System safety program plan for the Isotope Brayton Ground Demonstration System (phase I)

International Nuclear Information System (INIS)

1976-01-01

The safety engineering effort to be undertaken in achieving an acceptable level of safety in the Brayton Isotope Power System (BIPS) development program is discussed. The safety organizational relationships, the methods to be used, the tasks to be completed, and the documentation to be published are described. The plan will be updated periodically as the need arises

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...

Automated Flight Safety Inference Engine (AFSIE) System, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — We propose to develop an innovative Autonomous Flight Safety Inference Engine (AFSIE) system to autonomously and reliably terminate the flight of an errant launch...

Timing of Formal Phase Safety Reviews for Large-Scale Integrated Hazard Analysis

Science.gov (United States)

Massie, Michael J.; Morris, A. Terry

2010-01-01

Integrated hazard analysis (IHA) is a process used to identify and control unacceptable risk. As such, it does not occur in a vacuum. IHA approaches must be tailored to fit the system being analyzed. Physical, resource, organizational and temporal constraints on large-scale integrated systems impose additional direct or derived requirements on the IHA. The timing and interaction between engineering and safety organizations can provide either benefits or hindrances to the overall end product. The traditional approach for formal phase safety review timing and content, which generally works well for small- to moderate-scale systems, does not work well for very large-scale integrated systems. This paper proposes a modified approach to timing and content of formal phase safety reviews for IHA. Details of the tailoring process for IHA will describe how to avoid temporary disconnects in major milestone reviews and how to maintain a cohesive end-to-end integration story particularly for systems where the integrator inherently has little to no insight into lower level systems. The proposal has the advantage of allowing the hazard analysis development process to occur as technical data normally matures.

Engineering reliability in design phase: An application to AP-600 reactor passive safety system

International Nuclear Information System (INIS)

Majumdr, D.; Siahpush, A.S.; Hills, S.W.

1992-01-01

A computerized reliability enhancement methodology is described that can be used at the engineering design phase to help the designer achieve a desired reliability of the system. It can take into account the limitation imposed by a constraint such as budget, space, or weight. If the desired reliability of the system is known, it can determine the minimum reliabilities of the components, or how many redundant components are needed to achieve the desired reliability. This methodology is applied to examine the Automatic Depressurization System (ADS) of the new passively safe AP-600 reactor. The safety goal of a nuclear reactor dictates a certain reliability level of its components. It is found that a series parallel valve configuration instead of the parallel-series configuration of the four valves in one stage would improve the reliability of the ADS. Other valve characteristics and arrangements are explored to examine different reliability options for the system

Application of a two-phase injector in the safety systems of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Popov, E; Stanev, I [Energoproekt, Sofia (Bulgaria)

1996-12-31

A concept for simplification of the active part of the safety system (ASS) of nuclear power plants is presented. A two-phase injection jet device (IJD) is proposed to substitute the currently used IP-EM (impeller pumps -electric motors) couple. It is capable of sustaining a constant flow rate regardless of the variation in the system hydraulic resistance. The conditions for effective work of IJD are: development of the necessary head and flow rate, reliable supply of working medium and maintaining of the temperature of the injected water. IJD efficiency, steam and water flow rates have been calculated and compared with experimentally measured values. A short analysis of different typical accident regimes is carried out. It shows that IJD introduction brings significant advantages especially in the steam generator emergency feedwater system making it completely insensitive to loss of electricity supply accidents. 8 refs., 7 figs.

Application of a two-phase injector in the safety systems of nuclear power plants

International Nuclear Information System (INIS)

Popov, E.; Stanev, I.

1995-01-01

A concept for simplification of the active part of the safety system (ASS) of nuclear power plants is presented. A two-phase injection jet device (IJD) is proposed to substitute the currently used IP-EM (impeller pumps -electric motors) couple. It is capable of sustaining a constant flow rate regardless of the variation in the system hydraulic resistance. The conditions for effective work of IJD are: development of the necessary head and flow rate, reliable supply of working medium and maintaining of the temperature of the injected water. IJD efficiency, steam and water flow rates have been calculated and compared with experimentally measured values. A short analysis of different typical accident regimes is carried out. It shows that IJD introduction brings significant advantages especially in the steam generator emergency feedwater system making it completely insensitive to loss of electricity supply accidents. 8 refs., 7 figs

Phase 2 safety analysis report: National Synchrotron Light Source

International Nuclear Information System (INIS)

Stefan, P.

1989-06-01

The Phase II program was established in order to provide additional space for experiments, and also staging and equipment storage areas. It also provides additional office space and new types of advanced instrumentation for users. This document will deal with the new safety issues resulting from this extensive expansion program, and should be used as a supplement to BNL Report No. 51584 ''National Synchrotron Light Source Safety Analysis Report,'' July 1982 (hereafter referred to as the Phase I SAR). The initial NSLS facility is described in the Phase I SAR. It comprises two electron storage rings, an injection system common to both, experimental beam lines and equipment, and office and support areas, all of which are housed in a 74,000 sq. ft. building. The X-ray Ring provides for 28 primary beam ports and the VUV Ring, 16. Each port is capable of division into 2 or 3 separate beam lines. All ports receive their synchrotron light from conventional bending magnet sources, the magnets being part of the storage ring lattice. 4 refs

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.

Analysis of Aviation Safety Reporting System Incident Data Associated with the Technical Challenges of the System-Wide Safety and Assurance Technologies Project

Science.gov (United States)

Withrow, Colleen A.; Reveley, Mary S.

2015-01-01

The Aviation Safety Program (AvSP) System-Wide Safety and Assurance Technologies (SSAT) Project asked the AvSP Systems and Portfolio Analysis Team to identify SSAT-related trends. SSAT had four technical challenges: advance safety assurance to enable deployment of NextGen systems; automated discovery of precursors to aviation safety incidents; increasing safety of human-automation interaction by incorporating human performance, and prognostic algorithm design for safety assurance. This report reviews incident data from the NASA Aviation Safety Reporting System (ASRS) for system-component-failure- or-malfunction- (SCFM-) related and human-factor-related incidents for commercial or cargo air carriers (Part 121), commuter airlines (Part 135), and general aviation (Part 91). The data was analyzed by Federal Aviation Regulations (FAR) part, phase of flight, SCFM category, human factor category, and a variety of anomalies and results. There were 38 894 SCFM-related incidents and 83 478 human-factorrelated incidents analyzed between January 1993 and April 2011.

Health and Safety Management Plan for the Plutonium Stabilization and Packaging System

International Nuclear Information System (INIS)

1996-01-01

This Health and Safety Management Plan (HSMP) presents safety and health policies and a project health and safety organizational structure designed to minimize potential risks of harm to personnel performing activities associated with Plutonium Stabilization and Packaging System (Pu SPS). The objectives of the Pu SPS are to design, fabricate, install, and startup of a glovebox system for the safe repackaging of plutonium oxides and metals, with a requirement of a 50-year storage period. This HSMP is intended as an initial project health and safety submittal as part of a three phase effort to address health and safety issues related to personnel working the Pu SPS project. Phase 1 includes this HSMP and sets up the basic approach to health and safety on the project and addresses health and safety issues related to the engineering and design effort. Phase 2 will include the Site Specific Construction health and Safety Plan (SSCHSP). Phase 3 will include an additional addendum to this HSMP and address health and safety issues associated with the start up and on-site test phase of the project. This initial submittal of the HSMP is intended to address those activities anticipated to be performed during phase 1 of the project. This HSMP is intended to be a living document which shall be modified as information regarding the individual tasks associated with the project becomes available. These modifications will be in the form of addenda to be submitted prior to the initiation of each phase of the project. For additional work authorized under this project this HSMP will be modified as described in section 1.4

Reliability analysis of the recirculation phase of the safety injection system of Angra-1

International Nuclear Information System (INIS)

Rivera, R.R.J.M.

1981-09-01

The calculation of several reliability parameters-failure probability, unavailability and unreliability - of the recirculation phase of the safety injection system of Angra-1, was done. This system has two distinct modes of operation (short term and long term) which were fault tree analysed both separately and as a whole. To obtain quantitative results the computer codes SAMPLE and PRET-KITT were utilized. The former was used to consider the uncertainties in the failure data (drawn integrally from WASH-1400) and the latter to obtain time dependent unreliability values. Hardware failures and common-mode failures were considered. Altough the analysis methods employed here differ somewhat from those used in WASH-1400, the results which could be compared were found to have the order of magnitude. A viability study of some suggestions of system's modifications was performed, and it has shown that some significant reliability improvements can be achieved with reasonably simple changes. (Author) [pt

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)

Safety balance: Analysis of safety systems

International Nuclear Information System (INIS)

Delage, M.; Giroux, C.

1990-12-01

Safety analysis, and particularly analysis of exploitation of NPPs is constantly affected by EDF and by the safety authorities and their methodologies. Periodic safety reports ensure that important issues are not missed on daily basis, that incidents are identified and that relevant actions are undertaken. French safety analysis method consists of three principal steps. First type of safety balance is analyzed at the normal start-up phase for each unit including the final safety report. This enables analysis of behaviour of units ten years after their licensing. Second type is periodic operational safety analysis performed during a few years. Finally, the third step consists of safety analysis of the oldest units with the aim to improve the safety standards. The three steps of safety analysis are described in this presentation in detail with the aim to present the objectives and principles. Examples of most recent exercises are included in order to illustrate the importance of such analyses

Fluor Hanford Integrated Safety Management System Phase II Verification Vol 1 & Vol 2

Energy Technology Data Exchange (ETDEWEB)

PARSONS, J.E.

2000-07-15

The U.S. Department of Energy (DOE) is committed to conducting work efficiently and in a manner that ensures protection of the workers, public, and environment. DOE policy mandates that safety management systems be used to systematically integrate safety into management and work practices at all levels while accomplishing mission goals in an effective and efficient manner. The purpose of the Fluor Hanford (FH) Integrated Safety Management System (ISMS) verification was to determine whether FH's ISM system and processes are sufficiently implemented to accomplish the goal of ''Do work safely.'' The purpose of the DOE, Richland Operations Office (RL) verification was to determine whether RL has established processes that adequately describe RL's role in safety management and if those processes are sufficiently implemented.

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.

Phase separator safety valve blow-off.

CERN Multimedia

G. Perinic

2006-01-01

The fast discharge of the CMS solenoid leads to a pressure rise in the phase separator. On August 28th, a fast discharge was triggered at a current level of 19.1 kA. The pressure in the phase separator increased up to the set pressure of the safety valve and some helium was discharged. In consequence of this and prevoious similar observations the liquid helium level in the phase separator has been reduced from 60% to 50% and later to 45% in order to reduce the helium inventory in the magnet.

Model Transformation for a System of Systems Dependability Safety Case

Science.gov (United States)

Murphy, Judy; Driskell, Steve

2011-01-01

The presentation reviews the dependability and safety effort of NASA's Independent Verification and Validation Facility. Topics include: safety engineering process, applications to non-space environment, Phase I overview, process creation, sample SRM artifact, Phase I end result, Phase II model transformation, fault management, and applying Phase II to individual projects.

200-ZP-1 phase II and III IRM groundwater pump and treat site safety plan

International Nuclear Information System (INIS)

St. John, C.H.

1996-07-01

This safety plan covers operations, maintenance, and support activities related to the 200-ZP-1 Phase II and III Ground Water Pump- and-Treat Facility. The purpose of the facility is to extract carbon tetrachloride contaminated groundwater underlying the ZP-1 Operable Unit; separate the contaminant from the groundwater; and reintroduce the treated water to the aquifer. An air stripping methodology is employed to convert volatile organics to a vapor phase for absorption onto granular activated carbon. The automated process incorporates a variety of process and safety features that shut down the process system in the event that process or safety parameters are exceeded or compromised

Fluor Hanford Integrated Safety Management System Phase II Verification Vol 1 and Vol 2

CERN Document Server

Parsons, J E

2000-01-01

The U.S. Department of Energy (DOE) is committed to conducting work efficiently and in a manner that ensures protection of the workers, public, and environment. DOE policy mandates that safety management systems be used to systematically integrate safety into management and work practices at all levels while accomplishing mission goals in an effective and efficient manner. The purpose of the Fluor Hanford (FH) Integrated Safety Management System (ISMS) verification was to determine whether FH's ISM system and processes are sufficiently implemented to accomplish the goal of ''Do work safely.'' The purpose of the DOE, Richland Operations Office (RL) verification was to determine whether RL has established processes that adequately describe RL's role in safety management and if those processes are sufficiently implemented.

Fluor Hanford Integrated Safety Management System Phase II Verification Vol 1 and Vol 2

International Nuclear Information System (INIS)

PARSONS, J.E.

2000-01-01

The U.S. Department of Energy (DOE) is committed to conducting work efficiently and in a manner that ensures protection of the workers, public, and environment. DOE policy mandates that safety management systems be used to systematically integrate safety into management and work practices at all levels while accomplishing mission goals in an effective and efficient manner. The purpose of the Fluor Hanford (FH) Integrated Safety Management System (ISMS) verification was to determine whether FH's ISM system and processes are sufficiently implemented to accomplish the goal of ''Do work safely.'' The purpose of the DOE, Richland Operations Office (RL) verification was to determine whether RL has established processes that adequately describe RL's role in safety management and if those processes are sufficiently implemented

Occupational Safety. Hygiene Safety. Pre-Apprenticeship Phase 1 Training.

Science.gov (United States)

Lane Community Coll., Eugene, OR.

This self-paced student training module on hygiene safety is one of a number of modules developed for Pre-apprenticeship Phase 1 Training. Purpose of the module is to familiarize students with the different types of airborne contaminants--including noise--which may be health hazards and with the proper hygienic measures for dealing with them. The…

Towards the Verification of Safety-critical Autonomous Systems in Dynamic Environments

Directory of Open Access Journals (Sweden)

Adina Aniculaesei

2016-12-01

Full Text Available There is an increasing necessity to deploy autonomous systems in highly heterogeneous, dynamic environments, e.g. service robots in hospitals or autonomous cars on highways. Due to the uncertainty in these environments, the verification results obtained with respect to the system and environment models at design-time might not be transferable to the system behavior at run time. For autonomous systems operating in dynamic environments, safety of motion and collision avoidance are critical requirements. With regard to these requirements, Macek et al. [6] define the passive safety property, which requires that no collision can occur while the autonomous system is moving. To verify this property, we adopt a two phase process which combines static verification methods, used at design time, with dynamic ones, used at run time. In the design phase, we exploit UPPAAL to formalize the autonomous system and its environment as timed automata and the safety property as TCTL formula and to verify the correctness of these models with respect to this property. For the runtime phase, we build a monitor to check whether the assumptions made at design time are also correct at run time. If the current system observations of the environment do not correspond to the initial system assumptions, the monitor sends feedback to the system and the system enters a passive safe state.

Extension of CFD Codes Application to Two-Phase Flow Safety Problems - Phase 3

International Nuclear Information System (INIS)

Bestion, D.; Anglart, H.; Mahaffy, J.; Lucas, D.; Song, C.H.; Scheuerer, M.; Zigh, G.; Andreani, M.; Kasahara, F.; Heitsch, M.; Komen, E.; Moretti, F.; Morii, T.; Muehlbauer, P.; Smith, B.L.; Watanabe, T.

2014-11-01

The Writing Group 3 on the extension of CFD to two-phase flow safety problems was formed following recommendations made at the 'Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety Problems' held in Aix-en-Provence, in May 2002. Extension of CFD codes to two-phase flow is significant potentiality for the improvement of safety investigations, by giving some access to smaller scale flow processes which were not explicitly described by present tools. Using such tools as part of a safety demonstration may bring a better understanding of physical situations, more confidence in the results, and an estimation of safety margins. The increasing computer performance allows a more extensive use of 3D modelling of two-phase Thermal hydraulics with finer nodalization. However, models are not as mature as in single phase flow and a lot of work has still to be done on the physical modelling and numerical schemes in such two-phase CFD tools. The Writing Group listed and classified the NRS problems where extension of CFD to two-phase flow may bring real benefit, and classified different modelling approaches in a first report (Bestion et al., 2006). First ideas were reported about the specification and analysis of needs in terms of validation and verification. It was then suggested to focus further activity on a limited number of NRS issues with a high priority and a reasonable chance to be successful in a reasonable period of time. The WG3-step 2 was decided with the following objectives: - selection of a limited number of NRS issues having a high priority and for which two-phase CFD has a reasonable chance to be successful in a reasonable period of time; - identification of the remaining gaps in the existing approaches using two-phase CFD for each selected NRS issue; - review of the existing data base for validation of two-phase CFD application to the selected NRS problems

Developing a patient-led electronic feedback system for quality and safety within Renal PatientView.

Science.gov (United States)

Giles, Sally J; Reynolds, Caroline; Heyhoe, Jane; Armitage, Gerry

2017-03-01

It is increasingly acknowledged that patients can provide direct feedback about the quality and safety of their care through patient reporting systems. The aim of this study was to explore the feasibility of patients, healthcare professionals and researchers working in partnership to develop a patient-led quality and safety feedback system within an existing electronic health record (EHR), known as Renal PatientView (RPV). Phase 1 (inception) involved focus groups (n = 9) and phase 2 (requirements) involved cognitive walkthroughs (n = 34) and 1:1 qualitative interviews (n = 34) with patients and healthcare professionals. A Joint Services Expert Panel (JSP) was convened to review the findings from phase 1 and agree the core principles and components of the system prototype. Phase 1 data were analysed using a thematic approach. Data from phase 1 were used to inform the design of the initial system prototype. Phase 2 data were analysed using the components of heuristic evaluation, resulting in a list of core principles and components for the final system prototype. Phase 1 identified four main barriers and facilitators to patients feeding back on quality and safety concerns. In phase 2, the JSP agreed that the system should be based on seven core principles and components. Stakeholders were able to work together to identify core principles and components for an electronic patient quality and safety feedback system in renal services. Tensions arose due to competing priorities, particularly around anonymity and feedback. Careful consideration should be given to the feasibility of integrating a novel element with differing priorities into an established system with existing functions and objectives. © 2016 European Dialysis and Transplant Nurses Association/European Renal Care Association.

Project SAFE. Update of the SFR-1 safety assessment. Phase 1

International Nuclear Information System (INIS)

Andersson, Johan; Riggare, P.; Skagius, K.

1998-10-01

SFR-1 is a facility for disposal of low-level radioactive operational waste from the nuclear power plants in Sweden. Low-level radioactive waste from industry, medicine, and research is also disposed in SFR-1. The facility is situated in bedrock beneath the Baltic Sea, 1 km off the coast near the Forsmark nuclear power plant. SFR-1 was built between the years 1983 and 1988. An assessment of the long-term performance of the facility was included in the vast documentation that was a part of the application for an operational license. The assessment was presented in the form of a final safety report. In the operational licence for SFR-1 it is stated that renewed safety assessments should be carried out at least each ten years. In order to meet this demand SKB has launched a special project, SAFE (Safety Assessment of Final Disposal of Operational Radioactive Waste). The aim of the project is to update the safety analysis and to prepare a safety report that will be presented to the Swedish authorities not later than year 2000. Project SAFE is divided into three phases. The first phase is a prestudy, and the results of the prestudy are given in this report. The aim of the prestudy is to identify issues where additional studies would improve the basis for the updated safety analysis as well as to suggest how these studies should be carried out. The work has been divided into six different topics, namely the inventory, the near field, the far field, the biosphere, radionuclide transport calculations and scenarios. For each topic the former safety reports and regulatory reviews are scrutinised and needs for additional work is identified. The evaluations are given in appendices covering the respective topics. The main report is a summary of the appendices with a more stringent description of the repository system and the processes that are of interest and therefore should be addressed in an updated safety assessment. However, it should be pointed out that one of the

Routine testing on protective and safety systems and components

International Nuclear Information System (INIS)

Rysy, W.

1977-01-01

1) In-process inspection, tests during commissioning. 2) Tests during reactor operation. 2.1) Reactor protection system, for example: continuous auto-testing by a dynamic system, check of the output signals; 2.2) safety features: selected examples: functional tests on the ECCS, trial operation of the emergency diesels. 3) Tests during refuelling phase. 3.1) Containment: Leakage rate tests, leak testing; 3.2) coolant system: selected examples: inservice inspections of the pressure vessel, eddy current testing of the steam generator, functional tests of safety valves. (orig./HP) [de

Software for computer based systems important to safety in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2004-01-01

systems important to safety in nuclear power plants, for all phases of the system life cycle. The guidance is applicable to systems important to safety. Since at present the reliability of a computer based system cannot be predicted on the sole basis of, or built in by, the design process, it is difficult to define and to agree systematically on any possible relaxation in the guidance to apply to software for safety related systems. Whenever possible, recommendations which apply only to safety systems and not to safety related systems are explicitly identified. The guidance relates primarily to the software used in computer based systems important to safety. Guidance on the other aspects of computer based systems, such as those concerned with the design of the computer based system itself and its hardware, is limited to the issues raised by the development, verification and validation of software.The main focus of this Safety Guide is on the preparation of documentation that is used for an adequate demonstration of the safety and reliability of computer based systems important to safety.This Safety Guide applies to all types of software: pre-existing software or firmware (such as an operating system), software to be specifically developed for the project, or software to be developed from an existing pre developed equipment family of hardware or software modules. This Safety Guide is intended for use by those involved in the production, assessment and licensing of computer based systems, including plant system designers, software designers and programmers, verifiers, validators, certifiers and regulators, as well as plant operators. The various interfaces between those involved are considered. (author)

Software for computer based systems important to safety in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2005-01-01

systems important to safety in nuclear power plants, for all phases of the system life cycle. The guidance is applicable to systems important to safety. Since at present the reliability of a computer based system cannot be predicted on the sole basis of, or built in by, the design process, it is difficult to define and to agree systematically on any possible relaxation in the guidance to apply to software for safety related systems. Whenever possible, recommendations which apply only to safety systems and not to safety related systems are explicitly identified. The guidance relates primarily to the software used in computer based systems important to safety. Guidance on the other aspects of computer based systems, such as those concerned with the design of the computer based system itself and its hardware, is limited to the issues raised by the development, verification and validation of software.The main focus of this Safety Guide is on the preparation of documentation that is used for an adequate demonstration of the safety and reliability of computer based systems important to safety. This Safety Guide applies to all types of software: pre-existing software or firmware (such as an operating system), software to be specifically developed for the project, or software to be developed from an existing pre developed equipment family of hardware or software modules. This Safety Guide is intended for use by those involved in the production, assessment and licensing of computer based systems, including plant system designers, software designers and programmers, verifiers, validators, certifiers and regulators, as well as plant operators. The various interfaces between those involved are considered

Software for computer based systems important to safety in nuclear power plants. Safety guide

International Nuclear Information System (INIS)

2000-01-01

systems important to safety in nuclear power plants, for all phases of the system life cycle. The guidance is applicable to systems important to safety. Since at present the reliability of a computer based system cannot be predicted on the sole basis of, or built in by, the design process, it is difficult to define and to agree systematically on any possible relaxation in the guidance to apply to software for safety related systems. Whenever possible, recommendations which apply only to safety systems and not to safety related systems are explicitly identified. The guidance relates primarily to the software used in computer based systems important to safety. Guidance on the other aspects of computer based systems, such as those concerned with the design of the computer based system itself and its hardware, is limited to the issues raised by the development, verification and validation of software.The main focus of this Safety Guide is on the preparation of documentation that is used for an adequate demonstration of the safety and reliability of computer based systems important to safety.This Safety Guide applies to all types of software: pre-existing software or firmware (such as an operating system), software to be specifically developed for the project, or software to be developed from an existing pre developed equipment family of hardware or software modules. This Safety Guide is intended for use by those involved in the production, assessment and licensing of computer based systems, including plant system designers, software designers and programmers, verifiers, validators, certifiers and regulators, as well as plant operators. The various interfaces between those involved are considered

Criteria Document for B-plant's Surveillance and Maintenance Phase Safety Basis Document

International Nuclear Information System (INIS)

SCHWEHR, B.A.

1999-01-01

This document is required by the Project Hanford Managing Contractor (PHMC) procedure, HNF-PRO-705, Safety Basis Planning, Documentation, Review, and Approval. This document specifies the criteria that shall be in the B Plant surveillance and maintenance phase safety basis in order to obtain approval of the DOE-RL. This CD describes the criteria to be addressed in the S and M Phase safety basis for the deactivated Waste Fractionization Facility (B Plant) on the Hanford Site in Washington state. This criteria document describes: the document type and format that will be used for the S and M Phase safety basis, the requirements documents that will be invoked for the document development, the deactivated condition of the B Plant facility, and the scope of issues to be addressed in the S and M Phase safety basis document

System and safety studies of accelerator driven systems for transmutation. Annual report 2007

International Nuclear Information System (INIS)

Arzhanov, Vasily; Fokau, Andrei; Persson, Calle; Runevall, Odd; Sandberg, Nils; Tesinsky, Milan; Wallenius, Janne; Youpeng Zhang

2008-05-01

Within the project 'System and safety studies of accelerator driven systems for transmutation', research on design and safety of sub-critical reactors for recycling of minor actinides is performed. During 2007, the reactor physics division at KTH has calculated safety parameters for EFIT-400 with cermet fuel, permitting to start the transient safety analysis. The accuracy of different reactivity meters applied to the YALINA facility was assessed and neutron detection studies were performed. A model to address deviations from point kinetic behaviour was developed. Studies of basic radiation damage physics included calculations of vacancy formation and activation enthalpies in bcc niobium. In order to predict the oxygen potential of inert matrix fuels, a thermo-chemical model for mixed actinide oxides was implemented in a phase equilibrium code

Fault tree synthesis for software design analysis of PLC based safety-critical systems

International Nuclear Information System (INIS)

Koo, S. R.; Cho, C. H.; Seong, P. H.

2006-01-01

As a software verification and validation should be performed for the development of PLC based safety-critical systems, a software safety analysis is also considered in line with entire software life cycle. In this paper, we propose a technique of software safety analysis in the design phase. Among various software hazard analysis techniques, fault tree analysis is most widely used for the safety analysis of nuclear power plant systems. Fault tree analysis also has the most intuitive notation and makes both qualitative and quantitative analyses possible. To analyze the design phase more effectively, we propose a technique of fault tree synthesis, along with a universal fault tree template for the architecture modules of nuclear software. Consequently, we can analyze the safety of software on the basis of fault tree synthesis. (authors)

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.

Safety verification for the ECCS driven by the electrically 4 trains during LBLOCA reflood phase using ATLAS

International Nuclear Information System (INIS)

Park, Yusun; Park, Hyun-sik; Kang, Kyoung-ho; Choi, Nam-hyun; Min, Kyoung-ho; Choi, Ki-yong

2014-01-01

Highlights: • Safety improvement by adopting 4 train emergency core cooling system was validated experimentally. • General thermal hydraulic behaviors of the system during LBLOCA reflood phase were successfully demonstrated. • Key parameters such as the liquid levels, the PCTs, the quenching time, and the ECC bypass ratios were investigated. • Asymmetric effects of the different combination of safety injection were negligible during the reflood period. - Abstract: The APR1400 is equipped with four safety injection pumps driven by two emergency diesel generators. However, the design has been changed so that the four safety injection pumps are driven by 4 emergency diesel generators during the design certification process from the U.S. NRC. Thus, 4 safety injection pumps (SIPs) are completely independent electrically and mechanically and three safety injection pumps are available in a single failure condition. This design change could have a certain effects on the thermal-hydraulic phenomenon occurring in the downcomer region during the late reflood phase of a large break loss of coolant accident (LBLOCA). Thus, in this study, a verification experiment for the reflood phase of a LBLOCA was performed to evaluate the core cooling performance of the 4 train emergency core cooling system (ECCS) with an assumption of a single failure. And the different combinations of three SIPs positions were tested to investigate the asymmetric effects on the reactor core cooling performance. The overall experimental results revealed the typical thermal–hydraulic trends expected to occur during the reflood phase of a large-break LOCA scenario for the APR1400. Experiment with the injection of three SIPs showed a faster core quenching time and lower bypass ratio than that of the case in which two SIPs were injected. The RPV wall temperature distributions showed the similar trend in spite of the different SIP combinations

An approach for functional safety improvement of an existing automotive system

NARCIS (Netherlands)

Khabbaz Saberi, A.; Luo, Y.; Pawel Cichosz, F.; Brand, M. van den; Jansen, S.T.H.

2015-01-01

Safety of automotive systems is becoming more involved, specially for the case of autonomous vehicles. The ISO 26262 standard offers a systematic approach for designing a safe road vehicle (or subsystems of a car) from design phase through its production. However, providing functional safety

Integrated Safety Management System Phase I Verification for the Plutonium Finishing Plant (PFP) [VOL 1 & 2

Energy Technology Data Exchange (ETDEWEB)

SETH, S.S.

2000-01-10

U.S. Department of Energy (DOE) Policy 450.4, Safety Management System Policy commits to institutionalizing an Integrated Safety Management System (ISMS) throughout the DOE complex as a means of accomplishing its missions safely. DOE Acquisition Regulation 970.5204-2 requires that contractors manage and perform work in accordance with a documented safety management system.

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

Qualification of safety-critical software for digital reactor safety system in nuclear power plants

International Nuclear Information System (INIS)

Kwon, Kee-Choon; Park, Gee-Yong; Kim, Jang-Yeol; Lee, Jang-Soo

2013-01-01

This paper describes the software qualification activities for the safety-critical software of the digital reactor safety system in nuclear power plants. The main activities of the software qualification processes are the preparation of software planning documentations, verification and validation (V and V) of the software requirements specifications (SRS), software design specifications (SDS) and codes, and the testing of the integrated software and integrated system. Moreover, the software safety analysis and software configuration management are involved in the software qualification processes. The V and V procedure for SRS and SDS contains a technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, software safety analysis, and an evaluation of the software configuration management. The V and V processes for the code are a traceability analysis, source code inspection, test case and test procedure generation. Testing is the major V and V activity of the software integration and system integration phases. The software safety analysis employs a hazard operability method and software fault tree analysis. The software configuration management in each software life cycle is performed by the use of a nuclear software configuration management tool. Through these activities, we can achieve the functionality, performance, reliability, and safety that are the major V and V objectives of the safety-critical software in nuclear power plants. (author)

System and safety studies of accelerator driven systems for transmutation. Annual report 2007

Energy Technology Data Exchange (ETDEWEB)

Arzhanov, Vasily; Fokau, Andrei; Persson, Calle; Runevall, Odd; Sandberg, Nils; Tesinsky, Milan; Wallenius, Janne; Youpeng Zhang (Div. of Reactor Physics, Royal Institute of Technology, Stockholm (Sweden))

2008-05-15

Within the project 'System and safety studies of accelerator driven systems for transmutation', research on design and safety of sub-critical reactors for recycling of minor actinides is performed. During 2007, the reactor physics division at KTH has calculated safety parameters for EFIT-400 with cermet fuel, permitting to start the transient safety analysis. The accuracy of different reactivity meters applied to the YALINA facility was assessed and neutron detection studies were performed. A model to address deviations from point kinetic behaviour was developed. Studies of basic radiation damage physics included calculations of vacancy formation and activation enthalpies in bcc niobium. In order to predict the oxygen potential of inert matrix fuels, a thermo-chemical model for mixed actinide oxides was implemented in a phase equilibrium code

Simulation study of coal mine safety investment based on system dynamics

Institute of Scientific and Technical Information of China (English)

Tong Lei; Dou Yuanyuan

2014-01-01

To generate dynamic planning for coal mine safety investment, this study applies system dynamics to decision-making, classifying safety investments by accident type. It validates the relationship between safety investments and accident cost, by structurally analyzing the causality between safety investments and their influence factors. Our simulation model, based on Vensim software, conducts simulation anal-ysis on a series of actual data from a coalmine in Shanxi Province. Our results indicate a lag phase in safety investments, and that increasing pre-phase safety investment reduces accident costs. We found that a 24%increase in initial safety investment could help reach the target accident costs level 14 months earlier. Our simulation test included nine kinds of variation trends of accident costs brought by different investment ratios on accident prevention. We found an optimized ratio of accident prevention invest-ments allowing a mine to reach accident cost goals 4 months earlier, without changing its total investment.

Spent Nuclear Fuel (SNF) project Integrated Safety Management System phase I and II Verification Review Plan

International Nuclear Information System (INIS)

CARTER, R.P.

1999-01-01

The U.S. Department of Energy (DOE) commits to accomplishing its mission safely. To ensure this objective is met, DOE issued DOE P 450.4, Safety Management System Policy, and incorporated safety management into the DOE Acquisition Regulations ([DEAR] 48 CFR 970.5204-2 and 90.5204-78). Integrated Safety Management (ISM) requires contractors to integrate safety into management and work practices at all levels so that missions are achieved while protecting the public, the worker, and the environment. The contractor is required to describe the Integrated Safety Management System (ISMS) to be used to implement the safety performance objective

Spent Nuclear Fuel (SNF) project Integrated Safety Management System phase I and II Verification Review Plan

Energy Technology Data Exchange (ETDEWEB)

CARTER, R.P.

1999-11-19

The U.S. Department of Energy (DOE) commits to accomplishing its mission safely. To ensure this objective is met, DOE issued DOE P 450.4, Safety Management System Policy, and incorporated safety management into the DOE Acquisition Regulations ([DEAR] 48 CFR 970.5204-2 and 90.5204-78). Integrated Safety Management (ISM) requires contractors to integrate safety into management and work practices at all levels so that missions are achieved while protecting the public, the worker, and the environment. The contractor is required to describe the Integrated Safety Management System (ISMS) to be used to implement the safety performance objective.

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

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...

Engineering systems reliability, safety, and maintenance an integrated approach

CERN Document Server

Dhillon, B S

2017-01-01

Today, engineering systems are an important element of the world economy and each year billions of dollars are spent to develop, manufacture, operate, and maintain various types of engineering systems around the globe. Many of these systems are highly sophisticated and contain millions of parts. For example, a Boeing jumbo 747 is made up of approximately 4.5 million parts including fasteners. Needless to say, reliability, safety, and maintenance of systems such as this have become more important than ever before.　 Global competition and other factors are forcing manufacturers to produce highly reliable, safe, and maintainable engineering products. Therefore, there is a definite need for the reliability, safety, and maintenance professionals to work closely during design and other phases. Engineering Systems Reliability, Safety, and Maintenance: An Integrated Approach eliminates the need to consult many different and diverse sources in the hunt for the information required to design better engineering syste...

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.

Dependability analysis of proposed I and C architecture for safety systems of a large PWR

International Nuclear Information System (INIS)

Kabra, Ashutosh; Karmakar, G.; Tiwari, A.P.; Manoj Kumar; Marathe, P.P.

2014-01-01

Instrumentation and Control (I and C) systems in a reactor provide protection against unsafe operation during steady-state and transient power operations. Indian reactors traditionally adopted 2-out-of-3 (2oo3) architecture for safety systems. But, contemporary reactor safety systems are employing 2-out-of-4 (2oo4) architecture in spite of the increased cost due to the additional channel. This motivated us to carry out a comparative study of 2oo3 and 2oo4 architecture, especially for their dependability attributes - safety and availability. Quantitative estimation of safety and availability has been used to adjudge the worthiness of adopting 2oo4 architecture in I and C safety systems of a large PWR. Our analysis using Markov model shows that 2oo4 architecture, even with lower diagnostic coverage and longer proof test interval, can provide better safety and availability in comparison of 2oo3 architecture. This reduces total life cycle cost of system during development phase and complexity and frequency of surveillance test during operational phase. The paper also describes the proposed architecture for Reactor Protection System (RPS), a representative safety system, and determines its dependability using Markov analysis and Failure Mode Effect Analysis (FMEA). The proposed I and C safety system architecture also has been qualitatively analyzed for their effectiveness against common cause failures (CCFs). (author)

Angra-1 probabilistic safety study-phase B

International Nuclear Information System (INIS)

Fernandes Filho, T.L.; Gibelli, S.M.O.

1988-05-01

This study represents the Phase B of the Angra-1 Probabilistic Safety Study and is the the final report prepared for the IAEA under Research Contract No. 3423/R2/RB. The three main items covered in this report are the establishment of interim safety goals, analysis of Angra-1 operational experience and development of emergency procedures to address severe accidents. For establishment of interim safety goals a methodology for calculating consequences and risks associated to the Angra-1 operation was developed based on the available data and codes. The proposed safety goals refer to the individual risk of early fatality for people living in the vicinity of the plant, colective risk of cancer fatalities for people living near the plant, the propobability of core melt occurrence and the probability of dominant accident sequences. (author) [pt

Prediction of two-phase choked-flow through safety valves

International Nuclear Information System (INIS)

Arnulfo, G; Bertani, C; De Salve, M

2014-01-01

Different models of two-phase choked flow through safety valves are applied in order to evaluate their capabilities of prediction in different thermal-hydraulic conditions. Experimental data available in the literature for two-phase fluid and subcooled liquid upstream the safety valve have been compared with the models predictions. Both flashing flows and non-flashing flows of liquid and incondensable gases have been considered. The present paper shows that for flashing flows good predictions are obtained by using the two-phase valve discharge coefficient defined by Lenzing and multiplying it by the critical flow rate in an ideal nozzle evaluated by either Omega Method or the Homogeneous Non-equilibrium Direct Integration. In case of non-flashing flows of water and air, Leung/Darby formulation of the two-phase valve discharge coefficient together with the Omega Method is more suitable to the prediction of flow rate.

Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

Directory of Open Access Journals (Sweden)

Jaewoon Yoo

2016-10-01

Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

International Nuclear Information System (INIS)

Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook

2016-01-01

The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper

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.

Software for computers in the safety systems of nuclear power stations

International Nuclear Information System (INIS)

1987-08-01

This standard includes the safety actuation systems, the safety system support features and the protection systems. The standard provides requirements for each stage of software generation, including design, development, qualification and operation as well as the documentation for each stage of the software generation for the purpose of achieving highly reliable software. The principles applied in developing these requirements include: Best available practice; top-down design methods; modularity; verification of each phase; clear documentation; auditable documents and validation testing. (orig./HP)

Integrated Safety Management System Phase I Verification for the Plutonium Finishing Plant (PFP) [VOL 1 and 2

International Nuclear Information System (INIS)

SETH, S.S.

2000-01-01

U.S. Department of Energy (DOE) Policy 450.4, Safety Management System Policy commits to institutionalizing an Integrated Safety Management System (ISMS) throughout the DOE complex as a means of accomplishing its missions safely. DOE Acquisition Regulation 970.5204-2 requires that contractors manage and perform work in accordance with a documented safety management system

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

Occupational Safety. Hand Tools. Pre-Apprenticeship Phase 1 Training.

Science.gov (United States)

Lane Community Coll., Eugene, OR.

This self-paced student training module on safety when using hand tools is one of a number of modules developed for Pre-apprenticeship Phase 1 Training. Purpose of the module is to teach students the correct safety techniques for operating common hand- and arm-powered tools, including selection, maintenance, technique, and uses. The module may…

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

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.

Methodologies for verification and validation of expert systems as a function of component, criticality and life-cycle phase

International Nuclear Information System (INIS)

Miller, L.

1992-01-01

The review of verification and validation (V and V) methods presented here is based on results of the initial two tasks of a contract with the US Nuclear Regulatory Commission and the Electric Power Research Institute to Develop and Document Guidelines for Verifying and Validating Expert Systems. The first task was to review the applicability of conventional software techniques to expert systems; the second was to directly survey V and V practices associated with development of expert systems. Subsequent tasks will focus on selecting, synthesizing or developing V and V methods appropriate for the overall system, for specific expert systems components, and for different phases of the life-cycle. In addition, final guidelines will most likely be developed for each of three levels of expert systems: safety-related (systems whose functions directly relate to system safety, so-called safety-critical systems), important-to-safety (systems which support the critical safety functions), and non-safety (systems which are unrelated to safety functions). For the present purposes of categorizing and discussing various types of V and V methods, the authors simplify the life-cycle and consider only two aspects - systems validation phase. The authors identified a number of techniques for the first, combined, phase and two general classes of V and V techniques for the latter phase: static testing techniques, which do not involve execution of the system code, and dynamic testing techniques, which do. In the next two sections the author reviews first the applicability to expert systems of conventional V and V techniques and, second, the techniques expert system developers actually use. In the last section the authors make some general observations

Fluor Daniel Hanford Inc. integrated safety management system phase 1 verification final report

International Nuclear Information System (INIS)

PARSONS, J.E.

1999-01-01

The purpose of this review is to verify the adequacy of documentation as submitted to the Approval Authority by Fluor Daniel Hanford, Inc. (FDH). This review is not only a review of the Integrated Safety Management System (ISMS) System Description documentation, but is also a review of the procedures, policies, and manuals of practice used to implement safety management in an environment of organizational restructuring. The FDH ISMS should support the Hanford Strategic Plan (DOE-RL 1996) to safely clean up and manage the site's legacy waste; deploy science and technology while incorporating the ISMS theme to ''Do work safely''; and protect human health and the environment

European passive plant program preliminary safety analyses to support system design

International Nuclear Information System (INIS)

Saiu, Gianfranco; Barucca, Luciana; King, K.J.

1999-01-01

In 1994, a group of European Utilities, together with Westinghouse and its Industrial Partner GENESI (an Italian consortium including ANSALDO and FIAT), initiated a program designated EPP (European Passive Plant) to evaluate Westinghouse Passive Nuclear Plant Technology for application in Europe. In the Phase 1 of the European Passive Plant Program which was completed in 1996, a 1000 MWe passive plant reference design (EP1000) was established which conforms to the European Utility Requirements (EUR) and is expected to meet the European Safety Authorities requirements. Phase 2 of the program was initiated in 1997 with the objective of developing the Nuclear Island design details and performing supporting analyses to start development of Safety Case Report (SCR) for submittal to European Licensing Authorities. The first part of Phase 2, 'Design Definition' phase (Phase 2A) was completed at the end of 1998, the main efforts being design definition of key systems and structures, development of the Nuclear Island layout, and performing preliminary safety analyses to support design efforts. Incorporation of the EUR has been a key design requirement for the EP1000 form the beginning of the program. Detailed design solutions to meet the EUR have been defined and the safety approach has also been developed based on the EUR guidelines. The present paper describes the EP1000 approach to safety analysis and, in particular, to the Design Extension Conditions that, according to the EUR, represent the preferred method for giving consideration to the Complex Sequences and Severe Accidents at the design stage without including them in the design bases conditions. Preliminary results of some DEC analyses and an overview of the probabilistic safety assessment (PSA) are also presented. (author)

Safety Justification of Software Systems. Software Based Safety Systems. Regulatory Inspection Handbook

International Nuclear Information System (INIS)

Dahll, Gustav; Liwang, Bo; Wainwright, Norman

2006-01-01

The introduction of new software based technology in the safety systems in nuclear power plants also makes it necessary to develop new strategies for regulatory review and assessment of these new systems that is more focused on reviewing the processes at the different phases in design phases during the system life cycle. It is a general requirement that the licensee shall perform different kinds of reviews. From a regulatory point of view it is more cost effective to assess that the design activities at the suppliers and the review activities within the development project are performed with good quality. But the change from more technical reviews over to the development process oriented approach also cause problems. When reviewing development and quality aspects there are no 'hard facts' that can be judged against some specified criteria, the issues are more 'soft' and are more to build up structure of arguments and evidences that the requirements are met. The regulatory review strategy must therefore change to follow the development process over the whole life cycle from concept phase until installation and operation. Even if we know what factors that is of interest we need some guidance on how to interpret and judge the information.For that purpose SKl started research activities in this area at the end of the 1990s. In the first phase, in co-operation with Gustav Dahll at the Halden project, a life cycle model was selected. For the different phases a qualitative influence net was constructed of the type that is used in Bayesian Believe Network together with a discussion on different issues involved. In the second phase of the research work, in co-operation with Norman Wainwright, a former NII inspector, information from a selection of the most important sources as guidelines, IAEA and EC reports etc, was mapped into the influence net structure (the total list on used sources are in the report). The result is presented in the form of questions (Q) and a

Safety Justification of Software Systems. Software Based Safety Systems. Regulatory Inspection Handbook

Energy Technology Data Exchange (ETDEWEB)

Dahll, Gustav (OECD Halden Project, Halden (NO)); Liwaang, Bo (Swedish Nuclear Power Inspectorate, Stockholm (Sweden)); Wainwright, Norman (Wainwright Safety Advice (GB))

2006-07-01

The introduction of new software based technology in the safety systems in nuclear power plants also makes it necessary to develop new strategies for regulatory review and assessment of these new systems that is more focused on reviewing the processes at the different phases in design phases during the system life cycle. It is a general requirement that the licensee shall perform different kinds of reviews. From a regulatory point of view it is more cost effective to assess that the design activities at the suppliers and the review activities within the development project are performed with good quality. But the change from more technical reviews over to the development process oriented approach also cause problems. When reviewing development and quality aspects there are no 'hard facts' that can be judged against some specified criteria, the issues are more 'soft' and are more to build up structure of arguments and evidences that the requirements are met. The regulatory review strategy must therefore change to follow the development process over the whole life cycle from concept phase until installation and operation. Even if we know what factors that is of interest we need some guidance on how to interpret and judge the information.For that purpose SKl started research activities in this area at the end of the 1990s. In the first phase, in co-operation with Gustav Dahll at the Halden project, a life cycle model was selected. For the different phases a qualitative influence net was constructed of the type that is used in Bayesian Believe Network together with a discussion on different issues involved. In the second phase of the research work, in co-operation with Norman Wainwright, a former NII inspector, information from a selection of the most important sources as guidelines, IAEA and EC reports etc, was mapped into the influence net structure (the total list on used sources are in the report). The result is presented in the form of

Safety analysis report for the cold vacuum drying facility, phase 1, supporting civil/structural construction

International Nuclear Information System (INIS)

Pili-Vincens, C.

1998-01-01

The Cold Vacuum Drying Facility is a subproject of the overall Spent Nuclear Fuel Project. This Phase 2 Safety Analysis Report incorporates the CVD systems design and will update the SAR per DOE Order 5480.23 for manual and other Hanford infrastructure changes

Fluor Daniel Hanford Inc. integrated safety management system phase 1 verification final report

Energy Technology Data Exchange (ETDEWEB)

PARSONS, J.E.

1999-10-28

The purpose of this review is to verify the adequacy of documentation as submitted to the Approval Authority by Fluor Daniel Hanford, Inc. (FDH). This review is not only a review of the Integrated Safety Management System (ISMS) System Description documentation, but is also a review of the procedures, policies, and manuals of practice used to implement safety management in an environment of organizational restructuring. The FDH ISMS should support the Hanford Strategic Plan (DOE-RL 1996) to safely clean up and manage the site's legacy waste; deploy science and technology while incorporating the ISMS theme to ''Do work safely''; and protect human health and the environment.

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.

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.

Robotic and nuclear safety for an automated/teleoperated glove box system

International Nuclear Information System (INIS)

Domning, E.E.; McMahon, T.T.; Sievers, R.H.

1991-09-01

Lawrence Livermore National Laboratory (LLNL) is developing a fully automated system to handle the processing of special nuclear materials (SNM). This work is performed in response to the new goals at the Department of Energy (DOE) for hazardous waste minimization and radiation dose reduction. This fully automated system, called the automated test bed (ATB), consists of an IBM gantry robot and automated processing equipment sealed within a glove box. While the ATB is a cold system, we are designing it as a prototype of the future hot system. We recognized that identification and application of safety requirements early in the design phase will lead to timely installation and approval of the hot system. This paper identifies these safety issues as well as the general safety requirements necessary for the safe operation of the ATB. 4 refs., 2 figs

The unique safety challenges of space reactor systems

International Nuclear Information System (INIS)

Lanes, S.J.; Marshall, A.C.

1991-01-01

Compact reactor systems can provide high levels of power for extended periods in space environments. Their relatively low mass and their ability to operate independently of their proximity to the sun make reactor power systems high desirable for many civilian and military space missions. The US Department of Energy is developing reactor system technologies to provide electrical power for space applications. In addition, reactors are now being considered to provide thermal power to a hydrogen propellant for nuclear thermal rocketry. Space reactor safety issues differ from commercial reactor issues, in some areas, because of very different operating requirements and environments. Accidents similar to those postulated for commercial reactors must be considered for space reactors during their operational phase. Safety strategies will need to be established that account for the consequences of the loss of essential power

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?

Operational safety system performance alternative to the WANO's indicator

International Nuclear Information System (INIS)

Lyra, Moacir

2002-01-01

One of the operational safety performance indicators recommended by the World Association of Nuclear Operators (WANO) and adopted by Electronuclear is the reliability of the safety systems. The parameter selected to represent this indicator is the average unavailability of the trains of the concerned system. This parameter would be universally representative of the reliability for comparison purpose only if all nuclear power plants were designed within the same redundancy criteria. Considering the diversity of design criteria of the power plants in operation and based on a probabilistic approach, this paper proposes new performance indicators which are comparable regardless the redundancy criteria of the system. A case example applied to a system of the Angra 2 nuclear power plant shows that, even though with the plant in the infancy phase, the performance of the system in the period is very good. (author)

Electric capacitance tomography and two-phase flow for the nuclear reactor safety analysis

International Nuclear Information System (INIS)

Lee, Jae Young

2008-01-01

Recently electric capacitance tomography has been developed to be used in the analysis of two-phase flow. Although its electric field is not focused as the hard ray tomography such as the X-ray or gamma ray, its convenience of easy access to the system and easy maintenance due to no requirement of radiation shielding benefits us in its application in the two-phase flow study, one of important area in the nuclear safety analysis. In the present paper, the practical technologies in the electric capacitance tomography are represented in both parts of hardware and software. In the software part, both forward problem and inverse problem are discussed and the method of regularization. In the hardware part, the brief discussion of the electronics circuits is made which provides femto farad resolution with a reasonable speed (150 frame/sec for 16 electrodes). Some representative ideal cases are studied to demonstrate its potential capability for the two-phase flow analysis. Also, some variations of the tomography such as axial tomography, and three dimensional tomography are discussed. It was found that the present ECT is expected to become a useful tool to understand the complicated three dimensional two-phase flow which may be an important feature to be equipped by the safety analysis codes. (author)

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.

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

Radioactive wastes. Safety of storage facilities

International Nuclear Information System (INIS)

Devillers, Ch.

2001-01-01

A radioactive waste storage facility is designed in a way that ensures the isolation of wastes with respect to the biosphere. This function comprises the damping of the gamma and neutron radiations from the wastes, and the confinement of the radionuclides content of the wastes. The safety approach is based on two time scales: the safety of the insulation system during the main phase of radioactive decay, and the assessment of the radiological risks following this phase. The safety of a surface storage facility is based on a three-barrier concept (container, storage structures, site). The confidence in the safety of the facility is based on the quality assurance of the barriers and on their surveillance and maintenance. The safety of a deep repository will be based on the site quality, on the design and construction of structures and on the quality of the safety demonstration. This article deals with the safety approach and principles of storage facilities: 1 - recall of the different types of storage facilities; 2 - different phases of the life of a storage facility and regulatory steps; 3 - safety and radiation protection goals (time scales, radiation protection goals); 4 - safety approach and principles of storage facilities: safety of the isolation system (confinement system, safety analysis, scenarios, radiological consequences, safety principles), assessment of the radiation risks after the main phase of decay; 5 - safety of surface storage facilities: safety analysis of the confinement system of the Aube plant (barriers, scenarios, modeling, efficiency), evaluation of radiological risks after the main phase of decay; experience feedback of the Manche plant; variants of surface storage facilities in France and abroad (very low activity wastes, mine wastes, short living wastes with low and average activity); 6 - safety of deep geological disposal facilities: legal framework of the French research; international context; safety analysis of the confinement system

23 CFR 972.212 - Federal lands safety management system (SMS).

Science.gov (United States)

2010-04-01

... safety is considered and implemented as appropriate in all phases of transportation system planning... basic framework for a SMS: (1) An ongoing program for the collection, maintenance and reporting of a... correct the identified hazards and potential hazards. (3) A process for communication, coordination, and...

EMS helicopter incidents reported to the NASA Aviation Safety Reporting System

Science.gov (United States)

Connell, Linda J.; Reynard, William D.

1993-01-01

The objectives of this evaluation were to: Identify the types of safety-related incidents reported to the Aviation Safety Reporting System (ASRS) in Emergency Medical Service (EMS) helicopter operations; Describe the operational conditions surrounding these incidents, such as weather, airspace, flight phase, time of day; and Assess the contribution to these incidents of selected human factors considerations, such as communication, distraction, time pressure, workload, and flight/duty impact.

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)

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

What do implementers need in terms of regulatory safety criteria for the post-closure phase?

International Nuclear Information System (INIS)

Cahen, B.

2010-01-01

Bruno Cahen, Director Safety Division (ANDRA) presented the point of view of the NEA Integration Group for the Safety Case (IGSC) on 'What do implementers need in terms of regulatory safety criteria for the post-closure phase?' B. Cahen acknowledged that the national experience in siting and developing conceptual designs of geological disposal is growing rapidly. It implies increasing opportunities for interactions between implementers and regulators. There has been large development of international guidance in the recent years. Many regulators have already developed a regulatory framework. The implementers need practical, transparent and deliverable regulations. These regulations should draw on experiences gained from development of geological disposal projects. The IGSC has identified five key questions that the RF may focus on: 1. Over what time frame are the waste deemed to present a hazard? 2. Over what time frames are regulatory criteria applied and do they change over time? 3. Over what time frame(s) are safety assessments required to be conducted? 4. How do implementers have to address uncertainties in the long time frames? 5. What happens after cut-offs: are additional analyses needed? What types of arguments are to be used? Stable, understandable and practical criteria mean, namely, that they need to be developed on a strong scientific and societal basis, that there is consistency of safety options and requirements for different types of waste, that, in the longer time frames, the emphasis is given to robust systems, passive safety and multiple safety functions and that the criteria should fit the various phases of the project (siting, designing, operating, closure and post-closure). Experience feedback from safety cases shows that safety priorities depend very much on time frames. The derived safety criteria for the individual components should lead to measurable, verifiable specifications. The assessment of geological repository post-closure safety

Obtaining Valid Safety Data for Software Safety Measurement and Process Improvement

Science.gov (United States)

Basili, Victor r.; Zelkowitz, Marvin V.; Layman, Lucas; Dangle, Kathleen; Diep, Madeline

2010-01-01

We report on a preliminary case study to examine software safety risk in the early design phase of the NASA Constellation spaceflight program. Our goal is to provide NASA quality assurance managers with information regarding the ongoing state of software safety across the program. We examined 154 hazard reports created during the preliminary design phase of three major flight hardware systems within the Constellation program. Our purpose was two-fold: 1) to quantify the relative importance of software with respect to system safety; and 2) to identify potential risks due to incorrect application of the safety process, deficiencies in the safety process, or the lack of a defined process. One early outcome of this work was to show that there are structural deficiencies in collecting valid safety data that make software safety different from hardware safety. In our conclusions we present some of these deficiencies.

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)

Safety relevant failure mechanisms in the post-operational phase

International Nuclear Information System (INIS)

Mayer, Gerhard; Stiller, Jan Christopher; Roemer, Sarah

2017-03-01

When the 13"t"h amendment of the Atomic Energy Act came into force, eight Germ an nuclear power plant units had their power operating licences revoked and are now in the so-called post operation phase. Of the remaining nuclear power plants, one have by now also entered the post operation phase, with those left in operation bound for entering this phase sometime between now and the end of 2022. Therefore, failure mechanisms that are particularly relevant for post operation were to be identified and described in the frame of the present project. To do so, three major steps were taken: Firstly, recent national and international pertinent literature was evaluated to obtain indications of failure mechanisms in the post operation phase. It turned out that most of the national and international literature deals with the general procedure of the transition from power operation to decommissioning and dismantling. However, there were also some documents providing detailed indications of possible failure mechanisms in post operation. This includes e.g. the release of radioactive materials caused by the drop of containers, chemical impacts on systems important to safety in connection with decontamination work, and corrosion in connection with the storage of the core in the spent fuel pool, with the latter leading to the jamming of the fuel assemblies in the storage racks and a possible reduction of coolant circulation. In a second step, three safety analyses of pressurised water reactors prepared by the respective plant operators were evaluated to identify failure mechanisms based on systems engineering. The failure mechanisms that were found here include e.g. faults in the boric acid concentration of the reactor coolant, damage to the equipment airlock upon the unloading of Castor casks, leakages in connection with primary system decontamination, and the drop of packages holding radioactive residual materials or waste with subsequent mobilisation of radioactive aerosols

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

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

A study of leading indicators for occupational health and safety management systems in healthcare.

Science.gov (United States)

Almost, Joan M; VanDenKerkhof, Elizabeth G; Strahlendorf, Peter; Caicco Tett, Louise; Noonan, Joanna; Hayes, Thomas; Van Hulle, Henrietta; Adam, Ryan; Holden, Jeremy; Kent-Hillis, Tracy; McDonald, Mike; Paré, Geneviève C; Lachhar, Karanjit; Silva E Silva, Vanessa

2018-04-23

In Ontario, Canada, approximately $2.5 billion is spent yearly on occupational injuries in the healthcare sector. The healthcare sector has been ranked second highest for lost-time injury rates among 16 Ontario sectors since 2009 with female healthcare workers ranked the highest among all occupations for lost-time claims. There is a great deal of focus in Ontario's occupational health and safety system on compliance and fines, however despite this increased focus, the injury statistics are not significantly improving. One of the keys to changing this trend is the development of a culture of healthy and safe workplaces including the effective utilization of leading indicators within Occupational Health and Safety Management Systems (OHSMSs). In contrast to lagging indicators, which focus on outcomes retrospectively, a leading indicator is associated with proactive activities and consists of selected OHSMSs program elements. Using leading indicators to measure health and safety has been common practice in high-risk industries; however, this shift has not occurred in healthcare. The aim of this project is to conduct a longitudinal study implementing six elements of the Ontario Safety Association for Community and Healthcare (OSACH) system identified as leading indicators and evaluating the effectiveness of this intervention on improving selected health and safety workplace indicators. A quasi-experimental longitudinal research design will be used within two Ontario acute care hospitals. The first phase of the study will focus on assessing current OHSMSs using the leading indicators, determining potential facilitators and barriers to changing current OHSMSs, and identifying the leading indicators that could be added or changed to the existing OHSMS in place. Phase I will conclude with the development of an intervention designed to support optimizing current OHSMSs in participating hospitals based on identified gaps. Phase II will pilot test and evaluate the tailored

System Safety Program Plan for Project W-314, tank farm restoration and safe operations

International Nuclear Information System (INIS)

Boos, K.A.

1996-01-01

This System Safety Program Plan (SSPP) outlines the safety analysis strategy for project W-314, ''Tank Farm Restoration and Safe Operations.'' Project W-314 will provide capital improvements to Hanford's existing Tank Farm facilities, with particular emphasis on infrastructure systems supporting safe operation of the double-shell activities related to the project's conceptual Design Phase, but is planned to be updated and maintained as a ''living document'' throughout the life of the project to reflect the current safety analysis planning for the Tank Farm Restoration and Safe Operations upgrades. This approved W-314 SSPP provides the basis for preparation/approval of all safety analysis documentation needed to support the project

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.

Software safety analysis on the model specified by NuSCR and SMV input language at requirements phase of software development life cycle using SMV

International Nuclear Information System (INIS)

Koh, Kwang Yong; Seong, Poong Hyun

2005-01-01

Safety-critical software process is composed of development process, verification and validation (V and V) process and safety analysis process. Safety analysis process has been often treated as an additional process and not found in a conventional software process. But software safety analysis (SSA) is required if software is applied to a safety system, and the SSA shall be performed independently for the safety software through software development life cycle (SDLC). Of all the phases in software development, requirements engineering is generally considered to play the most critical role in determining the overall software quality. NASA data demonstrate that nearly 75% of failures found in operational software were caused by errors in the requirements. The verification process in requirements phase checks the correctness of software requirements specification, and the safety analysis process analyzes the safety-related properties in detail. In this paper, the method for safety analysis at requirements phase of software development life cycle using symbolic model verifier (SMV) is proposed. Hazard is discovered by hazard analysis and in other to use SMV for the safety analysis, the safety-related properties are expressed by computation tree logic (CTL)

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.

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.

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

Phase 1 space fission propulsion system design considerations

International Nuclear Information System (INIS)

Houts, Mike; Van Dyke, Melissa; Godfroy, Tom; Pedersen, Kevin; Martin, James; Dickens, Ricky; Salvail, Pat; Hrbud, Ivana; Carter, Robert

2002-01-01

Fission technology can enable rapid, affordable access to any point in the solar system. If fission propulsion systems are to be developed to their full potential; however, near-term customers must be identified and initial fission systems successfully developed, launched, and operated. Studies conducted in fiscal year 2001 (IISTP, 2001) show that fission electric propulsion (FEP) systems operating at 80 kWe or above could enhance or enable numerous robotic outer solar system missions of interest. At these power levels it is possible to develop safe, affordable systems that meet mission performance requirements. In selecting the system design to pursue, seven evaluation criteria were identified: safety, reliability, testability, specific mass, cost, schedule, and programmatic risk. A top-level comparison of three potential concepts was performed: an SP-100 based pumped liquid lithium system, a direct gas cooled system, and a heatpipe cooled system. For power levels up to at least 500 kWt (enabling electric power levels of 125-175 kWe, given 25-35% power conversion efficiency) the heatpipe system has advantages related to several criteria and is competitive with respect to all. Hardware-based research and development has further increased confidence in the heatpipe approach. Successful development and utilization of a 'Phase 1' fission electric propulsion system will enable advanced Phase 2 and Phase 3 systems capable of providing rapid, affordable access to any point in the solar system

Preparation of Phased and Merged Safety Analysis Reports for New DOE Nuclear Facilities

International Nuclear Information System (INIS)

BISHOP, G.E.

2000-01-01

The Spent Nuclear Fuels Project (SNFP) is charged with moving to storage 2,100 metric tons of spent nuclear fuel elements left over from plutonium production at DOE'S Hanford site in Washington state. Two new facilities, the Cold Vacuum Drying Facility (CVDF) and the Canister Storage Building (CSB) are in final construction. In order to meet aggressive schedule commitments, the SNFP chose to prepare the safety analysis reports (SAR's) in phases that covered only specific portions of each facility's design as it was built. Each SAR also merged the preliminary and final safety analysis reports into a single SAR, thereby covering all aspects of design, construction, and operation for that portion (phase) of the facility. A policy of ''NRC equivalency'' was also implemented in parallel with this effort, with the goal of achieving a rigor of safety analysis equivalent to that of NRC-licensed fuel processing facilities. DOE Order 5480.23. ''Nuclear Safety Analysis Reports'' allows preparation of both a phased and a merged SAR to accelerate construction schedules. However, project managers must be aware that such acceleration is not guaranteed. Managers considering this approach for their project should be cognizant of numerous obstacles that will be encountered. Merging and phasing SAR's will create new, unique, and unanticipated difficulties which may actually slow construction unless expeditiously and correctly managed. Pitfalls to be avoided and good practices to be implemented in preparing phased and merged SAR's are presented. The value of applying NRC requirements to the DOE safety analysis process is also discussed. As of December, 1999, the SNFP has completed and approved a SAR for the CVDF. Approval of the SAR for the CSB is pending

Safety and reliability of a multiphase pump system (MPA). Phase 1b; Sicherheit und Zuverlaessigkeit eines Mehrphasen-Pumpen Systems (MPA). Phase 1b

Energy Technology Data Exchange (ETDEWEB)

Wuersig, G.; Woehren, N.

2001-07-01

Since October 1997 Germanischer Lloyd, Johann Heinrich Bornemann GmbH (JHB) and other partners are working together to develop a multiphase pumping unit for sub sea application. Key component of the system is the twin screw pump developed by JHB. A very long running time between possible maintenance work is required due to the application of the system in very deep water. The research work reported are basics for the development of a real time condition monitoring system which currently is developed by Germanischer Lloyd within the scope of the next phase of the project. The following tasks have been part of Germanischer Lloyd work in MTK 616: (a) Evaluation of the mechanical system structure by use of a 3-dimentional coupled finite element model (Section 6.) of the pump. (b) First experiments on a test pump at JHB in Obernkirchen (Section 7.). (c) Systematic evaluation of possible failure modes (Section 8.1) to define the safety relevant modes and for the technical interpretation of system behaviour. (d) Evaluation of the fluid dynamic system behaviour (Section 8.2) as basis for the planned development of the simulation model which will be part of the condition monitoring system. The results are the scientific and technical bases for the work in the running project supported by BMBF under identification number MTK 623. (orig.) [German] Seit Oktober 1997 arbeitet der Germanische Lloyd zusammen mit der Firma Johann Heinrich Bornemann GmbH (JHB) und anderen Partnern an der Entwicklung von unter Wasser einsetzbaren Systemen zur Foerderung von Oel/Gas/Feststoff-Gemischen. Zentrales Element ist dabei die von JHB entwickelte Schraubenspindelpumpe mit sehr langen wartungsfreien Laufzeiten. Die Arbeiten, ueber die hier berichtet wird, sind Voraussetzung fuer die Entwicklung eines Systems zur Zustandsbeurteilung in einem anschliessenden Projekt und umfassen: (a) Mechanische konstruktionsbegleitende Stukturanalyse (vgl. Abschnitt 6) der Pumpe mit Hilfe von gekoppelten

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

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

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

NuSEE: an integrated environment of software specification and V and V for PLC based safety-critical systems

International Nuclear Information System (INIS)

Koo, Seo Ryong; Seong, Poong Hyun; Yoo, Jun Beom; Cha, Sung Deok; Youn, Cheong; Han, Hyun Chul

2006-01-01

As the use of digital systems becomes more prevalent, adequate techniques for software specification and analysis have become increasingly important in Nuclear Power Plant (NPP) safety-critical systems. Additionally, the importance of software Verification and Validation (V and V) based on adequate specification has received greater emphasis in view of improving software quality. For thorough V and V of safety-critical systems, V and V should be performed throughout the software lifecycle. However, systematic V and V is difficult as it involves many manual-oriented tasks. Tool support is needed in order to more conveniently perform software V and V. In response, we developed four kinds of Computer Aided Software Engineering (CASE) tools to support system specification for a formal-based analysis according to the software lifecycle. In this work, we achieved optimized integration of each tool. The toolset, NuSEE, is an integrated environment for software specification and V and V for PLC based safety-critical systems. In accordance with the software lifecycle, NuSEE consists of NuSISRT for the concept phase, NuSRS for the requirements phase, NuSDS for the design phase and NuSCM for configuration management. It is believed that after further development our integrated environment will be a unique and promising software specification and analysis toolset that will support the entire software lifecycle for the development of PLC based NPP safety-critical systems

Office of River Protection Integrated Safety Management System Phase 1 Verification Corrective Action Plan

International Nuclear Information System (INIS)

CLARK, D.L.

1999-01-01

The purpose of this Corrective Action Plan is to demonstrate the OW planned and/or completed actions to implement ISMS as well as prepare for the RPP ISMS Phase II Verification scheduled for August, 1999. This Plan collates implied or explicit ORP actions identified in several key ISMS documents and aligns those actions and responsibilities perceived necessary to appropriately disposition all ISM Phase II preparation activities specific to the ORP. The objective will be to complete or disposition the corrective actions prior to the commencement of the ISMS Phase II Verification. Improvement products/tasks not slated for completion prior to the RPP Phase II verification will be incorporated as corrective actions into the Strategic System Execution Plan (SSEP) Gap Analysis. Many of the business and management systems that were reviewed in the ISMS Phase I verification are being modified to support the ORP transition and are being assessed through the SSEP. The actions and processes identified in the SSEP will support the development of the ORP and continued ISMS implementation as committed to be complete by end of FY-2000

Electronic clinical safety reporting system: a benefits evaluation.

Science.gov (United States)

Elliott, Pamela; Martin, Desmond; Neville, Doreen

2014-06-11

Eastern Health, a large health care organization in Newfoundland and Labrador (NL), started a staged implementation of an electronic occurrence reporting system (used interchangeably with "clinical safety reporting system") in 2008, completing Phase One in 2009. The electronic clinical safety reporting system (CSRS) was designed to replace a paper-based system. The CSRS involves reporting on occurrences such as falls, safety/security issues, medication errors, treatment and procedural mishaps, medical equipment malfunctions, and close calls. The electronic system was purchased from a vendor in the United Kingdom that had implemented the system in the United Kingdom and other places, such as British Columbia. The main objective of the new system was to improve the reporting process with the goal of improving clinical safety. The project was funded jointly by Eastern Health and Canada Health Infoway. The objectives of the evaluation were to: (1) assess the CSRS on achieving its stated objectives (particularly, the benefits realized and lessons learned), and (2) identify contributions, if any, that can be made to the emerging field of electronic clinical safety reporting. The evaluation involved mixed methods, including extensive stakeholder participation, pre/post comparative study design, and triangulation of data where possible. The data were collected from several sources, such as project documentation, occurrence reporting records, stakeholder workshops, surveys, focus groups, and key informant interviews. The findings provided evidence that frontline staff and managers support the CSRS, identifying both benefits and areas for improvement. Many benefits were realized, such as increases in the number of occurrences reported, in occurrences reported within 48 hours, in occurrences reported by staff other than registered nurses, in close calls reported, and improved timelines for notification. There was also user satisfaction with the tool regarding ease of use

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

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)

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

An integrated environment of software development and V and V for PLC based safety-critical systems

International Nuclear Information System (INIS)

Koo, Seo Ryong

2005-02-01

To develop and implement a safety-critical system, the requirements of the system must be analyzed thoroughly during the phases of a software development's life cycle because a single error in the requirements can generate serious software faults. We therefore propose an Integrated Environment (IE) approach for requirements which is an integrated approach that enables easy inspection by combining requirement traceability and effective use of a formal method. For the V and V tasks of requirements phase, our approach uses software inspection, requirement traceability, and formal specification with structural decomposition. Software inspection and the analysis of requirements traceability are the most effective methods of software V and V. Although formal methods are also considered an effective V and V activity, they are difficult to use properly in nuclear fields, as well as in other fields, because of their mathematical nature. We also propose another Integrated Environment (IE) for the design and implementation of safety-critical systems. In this study, a nuclear FED-style design specification and analysis (NuFDS) approach was proposed for PLC based safety-critical systems. The NuFDS approach is suggested in a straightforward manner for the effective and formal specification and analysis of software designs. Accordingly, the proposed NuFDS approach comprises one technique for specifying the software design and another for analyzing the software design. In addition, with the NuFDS approach, we can analyze the safety of software on the basis of fault tree synthesis. To analyze the design phase more effectively, we propose a technique of fault tree synthesis, along with a universal fault tree template for the architecture modules of nuclear software. Various tools have been needed to make software V and V more convenient. We therefore developed four kinds of computer-aided software engineering tools that could be used in accordance with the software's life cycle to

Industrial safety and fire protection during the construction phase

International Nuclear Information System (INIS)

Zimmer, K.

1977-01-01

The questions and problems of industrial safety and fire prevention have to be treated like the activities planning, developing, assembly, etc. This statement is illustrated by statistics of the fire insurance companies, from which it can be seen that the number of fire accidents has decreased but that the damage caused has greatly increased. The sooner the fire prevention and industrial safety measures are integrated in the planning phase, the better for the total costs. Preventive measures that possibly have to be introduced at a later stage are not only generally much more expensive but are also seldom as effective. (orig./HK) [de

Non-Parametric, Closed-Loop Testing of Autonomy in Unmanned Aircraft Systems, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — The proposed Phase I program aims to develop new methods to support safety testing for integration of Unmanned Aircraft Systems into the National Airspace (NAS) with...

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 ...

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

Accelerated safety analyses - structural analyses Phase I - structural sensitivity evaluation of single- and double-shell waste storage tanks

International Nuclear Information System (INIS)

Becker, D.L.

1994-11-01

Accelerated Safety Analyses - Phase I (ASA-Phase I) have been conducted to assess the appropriateness of existing tank farm operational controls and/or limits as now stipulated in the Operational Safety Requirements (OSRs) and Operating Specification Documents, and to establish a technical basis for the waste tank operating safety envelope. Structural sensitivity analyses were performed to assess the response of the different waste tank configurations to variations in loading conditions, uncertainties in loading parameters, and uncertainties in material characteristics. Extensive documentation of the sensitivity analyses conducted and results obtained are provided in the detailed ASA-Phase I report, Structural Sensitivity Evaluation of Single- and Double-Shell Waste Tanks for Accelerated Safety Analysis - Phase I. This document provides a summary of the accelerated safety analyses sensitivity evaluations and the resulting findings

Nuclear criticality safety basics for personnel working with nuclear fissionable materials. Phase I

International Nuclear Information System (INIS)

Vausher, A.L.

1984-10-01

DOE order 5480.1A, Chapter V, ''Safety of Nuclear Facilities,'' establishes safety procedures and requirements for DOE nuclear facilities. The ''Nuclear Criticality Safety Basic Program - Phase I'' is documented in this report. The revised program has been developed to clearly illustrate the concept of nuclear safety and to help the individual employee incorporate safe behavior in his daily work performance. Because of this, the subject of safety has been approached through its three fundamentals: scientific basis, engineering criteria, and administrative controls. Only basics of these three elements were presented. 5 refs

Probabilistic safety goals. Phase 2 - Status report

International Nuclear Information System (INIS)

Holmberg, J.-E.; Bjoerkman, K.; Rossi, J.; Knochenhauer, M.; Xuhong He; Persson, A.; Gustavsson, H.

2008-07-01

The second phase of the project, the outcome of which is described in this project report has mainly dealt with four issues: 1) Consistency in the usage of safety goals 2) Criteria for assessment of results from PSA level 2 3) Overview of international safety goals and experiences from their use 4) Safety goals related to other man-made risks in society. Consistency in judgement over time has been perceived to be one of the main problems in the usage of safety goals. Safety goals defined in the 80ies were met in the beginning with PSA:s performed to the standards of that time, i.e., by PSA:s that were quite limited in scope and level of detail compared to today's state of the art. This issue was investigated by performing a comparative review was performed of three generations of the same PSA, focusing on the impact from changes over time in component failure data, IE frequency, and modelling of the plant, including plant changes and changes in success criteria. It proved to be very time-consuming and in some cases next to impossible to correctly identify the basic causes for changes in PSA results. A multitude of different sub-causes turned out to combined and difficult to differentiate. Thus, rigorous book-keeping is needed in order to keep track of how and why PSA results change. This is especially important in order to differentiate 'real' differences due to plant changes and updated component and IE data from differences that are due to general PSA development (scope, level of detail, modelling issues). (au)

Probabilistic safety goals. Phase 2 - Status report

Energy Technology Data Exchange (ETDEWEB)

Holmberg, J.-E.; Bjoerkman, K. Rossi, J. (VTT (Finland)); Knochenhauer, M.; Xuhong He; Persson, A.; Gustavsson, H. (Relcon Scandpower AB, Sundbyberg (Sweden))

2008-07-15

The second phase of the project, the outcome of which is described in this project report has mainly dealt with four issues: 1) Consistency in the usage of safety goals 2) Criteria for assessment of results from PSA level 2 3) Overview of international safety goals and experiences from their use 4) Safety goals related to other man-made risks in society. Consistency in judgement over time has been perceived to be one of the main problems in the usage of safety goals. Safety goals defined in the 80ies were met in the beginning with PSA:s performed to the standards of that time, i.e., by PSA:s that were quite limited in scope and level of detail compared to today's state of the art. This issue was investigated by performing a comparative review was performed of three generations of the same PSA, focusing on the impact from changes over time in component failure data, IE frequency, and modelling of the plant, including plant changes and changes in success criteria. It proved to be very time-consuming and in some cases next to impossible to correctly identify the basic causes for changes in PSA results. A multitude of different sub-causes turned out to combined and difficult to differentiate. Thus, rigorous book-keeping is needed in order to keep track of how and why PSA results change. This is especially important in order to differentiate 'real' differences due to plant changes and updated component and IE data from differences that are due to general PSA development (scope, level of detail, modelling issues). (au)

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

Safety in the Utilization and Modification of Research Reactors. Specific Safety Guide

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-07-15

This Safety Guide is a revision of Safety Series No. 35-G2 on safety in the utilization and modification of research reactors. It provides recommendations on meeting the requirements for the categorization, safety assessment and approval of research reactor experiments and modification projects. Specific safety considerations in different phases of utilization and modification projects are covered, including the pre-implementation, implementation and post-implementation phases. Guidance is also provided on the operational safety of experiments, including in the handling, dismantling, post-irradiation examination and disposal of experimental devices. Examples of the application of the safety categorization process for experiments and modification projects and of the content of the safety analysis report for an experiment are also provided. Contents: 1. Introduction; 2. Management system for the utilization and modification of a research reactor; 3. Categorization, safety assessment and approval of an experiment or modification; 4. Safety considerations for the design of an experiment or modification; 5. Pre-implementation phase of a modification or utilization project; 6. Implementation phase of a modification or utilization project; 7. Post-implementation phase of a utilization or modification project; 8. Operational safety of experiments at a research reactor; 9. Safety considerations in the handling, dismantling, post-irradiation examination and disposal of experimental devices; 10. Safety aspects of out-of-reactor-core installations; Annex I: Example of a checklist for the categorization of an experiment or modification at a research reactor; Annex II: Example of the content of the safety analysis report for an experiment at a research reactor; Annex III: Examples of reasons for a modification at a research reactor.

Safety in the Utilization and Modification of Research Reactors. Specific Safety Guide

International Nuclear Information System (INIS)

2012-01-01

This Safety Guide is a revision of Safety Series No. 35-G2 on safety in the utilization and modification of research reactors. It provides recommendations on meeting the requirements for the categorization, safety assessment and approval of research reactor experiments and modification projects. Specific safety considerations in different phases of utilization and modification projects are covered, including the pre-implementation, implementation and post-implementation phases. Guidance is also provided on the operational safety of experiments, including in the handling, dismantling, post-irradiation examination and disposal of experimental devices. Examples of the application of the safety categorization process for experiments and modification projects and of the content of the safety analysis report for an experiment are also provided. Contents: 1. Introduction; 2. Management system for the utilization and modification of a research reactor; 3. Categorization, safety assessment and approval of an experiment or modification; 4. Safety considerations for the design of an experiment or modification; 5. Pre-implementation phase of a modification or utilization project; 6. Implementation phase of a modification or utilization project; 7. Post-implementation phase of a utilization or modification project; 8. Operational safety of experiments at a research reactor; 9. Safety considerations in the handling, dismantling, post-irradiation examination and disposal of experimental devices; 10. Safety aspects of out-of-reactor-core installations; Annex I: Example of a checklist for the categorization of an experiment or modification at a research reactor; Annex II: Example of the content of the safety analysis report for an experiment at a research reactor; Annex III: Examples of reasons for a modification at a research reactor.

Intelligent Information Processing for Enhanced Safety in the NAS, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — Our Phase I work focused on how improved information flow between actors in a flight deck environment can improve safety performance. An operational prototype was...

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.

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)

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.

Dynamic modeling of the tradeoff between productivity and safety in critical engineering systems

International Nuclear Information System (INIS)

Cowing, Michelle M.; Elisabeth Pate-Cornell, M.; Glynn, Peter W.

2004-01-01

Short-term tradeoffs between productivity and safety often exist in the operation of critical facilities such as nuclear power plants, offshore oil platforms, or simply individual cars. For example, interruption of operations for maintenance on demand can decrease short-term productivity but may be needed to ensure safety. Operations are interrupted for several reasons: scheduled maintenance, maintenance on demand, response to warnings, subsystem failure, or a catastrophic accident. The choice of operational procedures (e.g. timing and extent of scheduled maintenance) generally affects the probabilities of both production interruptions and catastrophic failures. In this paper, we present and illustrate a dynamic probabilistic model designed to describe the long-term evolution of such a system through the different phases of operation, shutdown, and possibly accident. The model's parameters represent explicitly the effects of different components' performance on the system's safety and reliability through an engineering probabilistic risk assessment (PRA). In addition to PRA, a Markov model is used to track the evolution of the system and its components through different performance phases. The model parameters are then linked to different operations strategies, to allow computation of the effects of each management strategy on the system's long-term productivity and safety. Decision analysis is then used to support the management of the short-term trade-offs between productivity and safety in order to maximize long-term performance. The value function is that of plant managers, within the constraints set by local utility commissions and national (e.g. energy) agencies. This model is illustrated by the case of outages (planned and unplanned) in nuclear power plants to show how it can be used to guide policy decisions regarding outage frequency and plant lifetime, and more specifically, the choice of a reactor tripping policy as a function of the state of the

Thermal-hydraulic tests for reactor safety system

International Nuclear Information System (INIS)

Chun, Se Young; Chung, Moon Ki; Baek, Won Pil

2002-05-01

Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

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.

Kilowatt isotope power system. Phase II plan. Volume V. Safety, quality assurance and reliability

International Nuclear Information System (INIS)

1978-01-01

The development of a Kilowatt Isotope Power System (KIPS) was begun in 1975 for the purpose of satisfying the power requirements of satellites in the 1980's. The KIPS is a 238 PuO 2 -fueled organic Rankine cycle turbine power system to provide a design output of 500 to 2000 W. Included in this volume are: launch and flight safety considerations; quality assurance techniques and procedures to be followed through system fabrication, assembly and inspection; and the reliability program made up of reliability prediction analysis, failure mode analysis and criticality analysis

German risk study, phase (DRS-B)

International Nuclear Information System (INIS)

Werner, W.

1992-01-01

The first risk investigations were primarily intended to estimate the risk of accidents in nuclear power plants and to compare it with other natural risk and civilization risks. The American reactor safety study WASH 1400 and the German risk study phase A (DRS-A) gave a detailed analyses of the offsite consequences of accidents, especially the magnitude and frequency of health damage for the population. Risk investigations today are primarily used to examine the design of safety systems and to further develop the entire safety concept. Safety investigations have shown that nuclear power plants still possess safety reserves if safety systems do not operate as planned. These safety reserves can be exploited in the sense of a further development of safety by plant internal emergency measures. One purpose of risk analyses is to identify such measures and to evaluate their feasibility and effectiveness. The most important goals of the investigations in DRS-B were: Identification of vulnerabilities and possible safety improvements; determination of safety reserves during accident sequences exceeding the design limits; evaluation of plant internal emergency measures. Thus, goals in phase B compared with phase A have changed from investigations of the magnitude of damage to detailed analysis of the plant systems response under accident conditions. The magnitude of possible fission product releases is also determined in phase B. However, no new accident consequence calculations are performed. Figs and tabs

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.

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)

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.

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.

Phases, phase equilibria, and phase rules in low-dimensional systems

International Nuclear Information System (INIS)

Frolov, T.; Mishin, Y.

2015-01-01

We present a unified approach to thermodynamic description of one, two, and three dimensional phases and phase transformations among them. The approach is based on a rigorous definition of a phase applicable to thermodynamic systems of any dimensionality. Within this approach, the same thermodynamic formalism can be applied for the description of phase transformations in bulk systems, interfaces, and line defects separating interface phases. For both lines and interfaces, we rigorously derive an adsorption equation, the phase coexistence equations, and other thermodynamic relations expressed in terms of generalized line and interface excess quantities. As a generalization of the Gibbs phase rule for bulk phases, we derive phase rules for lines and interfaces and predict the maximum number of phases than may coexist in systems of the respective dimensionality

Status of the EU test blanket systems safety studies

International Nuclear Information System (INIS)

Panayotov, Dobromir; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

2015-01-01

Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

Status of the EU test blanket systems safety studies

Energy Technology Data Exchange (ETDEWEB)

Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

2015-10-15

Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

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

Probabilistic Safety Goals for Nuclear Power Plants; Phases 2-4 / Final Report

International Nuclear Information System (INIS)

Bengtsson, Lisa; Knochenhauer, Michael; Holmberg, Jan-Erik; Rossi, Jukka

2011-05-01

The outcome of a probabilistic safety assessment (PSA) for a nuclear power plant is a combination of qualitative and quantitative results. Quantitative results are typically presented as the Core Damage Frequency (CDF) and as the frequency of an unacceptable radioactive release. In order to judge the acceptability of PSA results, criteria for the interpretation of results and the assessment of their acceptability need to be defined. Safety goals are defined in different ways in different countries and also used differently. Many countries are presently developing them in connection to the transfer to risk-informed regulation of both operating nuclear power plants (NPP) and new designs. However, it is far from self-evident how probabilistic safety criteria should be defined and used. On one hand, experience indicates that safety goals are valuable tools for the interpretation of results from a probabilistic safety assessment (PSA), and they tend to enhance the realism of a risk assessment. On the other hand, strict use of probabilistic criteria is usually avoided. A major problem is the large number of different uncertainties in a PSA model, which makes it difficult to demonstrate the compliance with a probabilistic criterion. Further, it has been seen that PSA results can change a lot over time due to scope extensions, revised operating experience data, method development, changes in system requirements, or increases of level of detail, mostly leading to an increase of the frequency of the calculated risk. This can cause a problem of consistency in the judgments. The first phase of the project (2006) provided a general description of the issue of probabilistic safety goals for nuclear power plants, of important concepts related to the definition and application of safety goals, and of experiences in Finland and Sweden. The second, third and fourth phases (2007-2009) have been concerned with providing guidance related to the resolution of some of the problems

Safety, tolerability, and systemic absorption of dapivirine vaginal microbicide gel in healthy, HIV-negative women

NARCIS (Netherlands)

Nel, Annalene M.; Coplan, Paul; van de Wijgert, Janneke H.; Kapiga, Saidi H.; von Mollendorf, Claire; Geubbels, Eveline; Vyankandondera, Joseph; Rees, Helen V.; Masenga, Gileard; Kiwelu, Ireen; Moyes, Jocelyn; Smythe, Shanique C.

2009-01-01

To assess the local and systemic safety of dapivirine vaginal gel vs. placebo gel as well as the systemic absorption of dapivirine in healthy, HIV-negative women. Two prospective, randomized, double-blind, placebo-controlled phase I/II studies were conducted at five research centers, four in Africa

The human component in the safety of complex systems

International Nuclear Information System (INIS)

Wahlstroem, B.

1986-02-01

The safety of nuclear power and other complex processes requires that human actions are carried though on time and without error. Investigations indicate that human errors are the main or an important contributing cause in more than half of the incidents which occur. This makes it important to try understand the mechanisms behind the human errors and to investigate possibilities for decreasing their likelihood. The present report presents an overview of the Nordic cooperation in the field of human factors in nuclear safety, under the LIT-programme carried out 1981-1985. The work was divided into six different projects in the following fields: human reliability in test and maintenance work; safety oriented organizations and company structures; design of information and control systems; new approaches for information presentation; experimental validation of man-machine interfaces; planning and evaluation of operator training. The research topics were selected from the findings of an earlier phase of the Nordic cooperation. The results are described in more detail in separate reports

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)

Office of River Protection Integrated Safety Management System Phase 1 Verification Corrective Action Plan; FINAL

International Nuclear Information System (INIS)

CLARK, D.L.

1999-01-01

The purpose of this Corrective Action Plan is to demonstrate the OW planned and/or completed actions to implement ISMS as well as prepare for the RPP ISMS Phase II Verification scheduled for August, 1999. This Plan collates implied or explicit ORP actions identified in several key ISMS documents and aligns those actions and responsibilities perceived necessary to appropriately disposition all ISM Phase II preparation activities specific to the ORP. The objective will be to complete or disposition the corrective actions prior to the commencement of the ISMS Phase II Verification. Improvement products/tasks not slated for completion prior to the RPP Phase II verification will be incorporated as corrective actions into the Strategic System Execution Plan (SSEP) Gap Analysis. Many of the business and management systems that were reviewed in the ISMS Phase I verification are being modified to support the ORP transition and are being assessed through the SSEP. The actions and processes identified in the SSEP will support the development of the ORP and continued ISMS implementation as committed to be complete by end of FY-2000

A SIL quantification approach based on an operating situation model for safety evaluation in complex guided transportation systems

International Nuclear Information System (INIS)

Beugin, J.; Renaux, D.; Cauffriez, L.

2007-01-01

Safety analysis in guided transportation systems is essential to avoid rare but potentially catastrophic accidents. This article presents a quantitative probabilistic model that integrates Safety Integrity Levels (SIL) for evaluating the safety of such systems. The standardized SIL indicator allows the safety requirements of each safety subsystem, function and/or piece of equipment to be specified, making SILs pivotal parameters in safety evaluation. However, different interpretations of SIL exist, and faced with the complexity of guided transportation systems, the current SIL allocation methods are inadequate for the task of safety assessment. To remedy these problems, the model developed in this paper seeks to verify, during the design phase of guided transportation system, whether or not the safety specifications established by the transport authorities allow the overall safety target to be attained (i.e., if the SIL allocated to the different safety functions are sufficient to ensure the required level of safety). To meet this objective, the model is based both on the operating situation concept and on Monte Carlo simulation. The former allows safety systems to be formalized and their dynamics to be analyzed in order to show the evolution of the system in time and space, and the latter make it possible to perform probabilistic calculations based on the scenario structure obtained

Safety of Nuclear Power Plants: Design. Specific Safety Requirements

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

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.

Quality assurance of the modernized Dukovany I and C safety system software

International Nuclear Information System (INIS)

Karpeta, C.

2005-01-01

The approach to quality assurance of the software that implements the instrumentation and control functions for safety category A as per IEC 61226, which has been adopted within the 'NPP Dukovany I and C Refurbishment' project, is described. A survey of the requirements for software quality assurance of the systems that initiate protection interventions in the event of anticipated operational occurrences or accident conditions is given. The software development process applied by the system designers and manufacturers, from the software requirements specification phase to the software testing phase, is outlined. Basic information on technical audits of the software development process is also provided. (orig.)

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...

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.

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

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.

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

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.

[Road map for health and safety management systems in healthcare facilities, according to the OHSAS 18001:2007 standard].

Science.gov (United States)

Pugliese, F; Albini, E; Serio, O; Apostoli, P

2011-01-01

The 81/2008 Act has defined a model of a health and safety management system that can contribute to prevent the occupational health and safety risks. We have developed the structure of a health and safety management system model and the necessary tools for its implementation in health care facilities. The realization of a model is structured in various phases: initial review, safety policy, planning, implementation, monitoring, management review and continuous improvement. Such a model, in continuous evolution, is based on the responsibilities of the different corporate characters and on an accurate analysis of risks and involved norms.

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

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

YUCCA MOUNTAIN SITE CHARACTERIZATIONS PROJECT TUNNEL BORING MACHINE (TBM) SYSTEM SAFETY ANALYSIS

International Nuclear Information System (INIS)

1997-01-01

The purpose of this analysis is to systematically identify and evaluate hazards related to the tunnel boring machine (TBM) used in the Exploratory Studies Facility (ESF) at the Yucca Mountain Site Characterization Project. This process is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. Since the TBM is an ''as built'' system, the MandO is conducting the System Safety Analysis during the construction or assembly phase of the TBM. 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 TBM 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 system/subsystem/component design, (2) add safety features and capabilities to existing designs, and (3) 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. The scope of this analysis is limited to the TBM during normal operations, excluding hazards occurring during assembly and test of the TBM or maintenance of the TBM equipment

YUCCA MOUNTAIN SITE CHARACTERIZATIONS PROJECT TUNNEL BORING MACHINE (TBM) SYSTEM SAFETY ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

N/A

1997-02-19

The purpose of this analysis is to systematically identify and evaluate hazards related to the tunnel boring machine (TBM) used in the Exploratory Studies Facility (ESF) at the Yucca Mountain Site Characterization Project. This process is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. Since the TBM is an ''as built'' system, the M&O is conducting the System Safety Analysis during the construction or assembly phase of the TBM. 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 TBM 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 system/subsystem/component design, (2) add safety features and capabilities to existing designs, and (3) 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. The scope of this analysis is limited to the TBM during normal operations, excluding hazards occurring during assembly and test of the TBM or maintenance of the TBM equipment.

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...

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.

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

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

Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor

International Nuclear Information System (INIS)

Raisic, N.

1962-09-01

In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined

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.

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

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

NRC confirmatory safety system testing in support of AP600 design review

International Nuclear Information System (INIS)

Rhee, G.S.; Bessette, D.E.; Shotkin, L.M.

1994-01-01

Westinghouse Electric Corporation has submitted the Advanced Passive 600 MWe (AP600) nuclear power plant design to the NRC for design certification. The Office of Nuclear Regulatory Research is proceeding to conduct confirmatory testing to help the NRC staff evaluate the AP600 safety system design. For confirmatory testing, it was determined that the cost-effective route was to modify an existing full-height, full-pressure test facility rather than build a new one. Thus, all the existing integral effects test facilities, both in the US and abroad, were screened to select the best candidate. As a result, the ROSA-V (Rig of Safety Assessment-V) test facility located in the Japan Atomic Energy Research Institute (JAERI) was chosen. However, because of some differences in design between the existing ROSA-V facility and the AP600, the ROSA-V is being modified to conform to the AP600 safety system design. The modification work will be completed by the end of this year. A series of facility characterization tests will then be performed in January 1994 for the modified part of the facility before the main test series is initiated in February 1994. A total of 12 tests will be performed in 1994 under Phase I of this cooperative program with JAERI. Phase II testing is being considered to be conducted in 1995 mainly for beyond-design-basis accident evaluation

Intervention of French safety authorities during the design and construction phases of the Creys-Malville plant

International Nuclear Information System (INIS)

Orzoni, G.

1985-01-01

The intervention of French safety authorities during the design and construction phases of the Creys-Malville plant has been made by the different means of technical regulation, of several successive authorizations bound to different steps, and of numerous surveillance visits. Some safety-related problems have been met. Some of them are detailed, relating to the basis accident for containment design, decay heat removal, polar crane of reactor building, seismic resistance of main vessel internals, core cover plug, design and fabrication of steam generators. The main problems met during the design reviews and the construction phase of the plant have been solved in time; the safety level reached is provisionally judged acceptable by the French safety authorities

Efficacy and safety of tofacitinib in patients with active rheumatoid arthritis: review of key Phase 2 studies.

Science.gov (United States)

Fleischmann, Roy; Kremer, Joel; Tanaka, Yoshiya; Gruben, David; Kanik, Keith; Koncz, Tamas; Krishnaswami, Sriram; Wallenstein, Gene; Wilkinson, Bethanie; Zwillich, Samuel H; Keystone, Edward

2016-12-01

Tofacitinib is an oral Janus kinase inhibitor for the treatment of rheumatoid arthritis (RA). Here, the safety and efficacy data from five Phase 2 studies of tofacitinib in patients with RA are summarized. Tofacitinib 1-30 mg twice daily was investigated, as monotherapy and in combination with methotrexate, in patients with RA. Tofacitinib 20 mg once daily was investigated in one study. Tofacitinib 5 and 10 mg twice daily were selected for investigation in Phase 3 studies; therefore, the efficacy and safety of tofacitinib 5 and 10 mg twice daily in Phase 2 studies are the focus of this review. Tofacitinib ≥ 5 mg twice daily was efficacious in a dose-dependent manner, with statistically significant and clinically meaningful reductions in the signs and symptoms of RA and patient-reported outcomes. The safety profile was consistent across studies. The efficacy and safety profile of tofacitinib in Phase 2 studies supported its further investigation and the selection of tofacitinib 5 mg twice daily and tofacitinib 10 mg twice daily for evaluation in Phase 3 studies. © 2016 The Authors. International Journal of Rheumatic Diseases published by Asia Pacific League of Associations for Rheumatology and John Wiley & Sons Australia, Ltd.

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 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

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...

Comparative health and safety assessment of alternative future electrical-generation systems

International Nuclear Information System (INIS)

Habegger, L.J.; Gasper, J.R.; Brown, C.D.

1980-01-01

The report is an analysis of health and safety risks of seven alternative electrical generation systems, all of which have potential for commercial availability in the post-2000 timeframe. The systems are compared on the basis of expected public and occupational deaths and lost workdays per year associated with 1000 MWe average unit generation. Risks and their uncertainties are estimated for all phases of the energy production cycle, including fuel and raw material extraction and processing, direct and indirect component manufacture, on-site construction, and system operation and maintenance. Also discussed is the potential significance of related major health and safety issues that remain largely unquantifiable. The technologies include: the SPS; a low-Btu coal gasification system with an open-cycle gas turbine combined with a steam topping cycle (CG/CC); a light water fission reactor system without fuel reprocessing (LWR); a liquid metal fast breeder fission reactor system (LMFBR); a central station terrestrial photovoltaic system (CTPV); and a first generation fusion system with magnetic confinement. For comparison with the baseload technologies, risk from a decentralized roof-top photovoltaic system with 6 kWe peak capacity and battery storage (DTPV) was also evaluated

Verification and validation of the safety parameter display system for nuclear power plant

International Nuclear Information System (INIS)

Zhang Yuanfang

1993-05-01

During the design and development phase of the safety parameter display system for nuclear power plant, a verification and validation (V and V) plan has been implemented to improve the quality of system design. The V and V activities are briefly introduced, which were executed in four stages of feasibility research, system design, code development and system integration and regulation. The evaluation plan and the process of implementation as well as the evaluation conclusion of the final technical validation for this system are also presented in detail

Status of the design and safety project for the sodium-cooled fast reactor as a generation IV nuclear energy system

International Nuclear Information System (INIS)

Niwa, Hajime; Fiorini, Gian-Luigi; Sim, Yoon-Sub; Lennox, Tom; Cahalan, James E.

2005-01-01

The Design and Safety Project Management Board (DSPMB) was established under the Sodium Cooled Fast Reactor (SFR) System Steering Committee (SSC) in the Generation IV international Forum. The DSPMB will promote collaborative R and D activities on reactor core design, and safety assessment for candidate systems, and also integrate these results together with those from other PMBs such as advanced fuel and component to a whole fast reactor system in order to develop high performance systems that will satisfy the goals of Generation IV nuclear energy systems. The DSPMB has formulated the present R and D schedules for this purpose. Two SFR concepts were proposed: a loop-type system with primarily a MOX fuel core and a pool-type system with a metal fuel core. Study of innovative systems and their evaluation will also be included. The safety project will cover both the safety assessment of the design and the preparation of the methods/tools to be used for the assessment. After a rather short viability phase, the project will move to the performance phase for development of performance data and design optimization of conceptual designs. This paper describes the schedules, work packages and tasks for the collaborative studies of the member countries. (author)

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

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.

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)

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.)

Health and safety: Preliminary comparative assessment of the Satellite Power System (SPS) and other energy alternatives

Science.gov (United States)

Habegger, L. J.; Gasper, J. R.; Brown, C.

1980-01-01

Data readily available from the literature were used to make an initial comparison of the health and safety risks of a fission power system with fuel reprocessing; a combined-cycle coal power system with a low-Btu gasifier and open-cycle gas turbine; a central-station, terrestrial, solar photovoltaic power system; the satellite power system; and a first-generation fusion system. The assessment approach consists of the identification of health and safety issues in each phase of the energy cycle from raw material extraction through electrical generation, waste disposal, and system deactivation; quantitative or qualitative evaluation of impact severity; and the rating of each issue with regard to known or potential impact level and level of uncertainty.

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

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.

Safety balance: Analysis of safety systems; Bilans de surete: analyse par les organismes de surete

Energy Technology Data Exchange (ETDEWEB)

Delage, M; Giroux, C

1990-12-01

Safety analysis, and particularly analysis of exploitation of NPPs is constantly affected by EDF and by the safety authorities and their methodologies. Periodic safety reports ensure that important issues are not missed on daily basis, that incidents are identified and that relevant actions are undertaken. French safety analysis method consists of three principal steps. First type of safety balance is analyzed at the normal start-up phase for each unit including the final safety report. This enables analysis of behaviour of units ten years after their licensing. Second type is periodic operational safety analysis performed during a few years. Finally, the third step consists of safety analysis of the oldest units with the aim to improve the safety standards. The three steps of safety analysis are described in this presentation in detail with the aim to present the objectives and principles. Examples of most recent exercises are included in order to illustrate the importance of such analyses.

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 of Nuclear Power Plants: Design. Specific Safety Requirements (Chinese Ed.)

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Safety of Nuclear Power Plants: Design. Specific Safety Requirements (French Ed.)

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Arabic Ed.)

International Nuclear Information System (INIS)

2012-01-01

On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

International comparison of safety criteria applied to radwaste repositories. Safety aspects of the post-operational phase

International Nuclear Information System (INIS)

Baltes, B.

1994-01-01

There is a generally accepted system of framework safety conditions governing the construction, operation, and post-operational monitoring of radwaste repositories. Although the development of these framework conditions may vary from country to country, the resulting criteria are based on the commonly accepted system of priciples and purposes established for ultimate radioactive waste disposal. The experience accumulated by GRS in the course of the plan approval procedure for the Konrad mine site and the safety-relevant studies performed for the planned Morsleben repository clearly show demand for further development of the safety criteria. In Germany, it is especially the safety criteria and detailed requirements filling the framework safety conditions that need revision and in-depth definition, as well as comparison and harmonisation with internationally applied criteria. These activities will particularly consider the international convention on radioactive waste management currently in preparation under the auspieces of the IAEA. (orig.) [de

Safety features of subcritical fluid fueled systems

International Nuclear Information System (INIS)

Bell, C.R.

1995-01-01

Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible

Safety features of subcritical fluid fueled systems

International Nuclear Information System (INIS)

Bell, C.R.

1994-01-01

Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved in very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible

Safety features of subcritical fluid fueled systems

Energy Technology Data Exchange (ETDEWEB)

Bell, C.R. [Los Alamos National Laboratory, NM (United States)

1995-10-01

Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitative in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible.

77 FR 11120 - Patient Safety Organizations: Voluntary Relinquishment From UAB Health System Patient Safety...

Science.gov (United States)

2012-02-24

... Organizations: Voluntary Relinquishment From UAB Health System Patient Safety Organization AGENCY: Agency for... notification of voluntary relinquishment from the UAB Health System Patient Safety Organization of its status as a Patient Safety Organization (PSO). The Patient Safety and Quality Improvement Act of 2005...

Knowledge-Based Energy Damage Model for Evaluating Industrialised Building Systems (IBS Occupational Health and Safety (OHS Risk

Directory of Open Access Journals (Sweden)

Abas Nor Haslinda

2016-01-01

Full Text Available Malaysia’s construction industry has been long considered hazardous, owing to its poor health and safety record. It is proposed that one of the ways to improve safety and health in the construction industry is through the implementation of ‘off-site’ systems, commonly termed ‘industrialised building systems (IBS’ in Malaysia. This is deemed safer based on the risk concept of reduced exposure, brought about by the reduction in onsite workers; however, no method yet exists for determining the relative safety of various construction methods, including IBS. This study presents a comparative evaluation of the occupational health and safety (OHS risk presented by different construction approaches, namely IBS and traditional methods. The evaluation involved developing a model based on the concept of ‘argumentation theory’, which helps construction designers integrate the management of OHS risk into the design process. In addition, an ‘energy damage model’ was used as an underpinning framework. Development of the model was achieved through three phases, namely Phase I – knowledge acquisitaion, Phase II – argument trees mapping, and Phase III – validation of the model. The research revealed that different approaches/methods of construction projects carried a different level of energy damage, depending on how the activities were carried out. A study of the way in which the risks change from one construction process to another shows that there is a difference in the profile of OHS risk between IBS construction and traditional methods.Therefore, whether the option is an IBS or traditional approach, the fundamental idea of the model is to motivate construction designers or decision-makers to address safety in the design process and encourage them to examine carefully the probable OHS risk variables surrounding an action, thus preventing accidents in construction.

Design aspects of safety critical instrumentation of nuclear installations

Energy Technology Data Exchange (ETDEWEB)

Swaminathan, P. [Electronics Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)]. E-mail: swamy@igcar.ernet.in

2005-07-01

Safety critical instrumentation systems ensure safe shutdown/configuration of the nuclear installation when process status exceeds the safety threshold limits. Design requirements for safety critical instrumentation such as functional and electrical independence, fail-safe design, and architecture to ensure the specified unsafe failure rate and safe failure rate, human machine interface (HMI), etc., are explained with examples. Different fault tolerant architectures like 1/2, 2/2, 2/3 hot stand-by are compared for safety critical instrumentation. For embedded systems, software quality assurance is detailed both during design phase and O and M phase. Different software development models such as waterfall model and spiral model are explained with examples. The error distribution in embedded system is detailed. The usage of formal method is outlined to reduce the specification error. The guidelines for coding of application software are outlined. The interface problems of safety critical instrumentation with sensors, actuators, other computer systems, etc., are detailed with examples. Testability and maintainability shall be taken into account during design phase. Online diagnostics for safety critical instrumentation is detailed with examples. Salient details of design guides from Atomic Energy Regulatory Board, International Atomic Energy Agency and standards from IEEE, BIS are given towards the design of safety critical instrumentation systems. (author)

Design aspects of safety critical instrumentation of nuclear installations

International Nuclear Information System (INIS)

Swaminathan, P.

2005-01-01

Safety critical instrumentation systems ensure safe shutdown/configuration of the nuclear installation when process status exceeds the safety threshold limits. Design requirements for safety critical instrumentation such as functional and electrical independence, fail-safe design, and architecture to ensure the specified unsafe failure rate and safe failure rate, human machine interface (HMI), etc., are explained with examples. Different fault tolerant architectures like 1/2, 2/2, 2/3 hot stand-by are compared for safety critical instrumentation. For embedded systems, software quality assurance is detailed both during design phase and O and M phase. Different software development models such as waterfall model and spiral model are explained with examples. The error distribution in embedded system is detailed. The usage of formal method is outlined to reduce the specification error. The guidelines for coding of application software are outlined. The interface problems of safety critical instrumentation with sensors, actuators, other computer systems, etc., are detailed with examples. Testability and maintainability shall be taken into account during design phase. Online diagnostics for safety critical instrumentation is detailed with examples. Salient details of design guides from Atomic Energy Regulatory Board, International Atomic Energy Agency and standards from IEEE, BIS are given towards the design of safety critical instrumentation systems. (author)

Safety Justification and Safety Case for Safety-critical Software in Digital Reactor Protection System

International Nuclear Information System (INIS)

Kwon, Kee-Choon; Lee, Jang-Soo; Jee, Eunkyoung

2016-01-01

Nuclear safety-critical software is under strict regulatory requirements and these regulatory requirements are essential for ensuring the safety of nuclear power plants. The verification & validation (V and V) and hazard analysis of the safety-critical software are required to follow regulatory requirements through the entire software life cycle. In order to obtain a license from the regulatory body through the development and validation of safety-critical software, it is essential to meet the standards which are required by the regulatory body throughout the software development process. Generally, large amounts of documents, which demonstrate safety justification including standard compliance, V and V, hazard analysis, and vulnerability assessment activities, are submitted to the regulatory body during the licensing process. It is not easy to accurately read and evaluate the whole documentation for the development activities, implementation technology, and validation activities. The safety case methodology has been kwon a promising approach to evaluate the level and depth of the development and validation results. A safety case is a structured argument, supported by a body of evidence that provides a compelling, comprehensible, and valid case that a system is safe for a given application in a given operating environment. It is suggested to evaluate the level and depth of the results of development and validation by applying safety case methodology to achieve software safety demonstration. A lot of documents provided as evidence are connected to claim that corresponds to the topic for safety demonstration. We demonstrated a case study in which more systematic safety demonstration for the target system software is performed via safety case construction than simply listing the documents

Safety Justification and Safety Case for Safety-critical Software in Digital Reactor Protection System

Energy Technology Data Exchange (ETDEWEB)

Kwon, Kee-Choon; Lee, Jang-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jee, Eunkyoung [KAIST, Daejeon (Korea, Republic of)

2016-10-15

Nuclear safety-critical software is under strict regulatory requirements and these regulatory requirements are essential for ensuring the safety of nuclear power plants. The verification & validation (V and V) and hazard analysis of the safety-critical software are required to follow regulatory requirements through the entire software life cycle. In order to obtain a license from the regulatory body through the development and validation of safety-critical software, it is essential to meet the standards which are required by the regulatory body throughout the software development process. Generally, large amounts of documents, which demonstrate safety justification including standard compliance, V and V, hazard analysis, and vulnerability assessment activities, are submitted to the regulatory body during the licensing process. It is not easy to accurately read and evaluate the whole documentation for the development activities, implementation technology, and validation activities. The safety case methodology has been kwon a promising approach to evaluate the level and depth of the development and validation results. A safety case is a structured argument, supported by a body of evidence that provides a compelling, comprehensible, and valid case that a system is safe for a given application in a given operating environment. It is suggested to evaluate the level and depth of the results of development and validation by applying safety case methodology to achieve software safety demonstration. A lot of documents provided as evidence are connected to claim that corresponds to the topic for safety demonstration. We demonstrated a case study in which more systematic safety demonstration for the target system software is performed via safety case construction than simply listing the documents.

Control Method of Single-phase Inverter Based Grounding System in Distribution Networks

DEFF Research Database (Denmark)

Wang, Wen; Yan, L.; Zeng, X.

2016-01-01

of neutral-to-ground voltage is critical for the safety of distribution networks. An active grounding system based on single-phase inverter is proposed to achieve this objective. Relationship between output current of the system and neutral-to-ground voltage is derived to explain the principle of neutral......The asymmetry of the inherent distributed capacitances causes the rise of neutral-to-ground voltage in ungrounded system or high resistance grounded system. Overvoltage may occur in resonant grounded system if Petersen coil is resonant with the distributed capacitances. Thus, the restraint...

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.

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．

Probabilistic Safety Goals. Phase 1 Status and Experiences in Sweden and Finland

International Nuclear Information System (INIS)

Holmberg, Jan-Erik; Knochenhauer, Michael

2007-02-01

The outcome of a probabilistic safety assessment (PSA) for a nuclear power plant is a combination of qualitative and quantitative results. Quantitative results are typically presented as the Core Damage Frequency (CDF) and as the frequency of an unacceptable radioactive release. In order to judge the acceptability of PSA results, criteria for the interpretation of results and the assessment of their acceptability need to be defined. Ultimately, the goals are intended to define an acceptable level of risk from the operation of a nuclear facility. However, safety goals usually have a dual function, i.e., they define an acceptable safety level, but they also have a wider and more general use as decision criteria. The exact levels of the safety goals differ between organisations and between different countries. There are also differences in the definition of the safety goal, and in the formal status of the goals, i.e., whether they are mandatory or not. In this first phase of the project, the aim has been on providing a clear description of the issue of probabilistic safety goals for nuclear power plants, to define and describe important concepts related to the definition and application of safety goals, and to describe experiences in Finland and Sweden. Based on a series of interviews and on literature reviews as well as on a limited international over-view, the project has described the history and current status of safety goals in Sweden and Finland, and elaborated on a number of issues, including the following: The status of the safety goals in view of the fact that they have been exceeded for much of the time they have been in use, as well as the possible implications of these exceedances. Safety goals as informal or mandatory limits. Strategies for handling violations of safety goals, including various graded approaches, such as ALARP (As Low As Reasonably Practicable). Relation between safety goals defined on different levels, e.g., for core damage and for

Probabilistic safety goals. Phase 1 - Status and experiences in Sweden and Finland

International Nuclear Information System (INIS)

Holmberg, J.E.; Knochenhauer, M.

2007-03-01

The outcome of a probabilistic safety assessment (PSA) for a nuclear power plant is a combination of qualitative and quantitative results. Quantitative results are typically presented as the Core Damage Frequency (CDF) and as the frequency of an unacceptable radioactive release. In order to judge the acceptability of PSA results, criteria for the interpretation of results and the assessment of their acceptability need to be defined. Ultimately, the goals are intended to define an acceptable level of risk from the operation of a nuclear facility. However, safety goals usually have a dual function, i.e., they define an acceptable safety level, but they also have a wider and more general use as decision criteria. The exact levels of the safety goals differ between organisations and between different countries. There are also differences in the definition of the safety goal, and in the formal status of the goals, i.e., whether they are mandatory or not. In this first phase of the project, the aim has been on providing a clear description of the issue of probabilistic safety goals for nuclear power plants, to define and describe important concepts related to the definition and application of safety goals, and to describe experiences in Finland and Sweden. Based on a series of interviews and on literature reviews as well as on a limited international over-view, the project has described the history and current status of safety goals in Sweden and Finland, and elaborated on a number of issues, including the following: 1) The status of the safety goals in view of the fact that they have been exceeded for much of the time they have been in use, as well as the possible implications of these exceedances. 2) Safety goals as informal or mandatory limits. 3) Strategies for handling violations of safety goals, including various graded approaches, such as ALARP (As Low As Reasonably Practicable). 4) Relation between safety goals defined on different levels, e.g., for core damage

Probabilistic Safety Goals. Phase 1 Status and Experiences in Sweden and Finland

Energy Technology Data Exchange (ETDEWEB)

Holmberg, Jan-Erik (VTT, FI-02044 VTT (Finland)); Knochenhauer, Michael (Relcon Scandpower AB, SE-172 25 Sundbyberg (Sweden))

2007-02-15

The outcome of a probabilistic safety assessment (PSA) for a nuclear power plant is a combination of qualitative and quantitative results. Quantitative results are typically presented as the Core Damage Frequency (CDF) and as the frequency of an unacceptable radioactive release. In order to judge the acceptability of PSA results, criteria for the interpretation of results and the assessment of their acceptability need to be defined. Ultimately, the goals are intended to define an acceptable level of risk from the operation of a nuclear facility. However, safety goals usually have a dual function, i.e., they define an acceptable safety level, but they also have a wider and more general use as decision criteria. The exact levels of the safety goals differ between organisations and between different countries. There are also differences in the definition of the safety goal, and in the formal status of the goals, i.e., whether they are mandatory or not. In this first phase of the project, the aim has been on providing a clear description of the issue of probabilistic safety goals for nuclear power plants, to define and describe important concepts related to the definition and application of safety goals, and to describe experiences in Finland and Sweden. Based on a series of interviews and on literature reviews as well as on a limited international over-view, the project has described the history and current status of safety goals in Sweden and Finland, and elaborated on a number of issues, including the following: The status of the safety goals in view of the fact that they have been exceeded for much of the time they have been in use, as well as the possible implications of these exceedances. Safety goals as informal or mandatory limits. Strategies for handling violations of safety goals, including various graded approaches, such as ALARP (As Low As Reasonably Practicable). Relation between safety goals defined on different levels, e.g., for core damage and for

Probabilistic safety goals. Phase 1 - Status and experiences in Sweden and Finland

Energy Technology Data Exchange (ETDEWEB)

Holmberg, J.E. [VTT (Finland); Knochenhauer, M. [Relcon Scandpower AB (Sweden)

2007-03-15

The outcome of a probabilistic safety assessment (PSA) for a nuclear power plant is a combination of qualitative and quantitative results. Quantitative results are typically presented as the Core Damage Frequency (CDF) and as the frequency of an unacceptable radioactive release. In order to judge the acceptability of PSA results, criteria for the interpretation of results and the assessment of their acceptability need to be defined. Ultimately, the goals are intended to define an acceptable level of risk from the operation of a nuclear facility. However, safety goals usually have a dual function, i.e., they define an acceptable safety level, but they also have a wider and more general use as decision criteria. The exact levels of the safety goals differ between organisations and between different countries. There are also differences in the definition of the safety goal, and in the formal status of the goals, i.e., whether they are mandatory or not. In this first phase of the project, the aim has been on providing a clear description of the issue of probabilistic safety goals for nuclear power plants, to define and describe important concepts related to the definition and application of safety goals, and to describe experiences in Finland and Sweden. Based on a series of interviews and on literature reviews as well as on a limited international over-view, the project has described the history and current status of safety goals in Sweden and Finland, and elaborated on a number of issues, including the following: 1) The status of the safety goals in view of the fact that they have been exceeded for much of the time they have been in use, as well as the possible implications of these exceedances. 2) Safety goals as informal or mandatory limits. 3) Strategies for handling violations of safety goals, including various graded approaches, such as ALARP (As Low As Reasonably Practicable). 4) Relation between safety goals defined on different levels, e.g., for core damage

Software Reliability Issues Concerning Large and Safety Critical Software Systems

Science.gov (United States)

Kamel, Khaled; Brown, Barbara

1996-01-01

This research was undertaken to provide NASA with a survey of state-of-the-art techniques using in industrial and academia to provide safe, reliable, and maintainable software to drive large systems. Such systems must match the complexity and strict safety requirements of NASA's shuttle system. In particular, the Launch Processing System (LPS) is being considered for replacement. The LPS is responsible for monitoring and commanding the shuttle during test, repair, and launch phases. NASA built this system in the 1970's using mostly hardware techniques to provide for increased reliability, but it did so often using custom-built equipment, which has not been able to keep up with current technologies. This report surveys the major techniques used in industry and academia to ensure reliability in large and critical computer systems.

Deep Borehole Disposal Safety Analysis.

Energy Technology Data Exchange (ETDEWEB)

Freeze, Geoffrey A. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Price, Laura L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); MacKinnon, Robert J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Tillman, Jack Bruce [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

2016-10-01

This report presents a preliminary safety analysis for the deep borehole disposal (DBD) concept, using a safety case framework. A safety case is an integrated collection of qualitative and quantitative arguments, evidence, and analyses that substantiate the safety, and the level of confidence in the safety, of a geologic repository. This safety case framework for DBD follows the outline of the elements of a safety case, and identifies the types of information that will be required to satisfy these elements. At this very preliminary phase of development, the DBD safety case focuses on the generic feasibility of the DBD concept. It is based on potential system designs, waste forms, engineering, and geologic conditions; however, no specific site or regulatory framework exists. It will progress to a site-specific safety case as the DBD concept advances into a site-specific phase, progressing through consent-based site selection and site investigation and characterization.

System theory and safety models in Swedish, UK, Dutch and Australian road safety strategies.

Science.gov (United States)

Hughes, B P; Anund, A; Falkmer, T

2015-01-01

Road safety strategies represent interventions on a complex social technical system level. An understanding of a theoretical basis and description is required for strategies to be structured and developed. Road safety strategies are described as systems, but have not been related to the theory, principles and basis by which systems have been developed and analysed. Recently, road safety strategies, which have been employed for many years in different countries, have moved to a 'vision zero', or 'safe system' style. The aim of this study was to analyse the successful Swedish, United Kingdom and Dutch road safety strategies against the older, and newer, Australian road safety strategies, with respect to their foundations in system theory and safety models. Analysis of the strategies against these foundations could indicate potential improvements. The content of four modern cases of road safety strategy was compared against each other, reviewed against scientific systems theory and reviewed against types of safety model. The strategies contained substantial similarities, but were different in terms of fundamental constructs and principles, with limited theoretical basis. The results indicate that the modern strategies do not include essential aspects of systems theory that describe relationships and interdependencies between key components. The description of these strategies as systems is therefore not well founded and deserves further development. Copyright © 2014 Elsevier Ltd. All rights reserved.

Study on 'Safety qualification of process computers used in safety systems of nuclear power plants'

International Nuclear Information System (INIS)

Bertsche, K.; Hoermann, E.

1991-01-01

The study aims at developing safety standards for hardware and software of computer systems which are increasingly used also for important safety systems in nuclear power plants. The survey of the present state-of-the-art of safety requirements and specifications for safety-relevant systems and, additionally, for process computer systems has been compiled from national and foreign rules. In the Federal Republic of Germany the KTA safety guides and the BMI/BMU safety criteria have to be observed. For the design of future computer-aided systems in nuclear power plants it will be necessary to apply the guidelines in [DIN-880] and [DKE-714] together with [DIN-192]. With the aid of a risk graph the various functions of a system, or of a subsystem, can be evaluated with regard to their significance for safety engineering. (orig./HP) [de

Design an optimum safety policy for personnel safety management - A system dynamic approach

International Nuclear Information System (INIS)

Balaji, P.

2014-01-01

Personnel safety management (PSM) ensures that employee's work conditions are healthy and safe by various proactive and reactive approaches. Nowadays it is a complex phenomenon because of increasing dynamic nature of organisations which results in an increase of accidents. An important part of accident prevention is to understand the existing system properly and make safety strategies for that system. System dynamics modelling appears to be an appropriate methodology to explore and make strategy for PSM. Many system dynamics models of industrial systems have been built entirely for specific host firms. This thesis illustrates an alternative approach. The generic system dynamics model of Personnel safety management was developed and tested in a host firm. The model was undergone various structural, behavioural and policy tests. The utility and effectiveness of model was further explored through modelling a safety scenario. In order to create effective safety policy under resource constraint, DOE (Design of experiment) was used. DOE uses classic designs, namely, fractional factorials and central composite designs. It used to make second order regression equation which serve as an objective function. That function was optimized under budget constraint and optimum value used for safety policy which shown greatest improvement in overall PSM. The outcome of this research indicates that personnel safety management model has the capability for acting as instruction tool to improve understanding of safety management and also as an aid to policy making

Design an optimum safety policy for personnel safety management - A system dynamic approach

Energy Technology Data Exchange (ETDEWEB)

Balaji, P. [The Glocal University, Mirzapur Pole, Delhi- Yamuntori Highway, Saharanpur 2470001 (India)

2014-10-06

Personnel safety management (PSM) ensures that employee's work conditions are healthy and safe by various proactive and reactive approaches. Nowadays it is a complex phenomenon because of increasing dynamic nature of organisations which results in an increase of accidents. An important part of accident prevention is to understand the existing system properly and make safety strategies for that system. System dynamics modelling appears to be an appropriate methodology to explore and make strategy for PSM. Many system dynamics models of industrial systems have been built entirely for specific host firms. This thesis illustrates an alternative approach. The generic system dynamics model of Personnel safety management was developed and tested in a host firm. The model was undergone various structural, behavioural and policy tests. The utility and effectiveness of model was further explored through modelling a safety scenario. In order to create effective safety policy under resource constraint, DOE (Design of experiment) was used. DOE uses classic designs, namely, fractional factorials and central composite designs. It used to make second order regression equation which serve as an objective function. That function was optimized under budget constraint and optimum value used for safety policy which shown greatest improvement in overall PSM. The outcome of this research indicates that personnel safety management model has the capability for acting as instruction tool to improve understanding of safety management and also as an aid to policy making.

Design an optimum safety policy for personnel safety management - A system dynamic approach

Science.gov (United States)

Balaji, P.

2014-10-01

Personnel safety management (PSM) ensures that employee's work conditions are healthy and safe by various proactive and reactive approaches. Nowadays it is a complex phenomenon because of increasing dynamic nature of organisations which results in an increase of accidents. An important part of accident prevention is to understand the existing system properly and make safety strategies for that system. System dynamics modelling appears to be an appropriate methodology to explore and make strategy for PSM. Many system dynamics models of industrial systems have been built entirely for specific host firms. This thesis illustrates an alternative approach. The generic system dynamics model of Personnel safety management was developed and tested in a host firm. The model was undergone various structural, behavioural and policy tests. The utility and effectiveness of model was further explored through modelling a safety scenario. In order to create effective safety policy under resource constraint, DOE (Design of experiment) was used. DOE uses classic designs, namely, fractional factorials and central composite designs. It used to make second order regression equation which serve as an objective function. That function was optimized under budget constraint and optimum value used for safety policy which shown greatest improvement in overall PSM. The outcome of this research indicates that personnel safety management model has the capability for acting as instruction tool to improve understanding of safety management and also as an aid to policy making.

Meeting the maglev system's safety requirements

Energy Technology Data Exchange (ETDEWEB)

Pierick, K

1983-12-01

The author shows how the safety requirements of the maglev track system derive from the general legal conditions for the safety of tracked transport. It is described how their compliance beyond the so-called ''development-accompanying'' and ''acceptance-preparatory'' safety work can be assured for the Transrapid test layout (TVE) now building in Emsland and also for later application as public transport system in Germany within the meaning of the General Railway Act.

River Protection Project Integrated safety management system phase II verification review plan - 7/29/99

International Nuclear Information System (INIS)

SHOOP, D.S.

1999-01-01

The purpose of this review is to verify the implementation status of the Integrated Safety Management System (ISMS) for the River Protection Project (RPP) facilities managed by Fluor Daniel Hanford, Inc. (FDH) and operated by Lockheed Martin Hanford Company (LMHC). This review will also ascertain whether within RPP facilities and operations the work planning and execution processes are in place and functioning to effectively protect the health and safety of the workers, public, environment, and federal property over the RPP life cycle. The RPP ISMS should support the Hanford Strategic Plan (DOERL-96-92) to safely clean up and manage the site's legacy waste and deploy science and technology while incorporating the ISMS central theme to ''Do work safely'' and protect human health and the environment

Laboratory safety and the WHO World Alliance for Patient Safety.

Science.gov (United States)

McCay, Layla; Lemer, Claire; Wu, Albert W

2009-06-01

Laboratory medicine has been a pioneer in the field of patient safety; indeed, the College of American Pathology first called attention to the issue in 1946. Delivering reliable laboratory results has long been considered a priority, as the data produced in laboratory medicine have the potential to critically influence individual patients' diagnosis and management. Until recently, most attention on laboratory safety has focused on the analytic stage of laboratory medicine. Addressing this stage has led to significant and impressive improvements in the areas over which laboratories have direct control. However, recent data demonstrate that pre- and post-analytical phases are at least as vulnerable to errors; to further improve patient safety in laboratory medicine, attention must now be focused on the pre- and post-analytic phases, and the concept of patient safety as a multi-disciplinary, multi-stage and multi-system concept better understood. The World Alliance for Patient Safety (WAPS) supports improvement of patient safety globally and provides a potential framework for considering the total testing process.

MDEP Generic Common Position No DICWG-01. Common position on the treatment of common cause failure caused by software within digital safety systems

International Nuclear Information System (INIS)

2013-01-01

Common cause failures (CCF)2 have been a significant safety concern for nuclear power plant systems. The increasing dependence on software-in safety systems for nuclear power plants has increased the safety significance of CCF caused by software, when software in redundant channels or portions of safety systems has some common dependency. For example, the effect of systematic failures can lead to a loss of safety in many ways: unwanted actuations, a safety function is not provided when needed. Therefore, nuclear power plants should be systematically protected from the effects of common cause failures caused by software in DI and C safety systems. Software for nuclear power plant safety systems should be of the high quality necessary to help assure against the loss of safety (i.e. developed with high-quality engineering practices, commensurate quality assurance applied, with continuous improvement through corrective actions based on lessons learned from operating experience). However, demonstrating adequate software quality only through verification and validation activities and controls on the development process has proved to be problematic. Therefore, this common position provides guidance for the assessment of the potential for CCF for software. It is recognized that programmable logic devices do not execute software in the conventional sense; however, the application development process using these devices have many similarities with software development, and the deficiencies that may be introduced during the application development process may induce errors in the programmable logic devices that can result in common cause failures of these devices of a type similar to software common cause failure. Although deficiencies with the potential to give rise to software common cause failures can be introduced at all phases of the software life cycle, this common position will only consider the potential for software common cause failures within digital safety system

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 in wastewater treatment: the pure oxygen system

International Nuclear Information System (INIS)

Giagnoni, L.

1998-01-01

Though the active sludge process represent, nowadays, the main reference system referring to installations for wastewater treatments, nevertheless systems that exploit the pure oxygen properties constitute an alternative method to the traditional cycle. The following essay is divided into two parts: the first one deals with the fundamental concepts related to the active sludge process and to the alternative system proposed, mentioned before, and includes a short account of the functional characteristics and a brief comparison with traditional methods; the second part represents the head corpus of the work and deals with the problems related to the safety with particular reference to the risk of an explosion meanwhile the process. Moreover, it's drawn attention to the fundamental role of security systems that, nowadays, get frequently used in such kind of installations. On this subject, furthermore, it's pointed out the great importance of the whole preliminary treatments in the planning phase, with particular reference to the processes used for stripping [it

System safety education focused on system management

Science.gov (United States)

Grose, V. L.

1971-01-01

System safety is defined and characteristics of the system are outlined. Some of the principle characteristics include role of humans in hazard analysis, clear language for input and output, system interdependence, self containment, and parallel analysis of elements.

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...

Spent nuclear fuel project cold vacuum drying facility safety equipment list

International Nuclear Information System (INIS)

IRWIN, J.J.

1999-01-01

This document provides the safety equipment list (SEL) for the Cold Vacuum Drying Facility (CVDF). The SEL was prepared in accordance with the procedure for safety structures, systems, and components (SSCs) in HNF-PRO-516, ''Safety Structures, Systems, and Components,'' Revision 0 and HNF-PRO-097, Engineering Design and Evaluation, Revision 0. The SEL was developed in conjunction with HNF-SO-SNF-SAR-O02, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998). The SEL identifies the SSCs and their safety functions, the design basis accidents for which they are required to perform, the design criteria, codes and standards, and quality assurance requirements that are required for establishing the safety design basis of the SSCs. This SEL has been developed for the CVDF Phase 2 Safety Analysis Report (SAR) and shall be updated, expanded, and revised in accordance with future phases of the CVDF SAR until the CVDF final SAR is approved

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.

Usability Methods for Ensuring Health Information Technology Safety: Evidence-Based Approaches. Contribution of the IMIA Working Group Health Informatics for Patient Safety.

Science.gov (United States)

Borycki, E; Kushniruk, A; Nohr, C; Takeda, H; Kuwata, S; Carvalho, C; Bainbridge, M; Kannry, J

2013-01-01

Issues related to lack of system usability and potential safety hazards continue to be reported in the health information technology (HIT) literature. Usability engineering methods are increasingly used to ensure improved system usability and they are also beginning to be applied more widely for ensuring the safety of HIT applications. These methods are being used in the design and implementation of many HIT systems. In this paper we describe evidence-based approaches to applying usability engineering methods. A multi-phased approach to ensuring system usability and safety in healthcare is described. Usability inspection methods are first described including the development of evidence-based safety heuristics for HIT. Laboratory-based usability testing is then conducted under artificial conditions to test if a system has any base level usability problems that need to be corrected. Usability problems that are detected are corrected and then a new phase is initiated where the system is tested under more realistic conditions using clinical simulations. This phase may involve testing the system with simulated patients. Finally, an additional phase may be conducted, involving a naturalistic study of system use under real-world clinical conditions. The methods described have been employed in the analysis of the usability and safety of a wide range of HIT applications, including electronic health record systems, decision support systems and consumer health applications. It has been found that at least usability inspection and usability testing should be applied prior to the widespread release of HIT. However, wherever possible, additional layers of testing involving clinical simulations and a naturalistic evaluation will likely detect usability and safety issues that may not otherwise be detected prior to widespread system release. The framework presented in the paper can be applied in order to develop more usable and safer HIT, based on multiple layers of evidence.

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

Generic test platform for representative tests of safety I/C systems - 15546

International Nuclear Information System (INIS)

Fourestie, B.; Kuck, H.; Richter, J.; Rieche, S.; Waitz, M.

2015-01-01

In compliance with the IEC 61513 safety Instrumentation and Control (I/C) systems must be successfully validated in their final configuration prior to installation on site and commissioning. However the contingent need for modifications during system validation activities or subsequently during the commissioning phase may entail long and costly re-engineering of the I/C systems. With the view to ease these possible modifications, a Generic Test Platform has been developed by AREVA which allows combining a real I/C system subpart with an emulation server. This platform provides a faithful representation of the I/C System allowing crediting the validation test results carried out on this platform. (authors)

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

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)

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 assessment of high consequence robotics system

International Nuclear Information System (INIS)

Robinson, D.G.; Atcitty, C.B.

1996-01-01

This paper outlines the use of a failure modes and effects analysis for the safety assessment of a robotic system being developed at Sandia National Laboratories. The robotic system, the weigh and leak check system, is to replace a manual process for weight and leakage of nuclear materials at the DOE Pantex facility. Failure modes and effects analyses were completed for the robotics process to ensure that safety goals for the systems have been met. Due to the flexible nature of the robot configuration, traditional failure modes and effects analysis (FMEA) were not applicable. In addition, the primary focus of safety assessments of robotics systems has been the protection of personnel in the immediate area. In this application, the safety analysis must account for the sensitivities of the payload as well as traditional issues. A unique variation on the classical FMEA was developed that permits an organized and quite effective tool to be used to assure that safety was adequately considered during the development of the robotic system. The fundamental aspects of the approach are outlined in the paper

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

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

System principles, mathematical models and methods to ensure high reliability of safety systems

Science.gov (United States)

Zaslavskyi, V.

2017-04-01

Modern safety and security systems are composed of a large number of various components designed for detection, localization, tracking, collecting, and processing of information from the systems of monitoring, telemetry, control, etc. They are required to be highly reliable in a view to correctly perform data aggregation, processing and analysis for subsequent decision making support. On design and construction phases of the manufacturing of such systems a various types of components (elements, devices, and subsystems) are considered and used to ensure high reliability of signals detection, noise isolation, and erroneous commands reduction. When generating design solutions for highly reliable systems a number of restrictions and conditions such as types of components and various constrains on resources should be considered. Various types of components perform identical functions; however, they are implemented using diverse principles, approaches and have distinct technical and economic indicators such as cost or power consumption. The systematic use of different component types increases the probability of tasks performing and eliminates the common cause failure. We consider type-variety principle as an engineering principle of system analysis, mathematical models based on this principle, and algorithms for solving optimization problems of highly reliable safety and security systems design. Mathematical models are formalized in a class of two-level discrete optimization problems of large dimension. The proposed approach, mathematical models, algorithms can be used for problem solving of optimal redundancy on the basis of a variety of methods and control devices for fault and defects detection in technical systems, telecommunication networks, and energy systems.

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)

Formal verification and validation of the safety-critical software in a digital reactor protection system

International Nuclear Information System (INIS)

Kwon, K. C.; Park, G. Y.

2006-01-01

This paper describes the Verification and Validation (V and V) activities for the safety-critical software in a Digital Reactor Protection System (DRPS) that is being developed through the Korea nuclear instrumentation and control system project. The main activities of the DRPS V and V process are a preparation of the software planning documentation, a verification of the software according to the software life cycle, a software safety analysis and a software configuration management. The verification works for the Software Requirement Specification (SRS) of the DRPS consist of a technical evaluation, a licensing suitability evaluation, a inspection and traceability analysis, a formal verification, and preparing a test plan and procedure. Especially, the SRS is specified by the formal specification method in the development phase, and the formal SRS is verified by a formal verification method. Through these activities, we believe we can achieve the functionality, performance, reliability, and safety that are the major V and V objectives of the nuclear safety-critical software in a DRPS. (authors)

Safety methodology implementation in the conceptual design phase of a fusion reactor

International Nuclear Information System (INIS)

Rodriguez-Rodrigo, L.; Elbez-Uzan, J.

2007-01-01

The licensing of ITER in France represents the first process for licensing a fusion facility in the framework of an experimental device with a total Tritium inventory of 3 kg. The main ITER parameters are far from those expected in the future demonstration reactors where the fusion power will be at least 5 times higher and the additional heating power could also reach up to 5 times the one foreseen in ITER. Main safety requirements for these reactors are based, among other conditions, on their inherent features as low amount of fuel, very low impurity content of structural materials, minimum waste repository, no active systems for safe shut-down, and no need for evacuation of population after the most severe accident. The design of such reactors is at the stage of conceptual studies and is mainly dealing with plasma performances, tritium breeding, blanket/divertor designs and solution of engineering issues, as well as bounding accidents or classification of waste. The methodological approach for integrating safety analysis as a tool for optimizing the design of the overall fusion installation for future reactors in the conceptual design phase is sketched, including the machine itself and the different auxiliary nuclear buildings. (author)

Safety status system for operating room devices.

Science.gov (United States)

Guédon, Annetje C P; Wauben, Linda S G L; Overvelde, Marlies; Blok, Joleen H; van der Elst, Maarten; Dankelman, Jenny; van den Dobbelsteen, John J

2014-01-01

Since the increase of the number of technological aids in the operating room (OR), equipment-related incidents have come to be a common kind of adverse events. This underlines the importance of adequate equipment management to improve the safety in the OR. A system was developed to monitor the safety status (periodic maintenance and registered malfunctions) of OR devices and to facilitate the notification of malfunctions. The objective was to assess whether the system is suitable for use in an busy OR setting and to analyse its effect on the notification of malfunctions. The system checks automatically the safety status of OR devices through constant communication with the technical facility management system, informs the OR staff real-time and facilitates notification of malfunctions. The system was tested for a pilot period of six months in four ORs of a Dutch teaching hospital and 17 users were interviewed on the usability of the system. The users provided positive feedback on the usability. For 86.6% of total time, the localisation of OR devices was accurate. 62 malfunctions of OR devices were reported, an increase of 12 notifications compared to the previous year. The safety status system was suitable for an OR complex, both from a usability and technical point of view, and an increase of reported malfunctions was observed. The system eases monitoring the safety status of equipment and is a promising tool to improve the safety related to OR devices.

Plant air systems safety study: Portsmouth Gaseous Diffusion Plant

International Nuclear Information System (INIS)

1982-05-01

The Portsmouth Gaseous Diffusion Plant Air System facilities and operations are reviewed for potential safety problems not covered by standard industrial safety procedures. Information is presented under the following section headings: facility and process description (general); air plant equipment; air distribution system; safety systems; accident analysis; plant air system safety overview; and conclusion

Phase Control in Nonlinear Systems

Science.gov (United States)

Zambrano, Samuel; Seoane, Jesús M.; Mariño, Inés P.; Sanjuán, Miguel A. F.; Meucci, Riccardo

The following sections are included: * Introduction * Phase Control of Chaos * Description of the model * Numerical exploration of phase control of chaos * Experimental evidence of phase control of chaos * Phase Control of Intermittency in Dynamical Systems * Crisis-induced intermittency and its control * Experimental setup and implementation of the phase control scheme * Phase control of the laser in the pre-crisis regime * Phase control of the intermittency after the crisis * Phase control of the intermittency in the quadratic map * Phase Control of Escapes in Open Dynamical Systems * Control of open dynamical systems * Model description * Numerical simulations and heuristic arguments * Experimental implementation in an electronic circuit * Conclusions and Discussions * Acknowledgments * References

A philosophy for space nuclear systems safety

International Nuclear Information System (INIS)

Marshall, A.C.

1992-01-01

The unique requirements and contraints of space nuclear systems require careful consideration in the development of a safety policy. The Nuclear Safety Policy Working Group (NSPWG) for the Space Exploration Initiative has proposed a hierarchical approach with safety policy at the top of the hierarchy. This policy allows safety requirements to be tailored to specific applications while still providing reassurance to regulators and the general public that the necessary measures have been taken to assure safe application of space nuclear systems. The safety policy used by the NSPWG is recommended for all space nuclear programs and missions

Verification and validation process for the safety software in KNICS

International Nuclear Information System (INIS)

Kwon, Kee-Choon; Lee, Jang-Soo; Kim, Jang-Yeol

2004-01-01

This paper describes the Verification and Validation (V and V ) process for safety software of Programmable Logic Controller (PLC), Digital Reactor Protection System (DRPS), and Engineered Safety Feature-Component Control System (ESF-CCS) that are being developed in Korea Nuclear Instrumentation and Control System (KNICS) projects. Specifically, it presents DRPS V and V experience according to the software development life cycle. The main activities of DRPS V and V process are preparation of software planning documentation, verification of Software Requirement Specification (SRS), Software Design Specification (SDS) and codes, and testing of the integrated software and the integrated system. In addition, they include software safety analysis and software configuration management. SRS V and V of DRPS are technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, preparing integrated system test plan, software safety analysis, and software configuration management. Also, SDS V and V of RPS are technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, preparing integrated software test plan, software safety analysis, and software configuration management. The code V and V of DRPS are traceability analysis, source code inspection, test case and test procedure generation, software safety analysis, and software configuration management. Testing is the major V and V activity of software integration and system integration phase. Software safety analysis at SRS phase uses Hazard Operability (HAZOP) method, at SDS phase it uses HAZOP and Fault Tree Analysis (FTA), and at implementation phase it uses FTA. Finally, software configuration management is performed using Nu-SCM (Nuclear Software Configuration Management) tool developed by KNICS project. Through these activities, we believe we can achieve the functionality, performance, reliability and safety that are V

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

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.

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....

A management system integrating radiation protection and safety supporting safety culture in the hospital

International Nuclear Information System (INIS)

Almen, A.; Lundh, C.

2015-01-01

Quality assurance has been identified as an important part of radiation protection and safety for a considerable time period. A rational expansion and improvement of quality assurance is to integrate radiation protection and safety in a management system. The aim of this study was to explore factors influencing the implementing strategy when introducing a management system including radiation protection and safety in hospitals and to outline benefits of such a system. The main experience from developing a management system is that it is possible to create a vast number of common policies and routines for the whole hospital, resulting in a cost-efficient system. One of the key benefits is the involvement of management at all levels, including the hospital director. Furthermore, a transparent system will involve staff throughout the organisation as well. A management system supports a common view on what should be done, who should do it and how the activities are reviewed. An integrated management system for radiation protection and safety includes key elements supporting a safety culture. (authors)

Regulatory Oversight of Safety Culture in Finland: A Systemic Approach to Safety

International Nuclear Information System (INIS)

Oedewald, P.; Väisäsvaara, J.

2016-01-01

In Finland the Radiation and Nuclear Safety Authority STUK specifies detailed regulatory requirements for good safety culture. Both the requirements and the practical safety culture oversight activities reflect a systemic approach to safety: the interconnections between the technical, human and organizational factors receive special attention. The conference paper aims to show how the oversight of safety culture can be integrated into everyday oversight activities. The paper also emphasises that the scope of the safety culture oversight is not specific safety culture activities of the licencees, but rather the overall functioning of the licence holder or the new build project organization from safety point of view. The regulatory approach towards human and organizational factors and safety culture has evolved throughout the years of nuclear energy production in Finland. Especially the recent new build projects have highlighted the need to systematically pay attention to the non-technical aspects of safety as it has become obvious how the HOF issues can affect the design processes and quality of construction work. Current regulatory guides include a set of safety culture related requirements. The requirements are binding to the licence holders and they set both generic and specific demands on the licencee to understand, monitor and to develop safety culture of their own organization but also that of their supplier network. The requirements set for the licence holders has facilitated the need to develop the regulator’s safety culture oversight practices towards a proactive and systemic approach.

Safety Analysis For Evaluating (SAFE) sUAS, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — The primary goal of Air Traffic Management systems is to ensure safety of operations, in the air and on the ground. While they system have served the National...

New developments in two-phase flow heat transfer with emphasis on nuclear safety research

International Nuclear Information System (INIS)

Mayinger, F.

1987-01-01

The literature on two-phase flow - with and without heat transfer - shows an explosive-like growth of published papers within the last ten years. Many of these papers were published as a result of nuclear safety research. It is impossible to deal with all new developments reported in this extensive literature. So one has to ask: Are there trends of special interest, where this report could be concentrated on? Looking over the situation, there seem to be three very promising fields of research having high actuality, especially for nuclear safety, namely: fluiddynamic and thermodynamic nonequilibrium in steady state, transient conditions, and scaling. The discussion on new developments in two-phase flow heat transfer, therefore, is limited on these subjects

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)

KAERI software verification and validation guideline for developing safety-critical software in digital I and C system of NPP

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-01

This technical report is to present V and V guideline development methodology for safety-critical software in NPP safety system. Therefore it is to present V and V guideline of planning phase for the NPP safety system in addition to critical safety items, for example, independence philosophy, software safety analysis concept, commercial off the shelf (COTS) software evaluation criteria, inter-relationships between other safety assurance organizations, including the concepts of existing industrial standard, IEEE Std-1012, IEEE Std-1059. This technical report includes scope of V and V guideline, guideline framework as part of acceptance criteria, V and V activities and task entrance as part of V and V activity and exit criteria, review and audit, testing and QA records of V and V material and configuration management, software verification and validation plan production etc., and safety-critical software V and V methodology. (author). 11 refs.

KAERI software verification and validation guideline for developing safety-critical software in digital I and C system of NPP

International Nuclear Information System (INIS)

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

1997-07-01

This technical report is to present V and V guideline development methodology for safety-critical software in NPP safety system. Therefore it is to present V and V guideline of planning phase for the NPP safety system in addition to critical safety items, for example, independence philosophy, software safety analysis concept, commercial off the shelf (COTS) software evaluation criteria, inter-relationships between other safety assurance organizations, including the concepts of existing industrial standard, IEEE Std-1012, IEEE Std-1059. This technical report includes scope of V and V guideline, guideline framework as part of acceptance criteria, V and V activities and task entrance as part of V and V activity and exit criteria, review and audit, testing and QA records of V and V material and configuration management, software verification and validation plan production etc., and safety-critical software V and V methodology. (author). 11 refs

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...

Aging and malfunction of valves in CANDU special safety systems. Phase 1

International Nuclear Information System (INIS)

1989-03-01

Aging and wear related valve malfunctions have been reported in American nuclear generating systems. This report documents the first attempt to study these phenomena on a global basis in Canadian nuclear power plants. A general methodology outlines an approach to this type of study which is amenable to use within existing information structures. Nuclear regulatory requirements which influence the testing of valves in Canadian nuclear power plants are reviewed. The reporting systems which emanate from these requirements are discussed and sources of valve failure data are reviewed. It is determined that modifications to existing failure reporting systems are required before practical means of collecting data necessary for the analysis of age related valve malfunctions can be developed. In spite of limitations in reported failure data, a partial data base is compiled for valve failures in Special Safety Systems of domestic nuclear plants. Data are reported for the period 1982 to 1986. The valve population and basic parameters of each valve such as type, operator, function, etc., and the reported failures against this population are compiled and reviewed for evidence of time dependent versus random failure trends. Results suggest that there is no clear age related failure trend. In fact, some systems and stations, experienced a reduction in failure rates with years of servicing, suggesting that some earlier generic valve problems may have been solved. Present inspection, test, and maintenance practices are reviewed and their effectiveness for purposes of predicting or preventing incipient failures is assessed to be of moderate value. Modern failure prevention methods are highlighted and their applicability discussed

Reliability and safety program plan outline for the operational phase of a waste isolation facility

International Nuclear Information System (INIS)

Ammer, H.G.; Wood, D.E.

1977-01-01

A Reliability and Safety Program plan outline has been prepared for the operational phase of a Waste Isolation Facility. The program includes major functions of risk assessment, technical support activities, quality assurance, operational safety, configuration monitoring, reliability analysis and support and coordination meetings. Detailed activity or task descriptions are included for each function. Activities are time-phased and presented in the PERT format for scheduling and interactions. Task descriptions include manloading, travel, and computer time estimates to provide data for future costing. The program outlined here will be used to provide guidance from a reliability and safety standpoint to design, procurement, construction, and operation of repositories for nuclear waste. These repositories are to be constructed under the National Waste Terminal Storage program under the direction of the Office of Waste Isolation, Union Carbide Corp. Nuclear Division

Systems engineered health and safety criteria for safety analysis reports

International Nuclear Information System (INIS)

Beitel, G.A.; Morcos, N.

1993-01-01

The world of safety analysis is filled with ambiguous words: codes and standards, consequences and risks, hazard and accident, and health and safety. These words have been subject to disparate interpretations by safety analysis report (SAR) writers, readers, and users. open-quotes Principal health and safety criteriaclose quotes has been one of the most frequently misused phrases; rarely is it used consistently or effectively. This paper offers an easily understood definition for open-quotes principal health and safety criteriaclose quotes and uses systems engineering to convert an otherwise mysterious topic into the primary means of producing an integrated SAR. This paper is based on SARs being written for environmental restoration and waste management activities for the U.S. Department of Energy (DOE). Requirements for these SARs are prescribed in DOE Order 5480-23, open-quotes Nuclear Safety Analysis Reports.close quotes

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)

Seismic safety margins research program. Phase I final report - Overview

International Nuclear Information System (INIS)

Smith, P.D.; Dong, R.G.; Bernreuter, D.L.; Bohn, M.P.; Chuang, T.Y.; Cummings, G.E.; Johnson, J.J.; Mensing, R.W.; Wells, J.E.

1981-04-01

The Seismic Safety Margins Research Program (SSMRP) is a multiyear, multiphase program whose overall objective is to develop improved methods for seismic safety assessments of nuclear power plants, using a probabilistic computational procedure. The program is being carried out at the Lawrence Livermore National Laboratory and is sponsored by the U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. Phase I of the SSMRP was successfully completed in January 1981: A probabilistic computational procedure for the seismic risk assessment of nuclear power plants has been developed and demonstrated. The methodology is implemented by three computer programs: HAZARD, which assesses the seismic hazard at a given site, SMACS, which computes in-structure and subsystem seismic responses, and SEISIM, which calculates system failure probabilities and radioactive release probabilities, given (1) the response results of SMACS, (2) a set of event trees, (3) a family of fault trees, (4) a set of structural and component fragility descriptions, and (5) a curve describing the local seismic hazard. The practicality of this methodology was demonstrated by computing preliminary release probabilities for Unit 1 of the Zion Nuclear Power Plant north of Chicago, Illinois. Studies have begun aimed at quantifying the sources of uncertainty in these computations. Numerous side studies were undertaken to examine modeling alternatives, sources of error, and available analysis techniques. Extensive sets of data were amassed and evaluated as part of projects to establish seismic input parameters and to produce the fragility curves. (author)

Analyzing Software Requirements Errors in Safety-Critical, Embedded Systems

Science.gov (United States)

Lutz, Robyn R.

1993-01-01

This paper analyzes the root causes of safety-related software errors in safety-critical, embedded systems. The results show that software errors identified as potentially hazardous to the system tend to be produced by different error mechanisms than non- safety-related software errors. Safety-related software errors are shown to arise most commonly from (1) discrepancies between the documented requirements specifications and the requirements needed for correct functioning of the system and (2) misunderstandings of the software's interface with the rest of the system. The paper uses these results to identify methods by which requirements errors can be prevented. The goal is to reduce safety-related software errors and to enhance the safety of complex, embedded systems.

Study of reactor parameters on the critical systems. Phase I; Ispitivanje reaktorskih parametara na kriticnim sistemima, I faza

Energy Technology Data Exchange (ETDEWEB)

Raisic, N et al [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

1962-08-15

Phase 1 of the report on reactor parameters study describes the preparation of the RB reactor for operation including the following tasks: Completing and verification of reactor safety system; arranging dosimetry instruments; formation of fuel elements with 2% enriched fuel and aluminium holders; improvement of the heavy water level-meter; mounting the horizontal experimental channel; formation of reactor lattice with 16 cm pitch; testing the reactor system; filling the tank with heavy water and preparing the safety report.

Evaluation of Model Driven Development of Safety Critical Software in the Nuclear Power Plant I and C system

International Nuclear Information System (INIS)

Jung, Jae Cheon; Chang, Hoon Seon; Chang, Young Woo; Kim, Jae Hack; Sohn, Se Do

2005-01-01

The major issues of the safety critical software are formalism and V and V. Implementing these two characteristics in the safety critical software will greatly enhance the quality of software product. The structure based development requires lots of output documents from the requirements phase to the testing phase. The requirements analysis phase is open omitted. According to the Standish group report in 2001, 49% of software project is cancelled before completion or never implemented. In addition, 23% is completed and become operational, but over-budget, over the time estimation, and with fewer features and functions than initially specified. They identified ten success factors. Among them, firm basic requirements and formal methods are technically achievable factors while the remaining eight are management related. Misunderstanding of requirements due to lack of communication between the design engineer and verification engineer causes unexpected result such as functionality error of system. Safety critical software shall comply with such characteristics as; modularity, simplicity, minimizing the sub-routine, and excluding the interrupt routine. In addition, the crosslink fault and erroneous function shall be eliminated. The easiness of repairing work after the installation shall be achieved as well. In consideration of the above issues, we evaluate the model driven development (MDD) methods for nuclear I and C systems software. For qualitative analysis, the unified modeling language (UML), functional block language (FBL) and the safety critical application environment (SCADE) are tested for the above characteristics

Using system dynamics simulation for assessment of hydropower system safety

Science.gov (United States)

King, L. M.; Simonovic, S. P.; Hartford, D. N. D.

2017-08-01

Hydropower infrastructure systems are complex, high consequence structures which must be operated safely to avoid catastrophic impacts to human life, the environment, and the economy. Dam safety practitioners must have an in-depth understanding of how these systems function under various operating conditions in order to ensure the appropriate measures are taken to reduce system vulnerability. Simulation of system operating conditions allows modelers to investigate system performance from the beginning of an undesirable event to full system recovery. System dynamics simulation facilitates the modeling of dynamic interactions among complex arrangements of system components, providing outputs of system performance that can be used to quantify safety. This paper presents the framework for a modeling approach that can be used to simulate a range of potential operating conditions for a hydropower infrastructure system. Details of the generic hydropower infrastructure system simulation model are provided. A case study is used to evaluate system outcomes in response to a particular earthquake scenario, with two system safety performance measures shown. Results indicate that the simulation model is able to estimate potential measures of system safety which relate to flow conveyance and flow retention. A comparison of operational and upgrade strategies is shown to demonstrate the utility of the model for comparing various operational response strategies, capital upgrade alternatives, and maintenance regimes. Results show that seismic upgrades to the spillway gates provide the largest improvement in system performance for the system and scenario of interest.

Soft systems methodology as a systemic approach to nuclear safety management

International Nuclear Information System (INIS)

Vieira Neto, Antonio S.; Guilhen, Sabine N.; Rubin, Gerson A.; Caldeira Filho, Jose S.; Camargo, Iara M.C.

2017-01-01

Safety approach currently adopted by nuclear installations is built almost exclusively upon analytical methodologies based, mainly, on the belief that the properties of a system, such as its safety, are given by its constituent parts. This approach, however, does not properly address the complex dynamic interactions between technical, human and organizational factors occurring within and outside the organization. After the accident at Fukushima Daiichi nuclear power plant in March 2011, experts of the International Atomic Energy Agency (IAEA) recommended a systemic approach as a complementary perspective to nuclear safety. The aim of this paper is to present an overview of the systems thinking approach and its potential use for structuring socio technical problems involved in the safety of nuclear installations, highlighting the methodologies related to the soft systems thinking, in particular the Soft Systems Methodology (SSM). The implementation of a systemic approach may thus result in a more holistic picture of the system by the complex dynamic interactions between technical, human and organizational factors. (author)

Soft systems methodology as a systemic approach to nuclear safety management

Energy Technology Data Exchange (ETDEWEB)

Vieira Neto, Antonio S.; Guilhen, Sabine N.; Rubin, Gerson A.; Caldeira Filho, Jose S.; Camargo, Iara M.C., E-mail: asvneto@ipen.br, E-mail: snguilhen@ipen.br, E-mail: garubin@ipen.br, E-mail: jscaldeira@ipen.br, E-mail: icamargo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

2017-07-01

Safety approach currently adopted by nuclear installations is built almost exclusively upon analytical methodologies based, mainly, on the belief that the properties of a system, such as its safety, are given by its constituent parts. This approach, however, does not properly address the complex dynamic interactions between technical, human and organizational factors occurring within and outside the organization. After the accident at Fukushima Daiichi nuclear power plant in March 2011, experts of the International Atomic Energy Agency (IAEA) recommended a systemic approach as a complementary perspective to nuclear safety. The aim of this paper is to present an overview of the systems thinking approach and its potential use for structuring socio technical problems involved in the safety of nuclear installations, highlighting the methodologies related to the soft systems thinking, in particular the Soft Systems Methodology (SSM). The implementation of a systemic approach may thus result in a more holistic picture of the system by the complex dynamic interactions between technical, human and organizational factors. (author)

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 analysis and evaluation methodology for fusion systems

International Nuclear Information System (INIS)

Fujii-e, Y.; Kozawa, Y.; Namba, C.

1987-03-01

Fusion systems which are under development as future energy systems have reached a stage that the break even is expected to be realized in the near future. It is desirable to demonstrate that fusion systems are well acceptable to the societal environment. There are three crucial viewpoints to measure the acceptability, that is, technological feasibility, economy and safety. These three points have close interrelation. The safety problem is more important since three large scale tokamaks, JET, TFTR and JT-60, start experiment, and tritium will be introduced into some of them as the fusion fuel. It is desirable to establish a methodology to resolve the safety-related issues in harmony with the technological evolution. The promising fusion system toward reactors is not yet settled. This study has the objective to develop and adequate methodology which promotes the safety design of general fusion systems and to present a basis for proposing the R and D themes and establishing the data base. A framework of the methodology, the understanding and modeling of fusion systems, the principle of ensuring safety, the safety analysis based on the function and the application of the methodology are discussed. As the result of this study, the methodology for the safety analysis and evaluation of fusion systems was developed. New idea and approach were presented in the course of the methodology development. (Kako, I.)

Understanding Nuclear Safety Culture: A Systemic Approach

International Nuclear Information System (INIS)

Afghan, A.N.

2016-01-01

The Fukushima accident was a systemic failure (Report by Director General IAEA on the Fukushima Daiichi Accident). Systemic failure is a failure at system level unlike the currently understood notion which regards it as the failure of component and equipment. Systemic failures are due to the interdependence, complexity and unpredictability within systems and that is why these systems are called complex adaptive systems (CAS), in which “attractors” play an important role. If we want to understand the systemic failures we need to understand CAS and the role of these attractors. The intent of this paper is to identify some typical attractors (including stakeholders) and their role within complex adaptive system. Attractors can be stakeholders, individuals, processes, rules and regulations, SOPs etc., towards which other agents and individuals are attracted. This paper will try to identify attractors in nuclear safety culture and influence of their assumptions on safety culture behavior by taking examples from nuclear industry in Pakistan. For example, if the nuclear regulator is an attractor within nuclear safety culture CAS then how basic assumptions of nuclear plant operators and shift in-charges about “regulator” affect their own safety behavior?

Fluor Hanford Integrated Safety Management System Phase 1 Verification 04/12/2000 Thru 04/28/2000 Volume 1 and 2

International Nuclear Information System (INIS)

PARSONS, J.E.

2000-01-01

The U.S. Department of Energy (DOE) commits to accomplishing its mission safely. To ensure this objective is met, DOE issued DOE P 450.4, Safety Management System Policy, and incorporated safety management into the DOE Acquisition Regulations ([DEAR] 48 CFR 970.5204-2 and 90.5204-78)

Fluor Hanford Integrated Safety Management System Phase 1 Verification 04/12/2000 Thru 04/28/2000 Volume 1 and 2

Energy Technology Data Exchange (ETDEWEB)

PARSONS, J.E.

2000-03-01

The U.S. Department of Energy (DOE) commits to accomplishing its mission safely. To ensure this objective is met, DOE issued DOE P 450.4, Safety Management System Policy, and incorporated safety management into the DOE Acquisition Regulations ([DEAR] 48 CFR 970.5204-2 and 90.5204-78).

Fluor Hanford Integrated Safety Management System Phase 1 Verification 04/12/2000 Thru 04/28/2000 Volume 1 and 2

CERN Document Server

Parsons, J E

2000-01-01

The U.S. Department of Energy (DOE) commits to accomplishing its mission safely. To ensure this objective is met, DOE issued DOE P 450.4, Safety Management System Policy, and incorporated safety management into the DOE Acquisition Regulations ([DEAR] 48 CFR 970.5204-2 and 90.5204-78).

Krsko periodic safety review project prioritization process

International Nuclear Information System (INIS)

Basic, I.; Vrbanic, I.; Spiler, J.; Lambright, J.

2004-01-01

Definition of a Krsko Periodic Safety Review (PSR) project is a comprehensive safety review of a plant after last ten years of operation. The objective is a verification by means of a comprehensive review using current methods that Krsko NPP remains safety when judged against current safety objectives and practices and that adequate arrangements are in place to maintain plant safety. This objective encompasses the three main criteria or goals: confirmation that the plant is as safe as originally intended, determination if there are any structures, systems or components that could limit the life of the plant in the foreseeable future, and comparison the plant against modern safety standards and to identify where improvements would be beneficial at justifiable cost. Krsko PSR project is structured in the three phases: Phase 1: Preparation of Detailed 10-years PSR Program, Phase 2: Performing of 10-years PSR Program and preparing of associated documents (2001-2003), and Phase 3: Implementation of the prioritized compensatory measures and modifications (development of associated EEAR, DMP, etc.) after agreement with the SNSA on the design, procedures and time-scales (2004-2008). This paper presents the NEK PSR results of work performed under Phase 2 focused on the ranking of safety issues and prioritization of corrective measures needed for establishing an efficient action plan. Safety issues were identified in Phase 2 during the following review processes: Periodic Safety Review (PSR) task; Krsko NPP Regulatory Compliance Program (RCP) review; Westinghouse Owner Group (WOG) catalog items screening/review; SNSA recommendations (including IAEA RAMP mission suggestions/recommendations).(author)

Development of ABWR-2 and its safety design

International Nuclear Information System (INIS)

Takafumi, Anegawa; Kenji, Tateiwa

2002-01-01

This paper reports the current status of development project on ABWR-II, a next generation reactor design based on ABWR, and its safety design. This project was initiated over a decade ago and has completed three phases to date. In Phase I (1991-92), basic design requirements were discussed and several plant concepts were studied. In Phase II (1993-95), key design features were selected in order to establish a reference reactor concept. In Phase III (1996-2000), based on the reference reactor concept, modifications and improvements were made to fulfill the design requirements. By adopting large electric output (1 700 MW), large fuel bundle, modified ECCS, and passive heat removal systems, among other design features, we achieved a design concept capable of increasing both economic competitiveness and safety performance. Main focus of this paper will be on the safety design, safety performance, and further research needs related to safety. (authors)

Towards the Development of a Methodology for the Cyber Security Analysis of Safety Related Nuclear Digital I and C Systems

International Nuclear Information System (INIS)

Khand, Parvaiz Ahmed; Seong, Poong Hyun

2007-01-01

In nuclear power plants the redundant safety related systems are designed to take automatic action to prevent and mitigate accident conditions if the operators and the non-safety systems fail to maintain the plant within normal operating conditions. In case of an event, the failure of these systems has catastrophic consequences. The tendency in the industry over the past 10 years has been to use of commercial of the shelf (COTS) technologies in these systems. COTS software was written with attention to function and performance rather than security. COTS hardware usually designed to fail safe, but security vulnerabilities could be exploited by an attacker to disable the fail safe mechanisms. Moreover, the use of open protocols and operating systems in these technologies make the plants to become vulnerable to a host of cyber attacks. An effective security analysis process is required during all life cycle phases of these systems in order to ensure the security from cyber attacks. We are developing a methodology for the cyber security analysis of safety related nuclear digital I and C Systems. This methodology will cover all phases of development, operation and maintenance processes of software life cycle. In this paper, we will present a security analysis process for the concept stage of software development life cycle

Safety standards of IAEA for management systems

International Nuclear Information System (INIS)

Vincze, P.

2005-01-01

IAEA has developed a new series of safety standards which are assigned for constitution of the conditions and which give the instruction for setting up the management systems that integrate the aims of safety, health, life environment and quality. The new standard shall replace IAEA 50-C-Q - Requirements for security of the quality for safety in nuclear power plants and other nuclear facilities as well as 14 related safety instructions mentioned in the Safety series No. 50-C/SG-Q (1996). When developing of this complex, integrated set of requirements for management systems, the IAEA requirements 50-C-Q (1996) were taken into consideration as well as the publications developed within the International organisation for standardization (ISO) ISO 9001:2000 and ISO14001: 1996. The experience of European Union member states during the development, implementation and improvement of the management systems were also taken into consideration

Model-based safety architecture framework for complex systems

NARCIS (Netherlands)

Schuitemaker, Katja; Rajabali Nejad, Mohammadreza; Braakhuis, J.G.; Podofillini, Luca; Sudret, Bruno; Stojadinovic, Bozidar; Zio, Enrico; Kröger, Wolfgang

2015-01-01

The shift to transparency and rising need of the general public for safety, together with the increasing complexity and interdisciplinarity of modern safety-critical Systems of Systems (SoS) have resulted in a Model-Based Safety Architecture Framework (MBSAF) for capturing and sharing architectural

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

Environmental, Health and Safety Assessment: ATS 7H Program (Phase 3R) Test Activities at the GE Power Systems Gas Turbine Manufacturing Facility, Greenville, SC

Energy Technology Data Exchange (ETDEWEB)

None

1998-11-17

International Technology Corporation (IT) was contracted by General Electric Company (GE) to assist in the preparation of an Environmental, Health and Safety (HI&3) assessment of the implementation of Phase 3R of the Advanced Turbine System (ATS) 7H program at the GE Gas Turbines facility located in Greenville, South Carolina. The assessment was prepared in accordance with GE's contractual agreement with the U.S. Department of Energy (GE/DOE Cooperative Agreement DE-FC21-95MC3 1176) and supports compliance with the requirements of the National Environmental Policy Act of 1970. This report provides a summary of the EH&S review and includes the following: General description of current site operations and EH&S status, Description of proposed ATS 7H-related activities and discussion of the resulting environmental, health, safety and other impacts to the site and surrounding area. Listing of permits and/or licenses required to comply with federal, state and local regulations for proposed 7H-related activities. Assessment of adequacy of current and required permits, licenses, programs and/or plans.

Operation safety of complex industrial systems

International Nuclear Information System (INIS)

Zwingelstein, G.

1999-01-01

Zero fault or zero risk is an unreachable goal in industrial activities like nuclear activities. However, methods and techniques exist to reduce the risks to the lowest possible and acceptable level. The operation safety consists in the recognition, evaluation, prediction, measurement and mastery of technological and human faults. This paper analyses each of these points successively: 1 - evolution of operation safety; 2 - definitions and basic concepts: failure, missions and functions of a system and of its components, basic concepts and operation safety; 3 - forecasting analysis of operation safety: reliability data, data-banks, precautions for the use of experience feedback data; realization of an operation safety study: management of operation safety, quality assurance, critical review and audit of operation safety studies; 6 - conclusions. (J.S.)

The reliability of nuclear power plant safety systems

International Nuclear Information System (INIS)

Susnik, J.

1978-01-01

A criterion was established concerning the protection that nuclear power plant (NPP) safety systems should afford. An estimate of the necessary or adequate reliability of the total complex of safety systems was derived. The acceptable unreliability of auxiliary safety systems is given, provided the reliability built into the specific NPP safety systems (ECCS, Containment) is to be fully utilized. A criterion for the acceptable unreliability of safety (sub)systems which occur in minimum cut sets having three or more components of the analysed fault tree was proposed. A set of input MTBF or MTTF values which fulfil all the set criteria and attain the appropriate overall reliability was derived. The sensitivity of results to input reliability data values was estimated. Numerical reliability evaluations were evaluated by the programs POTI, KOMBI and particularly URSULA, the last being based on Vesely's kinetic fault tree theory. (author)

Antennas for Frequency Reconfigurable Phased Arrays

NARCIS (Netherlands)

Haider, S.N.

2015-01-01

Sensors such as phased array radars play a crucial role in public safety. They are unavoidable for surveillance, threat identification and post-disaster management. However, different scenarios impose immensely diverse requirements for these systems. Phased array systems occupy a large space. In

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)

The software safety analysis based on SFTA for reactor power regulating system in nuclear power plant

International Nuclear Information System (INIS)

Liu Zhaohui; Yang Xiaohua; Liao Longtao; Wu Zhiqiang

2015-01-01

The digitalized Instrumentation and Control (I and C) system of Nuclear power plants can provide many advantages. However, digital control systems induce new failure modes that differ from those of analog control systems. While the cost effectiveness and flexibility of software is widely recognized, it is very difficult to achieve and prove high levels of dependability and safety assurance for the functions performed by process control software, due to the very flexibility and potential complexity of the software itself. Software safety analysis (SSA) was one way to improve the software safety by identify the system hazards caused by software failure. This paper describes the application of a software fault tree analysis (SFTA) at the software design phase. At first, we evaluate all the software modules of the reactor power regulating system in nuclear power plant and identify various hazards. The SFTA was applied to some critical modules selected from the previous step. At last, we get some new hazards that had not been identified in the prior processes of the document evaluation which were helpful for our design. (author)

Safety of huge systems

International Nuclear Information System (INIS)

Kondo, Jiro.

1995-01-01

Recently accompanying the development of engineering technology, huge systems tend to be constructed. The disaster countermeasures of huge cities become large problems as the concentration of population into cities is conspicuous. To make the expected value of loss small, the knowledge of reliability engineering is applied. In reliability engineering, even if a part of structures fails, the safety as a whole system must be ensured, therefore, the design having margin is carried out. The degree of margin is called redundancy. However, such design concept makes the structure of a system complex, and as the structure is complex, the possibility of causing human errors becomes high. At the time of huge system design, the concept of fail-safe is effective, but simple design must be kept in mind. The accident in Mihama No. 2 plant of Kansai Electric Power Co. and the accident in Chernobyl nuclear power station, and the accident of Boeing B737 airliner and the fatigue breakdown are described. The importance of safety culture was emphasized as the method of preventing human errors. Man-system interface and management system are discussed. (K.I.)

System Safety in an IT Service Organization

Science.gov (United States)

Parsons, Mike; Scutt, Simon

Within Logica UK, over 30 IT service projects are considered safetyrelated. These include operational IT services for airports, railway infrastructure asset management, nationwide radiation monitoring and hospital medical records services. A recent internal audit examined the processes and documents used to manage system safety on these services and made a series of recommendations for improvement. This paper looks at the changes and the challenges to introducing them, especially where the service is provided by multiple units supporting both safety and non-safety related services from multiple locations around the world. The recommendations include improvements to service agreements, improved process definitions, routine safety assessment of changes, enhanced call logging, improved staff competency and training, and increased safety awareness. Progress is reported as of today, together with a road map for implementation of the improvements to the service safety management system. A proposal for service assurance levels (SALs) is discussed as a way forward to cover the wide variety of services and associated safety risks.

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.

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.

COMPRESS - a computerized reactor safety system

International Nuclear Information System (INIS)

Vegh, E.

1986-01-01

The computerized reactor safety system, called COMPRESS, provides the following services: scram initiation; safety interlockings; event recording. The paper describes the architecture of the system and deals with reliability problems. A self-testing unit checks permanently the correct operation of the independent decision units. Moreover the decision units are tested by short pulses whether they can initiate a scram. The self-testing is described in detail

Cascade Distillation System Design for Safety and Mission Assurance

Science.gov (United States)

Sarguisingh, Miriam; Callahan, Michael R.; Okon, Shira

2015-01-01

Per the NASA Human Health, Life Support and Habitation System Technology Area 06 report "crewed missions venturing beyond Low-Earth Orbit (LEO) will require technologies with improved reliability, reduced mass, self-sufficiency, and minimal logistical needs as an emergency or quick-return option will not be feasible".1 To meet this need, the development team of the second generation Cascade Distillation System (CDS 2.0) chose a development approach that explicitly incorporate consideration of safety, mission assurance, and autonomy. The CDS 2.0 preliminary design focused on establishing a functional baseline that meets the CDS core capabilities and performance. The critical design phase is now focused on incorporating features through a deliberative process of establishing the systems failure modes and effects, identifying mitigation strategies, and evaluating the merit of the proposed actions through analysis and test. This paper details results of this effort on the CDS 2.0 design.

IT safety in production data networks (PDN). Tracking, evaluation and elimination of safety threats; IT-Sicherheit in Produktionsnetzen (PDN). Aufspueren, einschaetzen und beseitigen von Sicherheitsbedrohungen

Energy Technology Data Exchange (ETDEWEB)

Neider, Ulrich [DETACK GmbH, Ludwigsburg (Germany)

2013-03-01

This contribution reports on the threats of production data networks. Within the context of the IT security, the risks of production data networks are not only based on a possible contamination by computer viruses. The author of this contribution increases awareness of dangers to whose production systems are faced. The author also presents a phase model from his own consulting practice in order to increase the safety of production data systems by implementation of a safety control. This phase model consists of the six following steps: (a) Stock taking of the state of the art; (b) Determination of the target (development of a safety concept); (c) Creation of an IT safety policy; (d) Application of IT safety policy; (e) Audit of the results (IT safety audit); (f) Regular safety tests.

Nitrogen-system safety study: Portsmouth Gaseous Diffusion Plant

International Nuclear Information System (INIS)

1982-07-01

The Department of Energy has primary responsibility for the safety of operations at DOE-owned nuclear facilities. The guidelines for the analysis of credible accidents are outlined in DOE Order 5481.1. DOE has requested that existing plant facilities and operations be reviewed for potential safety problems not covered by standard industrial safety procedures. This review is being conducted by investigating individual facilities and documenting the results in Safety Study Reports which will be compiled to form the Existing Plant Final Safety Analysis Report which is scheduled for completion in September, 1984. This Safety Study documents the review of the Plant Nitrogen System facilities and operations and consists of Section 4.0, Facility and Process Description, and Section 5.0, Accident Analysis, of the Final Safety Analysis Report format. The existing nitrogen system consists of a Superior Air Products Company Type D Nitrogen Plant, nitrogen storage facilities, vaporization facilities and a distribution system. The system is designed to generate and distribute nitrogen gas used in the cascade for seal feed, buffer systems, and for servicing equipment when exceptionally low dew points are required. Gaseous nitrogen is also distributed to various process auxiliary buildings. The average usage is approximately 130,000 standard cubic feet per day

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

From Safe Systems to Patient Safety

DEFF Research Database (Denmark)

Aarts, J.; Nøhr, C.

2010-01-01

for the third conference with the theme: The ability to design, implement and evaluate safe, useable and effective systems within complex health care organizations. The theme for this conference was "Designing and Implementing Health IT: from safe systems to patient safety". The contributions have reflected...... and implementation of safe systems and thus contribute to the agenda of patient safety? The contributions demonstrate how the health informatics community has contributed to the performance of significant research and to translating research findings to develop health care delivery and improve patient safety......This volume presents the papers from the fourth International Conference on Information Technology in Health Care: Socio-technical Approaches held in Aalborg, Denmark in June 2010. In 2001 the first conference was held in Rotterdam, The Netherlands with the theme: Sociotechnical' approaches...

Benefits of a systematic approach to maintenance for safety and safety related systems

International Nuclear Information System (INIS)

Dam, R.F.; Ayazzudin, S.; Nickerson, J.H.

2003-01-01

For safety and safety-related systems, nuclear plants have to balance the requirements of demonstrating the reliability of each system, while maintaining the system and plant availability. With the goal of demonstrating statistical reliability, these systems have extensive testing programs, which often results in system unavailability and this can impact the plant capacity. The inputs to the process are often safety and regulatory related, resulting in programs that provide a high level of scrutiny. In such cases, the value of the application of a Systematic Assessment of Maintenance (SAM) process, such as Reliability Centered Maintenance (RCM), is questioned. The special case of Standby-Safety systems was discussed in a previous paper, where it was demonstrated how SAM techniques provide useful insight into current system performance, the impact of testing on component and system reliability, and how PSA considerations can be integrated into a comprehensive Maintenance, Surveillance, and Inspection (MSI) strategy. Although the system reliability requirements are an important part of the strategy evaluation, SAM techniques provide a systematic assessment within a broader context. Testing is only one part of an overall strategy focused on ensuring that component function is maintained through a combination of monitoring technologies (including testing), predictive techniques, and intrusive maintenance strategies. Each strategy is targeted to known component degradation mechanisms. This thinking can be extended to safety and safety related systems in general. Over the past 6 years, AECL has been working with CANDU utilities in the development and implementation of a comprehensive and integrated Plant Life Management (PLiM) program. As part of developing a comprehensive plant asset management approach, SAM techniques are used to develop a technical basis that not only works towards ensuring reliable operation of plant systems, but also facilitates the optimization and

Analysis of gas-liquid metal two-phase flows using a reactor safety analysis code SIMMER-III

International Nuclear Information System (INIS)

Suzuki, Tohru; Tobita, Yoshiharu; Kondo, Satoru; Saito, Yasushi; Mishima, Kaichiro

2003-01-01

SIMMER-III, a safety analysis code for liquid-metal fast reactors (LMFRs), includes a momentum exchange model based on conventional correlations for ordinary gas-liquid flows, such as an air-water system. From the viewpoint of safety evaluation of core disruptive accidents (CDAs) in LMFRs, we need to confirm that the code can predict the two-phase flow behaviors with high liquid-to-gas density ratios formed during a CDA. In the present study, the momentum exchange model of SIMMER-III was assessed and improved using experimental data of two-phase flows containing liquid metal, on which fundamental information, such as bubble shapes, void fractions and velocity fields, has been lacking. It was found that the original SIMMER-III can suitably represent high liquid-to-gas density ratio flows including ellipsoidal bubbles as seen in lower gas fluxes. In addition, the employment of Kataoka-Ishii's correlation has improved the accuracy of SIMMER-III for gas-liquid metal flows with cap-shape bubbles as identified in higher gas fluxes. Moreover, a new procedure, in which an appropriate drag coefficient can be automatically selected according to bubble shape, was developed. Through this work, the reliability and the precision of SIMMER-III have been much raised with regard to bubbly flows for various liquid-to-gas density ratios

Declarative Rule-based Safety for Robotic Perception Systems

DEFF Research Database (Denmark)

Mogensen, Johann Thor Ingibergsson; Kraft, Dirk; Schultz, Ulrik Pagh

2017-01-01

Mobile robots are used across many domains from personal care to agriculture. Working in dynamic open-ended environments puts high constraints on the robot perception system, which is critical for the safety of the system as a whole. To achieve the required safety levels the perception system needs...... to be certified, but no specific standards exist for computer vision systems, and the concept of safe vision systems remains largely unexplored. In this paper we present a novel domain-specific language that allows the programmer to express image quality detection rules for enforcing safety constraints...

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.

Operation safety of complex industrial systems. Main concepts

International Nuclear Information System (INIS)

Zwingelstein, G.

2009-01-01

Operation safety consists in knowing, evaluating, foreseeing, measuring and mastering the technological system and human failures in order to avoid their impacts on health and people's safety, on productivity, and on the environment, and to preserve the Earth's resources. This article recalls the main concepts of operation safety: 1 - evolutions in the domain; 2 - failures, missions and functions of a system and of its components: functional failure, missions and functions, industrial processes, notions of probability; 3 - basic concepts and operation safety: reliability, unreliability, failure density, failure rate, relations between them, availability, maintainability, safety. (J.S.)

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

33 CFR 147.847 - Safety Zone; BW PIONEER Floating Production, Storage, and Offloading System Safety Zone.

Science.gov (United States)

2010-07-01

... Production, Storage, and Offloading System Safety Zone. 147.847 Section 147.847 Navigation and Navigable... ZONES § 147.847 Safety Zone; BW PIONEER Floating Production, Storage, and Offloading System Safety Zone. (a) Description. The BW PIONEER, a Floating Production, Storage and Offloading (FPSO) system, is in...

Development of a check sheet for collecting information necessary for occupational safety and health activities and building relevant systems in overseas business places.

Science.gov (United States)

Kajiki, Shigeyuki; Kobayashi, Yuichi; Uehara, Masamichi; Nakanishi, Shigemoto; Mori, Koji

2016-06-07

This study aimed to develop an information gathering check sheet to efficiently collect information necessary for Japanese companies to build global occupational safety and health management systems in overseas business places. The study group consisted of 2 researchers with occupational physician careers in a foreign-affiliated company in Japan and 3 supervising occupational physicians who were engaged in occupational safety and health activities in overseas business places. After investigating information and sources of information necessary for implementing occupational safety and health activities and building relevant systems, we conducted information acquisition using an information gathering check sheet in the field, by visiting 10 regions in 5 countries (first phase). The accuracy of the information acquired and the appropriateness of the information sources were then verified in study group meetings to improve the information gathering check sheet. Next, the improved information gathering check sheet was used in another setting (3 regions in 1 country) to confirm its efficacy (second phase), and the information gathering check sheet was thereby completed. The information gathering check sheet was composed of 9 major items (basic information on the local business place, safety and health overview, safety and health systems, safety and health staff, planning/implementation/evaluation/improvement, safety and health activities, laws and administrative organs, local medical care systems and public health, and medical support for resident personnel) and 61 medium items. We relied on the following eight information sources: the internet, company (local business place and head office in Japan), embassy/consulate, ISO certification body, university or other educational institutions, and medical institutions (aimed at Japanese people or at local workers). Through multiple study group meetings and a two-phased field survey (13 regions in 6 countries), an information

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

Evaluating Safety Culture Under the Socio-Technical Complex Systems Perspective

International Nuclear Information System (INIS)

Lemos, F. L. de

2016-01-01

Since the term “safety culture” was coined, it has gained more and more attention as an effort to achieve higher levels of system safety. A good deal of effort has been done in order to better define, evaluate and implement safety culture programs in organizations throughout all industries, and especially in the Nuclear Industry. Unfortunately, despite all those efforts, we continue to witness accidents that are, in great part, attributed to flaws in the safety culture of the organization. Fukushima nuclear accident is one example of a serious accident in which flaws in the safety culture has been pointed to as one of the main contributors. In general, the definitions of safety culture emphasise the social aspect of the system. While the definitions also include the relations with the technical aspects, it does so in a general sense. For example, the International Nuclear Safety Advisory Group (INSAG) defines safety culture as: “The assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receives the attention warranted by their significance.” By the way safety culture is defined we can infer that it represents a property of a social system, or a property of the social aspect of the system. In this sense, the social system is a component of the whole system. Where, “system” is understood to be comprised of a social (humans) and technical (equipment) aspects, as a Nuclear Power Plant, for example. Therefore, treating safety culture as an identity on its own right, finding and fixing flaws in the safety culture may not be enough to improve safety of the system. We also needed to evaluate all the interactions between the components that comprise all the aspects of the system. In some cases a flaw in the safety culture can easily be detected, such as an employee not wearing appropriate individual protection equipment, e.g., dosimeter, or when basic safety

Describing phase coexistence in systems with small phases

International Nuclear Information System (INIS)

Lovett, R

2007-01-01

Clusters of atoms can be studied in molecular beams and by computer simulation; 'liquid drops' provide elementary models for atomic nuclei and for the critical nuclei of nucleation theory. These clusters are often described in thermodynamic terms, but the behaviour of small clusters near a phase boundary is qualitatively different from the behaviour at a first order phase transition in idealized thermodynamics. In the idealized case the density and entropy show mathematically sharp discontinuities when the phase boundary is crossed. In large, but finite, systems, the phase boundaries become regions of state space wherein these properties vary rapidly but continuously. In small clusters with a large surface/volume ratio, however, the positive interfacial free energy makes it unlikely, even in states on phase boundaries, that a cluster will have a heterogeneous structure. What is actually seen in these states is a structure that fluctuates in time between homogeneous structures characteristic of the two sides of the phase boundary. That is, structural fluctuations are observed. Thermodynamics only predicts average properties; statistical mechanics is required to understand these fluctuations. Failure to distinguish thermodynamic properties and characterizations of fluctuations, particularly in the context of first order phase transitions, has led to suggestions that the classical rules for thermodynamic stability are violated in small systems and that classical thermodynamics provides an inconsistent description of these systems. Much of the confusion stems from taking statistical mechanical identifications of thermodynamic properties, explicitly developed for large systems, and applying them uncritically to small systems. There are no inconsistencies if thermodynamic properties are correctly identified and the distinction between thermodynamic properties and fluctuations is made clear

Intelligent monitoring-based safety system of massage robot

Institute of Scientific and Technical Information of China (English)

胡宁; 李长胜; 王利峰; 胡磊; 徐晓军; 邹雲鹏; 胡玥; 沈晨

2016-01-01

As an important attribute of robots, safety is involved in each link of the full life cycle of robots, including the design, manufacturing, operation and maintenance. The present study on robot safety is a systematic project. Traditionally, robot safety is defined as follows: robots should not collide with humans, or robots should not harm humans when they collide. Based on this definition of robot safety, researchers have proposed ex ante and ex post safety standards and safety strategies and used the risk index and risk level as the evaluation indexes for safety methods. A massage robot realizes its massage therapy function through applying a rhythmic force on the massage object. Therefore, the traditional definition of safety, safety strategies, and safety realization methods cannot satisfy the function and safety requirements of massage robots. Based on the descriptions of the environment of massage robots and the tasks of massage robots, the present study analyzes the safety requirements of massage robots; analyzes the potential safety dangers of massage robots using the fault tree tool; proposes an error monitoring-based intelligent safety system for massage robots through monitoring and evaluating potential safety danger states, as well as decision making based on potential safety danger states; and verifies the feasibility of the intelligent safety system through an experiment.

Development and implementation of setpoint tolerances for special safety systems