to servo amplifiers and drive motor status is received over the same network. To interface the dSPACE controller directly to absolute Heidenhain sensors with EnDat 2.2 protocol, a custom communication board was developed. This paper covers details of operational control software, the HFC, algorithms, tuning, debugging, testing, and lessons learned.
THE FORMATION OF THE CONTOUR OF THE DOCUMENTED AND REAL FLIGHT SAFETY IN THE SYSTEM OF THE INFORMATION PROVISION OF SAFETY OF FLIGHTS

Directory of Open Access Journals (Sweden)

B. I. Bachkalo

2015-01-01

Full Text Available The article discusses the principles and mechanisms of formation of the contour of the real safety of flights and contour of the documented safety, allowing us to obtain information to control fligh safety. The proposed approach can be used in the algorithms of active on-board flight safety management system for the implementation of information support to the crew in flight and automatic control of flight safety.
Safety aspect of digital reactor protection system in Japan

International Nuclear Information System (INIS)

Ogiso, Zen-Ichi

1998-01-01

It was early in 1980's that the digital controllers were first applied to nuclear power plant in japan. After that, their application area had been expanding gradually, reaching to the overall integrated digital system including the safety system in Kashiwazaki-Kariwa units 6 and 7. The software for computer-based systems has been produced using the graphical language ''POL'' in Japanese nuclear power plants. It is the fundamental principle that the reliability of the software should be assured through the properly managed quality assurance. The POL-based system is fitted to this principle. In applying POL-based systems to safety system, the MITI, Ministry of International Trade and Industry, identified the licensing issues as the regulatory body, while the utilities had developed the digital technology feasible to the safety application. Through the activities, a specific industrial design guide for the software important to safety was established and the adequacy of the technology was certified through the demonstration tests of the integrated system. In the safety examination of the digital reactor protection system of K-6/7, the application of POL were approved. The POL-based systems in nuclear power plants were successful design and production process of the POL-based systems. This paper describes the activities in licensing and maintaining the computer-based systems by the utilities and manufacturers as well as the MITI. (author)
Safety critical systems handbook a straightforward guide to functional safety : IEC 61508 (2010 edition) and related standards

CERN Document Server

Smith, David J

2010-01-01

Electrical, electronic and programmable electronic systems increasingly carry out safety functions to guard workers and the public against injury or death and the environment against pollution. The international functional safety standard IEC 61508 was revised in 2010, and this is the first comprehensive guide available to the revised standard. As functional safety is applicable to many industries, this book will have a wide readership beyond the chemical and process sector, including oil and gas, power generation, nuclear, aircraft, and automotive industries, plus project, instrumentation, design, and control engineers. * The only comprehensive guide to IEC 61508, updated to cover the 2010 amendments, that will ensure engineers are compliant with the latest process safety systems design and operation standards* Helps readers understand the process required to apply safety critical systems standards* Real-world approach helps users to interpret the standard, with case studies and best practice design examples...
RICIS Symposium 1992: Mission and Safety Critical Systems Research and Applications

Science.gov (United States)

1992-01-01

This conference deals with computer systems which control systems whose failure to operate correctly could produce the loss of life and or property, mission and safety critical systems. Topics covered are: the work of standards groups, computer systems design and architecture, software reliability, process control systems, knowledge based expert systems, and computer and telecommunication protocols.
SYSTEMS SAFETY ANALYSIS FOR FIRE EVENTS ASSOCIATED WITH THE ECRB CROSS DRIFT

International Nuclear Information System (INIS)

R. J. Garrett

2001-01-01

The purpose of this analysis is to systematically identify and evaluate fire hazards related to the Yucca Mountain Site Characterization Project (YMP) Enhanced Characterization of the Repository Block (ECRB) East-West Cross Drift (commonly referred to as the ECRB Cross-Drift). This analysis builds upon prior Exploratory Studies Facility (ESF) System Safety Analyses and incorporates Topopah Springs (TS) Main Drift fire scenarios and ECRB Cross-Drift fire scenarios. Accident scenarios involving the fires in the Main Drift and the ECRB Cross-Drift were previously evaluated in ''Topopah Springs Main Drift System Safety Analysis'' (CRWMS M and O 1995) and the ''Yucca Mountain Site Characterization Project East-West Drift System Safety Analysis'' (CRWMS M and O 1998). In addition to listing required mitigation/control features, this analysis identifies the potential need for procedures and training as part of defense-in-depth mitigation/control features. The inclusion of this information in the System Safety Analysis (SSA) is intended to assist the organization(s) (e.g., Construction, Environmental Safety and Health, Design) responsible for these aspects of the ECRB Cross-Drift in developing mitigation/control features for fire events, including Emergency Refuge Station(s). This SSA was prepared, in part, in response to Condition/Issue Identification and Reporting/Resolution System (CIRS) item 1966. The SSA is an integral part of the systems engineering process, whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach is used which incorporates operating experiences and recommendations from vendors, the constructor and the operating contractor. The risk assessment in this analysis characterizes the scenarios associated with fires in terms of relative risk and includes recommendations for mitigating all identified hazards. The priority for recommending and implementing mitigation control features is: (1) Incorporate
Design of reactor containment systems for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2008-01-01

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems
Design of reactor containment systems for nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems
Evaluating the effectiveness of active vehicle safety systems.

Science.gov (United States)

Jeong, Eunbi; Oh, Cheol

2017-03-01

Advanced vehicle safety systems have been widely introduced in transportation systems and are expected to enhance traffic safety. However, these technologies mainly focus on assisting individual vehicles that are equipped with them, and less effort has been made to identify the effect of vehicular technologies on the traffic stream. This study proposed a methodology to assess the effectiveness of active vehicle safety systems (AVSSs), which represent a promising technology to prevent traffic crashes and mitigate injury severity. The proposed AVSS consists of longitudinal and lateral vehicle control systems, which corresponds to the Level 2 vehicle automation presented by the National Highway Safety Administration (NHTSA). The effectiveness evaluation for the proposed technology was conducted in terms of crash potential reduction and congestion mitigation. A microscopic traffic simulator, VISSIM, was used to simulate freeway traffic stream and collect vehicle-maneuvering data. In addition, an external application program interface, VISSIM's COM-interface, was used to implement the AVSS. A surrogate safety assessment model (SSAM) was used to derive indirect safety measures to evaluate the effectiveness of the AVSS. A 16.7-km freeway stretch between the Nakdong and Seonsan interchanges on Korean freeway 45 was selected for the simulation experiments to evaluate the effectiveness of AVSS. A total of five simulation runs for each evaluation scenario were conducted. For the non-incident conditions, the rear-end and lane-change conflicts were reduced by 78.8% and 17.3%, respectively, under the level of service (LOS) D traffic conditions. In addition, the average delay was reduced by 55.5%. However, the system's effectiveness was weakened in the LOS A-C categories. Under incident traffic conditions, the number of rear-end conflicts was reduced by approximately 9.7%. Vehicle delays were reduced by approximately 43.9% with 100% of market penetration rate (MPR). These results
Safety critical application of fuzzy control

International Nuclear Information System (INIS)

Schildt, G.H.

1995-01-01

After an introduction into safety terms a short description of fuzzy logic will be given. Especially, for safety critical applications of fuzzy controllers a possible controller structure will be described. The following items will be discussed: Configuration of fuzzy controllers, design aspects like fuzzfiication, inference strategies, defuzzification and types of membership functions. As an example a typical fuzzy rule set will be presented. Especially, real-time behaviour a fuzzy controllers is mentioned. An example of fuzzy controlling for temperature control purpose within a nuclear reactor together with membership functions and inference strategy of such a fuzzy controller will be presented. (author). 4 refs, 17 figs
Safety assessment of computerized instrumentation and control for nuclear power plants

International Nuclear Information System (INIS)

Fride, B.; Henry, J.Y.; Manners, S.

1996-01-01

France's latest 1400 MWe 'N4' generation of Pressurised Water Reactors (PWR) use distributed programmable control systems interconnected by data networks. The protection system is also software based. IPSN have the task of evaluating the safety demonstration before the government safety authority (DSIN) give the licensee (EDF) permission to fuel the reactor and to raise power. Some of the different aspects of the evaluation carried out and the methodologies used for assessing the C and I are presented. (author)
The management system for the disposal of radioactive waste. Safety guide

International Nuclear Information System (INIS)

2008-01-01

The objective of this Safety Guide is to provide recommendations on developing and implementing management systems for all phases of facilities for the disposal of radioactive waste and related activities. It covers the management systems for managing the different stages of waste disposal facilities, such as siting, design and construction, operation (i.e. the activities, which can extend over several decades, involving receipt of the waste product in its final packaging (if it is to be disposed of in packaged form), waste emplacement in the waste disposal facility, backfilling and sealing, and any subsequent period prior to closure), closure and the period of institutional control (i.e. either active control - monitoring, surveillance and remediation; or passive control - restricted land use). The management systems apply to various types of disposal facility for different categories of radioactive waste, such as: near surface (for low level waste), geological (for low, intermediate and/or high level waste), boreholes (for sealed sources), surface impoundment (for mining and milling waste) and landfill (for very low level waste). It also covers management systems for related processes and activities, such as extended monitoring and surveillance during the period of active institutional control in the post-closure phase, safety and performance assessments and development of the safety case for the waste disposal facility and regulatory authorization (e.g. licensing). This Safety Guide is intended to be used by organizations that are directly involved in, or that regulate, the facilities and activities described in paras 1.15 and 1.16, and by the suppliers of nuclear safety related products that are required to meet some or all of the requirements established in IAEA Safety Standards Series No. GS-R-3 'The Management System for Facilities and Activities'. It will also be useful to legislators and to members of the public and other parties interested in the nuclear
The increased use of radiation requires enhanced activities regarding radiation safety control

International Nuclear Information System (INIS)

Lee, Yun Jong; Lee, Jin Woo; Jeong, Gyo Seong

2015-01-01

More recently, companies that have obtained permission to use radioactive materials or radiation device and registered radiation workers have increased by 10% and 4% respectively. The increased use of radiation could have an effect on radiation safety control. However, there is not nearly enough manpower and budget compared to the number of workers and facilities. This paper will suggest a counteroffer thought analyzing pending issues. The results of this paper indicate that there are 47 and 31.3 workers per radiation protection officer in educational and research institutes, respectively. There are 20.1 persons per RPO in hospitals, even though there are 2 RPOs appointed. Those with a special license as a radioisotope handler were ruled out as possible managers because medical doctors who have a special license for radioisotope handling normally have no experience with radiation safety. The number of staff members and budget have been insufficient for safety control at most educational and research institutes. It is necessary to build an optimized safety control system for effective Radiation Safety Control. This will reduce the risk factor of safety, and a few RPOs can be supplied for efficiency and convenience
The increased use of radiation requires enhanced activities regarding radiation safety control

Energy Technology Data Exchange (ETDEWEB)

Lee, Yun Jong; Lee, Jin Woo; Jeong, Gyo Seong [Korea Atomic Energy Research Institute, Jeongeup (Korea, Republic of)

2015-05-15

More recently, companies that have obtained permission to use radioactive materials or radiation device and registered radiation workers have increased by 10% and 4% respectively. The increased use of radiation could have an effect on radiation safety control. However, there is not nearly enough manpower and budget compared to the number of workers and facilities. This paper will suggest a counteroffer thought analyzing pending issues. The results of this paper indicate that there are 47 and 31.3 workers per radiation protection officer in educational and research institutes, respectively. There are 20.1 persons per RPO in hospitals, even though there are 2 RPOs appointed. Those with a special license as a radioisotope handler were ruled out as possible managers because medical doctors who have a special license for radioisotope handling normally have no experience with radiation safety. The number of staff members and budget have been insufficient for safety control at most educational and research institutes. It is necessary to build an optimized safety control system for effective Radiation Safety Control. This will reduce the risk factor of safety, and a few RPOs can be supplied for efficiency and convenience.
Passive components of NPP safety-related systems

International Nuclear Information System (INIS)

Ionaytis Romuald, R.; Bubnova Tatyana, A.

2005-01-01

This paper presents a new passive components with having drives: fast-response cutoff valves; modular actuators with opposite cocking pneumatic drives and actuation spring drives; voting electromagnetic valve units for control of pneumatic drives; passive initiators of actuation; visual diagnostics . All these devices have been developed and tested at mock-ups. This paper presents also the following direct-action passive safety components: modular pressure-relief safety valves; pilot safety valves with passive action; check valves with remote position indicator and after-tightening; modular inserts for limiting emergency coolant flow; vortex rectifier; critical weld fasteners; gas-liquid valves; fast-removable seal assembly; seal spring loaders; grooves for increasing hydraulic resistance. Replacement of active safety system components for passive ones improves the general reliability NPP by 1.5 or 2 orders of magnitudes. (authors)
The impact of the instrumentation and control systems in the safety of a nuclear plant: a general vision; El impacto de los sistemas de instrumentacion y control en la seguridad de una planta nuclear: una vision general

Energy Technology Data Exchange (ETDEWEB)

Celis del Angel, L.; Rivero, T., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2011-11-15

One of the fundamental components so much for the sure operation, like in emergency cases or accident are the equipment s and instrumentation and control systems. The nuclear industry has had some accidents where the instrumentation and control have played and important part: a wrong design, instrumentation lack, faulty systems of safety, etc. At the present time the necessity to modernize the instrumentation and control in a nuclear power plant is before the challenge of finding innovative forms to improve the competitiveness and readiness, reducing operation costs without put ing in risk the safety and reliability of the nuclear power plant. Most of the nuclear power plants require actualizing their instrumentation and control systems, here the digital systems represent a great alternative, improving the performance and the safety, increasing the readiness and reducing the maintenance s. However they require of strict tests that allow assuring their application in critical systems. It is also necessary, the development of modernization programs that allow the programmed substitution of the systems without affecting the readiness of the nuclear power plants. During this whole modernization process will be necessary to put special attention in the cyber-safety because the attacks every time they are more elaborated. Therefore will be necessary to go toward the modernization of the instrumentation and control with the challenge of making without detriment some in the safety of the normal operation and with response reliability in emergency conditions or accident that which represents an effort that should not be postponed in the case of the nuclear power plant of Laguna Verde. (Author)
Nitric Acid Revamp and Upgrading of the Alarm & Protection Safety System at Petrokemija, Croatia

Directory of Open Access Journals (Sweden)

Hoško, I.

2012-04-01

Full Text Available Every industrial production, particularly chemical processing, demands special attention in conducting the technological process with regard to the security requirements. For this reason, production processes should be continuously monitored by means of control and alarm safety instrumented systems. In the production of nitric acid at Petrokemija d. d., the original alarm safety system was designed as a combination of an electrical relay safety system and transistorized alarm module system. In order to increase safety requirements and modernize the technological process of nitric acid production, revamping and upgrading of the existing alarm safety system was initiated with a new microprocessor system. The newly derived alarm safety system, Simatic PCS 7, links the function of "classically" distributed control (DCS and logical systems in a common hardware and software platform with integrated engineering tools and operator interface to meet the minimum safety standards with safety integrity level 2 (SIL2 up to level 3 (SIL3, according to IEC 61508 and IEC 61511. This professional paper demonstrates the methodology of upgrading the logic of the alarm safety system in the production of nitric acid in the form of a logical diagram, which was the basis for a further step in its design and construction. Based on the mentioned logical diagram and defined security requirements, the project was implemented in three phases: analysis and testing, installation of the safety equipment and system, and commissioning. Developed also was a verification system of all safety conditions, which could be applied to other facilities for production of nitric acid. With the revamped and upgraded interlock alarm safety system, a new and improved safety boundary in the production of nitric acid was set, which created the foundation for further improvement of the production process in terms of improved analysis.
System architecture of Detector Control and safety for the ATLAS Inner Detector Upgrade

International Nuclear Information System (INIS)

Ferrere, D.; Kersten, S.

2011-01-01

In the current ATLAS Upgrade plan a new Inner Detector (ID) based upon silicon sensor technology is being considered. The operational monitoring and control of the ID will be very demanding. The Detector Control System (DCS) is a common tool that is essential for the operational safety of a system. Even at this early stage the DCS system architecture has to be defined such that it is well integrated and optimized for its later implementation and use. For example the DCS diagnostics for the front-end (FE) chips is a serious option being considered that needs an early requirement and specification definition. In addition one of the main constraints is the service reuse between the service patch panels of the ATLAS ID and the counting room that limits the number of electrical lines to be reused. Conceptual differences in terms of readout architecture and layout have been identified between the strip and the pixel detector that lead to two distinct architectures. Nevertheless, the limitation of available electrical lines going to the counting room as well as the low material budget requirements inside the ID volume are two major constraints that lead the ID to consider an on-detector radiation hard integrated circuitry for the slow control. At this stage of the project, the definitions of the logical actions and protocol for the ADCs of such a chip are still being specified. In addition the experience gained from the current ID will be essential for the guidance of tuning the future DCS architecture in the coming years.

Radiation safety system (RSS) backbones: Design, engineering, fabrication and installation

International Nuclear Information System (INIS)

Wilmarth, J.E.; Sturrock, J.C.; Gallegos, F.R.

1998-01-01

The Radiation Safety System (RSS) Backbones are part of an electrical/electronic/mechanical system insuring safe access and exclusion of personnel to areas at the Los Alamos Neutron Science Center (LANSCE) accelerator. The RSS Backbones control the safety fusible beam plugs which terminate transmission of accelerated ion beams in response to predefined conditions. Any beam or access fault of the backbone inputs will cause insertion of the beam plugs in the low energy beam transport. The Backbones serve the function of tying the beam plugs to the access control systems, beam spill monitoring systems and current-level limiting systems. In some ways the Backbones may be thought of as a spinal column with beam plugs at the head and nerve centers along the spinal column. The two Linac Backbone segments and experimental area segments form a continuous cable plant over 3,500 feet from beam plugs to the tip on the longest tail. The Backbones were installed in compliance with current safety standards, such as installation of the two segments in separate conduits or tray. Monitoring for ground-faults and input wiring verification was an added enhancement to the system. The system has the capability to be tested remotely
Nuclear power plant systems, structures and components and their safety classification

International Nuclear Information System (INIS)

2000-01-01

The assurance of a nuclear power plant's safety is based on the reliable functioning of the plant as well as on its appropriate maintenance and operation. To ensure the reliability of operation, special attention shall be paid to the design, manufacturing, commissioning and operation of the plant and its components. To control these functions the nuclear power plant is divided into structural and functional entities, i.e. systems. A systems safety class is determined by its safety significance. Safety class specifies the procedures to be employed in plant design, construction, monitoring and operation. The classification document contains all documentation related to the classification of the nuclear power plant. The principles of safety classification and the procedures pertaining to the classification document are presented in this guide. In the Appendix of the guide, examples of systems most typical of each safety class are given to clarify the safety classification principles
Reconstruction of instrumentation and control system (SKR)

International Nuclear Information System (INIS)

Wiening, K.-H.

2001-01-01

For the first time extensive upgrades have been performed in all safety related areas of units with WWER 440/230 reactors. One of the most important actions was the replacement of the safety and safety related instrumentation and control. The state of the art digital safety instrumentation and control system TELEPERM XS has been implemented in units 1 and 2 of the Bohunice V1 power plant. The requirements as deduced from safety assessments conducted by commissions of international experts have been fulfilled, so that Bohunice V1 after this gradual reconstruction has been upgraded to an internationally accepted safety level for the remainder of its service life. (author)
A Novel Control Algorithm for Integration of Active and Passive Vehicle Safety Systems in Frontal Collisions

Directory of Open Access Journals (Sweden)

Daniel Wallner

2010-10-01

Full Text Available The present paper investigates an approach to integrate active and passive safety systems of passenger cars. Worldwide, the introduction of Integrated Safety Systems and Advanced Driver Assistance Systems (ADAS is considered to continue the today
Automated hazard analysis of digital control systems

International Nuclear Information System (INIS)

Garrett, Chris J.; Apostolakis, George E.

2002-01-01

Digital instrumentation and control (I and C) systems can provide important benefits in many safety-critical applications, but they can also introduce potential new failure modes that can affect safety. Unlike electro-mechanical systems, whose failure modes are fairly well understood and which can often be built to fail in a particular way, software errors are very unpredictable. There is virtually no nontrivial software that will function as expected under all conditions. Consequently, there is a great deal of concern about whether there is a sufficient basis on which to resolve questions about safety. In this paper, an approach for validating the safety requirements of digital I and C systems is developed which uses the Dynamic Flowgraph Methodology to conduct automated hazard analyses. The prime implicants of these analyses can be used to identify unknown system hazards, prioritize the disposition of known system hazards, and guide lower-level design decisions to either eliminate or mitigate known hazards. In a case study involving a space-based reactor control system, the method succeeded in identifying an unknown failure mechanism
Digital control system of advanced reactor

International Nuclear Information System (INIS)

Peng Huaqing; Zhang Rui; Liu Lixin

2001-01-01

This article produced the Digital Control System For Advanced Reactor made by NPIC. This system uses Siemens SIMATIC PCS 7 process control system and includes five control system: reactor power control system, pressurizer level control system, pressurizer pressure control system, steam generator water level control system and dump control system. This system uses three automatic station to realize the function of five control system. Because the safety requisition of reactor is very strict, the system is redundant. The system configuration uses CFC and SCL. the human-machine interface is configured by Wincc. Finally the system passed the test of simulation by using RETRAN 02 to simulate the control object. The research solved the key technology of digital control system of reactor and will be very helpful for the nationalization of digital reactor control system
Improved safety in advanced control complexes, without side effects

International Nuclear Information System (INIS)

Harmon, D.L.

1997-01-01

If we only look for a moment at the world around us, it is obvious that advances in digital electronic equipment and Human-System Interface (HSI) technology are occurring at a phenomenal pace. This is evidenced from our home entertainment systems to the dashboard and computer-based operation of our new cars. Though the nuclear industry has less vigorously embraced these advances, their application is being implemented through individual upgrades to current generation nuclear plants and as plant-wide control complexes for advanced plants. In both venues modem technology possesses widely touted advantages for improving plant availability as well as safety. The well-documented safety benefits of digital Instrumentation and Controls (I ampersand C) include higher reliability resulting from redundancy and fault tolerance, inherent self-test and self-diagnostic capabilities which have replaced error-prone human tasks, resistance to setpoint drift increasing available operating margins, and the ability to run complex, real-time, computer-based algorithms directly supporting an operator's monitoring and control task requirements. 22 refs., 3 figs., 5 tabs
Role of computers in CANDU safety systems

International Nuclear Information System (INIS)

Hepburn, G.A.; Gilbert, R.S.; Ichiyen, N.M.

1985-01-01

Small digital computers are playing an expanding role in the safety systems of CANDU nuclear generating stations, both as active components in the trip logic, and as monitoring and testing systems. The paper describes three recent applications: (i) A programmable controller was retro-fitted to Bruce ''A'' Nuclear Generating Station to handle trip setpoint modification as a function of booster rod insertion. (ii) A centralized monitoring computer to monitor both shutdown systems and the Emergency Coolant Injection system, is currently being retro-fitted to Bruce ''A''. (iii) The implementation of process trips on the CANDU 600 design using microcomputers. While not truly a retrofit, this feature was added very late in the design cycle to increase the margin against spurious trips, and has now seen about 4 unit-years of service at three separate sites. Committed future applications of computers in special safety systems are also described. (author)
Safeguarding the functions and performance of instrumentation and control systems

International Nuclear Information System (INIS)

Koehler, M.; Schoerner, O.

1996-01-01

Based on an analysis of the existing nuclear power plant control technology, the necessity of providing in the medium-term advanced and future-oriented, digital control system, both for normal operation and for safety-relevant tasks of the reactor and safety control systems. Siemens KWU has been promoting the development, review and marketing of the digital instrumentation and control systems called TELEPERM XS and TELEPERM XP in addition to the measures taken for safeguarding the functions of existing, wired systems. The paper briefly explains the performance and advantages of digital systems and the progress in approval and pioneering of the TELEPERM XS safety control system. Many examples discussed show the diversity of applications of the systems both in new reactor plants and as retrofitting measures, for KWU power plants and those of other manufacturers. (orig.) [de
Safety assessment of HLW geological disposal system

International Nuclear Information System (INIS)

Naito, Morimasa

2006-01-01

In accordance with the Japanese nuclear program, the liquid waste with a high level of radioactivity arising from reprocessing is solidified in a stable glass matrix (vitrification) in stainless steel fabrication containers. The vitrified waste is referred to as high-level radioactive waste (HLW), and is characterized by very high initial radioactivity which, even though it decreases with time, presents a potential long-term risk. It is therefore necessary to thoroughly manage HLW from human and his environment. After vitrification, HLW is stored for a period of 30 to 50 years to allow cooling, and finally disposed of in a stable geological environment at depths greater than 300 m below surface. The deep underground environment, in general, is considered to be stable over geological timescales compared with surface environment. By selecting an appropriate disposal site, therefore, it is considered to be feasible to isolate the waste in the repository from man and his environment until such time as radioactivity levels have decayed to insignificance. The concept of geological disposal in Japan is similar to that in other countries, being based on a multibarrier system which combines the natural geological environment with engineered barriers. It should be noted that geological disposal concept is based on a passive safety system that does not require any institutional control for assuring long term environmental safety. To demonstrate feasibility of safe HLW repository concept in Japan, following technical steps are essential. Selection of a geological environment which is sufficiently stable for disposal (site selection). Design and installation of the engineered barrier system in a stable geological environment (engineering measures). Confirmation of the safety of the constructed geological disposal system (safety assessment). For site selection, particular consideration is given to the long-term stability of the geological environment taking into account the fact
Liquefied Gaseous Fuels Safety and Environmental Control Assessment Program: second status report

Energy Technology Data Exchange (ETDEWEB)

1980-10-01

The Assistant Secretary for Environment has responsibility for identifying, characterizing, and ameliorating the environmental, health, and safety issues and public concerns associated with commercial operation of specific energy systems. The need for developing a safety and environmental control assessment for liquefied gaseous fuels was identified by the Environmental and Safety Engineering Division as a result of discussions with various governmental, industry, and academic persons having expertise with respect to the particular materials involved: liquefied natural gas, liquefied petroleum gas, hydrogen, and anhydrous ammonia. This document is arranged in three volumes and reports on progress in the Liquefied Gaseous Fuels (LGF) Safety and Environmental Control Assessment Program made in Fiscal Year (FY)-1979 and early FY-1980. Volume 1 (Executive Summary) describes the background, purpose and organization of the LGF Program and contains summaries of the 25 reports presented in Volumes 2 and 3. Annotated bibliographies on Liquefied Natural Gas (LNG) Safety and Environmental Control Research and on Fire Safety and Hazards of Liquefied Petroleum Gas (LPG) are included in Volume 1.
MAPLE-X10 reactor digital control system

International Nuclear Information System (INIS)

Deverno, M.T.; Hinds, H.W.

1991-10-01

The MAPLE-X10 reactor, currently under construction at the Chalk River Laboratories of Atomic Energy of Canada Limited, is a 10 MW t , pool-type, light-water reactor. It will be used for radioisotope production and silicon neutron transmutation doping. The reactor is controlled by a Digital Control System (DCS) and protected against abnormal process events by two independent safety systems. The DCS is an integrated control system used to regulate the reactor power and process systems. The safety philosophy for the control system is to minimize unsafe events arising from system failures and operational errors. this is achieved through redundancy, fail-safe design, automatic fault detection, and the selection of highly reliable components. The DCS provides both computer-controlled reactor regulation from the shutdown state to full power and automated reactor shutdown if safe limits are exceeded or critical sensors malfunction. The use of commercially available hardware with enhanced quality assurance makes the system cost effective while providing a high degree of reliability
Safety design integrated in the Building Delivery System

DEFF Research Database (Denmark)

Jørgensen, Kirsten

2012-01-01

phases of the building delivery system by using the principle of the lean construction modelling. The method for the research was to go through the lean construction building delivery system step by step and create a normative description of what to do, when to do and how to do to fully integration...... of safety in each process. The group of participants who created the description had a high experience in a combination of research, safety and health in general and especial in construction and knowledge of the lean construction processes both from the clients perspective as well as from the designers...... and the consultants. The result is a concept and guideline including control schemes for how to integrate safety design in the lean construction building delivery system including what to do and when. The concept has been tested in an educational context and found useful by the designers. The practical value...
A holistic approach to control process safety risks: Possible ways forward

International Nuclear Information System (INIS)

Pasman, H.J.; Knegtering, B.; Rogers, W.J.

2013-01-01

system, the process industry is monitoring safety performance indicators. The critical intensity level upon which management must be alarmed is less simple. Risk assessment may be improved, made dynamic, and be a tool of process control by taking account of short-term risk fluctuations based on sensor signals and the influence of human factors with its long-term changes via indicators. Bayesian network can provide the infrastructure. The paper will describe various complexities when applying a holistic control of safety to a process plant in general, and it will more specifically focus on safeguarding measures such as barriers and other controls with some examples. -- Highlights: • Complexity of process installations makes risk control of a process challenging. • Erosive drift by cost pressure and efficiency increase may undermine safety level. • Resilience engineering in socio-psychological context analyzed this successfully. • There is prospect too to develop the technical side of process safety resilience. • Process safety performance indicator information may help to establish risk level
Safety assessment of computerized instrumentation and control for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Fride, B.; Henry, J.Y.; Manners, S.

1996-12-31

France`s latest 1400 MWe `N4` generation of Pressurised Water Reactors (PWR) use distributed programmable control systems interconnected by data networks. The protection system is also software based. IPSN have the task of evaluating the safety demonstration before the government safety authority (DSIN) give the licensee (EDF) permission to fuel the reactor and to raise power. Some of the different aspects of the evaluation carried out and the methodologies used for assessing the C and I are presented. (author). 3 refs.
Job Demands-Control-Support model and employee safety performance.

Science.gov (United States)

Turner, Nick; Stride, Chris B; Carter, Angela J; McCaughey, Deirdre; Carroll, Anthony E

2012-03-01

The aim of this study was to explore whether work characteristics (job demands, job control, social support) comprising Karasek and Theorell's (1990) Job Demands-Control-Support framework predict employee safety performance (safety compliance and safety participation; Neal and Griffin, 2006). We used cross-sectional data of self-reported work characteristics and employee safety performance from 280 healthcare staff (doctors, nurses, and administrative staff) from Emergency Departments of seven hospitals in the United Kingdom. We analyzed these data using a structural equation model that simultaneously regressed safety compliance and safety participation on the main effects of each of the aforementioned work characteristics, their two-way interactions, and the three-way interaction among them, while controlling for demographic, occupational, and organizational characteristics. Social support was positively related to safety compliance, and both job control and the two-way interaction between job control and social support were positively related to safety participation. How work design is related to employee safety performance remains an important area for research and provides insight into how organizations can improve workplace safety. The current findings emphasize the importance of the co-worker in promoting both safety compliance and safety participation. Crown Copyright © 2011. Published by Elsevier Ltd. All rights reserved.
Safety monitoring in process and control

International Nuclear Information System (INIS)

Esparza, V. Jr.; Sebo, D.E.

1984-01-01

Safety Functions provide a method of ensuring the safe operation of any large-scale processing plant. Successful implementation of safety functions requires continuous monitoring of safety function values and trends. Because the volume of information handled by a plant operator occassionally can become overwhelming, attention may be diverted from the primary concern of maintaining plant safety. With this in mind EG and G, Idaho developed various methods and techniques for use in a computerized Safety Function Monitoring System and tested the application of these techniques using a simulated nuclear power plant, the Loss-of-Fluid Test Facility (LOFT) at the Idaho National Engineering Laboratory (INEL). This paper presents the methods used in the development of a Safety Function Monitoring System
Safety concepts and their implications with respect to systems, instrumentation (automatic) control and hardware

International Nuclear Information System (INIS)

Paziaud, A.; Walther, M.

1982-01-01

This overview of instrumentation and control in the French Nuclear Power Plants sets out the importance of safety requirements. As a matter of fact, the amount of equipment increases proportionally to the increase in safety requirements, resulting in higher costs in spite of the decrease in the prices of each component owing to the advance in electronics. However the improved reliability should improve the plant capacity factor and, as a consequence, improve both the power output and the safety which is often endangered by minor failures starting severe accidents. (orig.)
Health management and controls for Earth-to-orbit propulsion systems

Science.gov (United States)

Bickford, R. L.

1995-03-01

Avionics and health management technologies increase the safety and reliability while decreasing the overall cost for Earth-to-orbit (ETO) propulsion systems. New ETO propulsion systems will depend on highly reliable fault tolerant flight avionics, advanced sensing systems and artificial intelligence aided software to ensure critical control, safety and maintenance requirements are met in a cost effective manner. Propulsion avionics consist of the engine controller, actuators, sensors, software and ground support elements. In addition to control and safety functions, these elements perform system monitoring for health management. Health management is enhanced by advanced sensing systems and algorithms which provide automated fault detection and enable adaptive control and/or maintenance approaches. Aerojet is developing advanced fault tolerant rocket engine controllers which provide very high levels of reliability. Smart sensors and software systems which significantly enhance fault coverage and enable automated operations are also under development. Smart sensing systems, such as flight capable plume spectrometers, have reached maturity in ground-based applications and are suitable for bridging to flight. Software to detect failed sensors has reached similar maturity. This paper will discuss fault detection and isolation for advanced rocket engine controllers as well as examples of advanced sensing systems and software which significantly improve component failure detection for engine system safety and health management.
Design concepts for a nuclear digital instrumentation and control system platform

International Nuclear Information System (INIS)

Ou, T. C.; Chen, C. K.; Chen, P. J.; Shyu, S. S.; Lee, C. L.; Hsieh, S. F.

2010-10-01

The objective of this paper is to present the development results of the nuclear instrumentation and control system in Taiwan. As the Taiwan nuclear power plants age, the need to consider upgrading of both their safety and non-safety-related instrumentation and control systems becomes more urgent. Meanwhile, the digital instrumentation and control system that is based on current fast evolving electronic and information technologies are difficult to maintain effectively. Therefore, Institute of Nuclear Energy Research was made a decision to promote the Taiwan Nuclear Instrumentation and Control System project to collaborate with domestic electronic industry to establish self-reliant capabilities on the design, manufacturing, and application of nuclear instrumentation and control systems with newer technology. In the case of safety-related applications like nuclear instrumentation and control, safety-oriented quality control is required. In order to establish a generic qualified digital platform, the world-wide licensing experience should be considered in the licensing process. This paper describes the qualification and certification tools by IEC 61508 for design and development of safety related equipment and explains the basis for many decisions made while performing the digital upgrade. (Author)

How could intelligent safety transport systems enhance safety ?

NARCIS (Netherlands)

Wiethoff, M. Heijer, T. & Bekiaris, E.

2017-01-01

In Europe, many deaths and injured each years are the cost of today's road traffic. Therefore, it is wise to look for possible solutions for enhancing traffic safety. Some Advanced Driver Assistance Systems (ADAS) are expected to increase safety, but they may also evoke new safety hazards. Only
Superconducting Coil Winding Machine Control System

Energy Technology Data Exchange (ETDEWEB)

Nogiec, J. M. [Fermilab; Kotelnikov, S. [Fermilab; Makulski, A. [Fermilab; Walbridge, D. [Fermilab; Trombly-Freytag, K. [Fermilab

2016-10-05

The Spirex coil winding machine is used at Fermilab to build coils for superconducting magnets. Recently this ma-chine was equipped with a new control system, which al-lows operation from both a computer and a portable remote control unit. This control system is distributed between three layers, implemented on a PC, real-time target, and FPGA, providing respectively HMI, operational logic and direct controls. The system controls motion of all mechan-ical components and regulates the cable tension. Safety is ensured by a failsafe, redundant system.
Research on Occupational Safety, Health Management and Risk Control Technology in Coal Mines.

Science.gov (United States)

Zhou, Lu-Jie; Cao, Qing-Gui; Yu, Kai; Wang, Lin-Lin; Wang, Hai-Bin

2018-04-26

This paper studies the occupational safety and health management methods as well as risk control technology associated with the coal mining industry, including daily management of occupational safety and health, identification and assessment of risks, early warning and dynamic monitoring of risks, etc.; also, a B/S mode software (Geting Coal Mine, Jining, Shandong, China), i.e., Coal Mine Occupational Safety and Health Management and Risk Control System, is developed to attain the aforementioned objectives, namely promoting the coal mine occupational safety and health management based on early warning and dynamic monitoring of risks. Furthermore, the practical effectiveness and the associated pattern for applying this software package to coal mining is analyzed. The study indicates that the presently developed coal mine occupational safety and health management and risk control technology and the associated software can support the occupational safety and health management efforts in coal mines in a standardized and effective manner. It can also control the accident risks scientifically and effectively; its effective implementation can further improve the coal mine occupational safety and health management mechanism, and further enhance the risk management approaches. Besides, its implementation indicates that the occupational safety and health management and risk control technology has been established based on a benign cycle involving dynamic feedback and scientific development, which can provide a reliable assurance to the safe operation of coal mines.
A concept of JAERI passive safety light water reactor system (JPSR)

Energy Technology Data Exchange (ETDEWEB)

Murao, Y.; Araya, F.; Iwamura, T. [Japan Atomic Energy Research Institute, Tokai-mura (Japan)

1995-09-01

The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.
Environmental safety of the disposal system for radioactive substance-contaminated wastes

International Nuclear Information System (INIS)

Oosako, Masahiro

2012-01-01

In accordance with the full-scale enforcement of 'The Act on Special Measures concerning the Handling of Radioactive Pollution' in 2012, the collective efforts of entire Japan for dealing with radioactive pollutants began. The most important item for dealing with radioactive pollution is to control radioactive substances that polluted the global environment and establish a contaminated waste treatment system for risk reduction. On the incineration system and landfill disposal system of radioactive waste, this paper arranges the scientific information up to now, and discusses the safety of the treatment / disposal systems of contaminated waste. As for 'The Act on Special Measures concerning the Handling of Radioactive Pollution,' this paper discusses the points of the Act and basic policy, roadmap for the installation of interim storage facilities, and enforcement regulations (Ordinance of the Ministry of the Environment). About the safety of waste treatment system, it discusses the safety level of technical standards at waste treatment facilities, safety of incineration facilities, and safety of landfill disposal sites. (O.A.)
Development of antilock braking system based on various intelligent control system

NARCIS (Netherlands)

Aparow, V.R.; Ahmad, F.; Hassan, M.Z.; Hudha, K.; Othman, M.

2012-01-01

This paper presents about the development of an Antilock Braking System (ABS) using quarter vehicle model and control the ABS using different type of controllers. Antilock braking system (ABS) is an important part in vehicle system to produce additional safety for drivers. In general, Antilock
Use of digital computing devices in systems important to safety

International Nuclear Information System (INIS)

1986-01-01

The incorporation of digital computing devices in systems important to safety now is progressing fast in several countries, including Canada, France, Federal Republic of Germany, Japan, USA. There are now reactors with microprocessors in some trip systems. The major functions of those systems are: reactor trip initiation, display, monitoring, testing, re-calibration of detectors. The benefits of moving to a fully computerized shut-down system should be improved reliability, greater flexibility, better man-machine interface, improved testing, higher reactor output and lower overall cost. With the introduction of computer devices in systems important to safety, plant availability and safety are improved because disturbances are treated before they lead to safety action, in this way helping the operator to avoid errors. The Meeting presentations were divided into sessions devoted to the following topics: Needs for the use of digital devices (DCD) in safety important systems (SIS) (5 papers); Problems raised by the integration SIS in the NPP control (7 papers); Description and presentation of DCD of SIS (6 papers); Results of experiences in engineering, manufacture, qualification operation of DCD hardware and software (5 papers). A separate abstract was prepared for each of these papers
Penerapan Safety Management System (SMS) Dan Kompetensi Pemandu Lalu Lintas Penerbangan

OpenAIRE

Melissa, Anaria Carini; Subagyo, Toto Hardiyanto; Suharno, Hadi; Majid, Suharto Abdul

2017-01-01

The objective of the research is to analyze the influence between Implementation of Safety Management System (SMS) and Competency of Air Traffic Controller towards Aviation Safety at Perum LPPNPI Main Branch JATSC. Type of research is quantitative research by means of surveys, where data obtained through a questionnaire that will be distributed On Air Traffic Controller Perum LPPNPI Main Branch JATSC. The population in this research is an Air Traffic Controller Perum LPPNPI Main Branch JAT...
Reactor limit control system

International Nuclear Information System (INIS)

Rubbel, F.E.

1982-01-01

The very extensive use of limitations in the operational field between protection system and closed-loop controls is an important feature of German understanding of operational safety. The design of limitations is based on very large activities in the computational field but mostly on the high level of the plant-wide own commissioning experience of a turnkey contractor. Limitations combine intelligence features of closed-loop controls with the high availability of protection systems. (orig.)
Safety parameter display system: an operator support system for enhancement of safety in Indian PHWRs

International Nuclear Information System (INIS)

Subramaniam, K.; Biswas, T.

1994-01-01

Ensuring operational safety in nuclear power plants is important as operator errors are observed to contribute significantly to the occurrence of accidents. Computerized operator support systems, which process and structure information, can help operators during both normal and transient conditions, and thereby enhance safety and aid effective response to emergency conditions. An important operator aid being developed and described in this paper, is the safety parameter display system (SPDS). The SPDS is an event-independent, symptom-based operator aid for safety monitoring. Knowledge-based systems can provide operators with an improved quality of information. An information processing model of a knowledge based operator support system (KBOSS) developed for emergency conditions using an expert system shell is also presented. The paper concludes with a discussion of the design issues involved in the use of a knowledge based systems for real time safety monitoring and fault diagnosis. (author). 8 refs., 4 figs., 1 tab
Artificial intelligence enhancements to safety parameter display systems

International Nuclear Information System (INIS)

Hajek, B.K.; Hashemi, S.; Sharma, D.; Chandrasekaran, B.; Miller, D.W.

1986-01-01

Two prototype knowledge based systems have been developed at The Ohio State University to be the basis of an operator aid that can be attached to an existing nuclear power plant Safety Parameter Display System. The first system uses improved sensor validation techniques to provide input to a fault diagnosis process. The second system would use the diagnostic system output to synthesize corrective procedures to aid the control room licensed operator in plant recovery
Modeling of requirement specification for safety critical real time computer system using formal mathematical specifications

International Nuclear Information System (INIS)

Sankar, Bindu; Sasidhar Rao, B.; Ilango Sambasivam, S.; Swaminathan, P.

2002-01-01

Full text: Real time computer systems are increasingly used for safety critical supervision and control of nuclear reactors. Typical application areas are supervision of reactor core against coolant flow blockage, supervision of clad hot spot, supervision of undesirable power excursion, power control and control logic for fuel handling systems. The most frequent cause of fault in safety critical real time computer system is traced to fuzziness in requirement specification. To ensure the specified safety, it is necessary to model the requirement specification of safety critical real time computer systems using formal mathematical methods. Modeling eliminates the fuzziness in the requirement specification and also helps to prepare the verification and validation schemes. Test data can be easily designed from the model of the requirement specification. Z and B are the popular languages used for modeling the requirement specification. A typical safety critical real time computer system for supervising the reactor core of prototype fast breeder reactor (PFBR) against flow blockage is taken as case study. Modeling techniques and the actual model are explained in detail. The advantages of modeling for ensuring the safety are summarized
Comprehensive Lifecycle for Assuring System Safety

Science.gov (United States)

Knight, John C.; Rowanhill, Jonathan C.

2017-01-01

CLASS is a novel approach to the enhancement of system safety in which the system safety case becomes the focus of safety engineering throughout the system lifecycle. CLASS also expands the role of the safety case across all phases of the system's lifetime, from concept formation to decommissioning. As CLASS has been developed, the concept has been generalized to a more comprehensive notion of assurance becoming the driving goal, where safety is an important special case. This report summarizes major aspects of CLASS and contains a bibliography of papers that provide additional details.
Development of FPGA-based safety-related I and C systems

Energy Technology Data Exchange (ETDEWEB)

Goto, Y.; Oda, N.; Miyazaki, T.; Hayashi, T.; Sato, T.; Igawa, S. [08, Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); 1, Toshiba-cho, Fuchu, Tokyo 183-8511 (Japan)

2006-07-01

Toshiba has developed Non-rewritable (NRW) Field Programmable Gate Array (FPGA)-based safety-related Instrumentation and Control (I and C) system [1]. Considering application to safety-related systems, nonvolatile and non-rewritable FPGA which is impossible to be changed after once manufactured has been adopted in Toshiba FPGA-based system. FPGA is a device which consists only of defined digital circuit: hardware, which performs defined processing. FPGA-based system solves issues existing both in the conventional systems operated by analog circuits (analog-based system) and the systems operated by central processing unit (CPU-based system). The advantages of applying FPGA are to keep the long-life supply of products, improving testability (verification), and to reduce the drift which may occur in analog-based system. The system which Toshiba developed this time is Power Range Monitor (PRM). Toshiba is planning to expand application of FPGA-based technology by adopting this development method to the other safety-related systems from now on. (authors)
Tritium system test assembly control system cost estimate

International Nuclear Information System (INIS)

Stutz, R.A.

1979-01-01

The principal objectives of the Tritium Systems Test Assembly (TSTA), which includes the development, demonstration and interfacing of technologies related to the deuterium--tritium fuel cycle for fusion reactor systems, are concisely stated. The various integrated subsystems comprising TSTA and their functions are discussed. Each of the four major subdivisions of TSTA, including the main process system, the environmental and safety systems, supporting systems and the physical plant are briefly discussed. An overview of the Master Data Acquisition and Control System, which will control all functional operation of TSTA, is provided
Development of Necessary Technology for Safety Grade I and C System

International Nuclear Information System (INIS)

Lee, Dong Young; Jang, Tong Il; Kim, Jung Tack

2010-08-01

Through KNICS and Nu-Tech 2012, a MMIS(Man-Machine Interface System) package for nuclear plants was localized and is expected to be applied to SUN 1,2 plants. This study is aimed to support the application of the MMIS package including the following technologies that meet strengthened regulation requirements and enhanced utility's performance requirements to nuclear plants. - Dedication to safety-grade computers - The broadband communication network for safety information - Application of the automatic test for safety systems - Application of programmable logic controllers to nuclear plant
Potential safety enhancements to nuclear plant control: proof testing at EBR-II

International Nuclear Information System (INIS)

Lindsay, R.W.; Chisholm, G.H.

1984-01-01

Future changes in nuclear plant control and protective systems will reflect an evolutionary improvement through increased use of computers coupled with a better integration of man and machine. Before improvements can be accepted into the licensed commercial plant environment, significant testing must be accomplished to answer safety questions and to prove the worth of new ideas. The Experimental Breeder Reactor-II (EBR-II) is being used as a test-bed for both in-house development and testing for others in a DOE sponsored Man-Machine Integration program. The ultimate result of the development and testing would be a control system for which safety credit could be taken in the licensing process
Radiation safety management system in a radioactive facility

International Nuclear Information System (INIS)

Amador, Zayda H.

2008-01-01

Full text: This paper illustrates the Cuban experience in implementing and promoting an effective radiation safety system for the Centre of Isotopes, the biggest radioactive facility of our country. Current management practice demands that an organization inculcate culture of safety in preventing radiation hazard. The aforementioned objectives of radiation protection can only be met when it is implemented and evaluated continuously. Commitment from the workforce to treat safety as a priority and the ability to turn a requirement into a practical language is also important to implement radiation safety policy efficiently. Maintaining and improving safety culture is a continuous process. There is a need to establish a program to measure, review and audit health and safety performance against predetermined standards. All those areas of the radiation protection program are considered (e.g. licensing and training of the staff, occupational exposure, authorization of the practices, control of the radioactive material, radiological occurrences, monitoring equipment, radioactive waste management, public exposure due to airborne effluents, audits and safety costs). A set of indicators designed to monitor key aspects of operational safety performance are used. Their trends over a period of time are analyzed with the modern information technologies, because this can provide an early warning to plant management for searching causes behind the observed changes. In addition to analyze the changes and trends, these indicators are compared against identified targets and goals to evaluate performance strengths and weaknesses. A structured and proper radiation self-auditing system is seen as a basic requirement to meet the current and future needs in sustainability of radiation safety. The integrated safety management system establishment has been identified as a goal and way for the continuous improvement. (author)
Safety controls according to the non-proliferation treaty in EC countries

International Nuclear Information System (INIS)

Pander, J. von.

1978-01-01

Above all, content and extent of the duty conferred upon the IAEA according to article III, paragraph 1 of the NP treaty which implies the conducting of safety controls and the consequences resulting here from are examined. Including the peaceful use of nuclear energy developing under international law the agreement on safety control signed on 5th April 1973 between IAEA and EURATOM as well as its seven non-nuclear-weapon member states is discussed, along with its technical and its implicit legal problems. In detail the manifold technical and judicial problems of IAEA safety controls are shown, their realization requiring a well-working cooperation between IAEA and the European Communities. As only the non-nuclear-weapon member states of the EC are subject to the IAEA safety control system within the frame of this agreement the following questions are discussed: 1. effects on the member status after the signing of the EURATOM contract and 2. granting the principle of equal treatment for all member states as against the nuclear-weapon member states of the EC, France and the United Kingdom. (orig./HP) [de
Safety systems I/C equipment reliability analyses of the Kozloduy NPP units 3 and 4

Energy Technology Data Exchange (ETDEWEB)

Halev, G; Christov, N [Risk Engineering Ltd., Sofia (Bulgaria)

1996-12-31

The purpose of the analysis is to assess the safety systems I/C equipment reliability. The assessment includes: quantification of the safety systems unavailability due to component failures; definition of the minimal cut sets leading to the analysed safety systems failure; quantification of the I/C equipment importance measures of the dominant contribution components. The safety systems I/C equipment reliability has been analysed using PSAPACK (a code for probabilistic safety assessment). Fault trees for the following safety systems of the Kozloduy-3 and Kozloduy-4 reactors have been constructed: neutron flow control equipment, reactor protection system, main coolant pumps, pressurizer safety valves `Sempell`, steam dump systems, spray system, low pressure injection system, emergency feeding water system, essential service water system. THree separate reports have been issued containing the performed analyses and results. 1 ref.

Incorporating Traffic Control and Safety Hardware Performance Functions into Risk-based Highway Safety Analysis

Directory of Open Access Journals (Sweden)

Zongzhi Li

2017-04-01

Full Text Available Traffic control and safety hardware such as traffic signs, lighting, signals, pavement markings, guardrails, barriers, and crash cushions form an important and inseparable part of highway infrastructure affecting safety performance. Significant progress has been made in recent decades to develop safety performance functions and crash modification factors for site-specific crash predictions. However, the existing models and methods lack rigorous treatments of safety impacts of time-deteriorating conditions of traffic control and safety hardware. This study introduces a refined method for computing the Safety Index (SI as a means of crash predictions for a highway segment that incorporates traffic control and safety hardware performance functions into the analysis. The proposed method is applied in a computation experiment using five-year data on nearly two hundred rural and urban highway segments. The root-mean square error (RMSE, Chi-square, Spearman’s rank correlation, and Mann-Whitney U tests are employed for validation.
Test to prove the resistance to incidents of components of electric and control systems in the safety containment of nuclear power plants

International Nuclear Information System (INIS)

1982-01-01

The marginal program for proving the suitability of safety-relevant components of electric and control systems in the safety containment during a loss-of-coolant incident is described. Variant test conditions are established in the component-specific test program. Special attention has been paid to the representation of the course of pressure and temperature for the performance test of the valve room of the Nuclear Power Plant Philippsburg 2. (DG) [de
Controlled versus automatic processes: which is dominant to safety? The moderating effect of inhibitory control.

Directory of Open Access Journals (Sweden)

Yaoshan Xu

Full Text Available This study explores the precursors of employees' safety behaviors based on a dual-process model, which suggests that human behaviors are determined by both controlled and automatic cognitive processes. Employees' responses to a self-reported survey on safety attitudes capture their controlled cognitive process, while the automatic association concerning safety measured by an Implicit Association Test (IAT reflects employees' automatic cognitive processes about safety. In addition, this study investigates the moderating effects of inhibition on the relationship between self-reported safety attitude and safety behavior, and that between automatic associations towards safety and safety behavior. The results suggest significant main effects of self-reported safety attitude and automatic association on safety behaviors. Further, the interaction between self-reported safety attitude and inhibition and that between automatic association and inhibition each predict unique variances in safety behavior. Specifically, the safety behaviors of employees with lower level of inhibitory control are influenced more by automatic association, whereas those of employees with higher level of inhibitory control are guided more by self-reported safety attitudes. These results suggest that safety behavior is the joint outcome of both controlled and automatic cognitive processes, and the relative importance of these cognitive processes depends on employees' individual differences in inhibitory control. The implications of these findings for theoretical and practical issues are discussed at the end.
Diversity for security: case assessment for FPGA-based safety-critical systems

Directory of Open Access Journals (Sweden)

Kharchenko Vyacheslav

2016-01-01

Full Text Available Industrial safety critical instrumentation and control systems (I&Cs are facing more with information (in general and cyber, in particular security threats and attacks. The application of programmable logic, first of all, field programmable gate arrays (FPGA in critical systems causes specific safety deficits. Security assessment techniques for such systems are based on heuristic knowledges and the expert judgment. Main challenge is how to take into account features of FPGA technology for safety critical I&Cs including systems in which are applied diversity approach to minimize risks of common cause failure. Such systems are called multi-version (MV systems. The goal of the paper is in description of the technique and tool for case-based security assessment of MV FPGA-based I&Cs.
Development of Necessary Technology for localizing of Nuclear Safety Grade I and C System

International Nuclear Information System (INIS)

Kwon, Kee Choon; Jang, Tong Il; Kim, Jung Tack

2010-08-01

Through KNICS and Nu-Tech 2012, a MMIS(Man-Machine Interface System) package for nuclear plants was localized and is expected to be applied to SUN 1,2 plants. This study is aimed to support the application of the MMIS package including the following technologies that meet strengthened regulation requirements and enhanced utility's performance requirements to nuclear plants. - Dedication to safety-grade computers and real time operating systems - The broadband communication network for safety information - Application of the automatic test for safety systems - Application of programmable logic controllers to nuclear plants - Development strategy for an integrated SW development tool for control device
Controls and Machine Protection Systems

CERN Document Server

Carrone, E.

2016-01-01

Machine protection, as part of accelerator control systems, can be managed with a 'functional safety' approach, which takes into account product life cycle, processes, quality, industrial standards and cybersafety. This paper will discuss strategies to manage such complexity and the related risks, with particular attention to fail-safe design and safety integrity levels, software and hardware standards, testing, and verification philosophy. It will also discuss an implementation of a machine protection system at the SLAC National Accelerator Laboratory's Linac Coherent Light Source (LCLS).
Ageing study of the engineered safety features actuation system of the Loviisa NPP

International Nuclear Information System (INIS)

Simola, K.; Maskuniitty, M.

1995-06-01

An ageing study of the engineered safety features actuation system of the Loviisa nuclear power plant has been performed. The operating experience, including failure and maintenance histories of analog measuring devices, logics for safety signal formation and individual control electronics of pumps and valves, has been collected and analysed. The safety importance of system components has been studied with a fault tree analysis of a selected safety function. Based on the results of the analysis of operating experiences and the fault tree analysis, some components were selected for deeper analyses. According to the operating experience, the amount of failures in the Loviisa plant safety system has been low and no increasing trend in the failure history can yet be observed. Only a few failures had prohibited the propagation of the safety signal, mostly the failures have caused a false alarm. The failures reported have concerned mainly limit signal units, transmitters, and priority units. According to the fault tree analysis of one safety function, the most important components of this subsystem are individual control units and pulse/DC converters. Failure modes and effect analyses were performed for priority and individual control unit, limit signal unit and comparator and pulse/DC converter in order to identify the critical failure modes of these devices. (orig.) (15 refs., 26 figs., 9 tabs.)
Development of automated controller system for controlling reactivity by using FPGA in research reactor application

International Nuclear Information System (INIS)

Mohd Sabri Minhat; Izhar Abu Hussin; Mohd Idris Taib

2012-01-01

The scope for this research paper is to produce a detail design for Development of Automated Controller System for Controlling Reactivity by using FPGA in Research Reactor Application for high safety nuclear operation. The development of this project including design, purchasing, fabrication, installation, testing and validation and verification for one prototype automated controller system for controlling reactivity in industry local technology for human capacity and capability development towards the first Nuclear Power Programme (NPP) in Malaysia. The specific objectives of this research paper are to Development of Automated Controller System for Controlling Reactivity (ACSCR) in Research Reactor Application (PUSPATI TRIGA Reactor) by using simultaneous movement method; To design, fabricate and produce the accuracy of Control Rods Drive Mechanism to 0.1 mm resolution using a stepper motor as an actuator; To design, install and produce the system response to be more faster by using Field Programmable Gate Array (FPGA) and High Speed Computer; and to improve the Safety Level of the Research Reactor in high safety nuclear operation condition. (author)
Multilayer robust control for safety enhancement of reactor operations

International Nuclear Information System (INIS)

Edwards, R.M.; Lee, K.Y.; Ray, A.

1991-01-01

A novel concept of reactor power and temperature control has been recently reported in which a conventional output feedback controller is embedded within a state feedback setting. The embedded output feedback controller at the inner layer largely compensates for plant modeling uncertainties and external disturbances, and the outer layer generates an optimal control signal via feedback of the estimated plant states. A major advantage of this embedded architecture is the robustness of the control system relative to parametric and nonparametric uncertainties and thus the opportunity for designing fault-accommodating control algorithms to improve reactor operations and plant safety. The paper illustrates the architecture of the state-feedback-assisted classical (SFAC) control, which utilizes an embedded output feedback controller designed via classical techniques. It demonstrates the difference between the performance of conventional state feedback control and SFAC by examining the sensitivity of the dominant eigenvalues of the individual closed-loop systems
Implementation of an Enhanced Measurement Control Program for handling nuclear safety samples at WSRC

International Nuclear Information System (INIS)

Boler-Melton, C.; Holland, M.K.

1991-01-01

In the separation and purification of nuclear material, nuclear criticality safety (NCS) is of primary concern. The primary nuclear criticality safety controls utilized by the Savannah River Site (SRS) Separations Facilities involve administrative and process equipment controls. Additional assurance of NCS is obtained by identifying key process hold points where sampling is used to independently verify the effectiveness of production control. Nuclear safety measurements of samples from these key process locations provide a high degree of assurance that processing conditions are within administrative and procedural nuclear safety controls. An enhanced procedure management system aimed at making improvements in the quality, safety, and conduct of operation was implemented for Nuclear Safety Sample (NSS) receipt, analysis, and reporting. All procedures with nuclear safety implications were reviewed for accuracy and adequate detail to perform the analytical measurements safely, efficiently, and with the utmost quality. Laboratory personnel worked in a ''Deliberate Operating'' mode (a systematic process requiring continuous expert oversight during all phases of training, testing, and implementation) to initiate the upgrades. Thus, the effort to revise and review nuclear safety sample procedures involved a team comprised of a supervisor, chemist, and two technicians for each procedure. Each NSS procedure was upgraded to a ''Use Every Time'' (UET) procedure with sign-off steps to ensure compliance with each step for every nuclear safety sample analyzed. The upgrade program met and exceeded both the long and short term customer needs by improving measurement reliability, providing objective evidence of rigid adherence to program principles and requirements, and enhancing the system for independent verification of representative sampling from designated NCS points
Regenerative braking strategies, vehicle safety and stability control systems: critical use-case proposals

Science.gov (United States)

Oleksowicz, Selim A.; Burnham, Keith J.; Southgate, Adam; McCoy, Chris; Waite, Gary; Hardwick, Graham; Harrington, Cian; McMurran, Ross

2013-05-01

The sustainable development of vehicle propulsion systems that have mainly focused on reduction of fuel consumption (i.e. CO2 emission) has led, not only to the development of systems connected with combustion processes but also to legislation and testing procedures. In recent years, the low carbon policy has made hybrid vehicles and fully electric vehicles (H/EVs) popular. The main virtue of these propulsion systems is their ability to restore some of the expended energy from kinetic movement, e.g. the braking process. Consequently new research and testing methods for H/EVs are currently being developed. This especially concerns the critical 'use-cases' for functionality tests within dynamic events for both virtual simulations, as well as real-time road tests. The use-case for conventional vehicles for numerical simulations and road tests are well established. However, the wide variety of tests and their great number (close to a thousand) creates a need for selection, in the first place, and the creation of critical use-cases suitable for testing H/EVs in both virtual and real-world environments. It is known that a marginal improvement in the regenerative braking ratio can significantly improve the vehicle range and, therefore, the economic cost of its operation. In modern vehicles, vehicle dynamics control systems play the principal role in safety, comfort and economic operation. Unfortunately, however, the existing standard road test scenarios are insufficient for H/EVs. Sector knowledge suggests that there are currently no agreed tests scenarios to fully investigate the effects of brake blending between conventional and regenerative braking as well as the regenerative braking interaction with active driving safety systems (ADSS). The paper presents seven manoeuvres, which are considered to be suitable and highly informative for the development and examination of H/EVs with regenerative braking capability. The critical manoeuvres presented are considered to be
Emergency Diesel: Safety-related instrumentation and control with programmable logic controllers

International Nuclear Information System (INIS)

Breidenich, G.; Luedtke, M.

2004-01-01

This report presents a new concept for the design of emergency diesel equipment protection circuits as a part of the safety related instrumentation in the nuclear power plant Biblis, units A and B. The concept was implemented with state of the art SIMATIC S7/316 programmable logic controllers (PLCs) and can be adapted to any system with high availability requirements (e.g. power plant turbines, aircraft engines, mining pumps etc). (orig.)
A Methodological Framework for Software Safety in Safety Critical Computer Systems

OpenAIRE

P. V. Srinivas Acharyulu; P. Seetharamaiah

2012-01-01

Software safety must deal with the principles of safety management, safety engineering and software engineering for developing safety-critical computer systems, with the target of making the system safe, risk-free and fail-safe in addition to provide a clarified differentaition for assessing and evaluating the risk, with the principles of software risk management. Problem statement: Prevailing software quality models, standards were not subsisting in adequately addressing the software safety ...
An analysis of safety control effectiveness

International Nuclear Information System (INIS)

Son, K.S.; Melchers, R.E.; Kal, W.M.

2000-01-01

The cost of injuries and 'accidents' to an organisation is very important in establishing how much it should spend on safety control. Despite the usefulness of information about the cost of a company's accidents, it is not customary accounting practice to make these data available. Of the two kinds of costs incurred by a company through occupational injuries and accidents, direct costs and indirect costs; the direct costs are much easier to estimate. However, the uninsured costs are usually more critical and should be estimated by each company. The authors investigate a general model to estimate the above costs and hence to establish efficient safety control. One construction company has been a pilot for this study. By analysing actual company data for three years, it is found that the efficient safety control cost should be 1.2-1.3% of total contract costs
DASHBOARDS AND CONTROL CHARTS. EXPERIENCES IN IMPROVING SAFETY AT HANFORD WASHINGTON

International Nuclear Information System (INIS)

PREVETTE, S.S.

2006-01-01

The aim of this paper is to demonstrate the integration of safety methodology, quality tools, leadership, and teamwork at Hanford and their significant positive impact on safe performance of work. Dashboards, Leading Indicators, Control charts, Pareto Charts, Dr. W. Edward Deming's Red Bead Experiment, and Dr. Deming's System of Profound Knowledge have been the principal tools and theory of an integrated management system. Coupled with involved leadership and teamwork, they have led to significant improvements in worker safety and protection, and environmental restoration at one of the nation's largest nuclear cleanup sites
Safety climate and self-reported injury: assessing the mediating role of employee safety control.

Science.gov (United States)

Huang, Yueng-Hsiang; Ho, Michael; Smith, Gordon S; Chen, Peter Y

2006-05-01

To further reduce injuries in the workplace, companies have begun focusing on organizational factors which may contribute to workplace safety. Safety climate is an organizational factor commonly cited as a predictor of injury occurrence. Characterized by the shared perceptions of employees, safety climate can be viewed as a snapshot of the prevailing state of safety in the organization at a discrete point in time. However, few studies have elaborated plausible mechanisms through which safety climate likely influences injury occurrence. A mediating model is proposed to link safety climate (i.e., management commitment to safety, return-to-work policies, post-injury administration, and safety training) with self-reported injury through employees' perceived control on safety. Factorial evidence substantiated that management commitment to safety, return-to-work policies, post-injury administration, and safety training are important dimensions of safety climate. In addition, the data support that safety climate is a critical factor predicting the history of a self-reported occupational injury, and that employee safety control mediates the relationship between safety climate and occupational injury. These findings highlight the importance of incorporating organizational factors and workers' characteristics in efforts to improve organizational safety performance.
Concept for creating program-technical complex of safety monitoring with system of safety parameters presentation functions on the basis of routine WWER-1000 systems

International Nuclear Information System (INIS)

Dunaev, V.G.; Tarasov, M. V.; Povarov, P.V.

2005-01-01

Prerequisites of creating the software-hardware complex for reactor safety monitoring on the Volgodonsk NPP are analyzed and generalized. The concept of this complex is based on functions of the safety parameters presentation system. It will serve as an interface between operator and technological process and give to operator a possibility to estimate quickly the state of the safety of the nuclear power unit. The complex will be created on the basis of routine reactor monitoring and control systems intended for the WWER-1000 reactor. In addition to existing soft- and hard-wares for reactor monitoring and for analysis of technological archive, it is proposed to create and connect in parallel the new software-hardware complex which ensures calculation and presentation of generalized factors of reactor safety [ru
Study of system safety evaluation on LTO of national project. NISA safety research project on system safety of nuclear power plants

International Nuclear Information System (INIS)

Takizawa, Masayuki; Sekimura, Naoto; Miyano, Hiroshi; Aoyama, Katsunobu

2012-01-01

Japanese safety regulatory body, that is, Nuclear and Industrial Safety Agency (NISA) started a 5-year national safety research project as 'the first stage' from 2006 FY to 2010 FY whose objective is 'Improve the technical information basis in order to utilize knowledge as well as information related to ageing management and maintenance of NPPs. Fukushima disaster happened in March 2011, and the priority of research needs for ageing management dramatically changed in Japan. The second-stage national project started in October 2011 with the concept of 'system safety' of NNPs where not only ageing management on degradation phenomena of important components but also safety management on total plant systems are paid attention to. The second-stage project is so called 'Japanese Ageing Management Program for System Safety (JAMPSS)'. (author)
Preliminary safety evaluation for CSR1000 with passive safety system

International Nuclear Information System (INIS)

Wu, Pan; Gou, Junli; Shan, Jianqiang; Zhang, Bo; Li, Xiang

2014-01-01

Highlights: • The basic information of a Chinese SCWR concept CSR1000 is introduced. • An innovative passive safety system is proposed for CSR1000. • 6 Transients and 3 accidents are analysed with system code SCTRAN. • The passive safety systems greatly mitigate the consequences of these incidents. • The inherent safety of CSR1000 is enhanced. - Abstract: This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities
Application of Risk Management for Control and Monitoring Systems

CERN Document Server

Grau, S; Balda, F; Chouvelon, A

2001-01-01

This paper presents an application of the state of the art and new trends for risk management of safety-related control and monitoring systems, currently applied in the industry. These techniques not only enable to manage safety and reliability issues but they also help in the control of quality and economic factors affected by the availability and maintenance of the system. The method includes an unambiguous definition of the system in terms of functions and a systematic analysis of hazardous situations, undesired events and possible malfunctions. It also includes the identification and quantification of the risk associated to the system. The required risk reduction is specified in terms of safety integrity levels. The safety integrity level results in requirements, preventive measures, possible improvements and recommendations to assure the satisfactory management of the risk.

The reliability of the software of the digital control system Nuclear Advantage

International Nuclear Information System (INIS)

Graae, T.; Engdahl, L.

1996-01-01

The ABB nuclear power control system Nuclear Advantage is a truly integrated control system. The integration of process control and safety control aims at achieving a common operator interface in order to simplify and thus improve control room ergonomics. The challenge is to design an integrated control system and at the same time ensure the functional separation between the independent safety subsystems as well as between the safety and the conventional sections. Software reliability is discussed and illustrated by statistical test results. It has proved to be a hundred times better than the reliability of the high-quality hardware. (orig.) [de
Independent V and V for Safety Grade I and C System

International Nuclear Information System (INIS)

Lee, Jang Soo; Kwon, Kee Choon; Lee, Dong Young

2010-08-01

Through KNICS and Nu-Tech 2012, a MMIS(Man-Machine Interface System) package for nuclear plants was localized and is expected to be applied to SUN 1,2 plants. This study is aimed to support the application of the MMIS package including the following technologies that meet strengthened regulation requirements and enhanced utility's performance requirements to nuclear plants. - Dedication to real time operating system for safety-grade computers - V and V of the broadband communication network for safety information - Independent V and V for application of programmable logic controllers to nuclear plants - Development strategy for an integrated SW development tool for control devices
The effect of using road safety equipment and systems and ...

African Journals Online (AJOL)

The effect of using road safety equipment and systems and determine their role on ... traffic control equipment situation and by multi-criteria weighting systems AHP ... The results have shown that indices median, lighting and panel type and the ...
Remote mobile communication in safety support system

International Nuclear Information System (INIS)

Inagaki, Kanji; Kobayashi, Hiroyuki; Hatanaka, Takahiro; Sakuma, Akira; Fukumoto, Akira; Ikeda, Jun

1999-01-01

Safety Support System (SSS) is a computerized operator support system for nuclear power plants, which is now under development. The concept of SSS covers 1) earlier detection of failure symptom and prediction of its influence to the plant operation, 2) improved transparency and robustness of plant control systems, 3) advanced human-machine interface and communication. The authors have been working on the third concept and proposed a remote mobile communication system called Plant Communication System (PCS). PCS aims to realize convenient communication between main control room and other areas such as plant local areas and site offices, using Personal Handyphone System (PHS) and wireless LAN (Local Area Network). PCS can transmit not only data but also graphic displays and dynamic video displays between the main control room and plant local areas. MPEG4 (Moving Picture Experts Group 4) technology is utilized in video data compression and decompression. The authors have developed the special multiplexing unit that connects PHS Cell Stations (CSs) and exiting coaxial cables. Voice recognition and announcement capability is also realized in the system, which enables verbal retrieval of information in the computer systems in the main control room from local areas. (author)
Metal food packaging design based on hazard analysis critical control point (HACCP system in canned food safety

Directory of Open Access Journals (Sweden)

Li Xingyi

2016-06-01

Full Text Available This study aims to design metal food packaging with hazard analysis critical control point (HACCP. First, theory of HACCP was introduced in detail. Taking empty cans provided by Wuxi Huapeng Food Packaging Company as an example, we studied migration of bisphenol compounds in coating of food can to food stimulant. Moreover, packaging design of luncheon meat can was taken as an example to confirm whether HACCP system could effectively control migration of phenolic substance. Results demonstrated that, coating of such empty were more likely to contain multiple bisphenol compounds such as bisphenol A (BPA, and bisphenol A diglycidyl ether (BADGE was considered as the leading bisphenol pollutant; food stimulant of different types, storage temperature and time could all impact migration of bisphenol compounds. HACCP system was proved to be effective in controlling hazards of phenolic substance in luncheon meat can and could reduce various phenolic substance indexes to an acceptable range. Therefore, HACCP can control migration of phenolic substance and recontamination of food and thus ensure food safety.
Survey and evaluation of inherent safety characteristics and passive safety systems for use in probabilistic safety analyses

International Nuclear Information System (INIS)

Wetzel, N.; Scharfe, A.

1998-01-01

The present report examines the possibilities and limits of a probabilistic safety analysis to evaluate passive safety systems and inherent safety characteristics. The inherent safety characteristics are based on physical principles, that together with the safety system lead to no damage. A probabilistic evaluation of the inherent safety characteristic is not made. An inventory of passive safety systems of accomplished nuclear power plant types in the Federal Republic of Germany was drawn up. The evaluation of the passive safety system in the analysis of the accomplished nuclear power plant types was examined. The analysis showed that the passive manner of working was always assumed to be successful. A probabilistic evaluation was not performed. The unavailability of the passive safety system was determined by the failure of active components which are necessary in order to activate the passive safety system. To evaluate the passive safety features in new concepts of nuclear power plants the AP600 from Westinghouse, the SBWR from General Electric and the SWR 600 from Siemens, were selected. Under these three reactor concepts, the SWR 600 is specially attractive because the safety features need no energy sources and instrumentation in this concept. First approaches for the assessment of the reliability of passively operating systems are summarized. Generally it can be established that the core melt frequency for the passive concepts AP600 and SBWR is advantageous in comparison to the probabilistic objectives from the European Pressurized Water Reactor (EPR). Under the passive concepts is the SWR 600 particularly interesting. In this concept the passive systems need no energy sources and instrumentation, and has active operational systems and active safety equipment. Siemens argues that with this concept the frequency of a core melt will be two orders of magnitude lower than for the conventional reactors. (orig.) [de
Nuclear safety risk control in the outage of CANDU unit

International Nuclear Information System (INIS)

Wu Mingliang; Zheng Jianhua

2014-01-01

Nuclear fuel remains in the core during the outage of CANDU unit, but there are still nuclear safety risks such as reactor accidental criticality, fuel element failure due to inability to properly remove residual heat. Furthermore, these risks are aggravated by the weakening plant system configuration and multiple cross operations during the outage. This paper analyzes the phases where there are potential nuclear safety risks on the basis of the typical critical path arrangement of the outage of Qinshan NPP 3 and introduces a series of CANDU-specific risk control measures taken during the past plant outages to ensure nuclear safety during the unit outage. (authors)
Practical Applications of Cosmic Ray Science: Spacecraft, Aircraft, Ground Based Computation and Control Systems and Human Health and Safety

Science.gov (United States)

Atwell, William; Koontz, Steve; Normand, Eugene

2012-01-01

In this paper we review the discovery of cosmic ray effects on the performance and reliability of microelectronic systems as well as on human health and safety, as well as the development of the engineering and health science tools used to evaluate and mitigate cosmic ray effects in earth surface, atmospheric flight, and space flight environments. Three twentieth century technological developments, 1) high altitude commercial and military aircraft; 2) manned and unmanned spacecraft; and 3) increasingly complex and sensitive solid state micro-electronics systems, have driven an ongoing evolution of basic cosmic ray science into a set of practical engineering tools (e.g. ground based test methods as well as high energy particle transport and reaction codes) needed to design, test, and verify the safety and reliability of modern complex electronic systems as well as effects on human health and safety. The effects of primary cosmic ray particles, and secondary particle showers produced by nuclear reactions with spacecraft materials, can determine the design and verification processes (as well as the total dollar cost) for manned and unmanned spacecraft avionics systems. Similar considerations apply to commercial and military aircraft operating at high latitudes and altitudes near the atmospheric Pfotzer maximum. Even ground based computational and controls systems can be negatively affected by secondary particle showers at the Earth's surface, especially if the net target area of the sensitive electronic system components is large. Accumulation of both primary cosmic ray and secondary cosmic ray induced particle shower radiation dose is an important health and safety consideration for commercial or military air crews operating at high altitude/latitude and is also one of the most important factors presently limiting manned space flight operations beyond low-Earth orbit (LEO).
Contribution to the evaluation of safety of software used in command control systems in nuclear plants: application to the SPIN N4

International Nuclear Information System (INIS)

Soubies, B.; Boulc'h, J.; Elsensohn, O.; Le Meur, M.; Henry, J.Y.

1994-06-01

The licensing procedures process of nuclear plants features compulsory steps which bring about a thorough exam of the commands control system. This analysis accounts for the aspects linked to technologies (integrated circuits, software packages) which have been chosen by the manufacturer for the programmed systems in charge of safety functions. Important innovations have been introduced in terms of design and manufacturing processes of safety systems of 1400 MWe pressurized water reactors, more precisely for the integrated numerical protection system (SPIN). The methodology used by the IPSN for the exam of the software of this system is presented in the communication. This methodology leads the IPSN to carry out studies and developments of tools keeping in sight as their main goal to bring substantial help to analysis. (authors). 2 refs
An approach to the efficient assessment of safety and usability of computer based control systems, VeNuS 2. Global final report

International Nuclear Information System (INIS)

Nelke, T.; Dlugosch, C.; Olaverri Monreal, C.; Sachse, K.; Thuering, M.

2015-01-01

Prior to the use of computer-based instrumentation and control the evidence of sufficient safety, development methods and the suitability of man-machine interface must be provided. For this purpose, validation methods must be available, if possible supported by appropriate tools. Based on the multitude of the data which has to be taken into account it is important to generate technical documentation, to realize efficient operation and to prevent human based errors. An approach for computer based generation of user manuals for the operation of technical systems was developed in the VeNuS 2 project. A second goal was to develop an approach to evaluate the usability of safety relevant digital human-machine-interfaces (e.g. for nuclear industries). Therefore a software tool has been developed to assess aspects of usability of user interfaces by considering safety-related priorities. Additionally new or well known methods for provision of evidence of sufficient safety and usability for computer based systems shall be developed in a prototyped way.
Contribution at the evaluation of safety softwares in nuclear power plants control systems

International Nuclear Information System (INIS)

Soubies, B.; Le Meur, M.; Henry, J.Y.; Boulc'h, J.

1993-06-01

The introduction of programmable systems such the SPIN (Numerical Integrated Protection System) has conducted at particular dispositions for the conception and the use of such systems. The utilization of such systems until 1983 has conducted at modifications in the maintenance procedures. The new methods used for the N4 project in the evaluation of safety softwares are given in this report
Preliminary design of safety and interlock system for indian test facility of diagnostic neutral beam

International Nuclear Information System (INIS)

Tyagi, Himanshu; Soni, Jignesh; Yadav, Ratnakar; Bandyopadhyay, Mainak; Rotti, Chandramouli; Gahlaut, Agrajit; Joshi, Jaydeep; Parmar, Deepak; Bansal, Gourab; Pandya, Kaushal; Chakraborty, Arun

2016-01-01

Highlights: • Indian Test Facility being built to characterize DNB for ITER delivery. • Interlock system required to safeguard the investment incurred in building the facility and protecting ITER deliverable components. • Interlock levels upto 3IL-3 identified. • Safety instrumented system for occupational safety being designed. Safety I&C functions of SIL-2 identified. • The systems are based on ITER PIS and PSS design guidelines. - Abstract: Indian Test Facility (INTF) is being built in Institute For Plasma Research to characterize Diagnostic Neutral Beam in co-operation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics, high voltage power supplies, high power RF generators, mechanical systems and diagnostics systems. Out of these, several INTF components are ITER deliverable, that is, beam source, beam line components and power supplies. To ensure successful operation of INTF involving integrated operation of all the constituent plant systems a matured Data Acquisition and Control System (DACS) is required. The INTF DACS is based on CODAC platform following on PCDH (Plant Control Design Handbook) guidelines. The experimental phases involve application of HV power supplies (100 KV) and High RF power (∼800 KW) which will produce energetic beam of maximum power 6MW within the facility for longer durations. Hence the entire facility will be exposed tohigh heat fluxes and RF radiations. To ensure investment protection and to provide occupational safety for working personnel a matured Safety and Interlock system is required for INTF. The Safety and Interlock systems are high-reliability I&C systems devoted completely to the specific functions. These systems will be separate from the conventional DACS of INTF which will handle the conventional control and acquisition functions. Both, the Safety and Interlock systems are based on IEC 61511 and IEC 61508 standards as
Preliminary design of safety and interlock system for indian test facility of diagnostic neutral beam

Energy Technology Data Exchange (ETDEWEB)

Tyagi, Himanshu, E-mail: htyagi@iter-india.org [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Soni, Jignesh [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Yadav, Ratnakar; Bandyopadhyay, Mainak; Rotti, Chandramouli [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Gahlaut, Agrajit [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Joshi, Jaydeep; Parmar, Deepak [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Bansal, Gourab; Pandya, Kaushal; Chakraborty, Arun [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India)

2016-11-15

Highlights: • Indian Test Facility being built to characterize DNB for ITER delivery. • Interlock system required to safeguard the investment incurred in building the facility and protecting ITER deliverable components. • Interlock levels upto 3IL-3 identified. • Safety instrumented system for occupational safety being designed. Safety I&C functions of SIL-2 identified. • The systems are based on ITER PIS and PSS design guidelines. - Abstract: Indian Test Facility (INTF) is being built in Institute For Plasma Research to characterize Diagnostic Neutral Beam in co-operation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics, high voltage power supplies, high power RF generators, mechanical systems and diagnostics systems. Out of these, several INTF components are ITER deliverable, that is, beam source, beam line components and power supplies. To ensure successful operation of INTF involving integrated operation of all the constituent plant systems a matured Data Acquisition and Control System (DACS) is required. The INTF DACS is based on CODAC platform following on PCDH (Plant Control Design Handbook) guidelines. The experimental phases involve application of HV power supplies (100 KV) and High RF power (∼800 KW) which will produce energetic beam of maximum power 6MW within the facility for longer durations. Hence the entire facility will be exposed tohigh heat fluxes and RF radiations. To ensure investment protection and to provide occupational safety for working personnel a matured Safety and Interlock system is required for INTF. The Safety and Interlock systems are high-reliability I&C systems devoted completely to the specific functions. These systems will be separate from the conventional DACS of INTF which will handle the conventional control and acquisition functions. Both, the Safety and Interlock systems are based on IEC 61511 and IEC 61508 standards as
Application of system safety engineering techniques for hazard prevention at the Superconducting Super Collider

International Nuclear Information System (INIS)

Hendrix, B.L.

1991-01-01

A primary goal of the Superconducting Super Collider Laboratory (SSCL) is to establish an exemplary safety program. Achieving this goal requires leadership, planning, coordination, and technical know-how. To ensure that safety is an inherent part of the design, the Environment, Safety and Health Office employs a systems engineering discipline and process known as System Safety. The goal of System Safety - hazard prevention - is accomplished by analyzing systems to identify hazards and to evaluate design and procedural options and countermeasures to prevent, eliminate, mitigate, or control hazards and risks. Establishment of safety and human factors design criteria at the outset of the project prevents unsafe designs and safety violations, reduces risks, and helps in avoiding costly design changes later. This process requires a considerable amount of coordination with a variety of technical disciplines and safety professionals to integrate methods of hazard prevention, mitigation, and risk reduction throughout the system life-cycle
How to Deal with Revolutions in Train Control Systems

Directory of Open Access Journals (Sweden)

Hideo Nakamura

2016-09-01

Full Text Available Train control systems ensure the safety of railways. This paper begins with a summary of the typical train control systems in Japan and Europe. Based on this summary, the author then raises the following question regarding current train control systems: What approach should be adopted in order to enhance the functionality, safety, and reliability of train control systems and assist in commercial operations on railways? Next, the author provides a desirable architecture that is likely to assist with the development of new train control systems based on current information and communication technologies. A new unified train control system (UTCS is proposed that is effective in enhancing the robustness and competitiveness of a train control system. The ultimate architecture of the UTCS will be only composed of essential elements such as point machines and level crossing control devices in the field. Finally, a processing method of the UTCS is discussed.
Does the concept of safety culture help or hinder systems thinking in safety?

Science.gov (United States)

Reiman, Teemu; Rollenhagen, Carl

2014-07-01

The concept of safety culture has become established in safety management applications in all major safety-critical domains. The idea that safety culture somehow represents a "systemic view" on safety is seldom explicitly spoken out, but nevertheless seem to linger behind many safety culture discourses. However, in this paper we argue that the "new" contribution to safety management from safety culture never really became integrated with classical engineering principles and concepts. This integration would have been necessary for the development of a more genuine systems-oriented view on safety; e.g. a conception of safety in which human, technological, organisational and cultural factors are understood as mutually interacting elements. Without of this integration, researchers and the users of the various tools and methods associated with safety culture have sometimes fostered a belief that "safety culture" in fact represents such a systemic view about safety. This belief is, however, not backed up by theoretical or empirical evidence. It is true that safety culture, at least in some sense, represents a holistic term-a totality of factors that include human, organisational and technological aspects. However, the departure for such safety culture models is still human and organisational factors rather than technology (or safety) itself. The aim of this paper is to critically review the various uses of the concept of safety culture as representing a systemic view on safety. The article will take a look at the concepts of culture and safety culture based on previous studies, and outlines in more detail the theoretical challenges in safety culture as a systems concept. The paper also presents recommendations on how to make safety culture more systemic. Copyright © 2013 Elsevier Ltd. All rights reserved.
The aviation safety reporting system

Science.gov (United States)

Reynard, W. D.

1984-01-01

The aviation safety reporting system, an accident reporting system, is presented. The system identifies deficiencies and discrepancies and the data it provides are used for long term identification of problems. Data for planning and policy making are provided. The system offers training in safety education to pilots. Data and information are drawn from the available data bases.
NASA Aviation Safety Reporting System (ASRS)

Science.gov (United States)

Connell, Linda J.

2017-01-01

The NASA Aviation Safety Reporting System (ASRS) collects, analyzes, and distributes de-identified safety information provided through confidentially submitted reports from frontline aviation personnel. Since its inception in 1976, the ASRS has collected over 1.4 million reports and has never breached the identity of the people sharing their information about events or safety issues. From this volume of data, the ASRS has released over 6,000 aviation safety alerts concerning potential hazards and safety concerns. The ASRS processes these reports, evaluates the information, and provides selected de-identified report information through the online ASRS Database at http:asrs.arc.nasa.gov. The NASA ASRS is also a founding member of the International Confidential Aviation Safety Systems (ICASS) group which is a collection of other national aviation reporting systems throughout the world. The ASRS model has also been replicated for application to improving safety in railroad, medical, fire fighting, and other domains. This presentation will discuss confidential, voluntary, and non-punitive reporting systems and their advantages in providing information for safety improvements.
Control strategy for power management, efficiency-optimization and operating-safety of a 5-kW solid oxide fuel cell system

International Nuclear Information System (INIS)

Zhang, Lin; Jiang, Jianhua; Cheng, Huan; Deng, Zhonghua; Li, Xi

2015-01-01

Highlights: • Efficiency optimization associated with simultaneous power and thermal management. • Fast load tracing, fuel starvation, high efficiency and operating safety are considered. • Open loop pre-conditioning current strategy is proposed for load step-up transients. • Feedback control scheme is proposed for load step-up transients. - Abstract: The slow power tracking, operating safety, especially the fuel exhaustion, and high efficiency considerations are the key issues for integrated solid oxide fuel cell (SOFC) systems during power step up transients, resulting in the relatively poor dynamic capabilities and make the transient load following very challenging and must be enhanced. To this end, this paper first focus on addressing the efficiency optimization associated with simultaneous power and thermal management of a 5-kW SOFC system. Particularly, a traverse optimization process including cubic convolution interpolation algorithm are proposed to obtain optimal operating points (OOPs) with the maximum efficiency. Then this paper investigate the current implications on system step-up transient performance, then a two stage pre-conditioning current strategy and a feedback power reference control scheme is proposed for load step-up transients to balance fast load following and fuel starvation, after that safe thermal transient is validated. Simulation results show the efficacy of the control design by demonstrating the fast load following ability while maintaining the safe operation, thus safe; efficient and fast load transition can be achieved
Jefferson Lab IEC 61508/61511 Safety PLC Based Safety System

International Nuclear Information System (INIS)

Mahoney, Kelly; Robertson, Henry

2009-01-01

This paper describes the design of the new 12 GeV Upgrade Personnel Safety System (PSS) at the Thomas Jefferson National Accelerator Facility (TJNAF). The new PSS design is based on the implementation of systems designed to meet international standards IEC61508 and IEC 61511 for programmable safety systems. In order to meet the IEC standards, TJNAF engineers evaluated several SIL 3 Safety PLCs before deciding on an optimal architecture. In addition to hardware considerations, software quality standards and practices must also be considered. Finally, we will discuss R and D that may lead to both high safety reliability and high machine availability that may be applicable to future accelerators such as the ILC.

Traction Control System for Motorcycles

Directory of Open Access Journals (Sweden)

Cardinale Pascal

2009-01-01

Full Text Available Traction control is a widely used control system to increase stability and safety of four wheel vehicles. Automatic stability control is used in the BMW K1200R motorcycle and in motoGP competition, but not in other motorcycles. This paper presents an algorithm and a low-cost real-time hardware implementation for motorcycles. A prototype has been developed, applied on a commercial motorcycle, and tested in a real track. The control system that can be tuned by the driver during the race has been appreciated by the test driver.
Integrating system safety into the basic systems engineering process

Science.gov (United States)

Griswold, J. W.

1971-01-01

The basic elements of a systems engineering process are given along with a detailed description of what the safety system requires from the systems engineering process. Also discussed is the safety that the system provides to other subfunctions of systems engineering.
A model of Occupational Safety and Health Management System (OSHMS) for promoting and controlling health and safety in textile industry.

Science.gov (United States)

Manimaran, S; Rajalakshmi, R; Bhagyalakshmi, K

2015-01-01

The development of Occupational Safety and Health Management System in textile industry will rejuvenate the workers and energize the economy as a whole. In India, especially in Tamil Nadu, approximately 1371 textile business is running with the help of 38,461 workers under Ginning, Spinning, Weaving, Garment and Dyeing sectors. Textile industry of contributes to the growth of Indian economy but it fails to foster education and health as key components of human development and help new democracies. The present work attempts to measure and develop OSHMS which reduce the hazards and risk involved in textile industry. Among all other industries textile industry is affected by enormous hazards and risk because of negligence by management and Government. It is evident that managements are not abiding by law when an accident has occurred. Managements are easily deceiving workers and least bothered about the Quality of Work Life (QWL). A detailed analysis of factors promoting safety and health to the workers has been done by performing confirmatory factor analysis, evaluating Risk Priority Number and the framework of OHMS has been conceptualized using Structural Equation Model. The data have been collected using questionnaire and interview method. The study finds occupation health for worker in Textile industry is affected not only by safety measure but also by technology and management. The work shows that difficulty in identifying the cause and effect of hazards, the influence of management in controlling and promoting OSHMS under various dimensions. One startling fact is existence of very low and insignificance correlation between health factors and outcome.
Safety analysis and review system: a Department of Energy safety assurance tool

International Nuclear Information System (INIS)

Rosenthal, H.B.

1981-01-01

The concept of the Safety Analysis and Review System is not new. It has been used within the Department and its predecessor agencies, Atomic Energy Commission (AEC) and Energy Research and Development Administration (ERDA), for over 20 years. To minimize the risks from nuclear reactor and power plants, the AEC developed a process to support management authorization of each operation through identification and analysis of potential hazards and the measures taken to control them. As the agency evolved from AEC through ERDA to the Department of Energy, its responsibilities were broadened to cover a diversity of technologies, including those associated with the development of fossil, solar, and geothermal energy. Because the safety analysis process had proved effective in a technology of high potential hazard, the Department investigated the applicability of the process to the other technologies. This paper describes the system and discusses how it is implemented within the Department
Applications of modern control systems in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Wilhelm, H [Brown, Boveri und Cie A.G., Mannheim (Germany, F.R.). Abt. GK/TE

1980-10-01

A new generation of automation and control systems are currently becoming commercially available in the power plant market which, because of their incorporation of microprocessors and bus data transmission systems, represent a major step forward in innovation. The application of these systems meets today's requirements and solutions, for the number of measurements to be performed has increased five or sixfold in the past few years, and the number of drive systems to be controlled has doubled or even tripled. Requirements to be met by process management systems have become vastly more complicated: peak load operation, short startup times, improved communication, and rising safety and reliability requirements, especially in nuclear power plants. Control concepts have been developed for the area relevant to reactor safety and for the whole of the plant, which make full use of the possibilities offered by plant systems. More stringent demands must be met especially in the areas of handling, communication, testing capability, improved function, and flexibility and modular design in the safety sector.
Safety culture as a matter of regulatory control and regulatory effectiveness

International Nuclear Information System (INIS)

Camargo, C.T.M.; Furieri, E.B.; Arrieta, L.A.I.; Almeida, C.U.C.

2002-01-01

More than 15 years have passed since the term 'safety culture' was introduced by the International Nuclear Safety Advisory Group (INSAG), and although the concept now is widely accepted, practical applications and characteristics have been disseminated mainly for nuclear power plant operating organizations. There is still a lack of international guidance on the use of safety culture as a regulatory matter and on the application of the concept within regulatory organizations. This work explores the meaning of safety culture in two different fields: as an element of safety management systems it shall be a matter of regulatory control; as a complementary tool for quality management it should be used to enhance regulatory effectiveness. Brazilian recent experience on regulating nuclear power reactors provide some examples on how the concept of safety culture may influence regulatory strategies and regulatory management. (author)
Programmable Electronic Safety Systems

International Nuclear Information System (INIS)

Parry, R.

1993-05-01

Traditionally safety systems intended for protecting personnel from electrical and radiation hazards at particle accelerator laboratories have made extensive use of electromechanical relays. These systems have the advantage of high reliability and allow the designer to easily implement failsafe circuits. Relay based systems are also typically simple to design, implement, and test. As systems, such as those presently under development at the Superconducting Super Collider Laboratory (SSCL), increase in size, and the number of monitored points escalates, relay based systems become cumbersome and inadequate. The move toward Programmable Electronic Safety Systems is becoming more widespread and accepted. In developing these systems there are numerous precautions the designer must be concerned with. Designing fail-safe electronic systems with predictable failure states is difficult at best. Redundancy and self-testing are prime examples of features that should be implemented to circumvent and/or detect failures. Programmable systems also require software which is yet another point of failure and a matter of great concern. Therefore the designer must be concerned with both hardware and software failures and build in the means to assure safe operation or shutdown during failures. This paper describes features that should be considered in developing safety systems and describes a system recently installed at the Accelerator Systems String Test (ASST) facility of the SSCL
A sensor monitoring system for telemedicine, safety and security applications

Science.gov (United States)

Vlissidis, Nikolaos; Leonidas, Filippos; Giovanis, Christos; Marinos, Dimitrios; Aidinis, Konstantinos; Vassilopoulos, Christos; Pagiatakis, Gerasimos; Schmitt, Nikolaus; Pistner, Thomas; Klaue, Jirka

2017-02-01

A sensor system capable of medical, safety and security monitoring in avionic and other environments (e.g. homes) is examined. For application inside an aircraft cabin, the system relies on an optical cellular network that connects each seat to a server and uses a set of database applications to process data related to passengers' health, safety and security status. Health monitoring typically encompasses electrocardiogram, pulse oximetry and blood pressure, body temperature and respiration rate while safety and security monitoring is related to the standard flight attendance duties, such as cabin preparation for take-off, landing, flight in regions of turbulence, etc. In contrast to previous related works, this article focuses on the system's modules (medical and safety sensors and associated hardware), the database applications used for the overall control of the monitoring function and the potential use of the system for security applications. Further tests involving medical, safety and security sensing performed in an real A340 mock-up set-up are also described and reference is made to the possible use of the sensing system in alternative environments and applications, such as health monitoring within other means of transport (e.g. trains or small passenger sea vessels) as well as for remotely located home users, over a wired Ethernet network or the Internet.
Safety evaluation report related to the preliminary design of the Standard Reference System, RESAR-414

International Nuclear Information System (INIS)

1978-11-01

The safety evaluation for the Westinghouse Standard Reactor includes information on general reactor characteristics; design criteria for systems and components; reactor coolant system; engineered safety systems; instrumentation and controls; electric power systems; auxiliary systems; steam and power conversion system; radioactive waste management; radiation protection; conduct of operations; accident analyses; and quality assurance
Life-critical digital flight control systems

Science.gov (United States)

Mcwha, James

1990-01-01

Digital autopilot systems were first used on commercial airplanes in the late 1970s. The A-320 airplane was the first air transport airplane with a fly-by-wire primary flight control system. On the 767-X (777) airplane Boeing will install all fly-by-wire flight controls. Activities related to safety, industry status and program phases are discussed.
An Integrated Approach of Model checking and Temporal Fault Tree for System Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Koh, Kwang Yong; Seong, Poong Hyun [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2009-10-15

Digitalization of instruments and control systems in nuclear power plants offers the potential to improve plant safety and reliability through features such as increased hardware reliability and stability, and improved failure detection capability. It however makes the systems and their safety analysis more complex. Originally, safety analysis was applied to hardware system components and formal methods mainly to software. For software-controlled or digitalized systems, it is necessary to integrate both. Fault tree analysis (FTA) which has been one of the most widely used safety analysis technique in nuclear industry suffers from several drawbacks as described in. In this work, to resolve the problems, FTA and model checking are integrated to provide formal, automated and qualitative assistance to informal and/or quantitative safety analysis. Our approach proposes to build a formal model of the system together with fault trees. We introduce several temporal gates based on timed computational tree logic (TCTL) to capture absolute time behaviors of the system and to give concrete semantics to fault tree gates to reduce errors during the analysis, and use model checking technique to automate the reasoning process of FTA.
Temperature Control System for Chromel-Alumel Thermocouple

International Nuclear Information System (INIS)

Piping Supriatna; Nurhanan; Riswan DJ; Heru K, B.; Edi Karyanta

2003-01-01

Nuclear Power Plan Operation Safety needs serious handling on temperature measurement and control. In this report has been done manufacturing Temperature Control System for Chromel-Alumel Thermocouple, accordance to material, equipment and human resource ability in the laboratory. Basic component for the Temperature Control System is LM-741 type of Operation Amplifier, which is functionalized as summer for voltage comparator. Function test for this Control System shown its ability for damping on temperature reference. The Temperature Control System will be implemented on PCB Processing Machine. (author)
The Management System for Nuclear Installations Safety Guide

International Nuclear Information System (INIS)

2009-01-01

This Safety Guide is applicable throughout the lifetime of a nuclear installation, including any subsequent period of institutional control, until there is no significant residual radiation hazard. For a nuclear installation, the lifetime includes site evaluation, design, construction, commissioning, operation and decommissioning. These stages in the lifetime of a nuclear installation may overlap. This Safety Guide may be applied to nuclear installations in the following ways: (a)To support the development, implementation, assessment and improvement of the management system of those organizations responsible for research, site evaluation, design, construction, commissioning, operation and decommissioning of a nuclear installation; (b)As an aid in the assessment by the regulatory body of the adequacy of the management system of a nuclear installation; (c)To assist an organization in specifying to a supplier, via contractual documentation, any specific element that should be included within the supplier's management system for the supply of products. This Safety Guide follows the structure of the Safety Requirements publication on The Management System for Facilities and Activities, whereby: (a)Section 2 provides recommendations on implementing the management system, including recommendations relating to safety culture, grading and documentation. (b)Section 3 provides recommendations on the responsibilities of senior management for the development and implementation of an effective management system. (c)Section 4 provides recommendations on resource management, including guidance on human resources, infrastructure and the working environment. (d)Section 5 provides recommendations on how the processes of the installation can be specified and developed, including recommendations on some generic processes of the management system. (e)Section 6 provides recommendations on the measurement, assessment and improvement of the management system of a nuclear installation. (f
Independent V and V for Safety Grade I and C System

Energy Technology Data Exchange (ETDEWEB)

Lee, Jang Soo; Kwon, Kee Choon; Lee, Dong Young [KAERI, Daejeon (Korea, Republic of)

2010-08-15

Through KNICS and Nu-Tech 2012, a MMIS(Man-Machine Interface System) package for nuclear plants was localized and is expected to be applied to SUN 1,2 plants. This study is aimed to support the application of the MMIS package including the following technologies that meet strengthened regulation requirements and enhanced utility's performance requirements to nuclear plants. - Dedication to real time operating system for safety-grade computers - V and V of the broadband communication network for safety information - Independent V and V for application of programmable logic controllers to nuclear plants - Development strategy for an integrated SW development tool for control devices
Seismic analysis of control and safety rod drive mechanism

International Nuclear Information System (INIS)

Meher Prasad, A.; Jaya, K.P.; Chellapandi, P.; Rajan Babu, V.; Selvaraj, T.

2003-01-01

Control rod and its driving mechanism for a Fast Breeder Reactor is to facilitate safe shutdown of the reactor in case of emergency. A theoretical study on the seismic qualification of control and safety rod driving mechanism is carried out. Earthquake excitations under Operational Basis (ORE) and Safe Shutdown condition (SSE) are considered. The time required for the control rod to reach the bottom position in order to shut down the reaction under excited condition is traced out. The maximum displaced positions and extreme stresses in various parts of the system under excitations are evaluated. The system modeled using beam elements. The connections between different parts are modeled through rigid elements. The interaction between various parts are modeled using GAP elements. (author)
REVIEW OF NRC APPROVED DIGITAL CONTROL SYSTEMS ANALYSIS

International Nuclear Information System (INIS)

Markman, D.W.

1999-01-01

Preliminary design concepts for the proposed Subsurface Repository at Yucca Mountain indicate extensive reliance on modern, computer-based, digital control technologies. The purpose of this analysis is to investigate the degree to which the U. S. Nuclear Regulatory Commission (NRC) has accepted and approved the use of digital control technology for safety-related applications within the nuclear power industry. This analysis reviews cases of existing digitally-based control systems that have been approved by the NRC. These cases can serve as precedence for using similar types of digitally-based control technologies within the Subsurface Repository. While it is anticipated that the Yucca Mountain Project (YMP) will not contain control systems as complex as those required for a nuclear power plant, the review of these existing NRC approved applications will provide the YMP with valuable insight into the NRCs review process and design expectations for safety-related digital control systems. According to the YMP Compliance Program Guidance, portions of various NUREGS, Regulatory Guidelines, and nuclear IEEE standards the nuclear power plant safety related concept would be applied to some of the designs on a case-by-case basis. This analysis will consider key design methods, capabilities, successes, and important limitations or problems of selected control systems that have been approved for use in the Nuclear Power industry. An additional purpose of this analysis is to provide background information in support of further development of design criteria for the YMP. The scope and primary objectives of this analysis are to: (1) Identify and research the extent and precedence of digital control and remotely operated systems approved by the NRC for the nuclear power industry. Help provide a basis for using and relying on digital technologies for nuclear related safety critical applications. (2) Identify the basic control architecture and methods of key digital control
46 CFR 61.30-20 - Automatic control and safety tests.

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Automatic control and safety tests. 61.30-20 Section 61... TESTS AND INSPECTIONS Tests and Inspections of Fired Thermal Fluid Heaters § 61.30-20 Automatic control and safety tests. Operational tests and checks of all safety and limit controls, combustion controls...
Seismic analysis of hydraulic control rod driving system

International Nuclear Information System (INIS)

Zheng, Yanhua; Bo, Hanliang; Dong, Duo

2002-01-01

A simplified mathematical model was developed for the Hydraulic Control Rod Driving System (HCRDS) of a 200 MW nuclear heating reactor, which incorporated the design of its chamfer-hole step cylinder, to analyze its seismic response characteristics. The control rod motion was analyzed for different sine-wave vibration loadings on platform vibrator. The vibration frequency domain and the minimum acceleration amplitude of the control rod needed to cause the control rod to step to its next setting were compared with the design acceleration amplitude spectrum. The system design was found to be safety within the calculated limits. The safety margin increased with increasing frequency. (author)
Safety precautions in atomic pile control (1962)

International Nuclear Information System (INIS)

Furet, J.

1962-01-01

We have been led to study the problem of safety in atomic pile control as a result of our participation on the one hand in the planning of C.E.A. atomic piles, and on the other hand in the pile safety sub omission considering atomic pile safety of operational or planned C.E.A. piles. We have thus had to consider the wishes occurring in piles during their operation and also their behaviour in the dynamic state The present work deals mainly with the importance of intrinsic safety devices, with the influence of reactivity variations on the power fluctuations during accidental operation, and with the development of robust and reliable safety appliances. The starting p accident has been especially studied both for low-flux piles where a compromise is necessary between the response time of the safety appliances and the statistical fluctuations and for high lux piles where xenon poisoning has an effect on the lower limit of the velocity of reactivity liberation. The desirability has been stressed of automation as a safety factor in atomic pile control. The details required for an understanding of the diagrams of the apparatus are given. (author) [fr
System safety engineering analysis handbook

Science.gov (United States)

Ijams, T. E.

1972-01-01

The basic requirements and guidelines for the preparation of System Safety Engineering Analysis are presented. The philosophy of System Safety and the various analytic methods available to the engineering profession are discussed. A text-book description of each of the methods is included.

Risk-based configuration control: Application of PSA in improving technical specifications and operational safety

International Nuclear Information System (INIS)

Samanta, P.K.; Kim, I.S.; Vesely, W.E.

1992-01-01

Risk-based configuration control is the management of component configurations using a risk perspective to control risk and assure safety. A configuration, as used here, is a set of component operability statuses that define the state of a nuclear power plant. If the component configurations that have high risk implications do not occur, then the risk from the operation of nuclear power plants would be minimal. The control of component configurations, i.e., the management of component statuses, to minimize the risk from components being unavailable, becomes difficult, because the status of a standby safety system component is often not apparent unless it is tested. Controlling plant configuration from a risk-perspective can provide more direct risk control and also more operational flexibility by allowing looser controls in areas unimportant to risk. Risk-based configuration control approaches can be used to replace parts of nuclear power plant Technical Specifications. With the advances in probabilistic safety assessment (PSA) technology, such approaches to improve Technical Specifications and operational safety are feasible. In this paper, we present an analysis of configuration risks, and a framework for risk-based configuration control to achieve the desired control of risk-significant configurations during plant operation
Safety performance monitoring of autonomous marine systems

International Nuclear Information System (INIS)

Thieme, Christoph A.; Utne, Ingrid B.

2017-01-01

The marine environment is vast, harsh, and challenging. Unanticipated faults and events might lead to loss of vessels, transported goods, collected scientific data, and business reputation. Hence, systems have to be in place that monitor the safety performance of operation and indicate if it drifts into an intolerable safety level. This article proposes a process for developing safety indicators for the operation of autonomous marine systems (AMS). The condition of safety barriers and resilience engineering form the basis for the development of safety indicators, synthesizing and further adjusting the dual assurance and the resilience based early warning indicator (REWI) approaches. The article locates the process for developing safety indicators in the system life cycle emphasizing a timely implementation of the safety indicators. The resulting safety indicators reflect safety in AMS operation and can assist in planning of operations, in daily operational decision-making, and identification of improvements. Operation of an autonomous underwater vehicle (AUV) exemplifies the process for developing safety indicators and their implementation. The case study shows that the proposed process leads to a comprehensive set of safety indicators. It is expected that application of the resulting safety indicators consequently will contribute to safer operation of current and future AMS. - Highlights: • Process for developing safety indicators for autonomous marine systems. • Safety indicators based on safety barriers and resilience thinking. • Location of the development process in the system lifecycle. • Case study on AUV demonstrating applicability of the process.
78 FR 29392 - Embedded Digital Devices in Safety-Related Systems, Systems Important to Safety, and Items Relied...

Science.gov (United States)

2013-05-20

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0098] Embedded Digital Devices in Safety-Related Systems, Systems Important to Safety, and Items Relied on for Safety AGENCY: Nuclear Regulatory Commission. ACTION... (NRC) is issuing for public comment Draft Regulatory Issue Summary (RIS) 2013-XX, ``Embedded Digital...
A Novel Series Connected Batteries State of High Voltage Safety Monitor System for Electric Vehicle Application

Directory of Open Access Journals (Sweden)

Qiang Jiaxi

2013-01-01

Full Text Available Batteries, as the main or assistant power source of EV (Electric Vehicle, are usually connected in series with high voltage to improve the drivability and energy efficiency. Today, more and more batteries are connected in series with high voltage, if there is any fault in high voltage system (HVS, the consequence is serious and dangerous. Therefore, it is necessary to monitor the electric parameters of HVS to ensure the high voltage safety and protect personal safety. In this study, a high voltage safety monitor system is developed to solve this critical issue. Four key electric parameters including precharge, contact resistance, insulation resistance, and remaining capacity are monitored and analyzed based on the equivalent models presented in this study. The high voltage safety controller which integrates the equivalent models and control strategy is developed. By the help of hardware-in-loop system, the equivalent models integrated in the high voltage safety controller are validated, and the online electric parameters monitor strategy is analyzed and discussed. The test results indicate that the high voltage safety monitor system designed in this paper is suitable for EV application.
A novel series connected batteries state of high voltage safety monitor system for electric vehicle application.

Science.gov (United States)

Jiaxi, Qiang; Lin, Yang; Jianhui, He; Qisheng, Zhou

2013-01-01

Batteries, as the main or assistant power source of EV (Electric Vehicle), are usually connected in series with high voltage to improve the drivability and energy efficiency. Today, more and more batteries are connected in series with high voltage, if there is any fault in high voltage system (HVS), the consequence is serious and dangerous. Therefore, it is necessary to monitor the electric parameters of HVS to ensure the high voltage safety and protect personal safety. In this study, a high voltage safety monitor system is developed to solve this critical issue. Four key electric parameters including precharge, contact resistance, insulation resistance, and remaining capacity are monitored and analyzed based on the equivalent models presented in this study. The high voltage safety controller which integrates the equivalent models and control strategy is developed. By the help of hardware-in-loop system, the equivalent models integrated in the high voltage safety controller are validated, and the online electric parameters monitor strategy is analyzed and discussed. The test results indicate that the high voltage safety monitor system designed in this paper is suitable for EV application.
The Evolution of System Safety at NASA

Science.gov (United States)

Dezfuli, Homayoon; Everett, Chris; Groen, Frank

2014-01-01

The NASA system safety framework is in the process of change, motivated by the desire to promote an objectives-driven approach to system safety that explicitly focuses system safety efforts on system-level safety performance, and serves to unify, in a purposeful manner, safety-related activities that otherwise might be done in a way that results in gaps, redundancies, or unnecessary work. An objectives-driven approach to system safety affords more flexibility to determine, on a system-specific basis, the means by which adequate safety is achieved and verified. Such flexibility and efficiency is becoming increasingly important in the face of evolving engineering modalities and acquisition models, where, for example, NASA will increasingly rely on commercial providers for transportation services to low-earth orbit. A key element of this objectives-driven approach is the use of the risk-informed safety case (RISC): a structured argument, supported by a body of evidence, that provides a compelling, comprehensible and valid case that a system is or will be adequately safe for a given application in a given environment. The RISC addresses each of the objectives defined for the system, providing a rational basis for making informed risk acceptance decisions at relevant decision points in the system life cycle.
77 FR 70409 - System Safety Program

Science.gov (United States)

2012-11-26

...-0060, Notice No. 2] 2130-AC31 System Safety Program AGENCY: Federal Railroad Administration (FRA... rulemaking (NPRM) published on September 7, 2012, FRA proposed regulations to require commuter and intercity passenger railroads to develop and implement a system safety program (SSP) to improve the safety of their...
Modelling safety of multistate systems with ageing components

Energy Technology Data Exchange (ETDEWEB)

Kołowrocki, Krzysztof; Soszyńska-Budny, Joanna [Gdynia Maritime University, Department of Mathematics ul. Morska 81-87, Gdynia 81-225 Poland (Poland)

2016-06-08

An innovative approach to safety analysis of multistate ageing systems is presented. Basic notions of the ageing multistate systems safety analysis are introduced. The system components and the system multistate safety functions are defined. The mean values and variances of the multistate systems lifetimes in the safety state subsets and the mean values of their lifetimes in the particular safety states are defined. The multi-state system risk function and the moment of exceeding by the system the critical safety state are introduced. Applications of the proposed multistate system safety models to the evaluation and prediction of the safty characteristics of the consecutive “m out of n: F” is presented as well.
Modelling safety of multistate systems with ageing components

International Nuclear Information System (INIS)

Kołowrocki, Krzysztof; Soszyńska-Budny, Joanna

2016-01-01

An innovative approach to safety analysis of multistate ageing systems is presented. Basic notions of the ageing multistate systems safety analysis are introduced. The system components and the system multistate safety functions are defined. The mean values and variances of the multistate systems lifetimes in the safety state subsets and the mean values of their lifetimes in the particular safety states are defined. The multi-state system risk function and the moment of exceeding by the system the critical safety state are introduced. Applications of the proposed multistate system safety models to the evaluation and prediction of the safty characteristics of the consecutive “m out of n: F” is presented as well.
Experience of creating a multifunctional safety system at the coal mining enterprise

Science.gov (United States)

Reshetnikov, V. V.; Davkaev, K. S.; Korolkov, M. V.; Lyakhovets, M. V.

2018-05-01

The principles of creating multifunctional safety systems (MFSS) based on mathematical models with Markov properties are considered. The applicability of such models for the analysis of the safety of the created systems and their effectiveness is substantiated. The method of this analysis and the results of its testing are discussed. The variant of IFSB implementation in the conditions of the operating coal-mining enterprise is given. The functional scheme, data scheme and operating modes of the MFSS are given. The automated workplace of the industrial safety controller is described.
Programmable electronic safety systems

International Nuclear Information System (INIS)

Parry, R.R.

1993-01-01

Traditionally safety systems intended for protecting personnel from electrical and radiation hazards at particle accelerator laboratories have made extensive use of electromechanical relays. These systems have the advantage of high reliability and allow the designer to easily implement fail-safe circuits. Relay based systems are also typically simple to design, implement, and test. As systems, such as those presently under development at the Superconducting Super Collider Laboratory (SSCL), increase in size, and the number of monitored points escalates, relay based systems become cumbersome and inadequate. The move toward Programmable Electronic Safety Systems is becoming more widespread and accepted. In developing these systems there are numerous precautions the designer must be concerned with. Designing fail-safe electronic systems with predictable failure states is difficult at best. Redundancy and self-testing are prime examples of features that should be implemented to circumvent and/or detect failures. Programmable systems also require software which is yet another point of failure and a matter of great concern. Therefore the designer must be concerned with both hardware and software failures and build in the means to assure safe operation or shutdown during failures. This paper describes features that should be considered in developing safety systems and describes a system recently installed at the Accelerator Systems String Test (ASST) facility of the SSCL
49 CFR 193.2619 - Control systems.

Science.gov (United States)

2010-10-01

... ADMINISTRATION, DEPARTMENT OF TRANSPORTATION (CONTINUED) PIPELINE SAFETY LIQUEFIED NATURAL GAS FACILITIES..., and control systems for internal shutoff valves for bottom penetration tanks must be inspected and...
Safety management of a complex R&D ground operating system

Science.gov (United States)

Connors, J.; Mauer, R. A.

1975-01-01

Report discusses safety program implementation for large R&D operating system. Analytical techniques are defined and suggested as tools for identifying potential hazards and determining means to effectively control or eliminate hazards.
System safety education focused on industrial engineering

Science.gov (United States)

Johnston, W. L.; Morris, R. S.

1971-01-01

An educational program, designed to train students with the specific skills needed to become safety specialists, is described. The discussion concentrates on application, selection, and utilization of various system safety analytical approaches. Emphasis is also placed on the management of a system safety program, its relationship with other disciplines, and new developments and applications of system safety techniques.
Perspective on Secure Development Activities and Features of Safety I and C Systems

International Nuclear Information System (INIS)

Kang, Youngdoo; Yu, Yeong Jin; Kim, Hyungtae; Kwon, Yong il; Park, Yeunsoo; Choo, Jaeyul; Son, Jun Young; Jeong, Choong Heui

2015-01-01

The Enforcement Decree of the Act on Physical Protection and Radiological Emergency (ED-APPRE) was revised December 2013 to include security requirements on computer systems at nuclear facilities to protect those systems against malicious cyber-attacks. It means Cyber-Security-related measures, controls and activities of safety I and C systems against cyber-attacks shall meet the requirements of ED-APPRE. Still regulation upon inadvertent access or non-malicious modifications to the safety I and C systems is covered under the Nuclear Safety Act. The objective of this paper is to propose KINS' regulatory perspective on secure development and features against non-malicious access or modification of safety I and C systems. Secure development activities and features aim to prevent inadvertent and non-malicious access, and to prevent unwanted action from personnel or connected systems for ensuring reliable operation of safety I and C systems. Secure development activities of safety I and C systems are life cycle activities to ensure unwanted, unneeded and undocumented code is not incorporated into the systems. Secure features shall be developed, verified and qualified throughout the development life cycle
Perspective on Secure Development Activities and Features of Safety I and C Systems

Energy Technology Data Exchange (ETDEWEB)

Kang, Youngdoo; Yu, Yeong Jin; Kim, Hyungtae; Kwon, Yong il; Park, Yeunsoo; Choo, Jaeyul; Son, Jun Young; Jeong, Choong Heui [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-05-15

The Enforcement Decree of the Act on Physical Protection and Radiological Emergency (ED-APPRE) was revised December 2013 to include security requirements on computer systems at nuclear facilities to protect those systems against malicious cyber-attacks. It means Cyber-Security-related measures, controls and activities of safety I and C systems against cyber-attacks shall meet the requirements of ED-APPRE. Still regulation upon inadvertent access or non-malicious modifications to the safety I and C systems is covered under the Nuclear Safety Act. The objective of this paper is to propose KINS' regulatory perspective on secure development and features against non-malicious access or modification of safety I and C systems. Secure development activities and features aim to prevent inadvertent and non-malicious access, and to prevent unwanted action from personnel or connected systems for ensuring reliable operation of safety I and C systems. Secure development activities of safety I and C systems are life cycle activities to ensure unwanted, unneeded and undocumented code is not incorporated into the systems. Secure features shall be developed, verified and qualified throughout the development life cycle.
Supplement to safety analysis report. 306-W building operations safety requirement

International Nuclear Information System (INIS)

Richey, C.R.

1979-08-01

The operations safety requirements (OSRs) presented in this report define the conditions, safe boundaries, and management control needed for safely conducting operations with radioactive materials in the Pacific Northwest Laboratory (PNL) 306-W building. The safety requirements are organized in five sections. Safety limits are safety-related process variables that are observable and measurable. Limiting conditions cover: equipment and technical conditions and characteristics of the facility and operations necessary for continued safe operation. Surveillance requirements prescribe the requirements for checking systems and components that are essential to safety. Equipment design controls require that changes to process equipment and systems be independently checked and approved to assure that the changes will have no adverse effect on safety. Administrative controls describe and discuss the organization and administrative systems and procedures to be used for safe operation of the facility. Details of the implementation of the operations safety requirements are prescribed by internal PNL documents such as criticality safety specifications and radiation work procedures
Radiation safety systems at the NSLS

International Nuclear Information System (INIS)

Dickinson, T.

1987-04-01

This report describes design principles that were used to establish the radiation safety systems at the National Synchrotron Light Source. The author described existing safety systems and the history of partial system failures. 1 fig
Development of Reactor Protection System (RPS) in Reactor Digital Instrumentation and Control System (ReDICS)

International Nuclear Information System (INIS)

Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Ridzuan Abdul Mutalib

2013-01-01

RTP Research Reactor are in the process upgraded from analogue control console system to a digital control console system . Upgrade process requires a statistical study to improve safety during reactor operation. RPS was developed to meet the needs of operational safety and at the same time comply with the guidelines set by the IAEA. RPS is in analog and hardware with industry standard interfaced with digital DAC (Data Acquisition and Control) and OWS (Operator Work Station). (author)
Conceptual Design Approach to Implementing Hardware-based Security Controls in Data Communication Systems

International Nuclear Information System (INIS)

Ibrahim, Ahmad Salah; Jung, Jaecheon

2016-01-01

In the Korean Advanced Power Reactor (APR1400), safety control systems network is electrically isolated and physically separated from non-safety systems data network. Unidirectional gateways, include data diode fiber-optic cabling and computer-based servers, transmit the plant safety critical parameters to the main control room (MCR) for control and monitoring processes. The data transmission is only one-way from safety to non-safety. Reverse communication is blocked so that safety systems network is protected from potential cyberattacks or intrusions from non-safety side. Most of commercials off-the-shelf (COTS) security devices are software-based solutions that require operating systems and processors to perform its functions. Field Programmable Gate Arrays (FPGAs) offer digital hardware solutions to implement security controls such as data packet filtering and deep data packet inspection. This paper presents a conceptual design to implement hardware-based network security controls for maintaining the availability of gateway servers. A conceptual design of hardware-based network security controls was discussed in this paper. The proposed design is aiming at utilizing the hardware-based capabilities of FPGAs together with filtering and DPI functions of COTS software-based firewalls and intrusion detection and prevention systems (IDPS). The proposed design implemented a network security perimeter between the DCN-I zone and gateway servers zone. Security control functions are to protect the gateway servers from potential DoS attacks that could affect the data availability and integrity

Conceptual Design Approach to Implementing Hardware-based Security Controls in Data Communication Systems

Energy Technology Data Exchange (ETDEWEB)

Ibrahim, Ahmad Salah; Jung, Jaecheon [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2016-10-15

In the Korean Advanced Power Reactor (APR1400), safety control systems network is electrically isolated and physically separated from non-safety systems data network. Unidirectional gateways, include data diode fiber-optic cabling and computer-based servers, transmit the plant safety critical parameters to the main control room (MCR) for control and monitoring processes. The data transmission is only one-way from safety to non-safety. Reverse communication is blocked so that safety systems network is protected from potential cyberattacks or intrusions from non-safety side. Most of commercials off-the-shelf (COTS) security devices are software-based solutions that require operating systems and processors to perform its functions. Field Programmable Gate Arrays (FPGAs) offer digital hardware solutions to implement security controls such as data packet filtering and deep data packet inspection. This paper presents a conceptual design to implement hardware-based network security controls for maintaining the availability of gateway servers. A conceptual design of hardware-based network security controls was discussed in this paper. The proposed design is aiming at utilizing the hardware-based capabilities of FPGAs together with filtering and DPI functions of COTS software-based firewalls and intrusion detection and prevention systems (IDPS). The proposed design implemented a network security perimeter between the DCN-I zone and gateway servers zone. Security control functions are to protect the gateway servers from potential DoS attacks that could affect the data availability and integrity.
General distributed control system for fusion experiments

International Nuclear Information System (INIS)

Klingner, P.L.; Levings, S.J.; Wilkins, R.W.

1986-01-01

A general control system using distributed LSI-11 microprocessors is being developed. Common software residues in each LSI-11 and is tailored to an application by control specifications downloaded from a host computer. The microprocessors, their control interfaces, and the micro-to-host communications are CAMAC based. The host computer also supports an operator interface, coordination of multiple microprocessors, and utilities to create and maintain the control specifications. Typical applications include monitoring safety interlocks as well as controlling vacuum systems, high voltage charging systems, and diagnostics
Assessment of Primary Production of Horticultural Safety Management Systems of Mushroom Farms in South Africa.

Science.gov (United States)

Dzingirayi, Garikayi; Korsten, Lise

2016-07-01

Growing global consumer concern over food safety in the fresh produce industry requires producers to implement necessary quality assurance systems. Varying effectiveness has been noted in how countries and food companies interpret and implement food safety standards. A diagnostic instrument (DI) for global fresh produce industries was developed to measure the compliancy of companies with implemented food safety standards. The DI is made up of indicators and descriptive grids for context factors and control and assurance activities to measure food safety output. The instrument can be used in primary production to assess food safety performance. This study applied the DI to measure food safety standard compliancy of mushroom farming in South Africa. Ten farms representing almost half of the industry farms and more than 80% of production were independently assessed for their horticultural safety management system (HSMS) compliance via in-depth interviews with each farm's quality assurance personnel. The data were processed using Microsoft Office Excel 2010 and are represented in frequency tables. The diagnosis revealed that the mushroom farming industry had an average food safety output. The farms were implementing an average-toadvanced HSMS and operating in a medium-risk context. Insufficient performance areas in HSMSs included inadequate hazard analysis and analysis of control points, low specificity of pesticide assessment, and inadequate control of suppliers and incoming materials. Recommendations to the industry and current shortcomings are suggested for realization of an improved industry-wide food safety assurance system.
Role of systems safety in maintaining affordable safety in the 1980's

International Nuclear Information System (INIS)

Hollister, H.; Trauth, C.A. Jr.

1979-01-01

Historically, the Department of Energy and its predecessors have used and supported the development of systems safety programs, practices, and principles, finding them by and large adequate, effective, and managerially efficient. Today, attempts are bing made to resolve increasingly complex environmental, safety, and health problems by turning to increasingly complex and detailed regulation as the primary governmental answer. It is increasingly doubtful that such an approach will provide management of these issues and problems that is either effective or efficient. Challenge is issued to those in systems safety to develop and apply systems safety principles and practices more broadly to total operational systems and not just to hardware and to environmental and health protection and not just to safety, so that the total universe of environmental, safety, and health can be managed effectively and efficiently with encouragement of innovation and creativity, using a relatively brief and concise, but adequate, regulatory base
Design of control and safety rod and its drive mechanism of PFBR

International Nuclear Information System (INIS)

Rajan Babu, V.; Govindarajan, S.; Chetal, S.C.

1997-01-01

Control and Safety Rod (CSR) is one of the two types of absorber rods in shutdown systems of PFBR. Control and Safety Rod Drive Mechanism (CSRDM) actuates CSR to have vertical translatory motion in reactor core. The dual responsibilities entrusted on CSR to control reactor power during normal operating condition and to shutdown the reactor by scram action during abnormal condition, necessitate highly reliable design, analysis, testing and surveillance of CSR and CSRDM. The paper discusses on the salient features of CSR and CSRDM and design and analysis of individual sub-assemblies, viz., gripper, scram-release electromagnet, hydraulic dash pot, seals. Also it discusses on the developmental activities proposed and surveillance test requirements. (author)
Operation and safety decision-making support expert system in NPP

International Nuclear Information System (INIS)

Wei Yanhui; Su Desong; Chen Weihua; Zhang Jianbo

2014-01-01

The article first reviewed three operation support systems currently used in NPP: real-time information surveillance system, important equipment surveillance system and plant process control and monitoring system, then presents the structure and function of three expert support sub-systems (intelligent alarm monitoring system, computer-based operating procedure support system, safety information expert decision support system). Finally the article discussed the meaning of a kind of operation decision making support system. (authors)
Systems Safety and Engineering Division

Data.gov (United States)

Federal Laboratory Consortium — Volpe's Systems Safety and Engineering Division conducts engineering, research, and analysis to improve transportation safety, capacity, and resiliency. We provide...
Improved safety of the system 80+TM standard plants design through increased diversity and redundancy of safety systems

International Nuclear Information System (INIS)

Matzie, Regis A.; Carpentino, Frederick L.; Robertson, James E.

1996-01-01

Safely systems in the System 80+ TM Standard Plant are designed with more redundancy, diversity and simplicity than earlier nuclear power plant designs. These gains were accomplished by an evolutionary process that preserved the desirable and proven features in currently operating nuclear plants, while improving reliability and defense-in-depth. The System 80+ safety systems are the primary contributors to a core damage frequency that is more than 100 times lower than 1980's vintage U. S. designs, including the predecessor System 80 R standard nuclear steam supply system (NSSS) design. The System 80+ design includes significant improvements to the safety injection system, emergency feedwater system, shutdown cooling system, containment spray system, reactor coolant gas vent system, and to their vital support systems. These improvements enhance performance for traditional design basis events and significantly reduce the probability of a severe accident. The System 80+ design also incorporates safety systems to mitigate a severe accident. The added systems include the rapid depressurization system, the in-containment refueling water storage tank, the cavity flooding system. These systems fully address the U. S. Nuclear Regulatory Commission's (US NRC) severe accident policy. The System 80+ safety systems are integrated with the System 80+ Nuclear Island (NI) design. The NI general arrangement provides quadrant separation of the safety systems for protection from fire and flooding, and large equipment pull spaces and lay down areas for maintenance. This paper will describe the System 80+ safety systems advanced design features, the improved accident prevention and mitigation capabilities, and startup, operating and maintenance benefits
Design on human supervisory control of safety management for advance NPP

International Nuclear Information System (INIS)

Nugroho, D.H.; Soentono, S.; Taryo, T.; Wahyon, P.I.

2006-01-01

Full text: Full text: Characteristics of an advance NPP related with economic and safety enhancement was represented on capabilities in intelligent control and diagnostic to provide nearly autonomous operation with anticipatory maintenance. An autonomous control system should enable automatic operation while adapting to component faults and system variable upsets. It needs therefore to have many intelligent capabilities, such as modeling, analysis, self-validation, diagnosis and decision. This paper describes a human supervisory control design for nuclear reactor fault management which collaborates between human and autonomous control. The beneficial of collaboration is provided based on belief of information value evaluated from Dempster's rule of evidence. The belief of the collaboration is better compared with single usage. The collaboration was conducted in which agent will autonomously and periodically be conducting surveillance by checking the component abnormalities in the plant if anomalies occur. The anomalies were determined using fault detection module. Thereby the system will be able to conduct preventive maintenance. In the situation of accident happened, hence the system will diagnose to determine the location of component failure autonomously. A human supervisor will then determine the action of decison making based on the prediction result. The decision making will be conducted based on the 4th Sheridan's autonomous level in which the autonomous control will execute the duty autonomously when the plant is in normal condition, or in the predictable accident range. But if the unpredictable accident occurs in the plant, the supervisor will then take over the role to control the plant, and the machine will do what commanded by the supervisor. Sistematically, the system explained before was represented by Traveling Salesman Problem-based surveillance, modified ART-2 artificial neural networks-based fault detection and Bayesian Networks-based fault
Applications of modern control systems in nuclear power plants

International Nuclear Information System (INIS)

Wilhelm, H.

1980-01-01

A new generation of automation and control systems are currently becoming commercially available in the power plant market which, because of their incorporation of microprocessors and bus data transmission systems, represent a major step forward in innovation. The application of these systems meets today's requirements and solutions, for the number of measurements to be performed has increased five or sixfold in the past few years, and the number of drive systems to be controlled has doubled or even tripled. Requirements to be met by process management systems have become vastly more complicated: peak load operation, short startup times, improved communication, and rising safety and reliability requirements, especially in nuclear power plants. Control concepts have been developed for the area relevant to reactor safety and for the whole of the plant, which make full use of the possibilities offered by plant systems. More stringent demands must be met especially in the areas of handling, communication, testing capability, improved function, and flexibility and modular design in the safety sector. (orig.) [de
Product Engineering Class in the Software Safety Risk Taxonomy for Building Safety-Critical Systems

Science.gov (United States)

Hill, Janice; Victor, Daniel

2008-01-01

When software safety requirements are imposed on legacy safety-critical systems, retrospective safety cases need to be formulated as part of recertifying the systems for further use and risks must be documented and managed to give confidence for reusing the systems. The SEJ Software Development Risk Taxonomy [4] focuses on general software development issues. It does not, however, cover all the safety risks. The Software Safety Risk Taxonomy [8] was developed which provides a construct for eliciting and categorizing software safety risks in a straightforward manner. In this paper, we present extended work on the taxonomy for safety that incorporates the additional issues inherent in the development and maintenance of safety-critical systems with software. An instrument called a Software Safety Risk Taxonomy Based Questionnaire (TBQ) is generated containing questions addressing each safety attribute in the Software Safety Risk Taxonomy. Software safety risks are surfaced using the new TBQ and then analyzed. In this paper we give the definitions for the specialized Product Engineering Class within the Software Safety Risk Taxonomy. At the end of the paper, we present the tool known as the 'Legacy Systems Risk Database Tool' that is used to collect and analyze the data required to show traceability to a particular safety standard
LANSCE personnel access control system (PACS)

International Nuclear Information System (INIS)

Sturrock, J.C.; Gallegos, F.R.; Hall, M.J.

1997-01-01

The Radiation Security System (RSS) at the Los Alamos Neutron Science Center (LANSCE) provides personnel protection from prompt radiation due to accelerated beam. The Personnel Access Control System (PACS) is a component of the RSS that is designed to prevent personnel access to areas where prompt radiation is a hazard. PACS was designed to replace several older personnel safety systems (PSS) with a single modem unified design. Lessons learned from the operation over the last 20 years were incorporated into a redundant sensor, single-point failure safe, fault tolerant, and tamper-resistant system that prevents access to the beam areas by controlling the access keys and beam stoppers. PACS uses a layered philosophy to the physical and electronic design. The most critical assemblies are battery backed up, relay logic circuits; less critical devices use Programmable Logic Controllers (PLCs) for timing functions and communications. Outside reviewers have reviewed the operational safety of the design. The design philosophy, lessons learned, hardware design, software design, operation, and limitations of the device are described
Resolution of Unresolved Safety Issue A-48, ''Hydrogen control measures and effects of hydrogen burns on safety equipment''

International Nuclear Information System (INIS)

Ferrell, C.M.; Soffer, L.

1989-09-01

Unresolved Safety Issue (USI) A-48 arose as a result of the large amount of hydrogen generated and burned within containment during the Three Mile Island accident. This issue covers hydrogen control measures for recoverable degraded-core accidents for all boiling-water reactors (BWRs) and those pressurized-water reactors (PWRs) with ice-condenser containments. The Commission and the nuclear industry have sponsored extensive research in this area, which has led to significant revision of the Commission's hydrogen control regulations, given in Title 10, Code of Federal Regulations, Part 50 (10 CFR 50), Section 50.44. BWRs having Mark I and II containments are presently required to operate with inerted containment atmospheres that effectively prevent hydrogen combustion. BWRs with Mark III containments and PWRs with ice-condenser containments are now required to be equipped with hydrogen control systems to protect containment integrity and safety systems inside containment. Industry has chosen to use hydrogen igniter systems to burn hydrogen produced in a controlled fashion to prevent damage. An independent review by a Committee of the National Research Council concluded that, for most accident scenarios, current regulatory requirements make it highly unlikely that hydrogen detonation would be the cause of containment failure. On the basis of the extensive research effort conducted and current regulatory requirements, including their implementation, the staff concludes that no new regulatory guidance on hydrogen control for recoverable degraded-core accidents for these types of plants is necessary and that USI A-48 is resolved
The Health and Safety Executive's regulatory framework for control of nuclear criticality safety

International Nuclear Information System (INIS)

Smith, K.; Simister, D.N.

1991-01-01

In the United Kingdom the Health and Safety at Work Act, 1974 is the main legal instrument under which risks to people from work activities are controlled. Certain sections of the Nuclear Installations Act, 1965 which deal with the licensing of nuclear sites and the regulatory control of risks arising from them, including the risk from accidental criticality, are relevant statutory provisions of the Health and Safety at Work Act. The responsibility for safety rests with the operator who has to make and implement arrangements to prevent accidental criticality. The adequacy of these arrangements must be demonstrated in a safety case to the regulatory authorities. Operators are encouraged to treat each plant on its own merits and develop the safety case accordingly. The Nuclear Installations Inspectorate (NII), for its part, assesses the adequacy of the operator's safety case against the industry's own standards and criteria, but more particularly against the NII's safety assessment principles and guides, and international standards. Risks should be made as low as reasonably practicable. Generally, the NII seeks improvements in safety using an enforcement policy which operates at a number of levels, ranging from persuasion through discussion to the ultimate deterrent of withdrawal of a site licence. This paper describes the role of the NII, which includes a specialist criticality expertise, within the Health and Safety Executive, in regulating the nuclear sites from the criticality safety viewpoint. (Author)
Protecting worker health and safety using remote handling systems

International Nuclear Information System (INIS)

Dennison, D.K.; Merrill, R.D.; Reed, R.K.

1995-03-01

Lawrence Livermore National Laboratory (LLNL) is currently developing and installing two large-scale, remotely controlled systems for use in improving worker health and safety by minimizing exposure to hazardous and radioactive materials. The first system is a full-scale liquid feed system for use in delivering chemical reagents to LLNL's existing aqueous low-level radioactive and mixed waste treatment facility (Tank Farm). The Tank Farm facility is used to remove radioactive and toxic materials in aqueous wastes prior to discharge to the City of Livermore Water Reclamation Plant (LWRP), in accordance with established discharge limits. Installation of this new reagent feed system improves operational safety and process efficiency by eliminating the need to manually handle reagents used in the treatment processes. This was done by installing a system that can inject precisely metered amounts of various reagents into the treatment tanks and can be controlled either remotely or locally via a programmable logic controller (PLC). The second system uses a robotic manipulator to remotely handle, characterize, process, sort, and repackage hazardous wastes containing tritium. This system uses an IBM-developed gantry robot mounted within a special glove box enclosure designed to isolate tritiated wastes from system operators and minimize the potential for release of tritium to the atmosphere. Tritiated waste handling is performed remotely, using the robot in a teleoperational mode for one-of-a-kind functions and in an autonomous mode for repetitive operations. The system is compatible with an existing portable gas cleanup unit designed to capture any gas-phase tritium inadvertently released into the glove box during waste handling
Optimal Design of Safety Instrumented Systems for Pressure Control of Methanol Separation Columns in the Bisphenol a Manufacturing Process

Directory of Open Access Journals (Sweden)

In-Bok Lee

2016-12-01

Full Text Available A bisphenol A production plant possesses considerable potential risks in the top of the methanol separation column, as pressurized acetone, methanol, and water are processed at an elevated temperature, especially in the event of an abnormal pressure increase due to a sudden power outage. This study assesses the potential risks in the methanol separation column through hazard and operability assessments and evaluates the damages in the case of fire and explosion accident scenarios. The study chooses three leakage scenarios: a 5-mm puncture on the methanol separation column, a 50-mm diameter fracture of a discharge pipe and a catastrophic rupture, and, simulated using Phast (Ver. 6.531, the concentration distribution of scattered methanol, thermal radiation distribution of fires, and overpressure distribution of vapor cloud explosions. Implementation of a safety-instrumented system equipped with two-out-of-three voting as a safety measure can detect overpressure at the top of the column and shut down the main control valve and the emergency shutoff valve simultaneously. By applying a safety integrity level of three, the maximal release volume of the safety relief valve can be reduced and, therefore, the design capacity of the flare stack can also be reduced. Such integration will lead to improved safety at a reduced cost.
Propose Reactor Control and Monitoring System for RTP

International Nuclear Information System (INIS)

Mohd Sabri Minhat; Izhar Abu Hussin; Mohd Idris Taib; Mohd Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha

2011-01-01

Reactor control and monitoring system is a one of the important features used in reactor. The control and monitoring must come together to provide safety, excellent performance and reliable in nuclear reactor technology application. Objectives of this technical paper are to design and propose reactor control system and reactor monitoring system in Research Reactor (RTP) for Reactor Upgrading Project. (author)
Formal development and verification of a distributed railway control system

DEFF Research Database (Denmark)

Haxthausen, Anne Elisabeth; Peleska, J.

2000-01-01

specifications which are transformed into directly implementable distributed control processes by applying a series of refinement and verification steps. Concrete safety requirements are derived from an abstract version that can be easily validated with respect to soundness and completeness. Complexity......The authors introduce the concept for a distributed railway control system and present the specification and verification of the main algorithm used for safe distributed control. Our design and verification approach is based on the RAISE method, starting with highly abstract algebraic...... is further reduced by separating the system model into a domain model and a controller model. The domain model describes the physical system in absence of control and the controller model introduces the safety-related control mechanisms as a separate entity monitoring observables of the physical system...
Software system safety

Science.gov (United States)

Uber, James G.

1988-01-01

Software itself is not hazardous, but since software and hardware share common interfaces there is an opportunity for software to create hazards. Further, these software systems are complex, and proven methods for the design, analysis, and measurement of software safety are not yet available. Some past software failures, future NASA software trends, software engineering methods, and tools and techniques for various software safety analyses are reviewed. Recommendations to NASA are made based on this review.
A history and overview of the safety parameter display system concept

International Nuclear Information System (INIS)

Joyce, J.P.; Lapinsky, G.W.

1983-01-01

Inquiries into the accident at the Three Mile Island Nuclear Power Plant Unit 2, on March 28, 1979 brought to public attention the need to improve operators' capabilities to interact with the systems under their control. Recommendations ran the full gamut of human/machine interaction, from improvements in training and procedures to improvements in control and display hardware in the control room. This presentation briefly traces the history and development of a display concept that evolved in the post-TMI era, the Safety Parameter Display System or SPDS. The SPDS is intended to function as a detection aid for control room operators, providing an integrated overview of significant plant parameters. The purpose of this report is to describe the general concept of SPDS, its history, and its current regulatory status. A review of NRC guidance documents is included, as well as a discussion of NRC requirements placed on the SPDS. The presentation concludes with an outline of the NRC staff review process for safety parameter display systems and a synopsis of the results of generic SPDS reviews performed thus far

Safety case development with SBVR-based controlled language

NARCIS (Netherlands)

Luo, Y.; van den Brand, M.G.J.; Kiburse, A.; Desfray, P.; Philipe, J.; Hammoudi, S.; Pires, L.F.

2015-01-01

Safety case development is highly recommended by some safety standards to justify the safety of a system. The Goal Structuring Notation (GSN) is a popular approach to construct a safety case. However, the content of the safety case elements, such as safety claims, is in natural language. Therefore,
Probabilistic safety criteria at the safety function/system level

International Nuclear Information System (INIS)

1989-09-01

A Technical Committee Meeting was held in Vienna, Austria, from 26-30 January 1987. The objectives of the meeting were: to review the national developments of PSC at the level of safety functions/systems including future trends; to analyse basic principles, assumptions, and objectives; to compare numerical values and the rationale for choosing them; to compile the experience with use of such PSC; to analyse the role of uncertainties in particular regarding procedures for showing compliance. The general objective of establishing PSC at the level of safety functions/systems is to provide a pragmatic tool to evaluate plant safety which is placing emphasis on the prevention principle. Such criteria could thus lead to a better understanding of the importance to safety of the various functions which have to be performed to ensure the safety of the plant, and the engineering means of performing these functions. They would reflect the state-of-the-art in modern PSAs and could contribute to a balance in system design. This report, prepared by the participants of the meeting, reviews the current status and future trends in the field and should assist Member States in developing their national approaches. The draft of this document was also submitted to INSAG to be considered in its work to prepare a document on safety principles for nuclear power plants. Five papers presented at the meeting are also included in this publication. A separate abstract was prepared for each of these papers. Refs, figs and tabs
INMACS: Operating experience of a mature, computer-assisted control system for nuclear material inventory and criticality safety

International Nuclear Information System (INIS)

Ross, A.M.

1983-01-01

This paper describes the operating experience of INMACS, the Integrated Nuclear Material Accounting and Control System used in the Recycle Fuel Fabrication Laboratories at Chalk River. Since commissioning was completed in 1977, INMACS has checked and recorded approximately 3000 inventory-related transactions involved in fabricating thermal-recycle fuels of (U,Pu)0 2 and (Th,Pu)0 2 . No changes have been necessary to INMACS programs that are used by laboratory staff when moving or processing nuclear material. The various utility programs have allowed efficient management and surveillance of the INMACS data base. Hardware failures and the nuisance of system unavailability at the laboratory terminals have been minimized by regular preventative maintenance. The original efforts in the design and rigorous testing of programs have helped INMACS to be accepted enthusiastically by old and new staff of the laboratories. The work required for nuclear material inventory control is done efficiently and in an atmosphere of safety
Safety Impacts of the Actuated Signal Control at Urban Intersections

Directory of Open Access Journals (Sweden)

Sang Hyuk Lee

2016-02-01

Full Text Available To reduce travel time, the actuated signal controls have been implemented at urban intersections. However, the safety impacts of actuated signal controls thus far have rarely been examined. In this assessment of the safety impact of urban intersections with semi-actuated signal controls, the safety performance functions and EB approaches were applied. The semi-actuated signal controls have increased injuries and total crashes in all crash types by around 5.9% and 3.8%, respectively. Regarding the most common crash types, such as angle, sideswipe & rear-end, and head-on crashes, semi-actuated signal controls have been seen to decrease injuries by 7.7%. Total crashes have been reduced by over 9.2% through the use of semi-actuated signal controls. This may be result of optimal signal timings considering traffic conditions during peak time periods. In conclusion, safety impact factors which have been established in this study can be used to improve safety and minimize travel times using semi-actuated signal controls.
Reactor Safety Assessment System

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category
Reactor safety assessment system

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category
Safety systems and safety analysis of the Qinshan phase III CANDU nuclear power plant

International Nuclear Information System (INIS)

Cai Jianping; Shen Sen; Barkman, N.

1999-01-01

The author introduces the Canadian nuclear reactor safety philosophy and the Qinshan Phase III CANDU NPP safety systems and safety analysis, which are designed and performed according to this philosophy. The concept of 'defence-in-depth' is a key element of the Canadian nuclear reactor safety philosophy. The design concepts of redundancy, diversity, separation, equipment qualification, quality assurance, and use of appropriate design codes and standards are adopted in the design. Four special safety systems as well as a set of reliable safety support systems are incorporated in the design of Qinshan phase III CANDU for accident mitigation. The assessment results for safety systems performance show that the fundamental safety criteria for public dose, and integrity of fuel, channels and the reactor building, are satisfied
Application of Safety Instrumented System (SIS) approach in older nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Nasimi, Elnara; Gabbar, Hossam A., E-mail: hossam.gabbar@uoit.ca

2016-05-15

Highlights: • Study Safety Instrumented System (SIS) design for older nuclear power plant. • Apply SIS on Reheater Drains (RD) system. • Apply IEC 61508/61511 to design safety system. • Evaluate risk reduction based on proposed SIS design. - Abstract: In order to remain economically effective and financially profitable, the modern industries have to take their safety culture to a higher level and consider production losses in addition to simple accident prevention techniques. Ideally, compliance with safety requirements start during early design stages, but in some older facilities provisions for Safety Instrumented Systems (SIS) may not have been originally included. In this paper, a case study of a Reheater Drains (RD) system is used to illustrate such an example. Frequent failures of tank level controller lead to transients where the operation of shutting down RD pumps requires operators to manually isolate the quenching water and to close the main steam admission valves. Water in this system is at saturation temperature for the reheater steam side pressure, and any manual operation of the system is highly undesirable due to hazards of working with wet steam at approximately 758 kPa(g) pressure, preheated to 237 °C. Additionally, losses of inventory are highly undesirable as well and challenge other systems in the plant. In this paper, it is suggested that RD system can benefit from installation of an independent SIS system in order to address current challenges. This idea is being explored using IEC 61508 framework for “Functional safety of electrical/electronic/programmable electronic safety-related systems” to provide assurance that the SIS will offer the necessary risk reduction required to achieve required safety for the equipment.
Reference to the Safety Engineering Undergraduate Courses to Improve the Subjects and Contents of the Certified Safety Engineer Qualification and Examination System of China

OpenAIRE

Haibin Qiu; Shanghong Shi; Tingdi Zhao; Yiwei Qiao; Jiangshi Zhang

2013-01-01

The aim of this paper is to recommend that the subjects and contents of certified safety engineers use safety engineering undergraduate curriculum system for reference. Human resources play an important role in accident prevention and loss control. Education on safety engineering develops quickly in China. Moreover, the State Administration of Work Safety and the National Human Resources and Social Security Ministry have implemented a certified safety engineer qualification and examination sy...
Food safety performance indicators to benchmark food safety output of food safety management systems.

Science.gov (United States)

Jacxsens, L; Uyttendaele, M; Devlieghere, F; Rovira, J; Gomez, S Oses; Luning, P A

2010-07-31

There is a need to measure the food safety performance in the agri-food chain without performing actual microbiological analysis. A food safety performance diagnosis, based on seven indicators and corresponding assessment grids have been developed and validated in nine European food businesses. Validation was conducted on the basis of an extensive microbiological assessment scheme (MAS). The assumption behind the food safety performance diagnosis is that food businesses which evaluate the performance of their food safety management system in a more structured way and according to very strict and specific criteria will have a better insight in their actual microbiological food safety performance, because food safety problems will be more systematically detected. The diagnosis can be a useful tool to have a first indication about the microbiological performance of a food safety management system present in a food business. Moreover, the diagnosis can be used in quantitative studies to get insight in the effect of interventions on sector or governmental level. Copyright 2010 Elsevier B.V. All rights reserved.
Safety-critical Java for embedded systems

DEFF Research Database (Denmark)

Schoeberl, Martin; Dalsgaard, Andreas Engelbredt; Hansen, René Rydhof

2016-01-01

This paper presents the motivation for and outcomes of an engineering research project on certifiable Javafor embedded systems. The project supports the upcoming standard for safety-critical Java, which defines asubset of Java and libraries aiming for development of high criticality systems....... The outcome of this projectinclude prototype safety-critical Java implementations, a time-predictable Java processor, analysis tools formemory safety, and example applications to explore the usability of safety-critical Java for this applicationarea. The text summarizes developments and key contributions...
Concept of safety related I and C and power supply systems in the passive safety concept of the HTR-module

International Nuclear Information System (INIS)

Juengst, U.

1990-01-01

The main motivation for the passive safety concepts is to gain a better quality of safety or at least to achieve higher public acceptance for nuclear power plants. This strategy has been introduced into the European Fast Reactor (EER), a common project of France, UK and Germany is applied stringently to the German high-temperature gas-cooled reactor ''HTR - Module''. The following fields are briefly described in the paper: Safety design features of the HTR - Module, overview of I and C concept, reactor protection system, emergency control room, power supply concept, system arrangement and protection against external hazards, accidents sequence of station black-out. (author). 3 figs
Retrofit of safety and control systems in nuclear power plants

International Nuclear Information System (INIS)

Keiper, J.T.; Fassett, G.B.

1986-01-01

The modularity, compactness, compatibility, and licensability of the microcontrol system make it a cost-effective approach to obtain the benefits of digital control technology in the retrofit of nuclear power plants. Retrofit of individual loops or complete systems can be scheduled to meet the operational needs of the plant. The existing racks, panels, and cable systems can be utilized to the maximum extent to minimize the installed cost. Future expansion to total plant control or plant management is supported by the network communication module or gateway. The microcontrol module provides benefits now in improved operation, and future benefits in planned, controlled upgrading
AEC controlled area safety program

Energy Technology Data Exchange (ETDEWEB)

Hendricks, D W [Nevada Operations Office, Atomic Energy Commission, Las Vegas, NV (United States)

1969-07-01

The detonation of underground nuclear explosives and the subsequent data recovery efforts require a comprehensive pre- and post-detonation safety program for workers within the controlled area. The general personnel monitoring and environmental surveillance program at the Nevada Test Site are presented. Some of the more unusual health-physics aspects involved in the operation of this program are also discussed. The application of experience gained at the Nevada Test Site is illustrated by description of the on-site operational and safety programs established for Project Gasbuggy. (author)
AEC controlled area safety program

International Nuclear Information System (INIS)

Hendricks, D.W.

1969-01-01

The detonation of underground nuclear explosives and the subsequent data recovery efforts require a comprehensive pre- and post-detonation safety program for workers within the controlled area. The general personnel monitoring and environmental surveillance program at the Nevada Test Site are presented. Some of the more unusual health-physics aspects involved in the operation of this program are also discussed. The application of experience gained at the Nevada Test Site is illustrated by description of the on-site operational and safety programs established for Project Gasbuggy. (author)
Analysis of automobile’s automatic control systems for the hill climbing start

Directory of Open Access Journals (Sweden)

Valeriy I. Klimenko

2014-12-01

Full Text Available To improve road safety while driving on the rise, facilitating the driver’s activity the automobile industry leaders are introducing automatic hill-hold control systems into the car design. This study purpose relates to the existing automatic start control systems’ design analysis. Analyzed are the existing design developments of automatic hill start assist control systems applied for driving at the start of the climbing. The effected research allows to select the scheme for further development of start driving automatic control systems. Further improvement of driving control systems and primarily the driver assistance hill-hold control systems is necessary to increase both the driving comfort and the traffic safety.
Compartmentalized safety coolant injection system

International Nuclear Information System (INIS)

Johnson, F.T.

1983-01-01

A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)
FOOD QUALITY MANAGEMENT AND SAFETY

OpenAIRE

Rizwana Khatoon; Debkumar Chakraborty; R.C. Chandni; Amar Sankar; A.V. Raghu

2017-01-01

Food safety system mainly focuses on identifying and preventing hazards that may lead product to deteriorate. The main important of manufacturing practice is a system that ensures that products meet food safety, quality and legal requirements. The hazard analysis and critical control point system, applies to food safety management, uses the approach of controlling critical points in food handling to prevent food safety problems. Besides enhancing food safety, other benefits of applying HACCP ...
ESSAA: Embedded system safety analysis assistant

Science.gov (United States)

Wallace, Peter; Holzer, Joseph; Guarro, Sergio; Hyatt, Larry

1987-01-01

The Embedded System Safety Analysis Assistant (ESSAA) is a knowledge-based tool that can assist in identifying disaster scenarios. Imbedded software issues hazardous control commands to the surrounding hardware. ESSAA is intended to work from outputs to inputs, as a complement to simulation and verification methods. Rather than treating the software in isolation, it examines the context in which the software is to be deployed. Given a specified disasterous outcome, ESSAA works from a qualitative, abstract model of the complete system to infer sets of environmental conditions and/or failures that could cause a disasterous outcome. The scenarios can then be examined in depth for plausibility using existing techniques.
Safety assessment for Generation IV nuclear systems

International Nuclear Information System (INIS)

Leahy, T.J.

2012-01-01

The Generation IV International Forum (GIF) Risk and Safety Working Group (RSWG) was created to develop an effective approach for the safety of Generation IV advanced nuclear energy systems. Recent RSWG work has focused on the definition of an integrated safety assessment methodology (ISAM) for evaluating the safety of Generation IV systems. ISAM is an integrated 'tool-kit' consisting of 5 analytical techniques that are available and matched to appropriate stages of Generation IV system concept development: 1) qualitative safety features review - QSR, 2) phenomena identification and ranking table - PIRT, 3) objective provision tree - OPT, 4) deterministic and phenomenological analyses - DPA, and 5) probabilistic safety analysis - PSA. The integrated methodology is intended to yield safety-related insights that help actively drive the evolving design throughout the technology development cycle, potentially resulting in enhanced safety, reduced costs, and shortened development time

[Establishment of Quality Control System of Nucleic Acid Detection for Ebola Virus in Sierra Leone-China Friendship Biological Safety Laboratory].

Science.gov (United States)

Wang, Qin; Zhang, Yong; Nie, Kai; Wang, Huanyu; Du, Haijun; Song, Jingdong; Xiao, Kang; Lei, Wenwen; Guo, Jianqiang; Wei, Hejiang; Cai, Kun; Wang, Yanhai; Wu, Jiang; Gerald, Bangura; Kamara, Idrissa Laybohr; Liang, Mifang; Wu, Guizhen; Dong, Xiaoping

2016-03-01

The quality control process throughout the Ebola virus nucleic acid detection in Sierra Leone-China Friendship Biological Safety Laboratory (SLE-CHN Biosafety Lab) was described in detail, in order to comprehensively display the scientific, rigorous, accurate and efficient practice in detection of Ebola virus of first batch detection team in SLE-CHN Biosafety Lab. Firstly, the key points of laboratory quality control system was described, including the managements and organizing, quality control documents and information management, instrument, reagents and supplies, assessment, facilities design and space allocation, laboratory maintenance and biosecurity. Secondly, the application of quality control methods in the whole process of the Ebola virus detection, including before the test, during the test and after the test, was analyzed. The excellent and professional laboratory staffs, the implementation of humanized management are the cornerstone of the success; High-level biological safety protection is the premise for effective quality control and completion of Ebola virus detection tasks. And professional logistics is prerequisite for launching the laboratory diagnosis of Ebola virus. The establishment and running of SLE-CHN Biosafety Lab has landmark significance for the friendship between Sierra Leone and China, and the lab becomes the most important base for Ebola virus laboratory testing in Sierra Leone.
Safety Supervisory Strategy for an Upper-Limb Rehabilitation Robot Based on Impedance Control

Directory of Open Access Journals (Sweden)

Lizheng Pan

2013-02-01

Full Text Available User security is an important consideration for robots that interact with humans, especially for upper-limb rehabilitation robots, during the use of which stroke patients are often more susceptible to injury. In this paper, a novel safety supervisory control method incorporating fuzzy logic is proposed so as to guarantee the impaired limb's safety should an emergency situation occur and the robustness of the upper-limb rehabilitation robot control system. Firstly, a safety supervisory fuzzy controller (SSFC was designed based on the impaired-limb's real-time physical state by extracting and recognizing the impaired-limb's tracking movement features. Then, the proposed SSFC was used to automatically regulate the desired force either to account for reasonable disturbance resulting from pose or position changes or to respond in adequate time to an emergency based on an evaluation of the impaired-limb's physical condition. Finally, a position-based impedance controller was implemented to achieve compliance between the robotic end-effector and the impaired limb during the robot-assisted rehabilitation training. The experimental results show the effectiveness and potential of the proposed method for achieving safety and robustness for the rehabilitation robot.
Control system security in nuclear power plant

International Nuclear Information System (INIS)

Li Jianghai; Huang Xiaojin

2012-01-01

The digitalization and networking of control systems in nuclear power plants has brought significant improvements in system control, operation and maintenance. However, the highly digitalized control system also introduces additional security vulnerabilities. Moreover, the replacement of conventional proprietary systems with common protocols, software and devices makes these vulnerabilities easy to be exploited. Through the interaction between control systems and the physical world, security issues in control systems impose high risks on health, safety and environment. These security issues may even cause damages of critical infrastructures and threaten national security. The importance of control system security by reviewing several control system security incidents that happened in nuclear power plants was showed in recent years. Several key difficulties in addressing these security issues were described. Finally, existing researches on control system security and propose several promising research directions were reviewed. (authors)
Final Technical Report on Quantifying Dependability Attributes of Software Based Safety Critical Instrumentation and Control Systems in Nuclear Power Plants

International Nuclear Information System (INIS)

Smidts, Carol; Huang, Fuqun; Li, Boyuan; Li, Xiang

2016-01-01

With the current transition from analog to digital instrumentation and control systems in nuclear power plants, the number and variety of software-based systems have significantly increased. The sophisticated nature and increasing complexity of software raises trust in these systems as a significant challenge. The trust placed in a software system is typically termed software dependability. Software dependability analysis faces uncommon challenges since software systems' characteristics differ from those of hardware systems. The lack of systematic science-based methods for quantifying the dependability attributes in software-based instrumentation as well as control systems in safety critical applications has proved itself to be a significant inhibitor to the expanded use of modern digital technology in the nuclear industry. Dependability refers to the ability of a system to deliver a service that can be trusted. Dependability is commonly considered as a general concept that encompasses different attributes, e.g., reliability, safety, security, availability and maintainability. Dependability research has progressed significantly over the last few decades. For example, various assessment models and/or design approaches have been proposed for software reliability, software availability and software maintainability. Advances have also been made to integrate multiple dependability attributes, e.g., integrating security with other dependability attributes, measuring availability and maintainability, modeling reliability and availability, quantifying reliability and security, exploring the dependencies between security and safety and developing integrated analysis models. However, there is still a lack of understanding of the dependencies between various dependability attributes as a whole and of how such dependencies are formed. To address the need for quantification and give a more objective basis to the review process -- therefore reducing regulatory uncertainty
Evaluating a Website to Teach Children Safety with Dogs: A Randomized Controlled Trial

Directory of Open Access Journals (Sweden)

David C. Schwebel

2016-12-01

Full Text Available Dog bites represent a significant threat to child health. Theory-driven interventions scalable for broad dissemination are sparse. A website was developed to teach children dog safety via increased knowledge, improved cognitive skills in relevant domains, and increased perception of vulnerability to bites. A randomized controlled trial was conducted with 69 children aged 4–5 randomly assigned to use the dog safety website or a control transportation safety website for ~3 weeks. Assessment of dog safety knowledge and behavior plus skill in three relevant cognitive constructs (impulse control, noticing details, and perspective-taking was conducted both at baseline and following website use. The dog safety website incorporated interactive games, instructional videos including testimonials, a motivational rewards system, and messaging to parents concerning child lessons. Our results showed that about two-thirds of the intervention sample was not adherent to website use at home, so both intent-to-treat and per-protocol analyses were conducted. Intent-to-treat analyses yielded mostly null results. Per-protocol analyses suggested children compliant to the intervention protocol scored higher on knowledge and recognition of safe behavior with dogs following the intervention compared to the control group. Adherent children also had improved scores post-intervention on the cognitive skill of noticing details compared to the control group. We concluded that young children’s immature cognition can lead to dog bites. Interactive eHealth training on websites shows potential to teach children relevant cognitive and safety skills to reduce risk. Compliance to website use is a challenge, and some relevant cognitive skills (e.g., noticing details may be more amenable to computer-based training than others (e.g., impulse control.
Almaraz ovation control system security

International Nuclear Information System (INIS)

Madronal Rodriguez, E.; Anderson, E.; Jimenez Diaz, J.; Carrasco Mateos, J. A.

2013-01-01

Improving the security of a plant's Distributed Control System (DCS) is an important consideration for plant safety and profitability, as well as the necessity to comply with the regulation. The U.S. Nuclear Regulatory Commission has produced Regulatory Guide (RG) 5.71, and the Nuclear Energy Institute (NEI) has produced NEI 08-09 to assist plants in meeting 10 CFR 73.54, Protection of digital computer and communication systems and networks. These requirements, which address the establishment, implementation and maintenance of a cyber security program, present challenges to ensure that safety, security and emergency preparedness functions of nuclear facilities are not negatively impacted by the vulnerability scanning and testing process.
Experience of digital control systems in Scandinavian BWRs

International Nuclear Information System (INIS)

Rydahl, I.

1989-01-01

Since 1984 digital control systems have been in operation in various Scandinavian BWRs. Examples of such digital control systems are: dual microprocessor based system for complete control of radwaste plant, three channel recirculation control system, and three channel feedwater control system. This paper describes Swedish development from one channel through three channel analog control systems to digital systems. The author describes experience of digital control systems during design, testing, commissioning and operation. The main benefits of digital compared with analog technology are discussed. Especially the outstanding facility of using a built-in process simulator for commissioning and tuning. The use of digital technology in nuclear safety system and future plans are dealt with
Evaluation of safety implications of control systems in LWR nuclear power plants

International Nuclear Information System (INIS)

Szukiewicz, A.J.

1989-06-01

An in-depth evaluation was performed on non-safety-related control systems (see Section 1) that are typically used during normal plant operation on four nuclear steam supply system plants: a General Electric Company boiling-water reactor, a Westinghouse 3-loop pressurized-water reactor (PWR), a Babcock ampersand Wilcox Co. (B ampersand W) once-through steam generator PWR, and a Combustion Engineering PWR design. A study was also conducted to determine the generic applicability of the results to the class of plants represented by the specific plants analyzed. Generic conclusions were then developed. Steam generator and reactor vessel overfill events and reactor vessel overcooling events were identified as major classes of events having the potential to be more severe than previously analyzed. Specific substasks of this issue were to study these events to determine the need for preventive and/or mitigating design measures. This report describes the technical studies performed by the laboratories, the NRC staff assessment of the results, the generic applicability of the evaluations, and the technical findings resulting from these studies. This final report contains the staff's responses to, and resolution of, the public comments that were solicited and received before September 16,1988, in response to the draft reports issued for public comment on May 27, 1988. 39 refs, 1 fig., 7 tabs
Industrial Personal Computer based Display for Nuclear Safety System

International Nuclear Information System (INIS)

Kim, Ji Hyeon; Kim, Aram; Jo, Jung Hee; Kim, Ki Beom; Cheon, Sung Hyun; Cho, Joo Hyun; Sohn, Se Do; Baek, Seung Min

2014-01-01

The safety display of nuclear system has been classified as important to safety (SIL:Safety Integrity Level 3). These days the regulatory agencies are imposing more strict safety requirements for digital safety display system. To satisfy these requirements, it is necessary to develop a safety-critical (SIL 4) grade safety display system. This paper proposes industrial personal computer based safety display system with safety grade operating system and safety grade display methods. The description consists of three parts, the background, the safety requirements and the proposed safety display system design. The hardware platform is designed using commercially available off-the-shelf processor board with back plane bus. The operating system is customized for nuclear safety display application. The display unit is designed adopting two improvement features, i.e., one is to provide two separate processors for main computer and display device using serial communication, and the other is to use Digital Visual Interface between main computer and display device. In this case the main computer uses minimized graphic functions for safety display. The display design is at the conceptual phase, and there are several open areas to be concreted for a solid system. The main purpose of this paper is to describe and suggest a methodology to develop a safety-critical display system and the descriptions are focused on the safety requirement point of view
Industrial Personal Computer based Display for Nuclear Safety System

Energy Technology Data Exchange (ETDEWEB)

Kim, Ji Hyeon; Kim, Aram; Jo, Jung Hee; Kim, Ki Beom; Cheon, Sung Hyun; Cho, Joo Hyun; Sohn, Se Do; Baek, Seung Min [KEPCO, Youngin (Korea, Republic of)

2014-08-15

The safety display of nuclear system has been classified as important to safety (SIL:Safety Integrity Level 3). These days the regulatory agencies are imposing more strict safety requirements for digital safety display system. To satisfy these requirements, it is necessary to develop a safety-critical (SIL 4) grade safety display system. This paper proposes industrial personal computer based safety display system with safety grade operating system and safety grade display methods. The description consists of three parts, the background, the safety requirements and the proposed safety display system design. The hardware platform is designed using commercially available off-the-shelf processor board with back plane bus. The operating system is customized for nuclear safety display application. The display unit is designed adopting two improvement features, i.e., one is to provide two separate processors for main computer and display device using serial communication, and the other is to use Digital Visual Interface between main computer and display device. In this case the main computer uses minimized graphic functions for safety display. The display design is at the conceptual phase, and there are several open areas to be concreted for a solid system. The main purpose of this paper is to describe and suggest a methodology to develop a safety-critical display system and the descriptions are focused on the safety requirement point of view.
Control system design considerations in a modern nuclear power plant

International Nuclear Information System (INIS)

Foster, P.; Raiskums, G.; Harber, J.; Tikku, S.

2010-01-01

Applying new technologies is a challenge for instrumentation and control (I and C) designers to ensure that the overall principles of defence-in-depth, the independence of safety functions (credited in the safety case), and modern human factors engineering principles are maintained. This paper describes the Advanced CANDU Reactor (ACR-1000) I and C architecture, including the display/control systems and the design approaches employed to ensure that the fundamental premise of independence between safety and process control is not compromised and that the reliability targets for each layer of protection are fulfilled to meet the overall plant safety goals. (author)
MULTIPLE ECH LAUNCHER CONTROL SYSTEM

International Nuclear Information System (INIS)

GREEN, M.T.; PONCE, D.; GRUNLOH, H.J.; ELLIS, R.A.; GROSNICKLE, W.H.; HUMPHREY, R.L.

2004-03-01

OAK-B135 The addition of new, high power gyrotrons to the heating and current drive arsenal at DIII-D, required a system upgrade for control of fully steerable ECH Launchers. Each launcher contains two pointing mirrors with two degrees of mechanical freedom. The two flavors of motion are called facet and tilt. Therefore up to four channels of motion per launcher need to be controlled. The system utilizes absolute encoders to indicate mirror position and therefore direction of the microwave beam. The launcher movement is primarily controlled by PLC, but future iterations of design, may require this control to be accomplished by a CPU on fast bus such as Compact PCI. This will be necessary to accomplish real time position control. Safety of equipment and personnel is of primary importance when controlling a system of moving parts. Therefore multiple interlocks and fault status enunciators have been implemented. This paper addresses the design of a Multiple ECH Launcher Control System, and characterizes the flexibility needed to upgrade to a real time position control system in the future
Enhancement of a radiation safety system through the use of a microprocessor-controlled speech synthesizer

International Nuclear Information System (INIS)

Keefe, D.J.; McDowell, W.P.

1980-01-01

A speech synthesizer is being used to differentiate eight separate safety alarms on a high energy accelerator at Argonne National Laboratory. A single board microcomputer monitors eight signals from an existing radiation safety logic circuit. The microcomputer is programmed to output the proper code at the proper time and sequence to a speech synthesizer which supplies the audio input to a local public address system. This eliminates the requirement for eight different alarm tones and the personnel training required to differentiate among them. A twenty-word vocabulary was found adequate to supply the necessary safety announcements. The article describes the techniques used to interface the speech synthesizer into the existing safety logic circuit
USAEC Controls for Nuclear Criticality Safety

Energy Technology Data Exchange (ETDEWEB)

McCluggage, W. C. [Division of Operational Safety, United States Atomic Energy Commission Washington, DC (United States)

1966-05-15

This is a paper written to provide a broad general view of the United States Atomic Energy Commission's controls for nuclear criticality safety within its own facilities. Included also is a brief' discussion of the USAEC's methods of obtaining assurance that the controls are being applied. The body of the document contains three sections. The first two describe the functions of the USAEC; the third deals with the contractors. The provisions of the Atomic Energy Act applicable to health and safety are discussed in relation to nuclear criticality safety. The use of United States Atomic Energy Commission manual chapters and Federal regulations is described. The functions of the USAEC Headquarters' offices and the operations offices are briefly outlined. Comments regarding the USAEC's inspection, auditing and appraisal programmes are included. Also briefly mentioned are the basic qualifications which must be met to become a contractor to possess and process or use fissionable materials. On the plant, factory or facility level the duties and responsibilities of industrial management are briefly outlined. The fundamental standards and their origin, together with the principal documents and guides are mentioned. The chief methods of control used by contractors operating large USAEC facilities and plants are described and compared. These include diagrams of how a typical nuclear criticality safety problem is handled from inception, design, construction and finally plant operation. Also included is a brief discussion of the contractors' methods of assuring strict employee compliance with the operating rules and limits. (author)
The CEBAF control system for the CHL

International Nuclear Information System (INIS)

Keesee, M.S.; Bevins, B.S.

1996-01-01

The CEBAF Central Helium Liquefier (CHL) control system consists of independent safety controls located at each subsystem, CAMAC computer interface hardware, and a CEBAF-designed control software called Thaumaturgic Automated Control Logic (TACL). The paper describes how control software was interfaced with the subsystems of the CHL. Topics of configuration, editing, operator interface, datalogging, and internal logic functions are presented as they relate to the operational needs of the helium plant. The paper also describes the effort underway to convert from TACL to the Experimental Physics and Industrial Control System (EPICS), the new control system for the CEBAF accelerator. This software change will require customizing EPICS software to cryogenic process control
Handbook of driver assistance systems basic information, components and systems for active safety and comfort

CERN Document Server

Hakuli, Stephan; Lotz, Felix; Singer, Christina

2016-01-01

This fundamental work explains in detail systems for active safety and driver assistance, considering both their structure and their function. These include the well-known standard systems such as Anti-lock braking system (ABS), Electronic Stability Control (ESC) or Adaptive Cruise Control (ACC). But it includes also new systems for protecting collisions protection, for changing the lane, or for convenient parking. The book aims at giving a complete picture focusing on the entire system. First, it describes the components which are necessary for assistance systems, such as sensors, actuators, mechatronic subsystems, and control elements. Then, it explains key features for the user-friendly design of human-machine interfaces between driver and assistance system. Finally, important characteristic features of driver assistance systems for particular vehicles are presented: Systems for commercial vehicles and motorcycles.
Safer Systems: A NextGen Aviation Safety Strategic Goal

Science.gov (United States)

Darr, Stephen T.; Ricks, Wendell R.; Lemos, Katherine A.

2008-01-01

The Joint Planning and Development Office (JPDO), is charged by Congress with developing the concepts and plans for the Next Generation Air Transportation System (NextGen). The National Aviation Safety Strategic Plan (NASSP), developed by the Safety Working Group of the JPDO, focuses on establishing the goals, objectives, and strategies needed to realize the safety objectives of the NextGen Integrated Plan. The three goal areas of the NASSP are Safer Practices, Safer Systems, and Safer Worldwide. Safer Practices emphasizes an integrated, systematic approach to safety risk management through implementation of formalized Safety Management Systems (SMS) that incorporate safety data analysis processes, and the enhancement of methods for ensuring safety is an inherent characteristic of NextGen. Safer Systems emphasizes implementation of safety-enhancing technologies, which will improve safety for human-centered interfaces and enhance the safety of airborne and ground-based systems. Safer Worldwide encourages coordinating the adoption of the safer practices and safer systems technologies, policies and procedures worldwide, such that the maximum level of safety is achieved across air transportation system boundaries. This paper introduces the NASSP and its development, and focuses on the Safer Systems elements of the NASSP, which incorporates three objectives for NextGen systems: 1) provide risk reducing system interfaces, 2) provide safety enhancements for airborne systems, and 3) provide safety enhancements for ground-based systems. The goal of this paper is to expose avionics and air traffic management system developers to NASSP objectives and Safer Systems strategies.
Development of Real Time Operating System for Safety Grade PLC (POSAFE-Q) for Nuclear Power Plants

International Nuclear Information System (INIS)

Son, Han Seong; Hwang, Sung Jae; Lee, Young Joon; Kim, Chang Hwoi; Lee, Dong Young

2006-01-01

POSAFE-Q is a newly developed programmable logic controller (PLC) in order to apply to digital safety system of nuclear power plants (NPP) according to Nuclear Power Plant safety requirements. POSAFE-Q hardware and software development process, including design, review, verification and validation, and configuration control and quality assurance, satisfies the requirements imposed by 10CFR50, Appendix B. This article introduces a real time operating system pCOS, which is the core of POSAFE-Q. Section 2 describes the structure of pCOS. Section 3 describes a few important features of pCOS, which are necessary to the application for the digital safety system of NPP
Development of Real Time Operating System for Safety Grade PLC (POSAFE-Q) for Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Son, Han Seong [ENESYS, Taejon (Korea, Republic of); Hwang, Sung Jae [POSCON, Seoul (Korea, Republic of); Lee, Young Joon; Kim, Chang Hwoi; Lee, Dong Young [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2006-07-01

POSAFE-Q is a newly developed programmable logic controller (PLC) in order to apply to digital safety system of nuclear power plants (NPP) according to Nuclear Power Plant safety requirements. POSAFE-Q hardware and software development process, including design, review, verification and validation, and configuration control and quality assurance, satisfies the requirements imposed by 10CFR50, Appendix B. This article introduces a real time operating system pCOS, which is the core of POSAFE-Q. Section 2 describes the structure of pCOS. Section 3 describes a few important features of pCOS, which are necessary to the application for the digital safety system of NPP.0.
Final Technical Report on Quantifying Dependability Attributes of Software Based Safety Critical Instrumentation and Control Systems in Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Smidts, Carol [The Ohio State Univ., Columbus, OH (United States); Huang, Funqun [The Ohio State Univ., Columbus, OH (United States); Li, Boyuan [The Ohio State Univ., Columbus, OH (United States); Li, Xiang [The Ohio State Univ., Columbus, OH (United States)

2016-03-25

With the current transition from analog to digital instrumentation and control systems in nuclear power plants, the number and variety of software-based systems have significantly increased. The sophisticated nature and increasing complexity of software raises trust in these systems as a significant challenge. The trust placed in a software system is typically termed software dependability. Software dependability analysis faces uncommon challenges since software systems’ characteristics differ from those of hardware systems. The lack of systematic science-based methods for quantifying the dependability attributes in software-based instrumentation as well as control systems in safety critical applications has proved itself to be a significant inhibitor to the expanded use of modern digital technology in the nuclear industry. Dependability refers to the ability of a system to deliver a service that can be trusted. Dependability is commonly considered as a general concept that encompasses different attributes, e.g., reliability, safety, security, availability and maintainability. Dependability research has progressed significantly over the last few decades. For example, various assessment models and/or design approaches have been proposed for software reliability, software availability and software maintainability. Advances have also been made to integrate multiple dependability attributes, e.g., integrating security with other dependability attributes, measuring availability and maintainability, modeling reliability and availability, quantifying reliability and security, exploring the dependencies between security and safety and developing integrated analysis models. However, there is still a lack of understanding of the dependencies between various dependability attributes as a whole and of how such dependencies are formed. To address the need for quantification and give a more objective basis to the review process -- therefore reducing regulatory uncertainty

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