Manpower analysis in transportation safety. Final report

Energy Technology Data Exchange (ETDEWEB)

Bauer, C.S.; Bowden, H.M.; Colford, C.A.; DeFilipps, P.J.; Dennis, J.D.; Ehlert, A.K.; Popkin, H.A.; Schrader, G.F.; Smith, Q.N.

1977-05-01

The project described provides a manpower review of national, state and local needs for safety skills, and projects future manning levels for transportation safety personnel in both the public and private sectors. Survey information revealed that there are currently approximately 121,000 persons employed directly in transportation safety occupations within the air carrier, highway and traffic safety, motor carrier, pipeline, rail carrier, and marine carrier transportation industry groups. The projected need for 1980 is over 145,000 of which over 80 percent will be in highway safety. An analysis of transportation tasks is included, and shows ten general categories about which the majority of safety activities are focused. A skills analysis shows a generally high level of educational background and several years of experience are required for most transportation safety jobs. An overall review of safety programs in the transportation industry is included, together with chapters on the individual transportation modes.

SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

Energy Technology Data Exchange (ETDEWEB)

E.N. Lindner

2004-12-03

The purpose of this seismic preclosure safety analysis is to identify the potential seismically-initiated event sequences associated with preclosure operations of the repository at Yucca Mountain and assign appropriate design bases to provide assurance of achieving the performance objectives specified in the Code of Federal Regulations (CFR) 10 CFR Part 63 for radiological consequences. This seismic preclosure safety analysis is performed in support of the License Application for the Yucca Mountain Project. In more detail, this analysis identifies the systems, structures, and components (SSCs) that are subject to seismic design bases. This analysis assigns one of two design basis ground motion (DBGM) levels, DBGM-1 or DBGM-2, to SSCs important to safety (ITS) that are credited in the prevention or mitigation of seismically-initiated event sequences. An application of seismic margins approach is also demonstrated for SSCs assigned to DBGM-2 by showing a high confidence of a low probability of failure at a higher ground acceleration value, termed a beyond-design basis ground motion (BDBGM) level. The objective of this analysis is to meet the performance requirements of 10 CFR 63.111(a) and 10 CFR 63.111(b) for offsite and worker doses. The results of this calculation are used as inputs to the following: (1) A classification analysis of SSCs ITS by identifying potential seismically-initiated failures (loss of safety function) that could lead to undesired consequences; (2) An assignment of either DBGM-1 or DBGM-2 to each SSC ITS credited in the prevention or mitigation of a seismically-initiated event sequence; and (3) A nuclear safety design basis report that will state the seismic design requirements that are credited in this analysis. The present analysis reflects the design information available as of October 2004 and is considered preliminary. The evolving design of the repository will be re-evaluated periodically to ensure that seismic hazards are properly

SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

International Nuclear Information System (INIS)

E.N. Lindner

2004-01-01

The purpose of this seismic preclosure safety analysis is to identify the potential seismically-initiated event sequences associated with preclosure operations of the repository at Yucca Mountain and assign appropriate design bases to provide assurance of achieving the performance objectives specified in the Code of Federal Regulations (CFR) 10 CFR Part 63 for radiological consequences. This seismic preclosure safety analysis is performed in support of the License Application for the Yucca Mountain Project. In more detail, this analysis identifies the systems, structures, and components (SSCs) that are subject to seismic design bases. This analysis assigns one of two design basis ground motion (DBGM) levels, DBGM-1 or DBGM-2, to SSCs important to safety (ITS) that are credited in the prevention or mitigation of seismically-initiated event sequences. An application of seismic margins approach is also demonstrated for SSCs assigned to DBGM-2 by showing a high confidence of a low probability of failure at a higher ground acceleration value, termed a beyond-design basis ground motion (BDBGM) level. The objective of this analysis is to meet the performance requirements of 10 CFR 63.111(a) and 10 CFR 63.111(b) for offsite and worker doses. The results of this calculation are used as inputs to the following: (1) A classification analysis of SSCs ITS by identifying potential seismically-initiated failures (loss of safety function) that could lead to undesired consequences; (2) An assignment of either DBGM-1 or DBGM-2 to each SSC ITS credited in the prevention or mitigation of a seismically-initiated event sequence; and (3) A nuclear safety design basis report that will state the seismic design requirements that are credited in this analysis. The present analysis reflects the design information available as of October 2004 and is considered preliminary. The evolving design of the repository will be re-evaluated periodically to ensure that seismic hazards are properly

Safety analysis fundamentals

International Nuclear Information System (INIS)

Wright, A.C.D.

2002-01-01

This paper discusses the safety analysis fundamentals in reactor design. This study includes safety analysis done to show consequences of postulated accidents are acceptable. Safety analysis is also used to set design of special safety systems and includes design assist analysis to support conceptual design. safety analysis is necessary for licensing a reactor, to maintain an operating license, support changes in plant operations

Analysis on safety production in coal mines Henan Province

Institute of Scientific and Technical Information of China (English)

KONG Liu-an; ZHANG Wen-yong

2006-01-01

Based on the rigorous situation of safety production in coal mines, the paper analyzed the statistical data of recent accidents indexes in Henan's coal mines. Using investigation and comparison analysis methods, a specified analysis on mining conditions, technical facility level, safety input and vocational quality of workers in Henan's coal mines was conducted. The result indicates that there have been existing such main safety production problems as weak safety management, low-level facilities, inadequate safety input and poor vocational quality and so on. Finally it proposes such reference solutions as to establish and perfect coal mining supervision and management system, to increase safety investment into techniques and facilities and to strengthen workers' safety education and introduction of more high-level professional talents.

41 CFR 102-80.105 - What information must be included in an equivalent level of safety analysis?

Science.gov (United States)

2010-07-01

... REGULATION REAL PROPERTY 80-SAFETY AND ENVIRONMENTAL MANAGEMENT Accident and Fire Prevention Equivalent Level.... Each analysis should describe potential reasonable worst case fire scenarios and their impact on the... and location of fuel items, space layout, building construction, openings and ventilation, suppression...

The practical implementation of integrated safety management for nuclear safety analysis and fire hazards analysis documentation

International Nuclear Information System (INIS)

COLLOPY, M.T.

1999-01-01

In 1995 Mr. Joseph DiNunno of the Defense Nuclear Facilities Safety Board issued an approach to describe the concept of an integrated safety management program which incorporates hazard and safety analysis to address a multitude of hazards affecting the public, worker, property, and the environment. Since then the U S . Department of Energy (DOE) has adopted a policy to systematically integrate safety into management and work practices at all levels so that missions can be completed while protecting the public, worker, and the environment. While the DOE and its contractors possessed a variety of processes for analyzing fire hazards at a facility, activity, and job; the outcome and assumptions of these processes have not always been consistent for similar types of hazards within the safety analysis and the fire hazard analysis. Although the safety analysis and the fire hazard analysis are driven by different DOE Orders and requirements, these analyses should not be entirely independent and their preparation should be integrated to ensure consistency of assumptions, consequences, design considerations, and other controls. Under the DOE policy to implement an integrated safety management system, identification of hazards must be evaluated and agreed upon to ensure that the public. the workers. and the environment are protected from adverse consequences. The DOE program and contractor management need a uniform, up-to-date reference with which to plan. budget, and manage nuclear programs. It is crucial that DOE understand the hazards and risks necessarily to authorize the work needed to be performed. If integrated safety management is not incorporated into the preparation of the safety analysis and the fire hazard analysis, inconsistencies between assumptions, consequences, design considerations, and controls may occur that affect safety. Furthermore, confusion created by inconsistencies may occur in the DOE process to grant authorization of the work. In accordance with

Technical safety requirements control level verification

International Nuclear Information System (INIS)

STEWART, J.L.

1999-01-01

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

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.

THE CONDITION AND THE DYNAMICS OF CHANGES OF REGIONAL ENERGETIC SAFETY LEVEL

Directory of Open Access Journals (Sweden)

A.L. Myzin

2006-12-01

Full Text Available On the basis of indicative analysis method use, the dynamic processes of changes of energetic safety condition of federal districts and subjects of Russian Federation for last 5 years are investigated. The results of diagnosing safety levels for separate indicators, their blocks and the results of situation evaluation as a whole are discussed. The comparison of regions’ energetic safety condition is given, the causes of crisis situations appearance are discovered, and on this basis the suggestions for regions’ safety levels increasing are formulated.

Dynamic probability evaluation of safety levels of earth-rockfill dams using Bayesian approach

Directory of Open Access Journals (Sweden)

Zi-wu Fan

2009-06-01

Full Text Available In order to accurately predict and control the aging process of dams, new information should be collected continuously to renew the quantitative evaluation of dam safety levels. Owing to the complex structural characteristics of dams, it is quite difficult to predict the time-varying factors affecting their safety levels. It is not feasible to employ dynamic reliability indices to evaluate the actual safety levels of dams. Based on the relevant regulations for dam safety classification in China, a dynamic probability description of dam safety levels was developed. Using the Bayesian approach and effective information mining, as well as real-time information, this study achieved more rational evaluation and prediction of dam safety levels. With the Bayesian expression of discrete stochastic variables, the a priori probabilities of the dam safety levels determined by experts were combined with the likelihood probability of the real-time check information, and the probability information for the evaluation of dam safety levels was renewed. The probability index was then applied to dam rehabilitation decision-making. This method helps reduce the difficulty and uncertainty of the evaluation of dam safety levels and complies with the current safe decision-making regulations for dams in China. It also enhances the application of current risk analysis methods for dam safety levels.

On Optimum Safety Levels of Breakwaters

DEFF Research Database (Denmark)

Burcharth, Hans F.; Sørensen, John Dalsgaard

2006-01-01

The paper presents results from numerical simulations performed with the objective of identifying optimum design safety levels of conventional rubble mound and caisson breakwaters, corresponding to the lowest costs over the service life of the structures. The work is related to the PIANC Working...... Group 47 on "Selection of type of breakwater structures". The paper summaries results given in Burcharth and Sorensen (2005) related to outer rubble mound breakwaters but focus on optimum safety levels for outer caisson breakwaters on low and high rubble foundations placed on sea beds strong enough...... to resist geotechnical slip failures. Optimum safety levels formulated for use both in deterministic and probabilistic design procedures are given. Results obtained so far indicate that the optimum safety levels for caisson breakwaters are much higher than for rubble mound breakwaters....

NPP Temelin safety analysis reports and PSA status

International Nuclear Information System (INIS)

Mlady, O.

1999-01-01

To enhance the safety level of Temelin NPP, recommendations of the international reviews were implemented into the design as well as into organization of the plant construction and preparation for operation. The safety assessment of these design changes has been integrated and reflected in the Safety Analysis Reports, which follow the internationally accepted guidelines. All safety analyses within Safety Analysis Reports were repeated carefully considering technical improvements and replacements to complement preliminary safety documentation. These analyses were performed by advanced western computer codes to the depth and in the structure required by western standards. The Temelin NPP followed a systematic approach in the functional design of the Reactor Protection System and related safety analyses. Modifications of reactor protection system increase defense in depth and facilitate demonstrating that LOCA and radiological limits are met for non-LOCA events. The rigorous safety analysis methodology provides assurance that LOCA and radiological limits are met. Established and accepted safety analysis methodology and accepted criteria were applied to Temelin NPP meeting US NRC and Czech Republic requirements. IAEA guidelines and recommendations

Software FMEA analysis for safety-related application software

International Nuclear Information System (INIS)

Park, Gee-Yong; Kim, Dong Hoon; Lee, Dong Young

2014-01-01

Highlights: • We develop a modified FMEA analysis suited for applying to software architecture. • A template for failure modes on a specific software language is established. • A detailed-level software FMEA analysis on nuclear safety software is presented. - Abstract: A method of a software safety analysis is described in this paper for safety-related application software. The target software system is a software code installed at an Automatic Test and Interface Processor (ATIP) in a digital reactor protection system (DRPS). For the ATIP software safety analysis, at first, an overall safety or hazard analysis is performed over the software architecture and modules, and then a detailed safety analysis based on the software FMEA (Failure Modes and Effect Analysis) method is applied to the ATIP program. For an efficient analysis, the software FMEA analysis is carried out based on the so-called failure-mode template extracted from the function blocks used in the function block diagram (FBD) for the ATIP software. The software safety analysis by the software FMEA analysis, being applied to the ATIP software code, which has been integrated and passed through a very rigorous system test procedure, is proven to be able to provide very valuable results (i.e., software defects) that could not be identified during various system tests

IAEA Review for Gap Analysis of Safety Analysis Capability

International Nuclear Information System (INIS)

Basic, Ivica; Kim, Manwoong; Huges, Peter; Lim, B-K; D'Auria, Francesco; Louis, Vidard Michael

2014-01-01

The IAEA Asian Nuclear Safety Network (ANSN) was launched in 2002 in the framework of the Extra Budgetary Programme (EBP) on the Safety of Nuclear Installations in the South East Asia, Pacific and Far East Countries. The main objective is to strengthen and expand human and advanced Information Technology (IT) network to pool, analyse and share nuclear safety knowledge and practical experience for peaceful uses in this region. Under the ANSN framework, a technical group on Safety Analysis (SATG) was established in 2004 aimed to providing a forum for the exchange of experience in the following areas of safety analysis: · To provide a forum for an exchange of experience in the area of safety analysis, · To maintain and improve the knowledge on safety analysis method, · To enhance the utilization of computer codes, · To pool and analyse the issues related with safety analysis of research reactor, and · To facilitate mutual interested on safety analysis among member countries. A sustainable and successful nuclear energy programme requires a strong technical infrastructure, including a workforce made up of highly specialized and well-educated professionals. A significant portion of this technical capacity must be dedicated to safety- especially to safety analysis- as only then can it serve as the basis for making the right decisions during the planning, licensing, construction and operation of new nuclear facilities. In this regard, the IAEA has provided ANSN member countries with comprehensive training opportunities for capacity building in safety analysis. Nevertheless, the SATG recognizes that it is difficult to achieve harmonization in this area among all member countries because of their different competency levels. Therefore, it is necessary to quickly identify the most obvious gaps in safety analysis capability and then to use existing resources to begin to fill those gaps. The goal of this Expert Mission (EM) for gap finding service is to facilitate

Linking Safety Analysis to Safety Requirements

DEFF Research Database (Denmark)

Hansen, Kirsten Mark

Software for safety critical systems must deal with the hazards identified by safety analysistechniques: Fault trees, event trees,and cause consequence diagrams can be interpreted as safety requirements and used in the design activity. We propose that the safety analysis and the system design use...

National Waste Repository Novi Han operational safety analysis report. Safety assessment methodology

International Nuclear Information System (INIS)

2003-01-01

The scope of the safety assessment (SA), presented includes: waste management functions (acceptance, conditioning, storage, disposal), inventory (current and expected in the future), hazards (radiological and non-radiological) and normal and accidental modes. The stages in the development of the SA are: criteria selection, information collection, safety analysis and safety assessment documentation. After the review the facilities functions and the national and international requirements, the criteria for safety level assessment are set. As a result from the 2nd stage actual parameters of the facility, necessary for safety analysis are obtained.The methodology is selected on the base of the comparability of the results with the results of previous safety assessments and existing standards and requirements. The procedure and requirements for scenarios selection are described. A radiological hazard categorisation of the facilities is presented. Qualitative hazards and operability analysis is applied. The resulting list of events are subjected to procedure for prioritization by method of 'criticality analysis', so the estimation of the risk is given for each event. The events that fall into category of risk on the boundary of acceptability or are unacceptable are subjected to the next steps of the analysis. As a result the lists with scenarios for PSA and possible design scenarios are established. PSA logical modeling and quantitative calculations of accident sequences are presented

A risk-informed perspective on deterministic safety analysis of nuclear power plants

International Nuclear Information System (INIS)

Wan, P.T.

2009-01-01

In this work, the deterministic safety analysis (DSA) approach to nuclear safety is examined from a risk-informed perspective. One objective of safety analysis of a nuclear power plant is to demonstrate via analysis that the risks to the public from events or accidents that are within the design basis of the power plant are within acceptable levels with a high degree of assurance. This nuclear safety analysis objective can be translated into two requirements on the risk estimates of design basis events or accidents: the nominal risk estimate to the public must be shown to be within acceptable levels, and the uncertainty in the risk estimates must be shown to be small on an absolute or relative basis. The DSA approach combined with the defense-in-depth (DID) principle is a simplified safety analysis approach that attempts to achieve the above safety analysis objective in the face of potentially large uncertainties in the risk estimates of a nuclear power plant by treating the various uncertainty contributors using a stylized conservative binary (yes-no) approach, and applying multiple overlapping physical barriers and defense levels to protect against the release of radioactivity from the reactor. It is shown that by focusing on the consequence aspect of risk, the previous two nuclear safety analysis requirements on risk can be satisfied with the DSA-DID approach to nuclear safety. It is also shown the use of multiple overlapping physical barriers and defense levels in the traditional DSA-DID approach to nuclear safety is risk-informed in the sense that it provides a consistently high level of confidence in the validity of the safety analysis results for various design basis events or accidents with a wide range of frequency of occurrence. It is hoped that by providing a linkage between the consequence analysis approach in DSA with a risk-informed perspective, greater understanding of the limitation and capability of the DSA approach is obtained. (author)

Safety analysis for research reactors

International Nuclear Information System (INIS)

2008-01-01

The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of

Technical safety requirements control level verification; TOPICAL

International Nuclear Information System (INIS)

STEWART, J.L.

1999-01-01

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

Support analysis for safety analysis development for CANDU nuclear power plant

International Nuclear Information System (INIS)

Bedreaga, L.; Florescu, Gh.; Apostol, M.; Nitoi, M.

2004-01-01

Probabilistic Safety Assessment analysis (PSA) is a technique used to assess the safety of a nuclear power plant. Assessments of the nuclear plant systems/components from safety point of view consist in accomplishment of a lot of support analyses that are the base for the main analysis, in order to evaluate the impact of occurrences of abnormal states for these systems. Evaluation of initiating events frequency and components failure rate is based on underlying probabilistic theory and mathematic statistics. Some of these analyses are detailed analyses and are known very well in PSA. There are also some analyses, named support analyses for PSA, which are very important but less applicable because they involve a huge human effort and hardware facilities to accomplish. The usual methods applicable in PSA such as input data extracted from the specific documentation (operation procedures, testing procedures, maintenance procedures and so on) or conservative evaluation provide a high level of uncertainty for both input and output data. The paper describes support analysis required to improve the certainty level in evaluation of reliability parameters and also in the final results (either risk, reliability or safety assessment). (author)

Preparing a Safety Analysis Report using the building block approach

International Nuclear Information System (INIS)

Herrington, C.C.

1990-01-01

The credibility of the applicant in a licensing proceeding is severely impacted by the quality of the license application, particularly the Safety Analysis Report. To ensure the highest possible credibility, the building block approach was devised to support the development of a quality Safety Analysis Report. The approach incorporates a comprehensive planning scheme that logically ties together all levels of the investigation and provides the direction necessary to prepare a superior Safety Analysis Report

HANFORD SAFETY ANALYSIS & RISK ASSESSMENT HANDBOOK (SARAH)

Energy Technology Data Exchange (ETDEWEB)

EVANS, C B

2004-12-21

The purpose of the Hanford Safety Analysis and Risk Assessment Handbook (SARAH) is to support the development of safety basis documentation for Hazard Category 2 and 3 (HC-2 and 3) U.S. Department of Energy (DOE) nuclear facilities to meet the requirements of 10 CFR 830, ''Nuclear Safety Management''. Subpart B, ''Safety Basis Requirements.'' Consistent with DOE-STD-3009-94, Change Notice 2, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'' (STD-3009), and DOE-STD-3011-2002, ''Guidance for Preparation of Basis for Interim Operation (BIO) Documents'' (STD-3011), the Hanford SARAH describes methodology for performing a safety analysis leading to development of a Documented Safety Analysis (DSA) and derivation of Technical Safety Requirements (TSR), and provides the information necessary to ensure a consistently rigorous approach that meets DOE expectations. The DSA and TSR documents, together with the DOE-issued Safety Evaluation Report (SER), are the basic components of facility safety basis documentation. For HC-2 or 3 nuclear facilities in long-term surveillance and maintenance (S&M), for decommissioning activities, where source term has been eliminated to the point that only low-level, residual fixed contamination is present, or for environmental remediation activities outside of a facility structure, DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities'' (STD-1120), may serve as the basis for the DSA. HC-2 and 3 environmental remediation sites also are subject to the hazard analysis methodologies of this standard.

Individual employee's perceptions of " Group-level Safety Climate" (supervisor referenced) versus " Organization-level Safety Climate" (top management referenced): Associations with safety outcomes for lone workers.

Science.gov (United States)

Huang, Yueng-Hsiang; Lee, Jin; McFadden, Anna C; Rineer, Jennifer; Robertson, Michelle M

2017-01-01

Research has shown that safety climate is among the strongest predictors of safety behavior and safety outcomes in a variety of settings. Previous studies have established that safety climate is a multi-faceted construct referencing multiple levels of management within a company, most generally: the organization level (employee perceptions of top management's commitment to and prioritization of safety) and group level (employee perceptions of direct supervisor's commitment to and prioritization of safety). Yet, no research to date has examined the potential interaction between employees' organization-level safety climate (OSC) and group-level safety climate (GSC) perceptions. Furthermore, prior research has mainly focused on traditional work environments in which supervisors and workers interact in the same location throughout the day. Little research has been done to examine safety climate with regard to lone workers. The present study aims to address these gaps by examining the relationships between truck drivers' (as an example of lone workers) perceptions of OSC and GSC, both potential linear and non-linear relationships, and how these predict important safety outcomes. Participants were 8095 truck drivers from eight trucking companies in the United States with an average response rate of 44.8%. Results showed that employees' OSC and GSC perceptions are highly correlated (r= 0.78), but notable gaps between the two were observed for some truck drivers. Uniquely, both OSC and GSC scores were found to have curvilinear relationships with safe driving behavior, and both scores were equally predictive of safe driving behavior. Results also showed the two levels of climate significantly interacted with one another to predict safety behavior such that if either the OSC or GSC scores were low, the other's contribution to safety behavior became stronger. These findings suggest that OSC and GSC may function in a compensatory manner and promote safe driving behavior even

2005 dossier: granite. Tome: safety analysis of the geologic disposal

International Nuclear Information System (INIS)

2005-01-01

This document makes a status of the researches carried out by the French national agency of radioactive wastes (ANDRA) about the safety aspects of the geologic disposal of high-level and long-lived (HLLL) radioactive wastes in granite formations. Content: 1 - safety approach: context and general goal, references, design approach by safety functions, safety approach during the construction-exploitation-observation-closure phase, safety analysis during the post-closure phase; 2 - general description: HLLL wastes, granitic environment, general structure of the architecture of a disposal facility; 3 - safety functions and disposal design: general context, safety functions of the long-term disposal, design dispositions retained to answer the functions; 4 - operational safety: people's protection, radiological risks during exploitation, risk analysis in accident situation; 5 - qualitative safety analysis: methodology, main results of the analysis of the features, events and processes (FEP) database; 6 - disposal efficiency evaluation during post-closure phase: calculation models, calculation tools used for the modeling of radionuclides transport, calculation results and main lessons. (J.S.)

Time Based Workload Analysis Method for Safety-Related Operator Actions in Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

During the design basis event, the safety system performs safety functions to mitigate the event. The most of safety system is actuated by automatic system however, there are operator manual actions that are needed for the plant safety. These operator actions are classified as important human actions in human factors engineering design. The human factors engineering analysis and evaluation is needed for these important human actions to assure that operator successfully perform their tasks for plant safety and operational goals. The work load analysis is one of the required analysis for the important human actions.

Time Based Workload Analysis Method for Safety-Related Operator Actions in Safety Analysis

International Nuclear Information System (INIS)

Kim, Yun Goo; Oh, Eung Se

2016-01-01

During the design basis event, the safety system performs safety functions to mitigate the event. The most of safety system is actuated by automatic system however, there are operator manual actions that are needed for the plant safety. These operator actions are classified as important human actions in human factors engineering design. The human factors engineering analysis and evaluation is needed for these important human actions to assure that operator successfully perform their tasks for plant safety and operational goals. The work load analysis is one of the required analysis for the important human actions.

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

Nuclear power: levels of safety

International Nuclear Information System (INIS)

Lidsky, L.M.

1988-01-01

The rise and fall of the nuclear power industry in the United States is a well-documented story with enough socio-technological conflict to fill dozens of scholarly, and not so scholarly, books. Whatever the reasons for the situation we are now in, and no matter how we apportion the blame, the ultimate choice of whether to use nuclear power in this country is made by the utilities and by the public. Their choices are, finally, based on some form of risk-benefit analysis. Such analysis is done in well-documented and apparently logical form by the utilities and in a rather more inchoate but not necessarily less accurate form by the public. Nuclear power has failed in the United States because both the real and perceived risks outweigh the potential benefits. The national decision not to rely upon nuclear power in its present form is not an irrational one. A wide ranging public balancing of risk and benefit requires a classification of risk which is clear and believable for the public to be able to assess the risks associated with given technological structures. The qualitative four-level safety ladder provides such a framework. Nuclear reactors have been designed which fit clearly and demonstrably into each of the possible qualitative safety levels. Surprisingly, it appears that safer may also mean cheaper. The intellectual and technical prerequisites are in hand for an important national decision. Deployment of a qualitatively different second generation of nuclear reactors can have important benefits for the United States. Surprisingly, it may well be the nuclear establishment itself, with enormous investments of money and pride in the existing nuclear systems, that rejects second generation reactors. It may be that we will not have a second generation of reactors until the first generation of nuclear engineers and nuclear power advocates has retired

A multilevel model of patient safety culture: cross-level relationship between organizational culture and patient safety behavior in Taiwan's hospitals.

Science.gov (United States)

Chen, I-Chi; Ng, Hui-Fuang; Li, Hung-Hui

2012-01-01

As health-care organizations endeavor to improve their quality of care, there is a growing recognition of the importance of establishing a culture of patient safety. The main objective of this study was to investigate the cross-level influences of organizational culture on patient safety behavior in Taiwan's hospitals. The authors measured organizational culture (bureaucratic, supportive and innovative culture), patient safety culture and behavior from 788 hospital workers among 42 hospitals in Taiwan. Multilevel analysis was applied to explore the relationship between organizational culture (group level) and patient safety behavior (individual level). Patient safety culture had positive impact on patient safety behavior in Taiwan's hospitals. The results also indicated that bureaucratic, innovative and supportive organizational cultures all had direct influence on patient safety behavior. However, only supportive culture demonstrated significant moderation effect on the relationship between patient safety culture and patient safety behavior. Furthermore, organizational culture strength was shown correlated negatively with patient safety culture variability. Overall, organizational culture plays an important role in patient safety activities. Safety behaviors of hospital staff are partly influenced by the prevailing cultural norms in their organizations and work groups. For management implications, constructed patient priority from management commitment to leadership is necessary. For academic implications, research on patient safety should consider leadership, group dynamics and organizational learning. These factors are important for understanding the barriers and the possibilities embedded in patient safety. Copyright © 2011 John Wiley & Sons, Ltd.

HANFORD SAFETY ANALYSIS and RISK ASSESSMENT HANDBOOK (SARAH)

International Nuclear Information System (INIS)

EVANS, C.B.

2004-01-01

The purpose of the Hanford Safety Analysis and Risk Assessment Handbook (SARAH) is to support the development of safety basis documentation for Hazard Category 2 and 3 (HC-2 and 3) U.S. Department of Energy (DOE) nuclear facilities to meet the requirements of 10 CFR 830, ''Nuclear Safety Management''. Subpart B, ''Safety Basis Requirements.'' Consistent with DOE-STD-3009-94, Change Notice 2, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'' (STD-3009), and DOE-STD-3011-2002, ''Guidance for Preparation of Basis for Interim Operation (BIO) Documents'' (STD-3011), the Hanford SARAH describes methodology for performing a safety analysis leading to development of a Documented Safety Analysis (DSA) and derivation of Technical Safety Requirements (TSR), and provides the information necessary to ensure a consistently rigorous approach that meets DOE expectations. The DSA and TSR documents, together with the DOE-issued Safety Evaluation Report (SER), are the basic components of facility safety basis documentation. For HC-2 or 3 nuclear facilities in long-term surveillance and maintenance (S and M), for decommissioning activities, where source term has been eliminated to the point that only low-level, residual fixed contamination is present, or for environmental remediation activities outside of a facility structure, DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities'' (STD-1120), may serve as the basis for the DSA. HC-2 and 3 environmental remediation sites also are subject to the hazard analysis methodologies of this standard

Systems Analysis of NASA Aviation Safety Program: Final Report

Science.gov (United States)

Jones, Sharon M.; Reveley, Mary S.; Withrow, Colleen A.; Evans, Joni K.; Barr, Lawrence; Leone, Karen

2013-01-01

A three-month study (February to April 2010) of the NASA Aviation Safety (AvSafe) program was conducted. This study comprised three components: (1) a statistical analysis of currently available civilian subsonic aircraft data from the National Transportation Safety Board (NTSB), the Federal Aviation Administration (FAA), and the Aviation Safety Information Analysis and Sharing (ASIAS) system to identify any significant or overlooked aviation safety issues; (2) a high-level qualitative identification of future safety risks, with an assessment of the potential impact of the NASA AvSafe research on the National Airspace System (NAS) based on these risks; and (3) a detailed, top-down analysis of the NASA AvSafe program using an established and peer-reviewed systems analysis methodology. The statistical analysis identified the top aviation "tall poles" based on NTSB accident and FAA incident data from 1997 to 2006. A separate examination of medical helicopter accidents in the United States was also conducted. Multiple external sources were used to develop a compilation of ten "tall poles" in future safety issues/risks. The top-down analysis of the AvSafe was conducted by using a modification of the Gibson methodology. Of the 17 challenging safety issues that were identified, 11 were directly addressed by the AvSafe program research portfolio.

Qualitative uncertainty analysis in probabilistic safety assessment context

International Nuclear Information System (INIS)

Apostol, M.; Constantin, M; Turcu, I.

2007-01-01

In Probabilistic Safety Assessment (PSA) context, an uncertainty analysis is performed either to estimate the uncertainty in the final results (the risk to public health and safety) or to estimate the uncertainty in some intermediate quantities (the core damage frequency, the radionuclide release frequency or fatality frequency). The identification and evaluation of uncertainty are important tasks because they afford credit to the results and help in the decision-making process. Uncertainty analysis can be performed qualitatively or quantitatively. This paper performs a preliminary qualitative uncertainty analysis, by identification of major uncertainty in PSA level 1- level 2 interface and in the other two major procedural steps of a level 2 PSA i.e. the analysis of accident progression and of the containment and analysis of source term for severe accidents. One should mention that a level 2 PSA for a Nuclear Power Plant (NPP) involves the evaluation and quantification of the mechanisms, amount and probabilities of subsequent radioactive material releases from the containment. According to NUREG 1150, an important task in source term analysis is fission products transport analysis. The uncertainties related to the isotopes distribution in CANDU NPP primary circuit and isotopes' masses transferred in the containment, using SOPHAEROS module from ASTEC computer code will be also presented. (authors)

Statistical considerations on safety analysis

International Nuclear Information System (INIS)

Pal, L.; Makai, M.

2004-01-01

The authors have investigated the statistical methods applied to safety analysis of nuclear reactors and arrived at alarming conclusions: a series of calculations with the generally appreciated safety code ATHLET were carried out to ascertain the stability of the results against input uncertainties in a simple experimental situation. Scrutinizing those calculations, we came to the conclusion that the ATHLET results may exhibit chaotic behavior. A further conclusion is that the technological limits are incorrectly set when the output variables are correlated. Another formerly unnoticed conclusion of the previous ATHLET calculations that certain innocent looking parameters (like wall roughness factor, the number of bubbles per unit volume, the number of droplets per unit volume) can influence considerably such output parameters as water levels. The authors are concerned with the statistical foundation of present day safety analysis practices and can only hope that their own misjudgment will be dispelled. Until then, the authors suggest applying correct statistical methods in safety analysis even if it makes the analysis more expensive. It would be desirable to continue exploring the role of internal parameters (wall roughness factor, steam-water surface in thermal hydraulics codes, homogenization methods in neutronics codes) in system safety codes and to study their effects on the analysis. In the validation and verification process of a code one carries out a series of computations. The input data are not precisely determined because measured data have an error, calculated data are often obtained from a more or less accurate model. Some users of large codes are content with comparing the nominal output obtained from the nominal input, whereas all the possible inputs should be taken into account when judging safety. At the same time, any statement concerning safety must be aleatory, and its merit can be judged only when the probability is known with which the

Deterministic Safety Analysis for Nuclear Power Plants. Specific Safety Guide (Russian Edition)

International Nuclear Information System (INIS)

2014-01-01

The objective of this Safety Guide is to provide harmonized guidance to designers, operators, regulators and providers of technical support on deterministic safety analysis for nuclear power plants. It provides information on the utilization of the results of such analysis for safety and reliability improvements. The Safety Guide addresses conservative, best estimate and uncertainty evaluation approaches to deterministic safety analysis and is applicable to current and future designs. Contents: 1. Introduction; 2. Grouping of initiating events and associated transients relating to plant states; 3. Deterministic safety analysis and acceptance criteria; 4. Conservative deterministic safety analysis; 5. Best estimate plus uncertainty analysis; 6. Verification and validation of computer codes; 7. Relation of deterministic safety analysis to engineering aspects of safety and probabilistic safety analysis; 8. Application of deterministic safety analysis; 9. Source term evaluation for operational states and accident conditions; References

The condition and the dynamics of changes of regional energetic safety level

OpenAIRE

Anatoliy Myzin; Aleksey Kalina; Andrey Kozitsyn; Pavel Pykhov

2006-01-01

On the basis of indicative analysis method use, the dynamic processes of changes of energetic safety condition of federal districts and subjects of Russian Federation for last 5 years are investigated. The results of diagnosing safety levels for separate indicators, their blocks and the results of situation evaluation as a whole are discussed. The comparison of regions’ energetic safety condition is given, the causes of crisis situations appearance are discovered, and on this basis the sugg...

K Basin safety analysis

International Nuclear Information System (INIS)

Porten, D.R.; Crowe, R.D.

1994-01-01

The purpose of this accident safety analysis is to document in detail, analyses whose results were reported in summary form in the K Basins Safety Analysis Report WHC-SD-SNF-SAR-001. The safety analysis addressed the potential for release of radioactive and non-radioactive hazardous material located in the K Basins and their supporting facilities. The safety analysis covers the hazards associated with normal K Basin fuel storage and handling operations, fuel encapsulation, sludge encapsulation, and canister clean-up and disposal. After a review of the Criticality Safety Evaluation of the K Basin activities, the following postulated events were evaluated: Crane failure and casks dropped into loadout pit; Design basis earthquake; Hypothetical loss of basin water accident analysis; Combustion of uranium fuel following dryout; Crane failure and cask dropped onto floor of transfer area; Spent ion exchange shipment for burial; Hydrogen deflagration in ion exchange modules and filters; Release of Chlorine; Power availability and reliability; and Ashfall

Safety analysis SFR 1. Long-term safety

Energy Technology Data Exchange (ETDEWEB)

2008-12-15

An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the regulatory

Safety analysis SFR 1. Long-term safety

International Nuclear Information System (INIS)

2008-12-01

An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the regulatory

Safety analysis SFR 1. Long-term safety

Energy Technology Data Exchange (ETDEWEB)

2008-12-15

An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the

Safety analysis for 'Fugen'

International Nuclear Information System (INIS)

1997-10-01

The improvement of safety in nuclear power stations is an important proposition. Therefore also as to the safety evaluation, it is important to comprehensively and systematically execute it by referring to the operational experience and the new knowledge which is important for the safety throughout the period of use as well as before the construction and the start of operation of nuclear power stations. In this report, the results when the safety analysis for ''Fugen'' was carried out by referring to the newest technical knowledge are described. As the result, it was able to be confirmed that the safety of ''Fugen'' has been secured by the inherent safety and the facilities which were designed for securing the safety. The basic way of thinking on the safety analysis including the guidelines to be conformed to is mentioned. As to the abnormal transient change in operation and accidents, their definition, the events to be evaluated and the standards for judgement are reported. The matters which were taken in consideration at the time of the analysis are shown. The computation programs used for the analysis were REACT, HEATUP, LAYMON, FATRAC, SENHOR, LOTRAC, FLOOD and CONPOL. The analyses of the abnormal transient change in operation and accidents are reported on the causes, countermeasures, protective functions and results. (K.I.)

Reactor Safety Analysis

International Nuclear Information System (INIS)

Arien, B.

2000-01-01

The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported

Operating safety requirements for the intermediate level liquid waste system

International Nuclear Information System (INIS)

1980-07-01

The operation of the Intermediate Level Liquid Waste (ILW) System, which is described in the Final Safety Analysis, consists of two types of operations, namely: (1) the operation of a tank farm which involves the storage and transportation through pipelines of various radioactive liquids; and (2) concentration of the radioactive liquids by evaporation including rejection of the decontaminated condensate to the Waste Treatment Plant and retention of the concentrate. The following safety requirements in regard to these operations are presented: safety limits and limiting control settings; limiting conditions for operation; and surveillance requirements. Staffing requirements, reporting requirements, and steps to be taken in the event of an abnormal occurrence are also described

Preclosure Safety Analysis Guide

International Nuclear Information System (INIS)

D.D. Orvis

2003-01-01

A preclosure safety analysis (PSA) is a required element of the License Application (LA) for the high- level radioactive waste repository at Yucca Mountain. This guide provides analysts and other Yucca Mountain Repository Project (the Project) personnel with standardized methods for developing and documenting the PSA. The definition of the PSA is provided in 10 CFR 63.2, while more specific requirements for the PSA are provided in 10 CFR 63.112, as described in Sections 1.2 and 2. The PSA requirements described in 10 CFR Part 63 were developed as risk-informed performance-based regulations. These requirements must be met for the LA. The PSA addresses the safety of the Geologic Repository Operations Area (GROA) for the preclosure period (the time up to permanent closure) in accordance with the radiological performance objectives of 10 CFR 63.111. Performance objectives for the repository after permanent closure (described in 10 CFR 63.113) are not mentioned in the requirements for the PSA and they are not considered in this guide. The LA will be comprised of two phases: the LA for construction authorization (CA) and the LA amendment to receive and possess (R and P) high-level radioactive waste (HLW). PSA methods must support the safety analyses that will be based on the differing degrees of design detail in the two phases. The methods described herein combine elements of probabilistic risk assessment (PRA) and deterministic analyses that comprise a risk-informed performance-based safety analysis. This revision to the PSA guide was prepared for the following objectives: (1) To correct factual and typographical errors. (2) To provide additional material suggested from reviews by the Project, the U.S. Department of Energy (DOE), and U.S. Nuclear Regulatory Commission (NRC) Staffs. (3) To update material in accordance with approaches and/or strategies adopted by the Project. In addition, a principal objective for the planned revision was to ensure that the methods and

C-Band Airport Surface Communications System Engineering-Initial High-Level Safety Risk Assessment and Mitigation

Science.gov (United States)

Zelkin, Natalie; Henriksen, Stephen

2011-01-01

This document is being provided as part of ITT's NASA Glenn Research Center Aerospace Communication Systems Technical Support (ACSTS) contract: "New ATM Requirements--Future Communications, C-Band and L-Band Communications Standard Development." ITT has completed a safety hazard analysis providing a preliminary safety assessment for the proposed C-band (5091- to 5150-MHz) airport surface communication system. The assessment was performed following the guidelines outlined in the Federal Aviation Administration Safety Risk Management Guidance for System Acquisitions document. The safety analysis did not identify any hazards with an unacceptable risk, though a number of hazards with a medium risk were documented. This effort represents an initial high-level safety hazard analysis and notes the triggers for risk reassessment. A detailed safety hazards analysis is recommended as a follow-on activity to assess particular components of the C-band communication system after the profile is finalized and system rollout timing is determined. A security risk assessment has been performed by NASA as a parallel activity. While safety analysis is concerned with a prevention of accidental errors and failures, the security threat analysis focuses on deliberate attacks. Both processes identify the events that affect operation of the system; and from a safety perspective the security threats may present safety risks.

Reliability analysis of PLC safety equipment

Energy Technology Data Exchange (ETDEWEB)

Yu, J.; Kim, J. Y. [Chungnam Nat. Univ., Daejeon (Korea, Republic of)

2006-06-15

FMEA analysis for Nuclear Safety Grade PLC, failure rate prediction for nuclear safety grade PLC, sensitivity analysis for components failure rate of nuclear safety grade PLC, unavailability analysis support for nuclear safety system.

Reliability analysis of PLC safety equipment

International Nuclear Information System (INIS)

Yu, J.; Kim, J. Y.

2006-06-01

FMEA analysis for Nuclear Safety Grade PLC, failure rate prediction for nuclear safety grade PLC, sensitivity analysis for components failure rate of nuclear safety grade PLC, unavailability analysis support for nuclear safety system

Disentangling the roles of safety climate and safety culture: Multi-level effects on the relationship between supervisor enforcement and safety compliance.

Science.gov (United States)

Petitta, Laura; Probst, Tahira M; Barbaranelli, Claudio; Ghezzi, Valerio

2017-02-01

Despite increasing attention to contextual effects on the relationship between supervisor enforcement and employee safety compliance, no study has yet explored the conjoint influence exerted simultaneously by organizational safety climate and safety culture. The present study seeks to address this literature shortcoming. We first begin by briefly discussing the theoretical distinctions between safety climate and culture and the rationale for examining these together. Next, using survey data collected from 1342 employees in 32 Italian organizations, we found that employee-level supervisor enforcement, organizational-level safety climate, and autocratic, bureaucratic, and technocratic safety culture dimensions all predicted individual-level safety compliance behaviors. However, the cross-level moderating effect of safety climate was bounded by certain safety culture dimensions, such that safety climate moderated the supervisor enforcement-compliance relationship only under the clan-patronage culture dimension. Additionally, the autocratic and bureaucratic culture dimensions attenuated the relationship between supervisor enforcement and compliance. Finally, when testing the effects of technocratic safety culture and cooperative safety culture, neither safety culture nor climate moderated the relationship between supervisor enforcement and safety compliance. The results suggest a complex relationship between organizational safety culture and safety climate, indicating that organizations with particular safety cultures may be more likely to develop more (or less) positive safety climates. Moreover, employee safety compliance is a function of supervisor safety leadership, as well as the safety climate and safety culture dimensions prevalent within the organization. Copyright © 2016 Elsevier Ltd. All rights reserved.

Regulatory requirements for demonstration of the achieved safety level at the Mochovce NPP before commissioning

International Nuclear Information System (INIS)

Lipar, M.

1997-01-01

A review of regulatory requirements for demonstration of the achieved safety level at the Mochovce NPP before commissioning is given. It contains licensing steps in Slovakia during commissioning; Status and methodology of Mochovce safety analysis report; Mochovce NPP safety enhancement program; Regulatory body policy towards Mochovce NPP safety enhancement; Recent development in Mochovce pre-operational safety enhancement program review and assessment process; Licensing steps in Slovakia during commissioning

Reactor safety analysis

International Nuclear Information System (INIS)

Arien, B.

1998-01-01

Risk assessments of nuclear installations require accurate safety and reliability analyses to estimate the consequences of accidental events and their probability of occurrence. The objective of the work performed in this field at the Belgian Nuclear Research Centre SCK-CEN is to develop expertise in probabilistic and deterministic reactor safety analysis. The four main activities of the research project on reactor safety analysis are: (1) the development of software for the reliable analysis of large systems; (2) the development of an expert system for the aid to diagnosis; (3) the development and the application of a probabilistic reactor-dynamics method, and (4) to participate in the international PHEBUS-FP programme for severe accidents. Progress in research during 1997 is described

Rad waste disposal safety analysis / Integrated safety assessment of a waste repository

International Nuclear Information System (INIS)

Jeong, Jongtae; Choi, Jongwon; Kang, Chulhyung

2012-04-01

We developed CYPRUS+and adopted PID and RES method for the development of scenario. Safety performance assessment program was developed using GoldSim for the safety assessment of disposal system for the disposal of spnet fuels and wastes resulting from the pyrpoprocessing. Biosphere model was developed and verified in cooperation with JAEA. The capability to evaluate post-closure performance and safety was added to the previously developed program. And, nuclide migration and release to the biosphere considering site characteristics was evaluated by using deterministic and probabilistic approach. Operational safety assessment for drop, fire, and earthquake was also statistically evaluated considering well-established input parameter distribution. Conservative assessment showed that dose rate is below the limit value of low- and intermediate-level repository. Gas generation mechanism within engineered barrier was defined and its influence on safety was evaluated. We made probabilistic safety assessment by obtaining the probability distribution functions of important input variables and also made a sensitivity analysis. The maximum annual dose rate was shown to be below the safety limit value of 10 mSv/yr. The structure and element of safety case was developed to increase reliability of safety assessment methodology for a deep geological repository. Finally, milestone for safety case development and implementation strategy for each safety case element was also proposed

Analysis respons to the implementation of nuclear installations safety culture using AHP-TOPSIS

Science.gov (United States)

Situmorang, J.; Kuntoro, I.; Santoso, S.; Subekti, M.; Sunaryo, G. R.

2018-02-01

An analysis of responses to the implementation of nuclear installations safety culture has been done using AHP (Analitic Hierarchy Process) - TOPSIS (Technique for Order of Preference by Similarity to Ideal Solution). Safety culture is considered as collective commitments of the decision-making level, management level, and individual level. Thus each level will provide a subjective perspective as an alternative approach to implementation. Furthermore safety culture is considered by the statement of five characteristics which in more detail form consist of 37 attributes, and therefore can be expressed as multi-attribute state. Those characteristics and or attributes will be a criterion and its value is difficult to determine. Those criteria of course, will determine and strongly influence the implementation of the corresponding safety culture. To determine the pattern and magnitude of the influence is done by using a TOPSIS that is based on decision matrix approach and is composed of alternatives and criteria. The weight of each criterion is determined by AHP technique. The data used are data collected through questionnaires at the workshop on safety and health in 2015. .Reliability test of data gives Cronbah Alpha value of 95.5% which according to the criteria is stated reliable. Validity test using bivariate correlation analysis technique between each attribute give Pearson correlation for all attribute is significant at level 0,01. Using confirmatory factor analysis gives Kaise-Meyer-Olkin of sampling Adequacy (KMO) is 0.719 and it is greater than the acceptance criterion 0.5 as well as the 0.000 significance level much smaller than 0.05 and stated that further analysis could be performed. As a result of the analysis it is found that responses from the level of decision maker (second echelon) dominate the best order preference rank to be the best solution in strengthening the nuclear installation safety culture, except for the first characteristics, safety is a

Use of standard reliability levels in design and safety assessment of in-pile loops

International Nuclear Information System (INIS)

Bogani, G.; Verre, A.; Balestreri, S.; Colombo, A.G.; Luisi, T.

1975-01-01

This paper describes a logic-probabilistic analysis technique for a critical design review and safety assessment of in-pile loops. The examples in this paper refer to the analysis performed for the experimental loops already constructed or under construction in the ESSOR reactor of the Joint Research Centre of Ispra, as irradiation facilities for fuel element research and development tests. The proposed technique is based on the classification into categories of components and protective device malfunctions. Such subdivision into categories was agreed upon by the Italian Safety Authority and Euratom JRC, and adopted for the safety assessment of the ESSOR reactor in-pile loops. For each category, the method makes a link with a corresponding malfunction probability range (probability level). This probability level is defined taking into account design, construction, inspection and maintenance criteria as well as periodic controls; therefore the quality level and consequently the reliability level are thus also defined. The analysis is developed in the following stages: (1) definition of the analysis object (top event) and drawing of the relative fault-tree; (2) loop design analysis and preliminary optimization based on logic criteria; (3) classification into categories of the fault-tree primary events; (4) final loop design analysis and optimization based on defined component quality requirements. Stages 2 and 4 are quite different since stage 2 mainly consists of a redundance optimization, while stage 4 acts on the component quality level in such a way that each minimum cut-set leading to the top has an acceptable probability level. During analysis development, use is made of computer codes which, among other things enable the verification of fault-tree logic makeup, the listing of the minimum cut-sets with and without event categorization, and the evaluation of each cut-set order. (author)

Safety analysis and inventory control of transuranic and low-level waste in common storage

International Nuclear Information System (INIS)

Porten, D.R.; Bonner, A.L.; Joyce, J.P.

1993-01-01

This paper describes a methodology developed For the inventory control of low-level waste (LLW) and transuranic (TRU) waste, when both are stored in the same location, and both contribute to an inventory constrained by safety considerations. Development of the method arose from the necessity to make safety analysis calculations for the addition of LLW, in quantities greater than existing inventory limits would allow when stored with TRU waste, in the Hanford Central Waste Complex (CWC)-Ensuring that the dose consequences of credible releases are maintained at low-hazard limits or less, was used to allow greater than Type A quantities of LLW into the CWC. Basically, what happens is the original limited amount of TRU allowed is reduced by some equivalent amount of LLW introduced. The total quantity of TRU, and LLW in excess of Type A quantities, must be administratively maintained via curie equivalency Factors to ensure operation as a low-hazard Facility. The ''equivalency'' between TRU and LLW proposed here is specific only to the CWC, but the methodology can be used for other specific applications, such as TRU and LLW storage or handling facilities where inventory limits must be enforced or where a simplified inventory system is required

Screening-Level Safety Assessment of Personal Care Product Constituents Using Publicly Available Data

Directory of Open Access Journals (Sweden)

Ernest S. Fung

2018-06-01

Full Text Available Organizations recommend evaluating individual ingredients when assessing the safety of personal care or cosmetic products. The goal of this study was to present a screening-level safety assessment methodology to evaluate the safety of a product by identifying individual ingredients, determining their frequency of use in on-market products, and examining published safe-level-of-use information for each ingredient. As a case study, we evaluated WEN by Chaz Dean (WCD cleansing conditioners since there have been claims of adverse health effects associated with product use. We evaluated 30 ingredients in three on-market WCD cleansing conditioners. We then analyzed the National Library of Medicine’s Household Products Database and the Environmental Working Group’s (EWG Skin Deep Cosmetic Database, two of the largest publicly available databases, for other on-market personal care and cosmetic products that contained these ingredients. Safe-level-of-use information for each ingredient was obtained by reviewing peer-reviewed literature, the Food and Drug Administration’s (FDA generally recognized as safe (GRAS database, available Cosmetic Ingredient Review (CIR publications, and available product safety publications. The results of this analysis showed that more than 20,000 personal care and cosmetic products contained one or more of the evaluated ingredients used in WCD cleaning conditioners. Published safety information was available for 21 of the 30 evaluated ingredients: seven identified ingredients were designated as GRAS by the FDA and 16 ingredients had safe-level-of-use information available from the CIR. This study presents a screening-level safety assessment methodology that can serve as an initial screening tool to evaluate the safety of an ingredient intended for use in personal care and cosmetic products before a product is launched onto the market. This study provides evidence that the evaluated WCD cleansing conditioner ingredients

Periodic safety review of the HTR-10 safety analysis

International Nuclear Information System (INIS)

Chen Fubing; Zheng Yanhua; Shi Lei; Li Fu

2015-01-01

Designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is the first modular High Temperature Gas-cooled Reactor (HTGR) in China. According to the nuclear safety regulations of China, the periodic safety review (PSR) of the HTR-10 was initiated by INET after approved by the National Nuclear Safety Administration (NNSA) of China. Safety analysis of the HTR-10 is one of the key safety factors of the PSR. In this paper, the main contents in the review of safety analysis are summarized; meanwhile, the internal evaluation on the review results is presented by INET. (authors)

Safety analysis of nuclear power plants

International Nuclear Information System (INIS)

Selvatici, E.

1981-01-01

A study about the safety analysis of nuclear power plant, giving emphasis to how and why to do is presented. The utilization of the safety analysis aiming to perform the licensing requirements is discussed, and an example of the Angra 2 and 3 safety analysis is shown. Some presented tendency of the safety analysis are presented and examples are shown.(E.G.) [pt

Gap Analysis Approach for Construction Safety Program Improvement

Directory of Open Access Journals (Sweden)

Thanet Aksorn

2007-06-01

Full Text Available To improve construction site safety, emphasis has been placed on the implementation of safety programs. In order to successfully gain from safety programs, factors that affect their improvement need to be studied. Sixteen critical success factors of safety programs were identified from safety literature, and these were validated by safety experts. This study was undertaken by surveying 70 respondents from medium- and large-scale construction projects. It explored the importance and the actual status of critical success factors (CSFs. Gap analysis was used to examine the differences between the importance of these CSFs and their actual status. This study found that the most critical problems characterized by the largest gaps were management support, appropriate supervision, sufficient resource allocation, teamwork, and effective enforcement. Raising these priority factors to satisfactory levels would lead to successful safety programs, thereby minimizing accidents.

Preliminary Safety Design Report for Remote Handled Low-Level Waste Disposal Facility

Energy Technology Data Exchange (ETDEWEB)

Timothy Solack; Carol Mason

2012-03-01

A new onsite, remote-handled low-level waste disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled low-level waste disposal for remote-handled low-level waste from the Idaho National Laboratory and for nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled low-level waste in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This preliminary safety design report supports the design of a proposed onsite remote-handled low-level waste disposal facility by providing an initial nuclear facility hazard categorization, by discussing site characteristics that impact accident analysis, by providing the facility and process information necessary to support the hazard analysis, by identifying and evaluating potential hazards for processes associated with onsite handling and disposal of remote-handled low-level waste, and by discussing the need for safety features that will become part of the facility design.

10 CFR 70.62 - Safety program and integrated safety analysis.

Science.gov (United States)

2010-01-01

...; (iv) Potential accident sequences caused by process deviations or other events internal to the... have experience in nuclear criticality safety, radiation safety, fire safety, and chemical process... this safety program; namely, process safety information, integrated safety analysis, and management...

Temperature and level measurements realized for Nuclear Safety Level Improvement of Slovak NPPs

International Nuclear Information System (INIS)

Badiar, S.; Slanina, M.; Stanc, S.; Golan, P.; Krupa, J.

2001-01-01

Process of continual safety improvement in the individual Slovak nuclear power plants has been in progress since the beginning of nineties with the objective to upgrade the safety level of units in operation up to the European standards. In the framework of these activities, safety instrumentation systems with 1E qualification for the control of WWER reactor coolant systems were built and added. Methods for implementation of safety instrumentation systems for monitoring temperature and level in reactor coolant systems in the particular plants in Slovakia are presented showing the objectives and methods of their implementation. (Authors)

Safety analysis of the VLJ repository

International Nuclear Information System (INIS)

Vieno, T.; Nordman, H.

1991-05-01

The VLJ repository is an underground disposal facility for the low and medium level waste generated at the Olkiluoto nuclear power plant. The repository is located within 1 km from TVO I and TVO II (2 x 710 MWe) BWR's on the Olkiluoto island at the west coast of Finland. It contains two rock silos excavated at the depth of 60...100 meters in the bedrock. Low level waste will be disposed of in a shotcreted rock silo. For bituminized medium level waste, a separate silo of reinforced concrete has been built inside the shotcreted rock silo. The post-closure safety analysis has been done for the Final Safety Analysis Report (FSAR) of the VLJ repository. In addition to the normal evolution scenario, several disturbed evolution and accident scenarios have been analysed. In the reference scenario, radio-nuclides are assumed to be released from the bituminized waste within 500 years, the concrete silo is assumed to gradually disintegrate and finally to collapse at 5 000 years, all concrete in the silo is assumed to be also chemically depleted within 6 000 years, and all the seals of the repository are assumed to deteriorate within 12 000 years. The ability of alone natural barriers to restrict the release of radionuclides into the biosphere has been evaluated by means of scenarios where the degradation of engineered barriers has been assumed to take place at a still faster rate. In one of the disturbed evolution scenarios it has been assumed that the concrete silo for medium level waste is severely impaired immediately after sealing of the repository. Effects of gas generation and consequences of human intrusion have been evaluated, too. The results of the safety analysis show that radiation doses of any significance are caused only if a well is bored in the vicinity of the repository or if the groundwater discharge spot is inhabited and used for cultivation. In the reference scenario the maximum expectation value of the individual dose rate is 0.3 mSv/a

Occupational Health and Safety. Numeracy. Level 1. Level 2. Level 3. Support Materials for Agricultural Training.

Science.gov (United States)

Batman, Kangan; Tully, Chris

This publication contains the three numeracy units of the three levels of Support Materials for Agricultural Training (SMAT) in the area of occupational health and safety: Level 1 (starting), 2 (continuing), and 3 (completing). The units are designed to help the learner improve his or her numeracy skills needed to deal with occupational safety and…

Reactor Safety Analysis

International Nuclear Information System (INIS)

Arien, B.

1998-01-01

The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

PWR core safety analysis with 3-dimensional methods

International Nuclear Information System (INIS)

Gensler, A.; Kühnel, K.; Kuch, S.

2015-01-01

Highlights: • An overview of AREVA’s safety analysis codes their coupling is provided. • The validation base and licensing applications of these codes are summarized. • Coupled codes and methods provide improved margins and non-conservative results. • Examples for REA and inadvertent opening of the pressurizer safety valve are given. - Abstract: The main focus of safety analysis is to demonstrate the required safety level of the reactor core. Because of the demanding requirements, the quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools, an extensive validation base, and last but not least highly educated engineers applying the methodology. The sophisticated 3-dimensional core models applied by AREVA ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. Presently AREVA employs SCIENCE, CASMO/NEMO and CASCADE-3D for pressurized water reactors. These codes are currently being consolidated into the next generation 3D code system ARCADIA®. AREVA continuously extends the validation base, including measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models provide reliable and comprehensive results for a wide range of applications. For the application of these powerful tools, AREVA is taking benefit of its interdisciplinary know-how and international teamwork. Experienced engineers of different technical backgrounds are working together to ensure an appropriate interpretation of the calculation results, uncertainty analysis, along with continuously maintaining and enhancing the quality of the analysis methodologies. In this paper, an overview of AREVA’s broad application experience as well as the broad validation

Safety Analysis Report for Ignalina NPP

International Nuclear Information System (INIS)

Negrivoda, G.

1997-01-01

In December 1994 an agreement was signed between the European Bank for Reconstruction and Development and the Republic of Lithuania for the grant of 32.86 MECU for the safety Improvement at Ignalina NPP. One of the conditions for the provision of the grant, was a requirement for an in-depth analysis of the safety level at Ignalina NPP in the scope and according to the standards acceptable for a western nuclear power plant, and to publish a Safety Analysis Report (SAR). The report should investigate and analyze any factor that could limit a safe operation of the plant, and provide recommendations for actual safety improvements. According to the agreement, Lithuania had to finalize the SAR until 31 December, 1995. The bank has also organized and financed investigation of safety at Ignalina NPP and preparation of the SAR. EBRD made an agreement with Sweden's Vattenfall, which subcontracted well-known companies from Canada, USA, Germany, etc., and also the Russian Research and Development Institute of Power Engineering (NIKIET), reactor designer of Ignalina NPP. The SAR is a very comprehensive document and contains about 8000 pages of text, diagrams and tables. The main findings of the SAR are provided in the article. A large number of discrepancies with modern rules and western practices was detected, but they were not proved to be serious enough to require reactors shutdown. Based on the recommendations of the SAR Ignalina NPP has worked out Safety Improvement Program No. 2 (SIP-2), which is planned for three years and will cost 486 MLT. (author)

Analysis of Correlations between the Level of Partnering Relations and their Influence on the Time, Cost, Quality and Safety of Implementation of Construction Projects

Directory of Open Access Journals (Sweden)

Radziszewska-Zielina Elżbieta

2014-11-01

Full Text Available The present paper uses the developed model of the influence of partnering relations on the time, cost, quality and safety of implementation of construction projects. On its basis, a questionnaire has been created and a preliminary survey has been conducted. The paper presents an analysis of correlations between the level of partnering relations in the context of the partnering measures indicated in the model and their influence on the time, cost, quality and safety of implementation of construction projects. The analysis was conducted based on the data collected in 52 construction projects. The values of the Spearman rank correlation coefficient and the Pearson product-moment correlation coefficient have been calculated for the examined relations. The analysis allowed for indicating the measures of partnering whose improvement most often brings benefits with regard to the time, cost, quality and safety of implementation of construction projects. Among the 80 analysed correlations, the ones identified as strong were: 15 relations connected with the time, 8 with the cost, 5 with the quality and 1 with the safety of implementation of construction projects.

Trending of low level events and near misses to enhance safety performance in nuclear power plants

International Nuclear Information System (INIS)

2005-11-01

The IAEA Safety Fundamentals publication, Safety of Nuclear Installations, Safety Series No. 110, states the need for operating organizations to establish a programme for the collection and analysis of operating experience in nuclear power plants. Such a programme ensures that operating experience is analysed, events important to safety are reviewed in depth, and lessons learned are disseminated to the staff of the organization and to relevant national and international organizations. As a result of the effort to enhance safety in operating organizations, incidents are progressively decreasing in number and significance. This means that in accordance with international reporting requirements the amount of collected data becomes less sufficient to draw meaningful statistical conclusions. This is where the collection and trend analysis of low level events and near misses can prove to be very useful. These trends can show which of the safety barriers are weak or failing more frequently. Evaluation and trending of low level events and near misses will help to prevent major incidents because latent weaknesses have been identified and corrective actions taken to prevent recurrence. This leads to improved safety and production. Low level events and near misses, which may reach several thousand per reactor operating year, need to be treated by the organizations as learning opportunities. A system for capturing these low level events and near misses truly needs to be an organization-wide system in which all levels of the organization, including contractors, participate. It is desirable that the overall operational experience feedback (OEF) process should integrate the lessons learned and the associated data from significant events with those of lower level events and near misses. To be able to effectively implement a process dealing with low level events and near misses, it is necessary that the organization have a well established OEF process for significant events

Confidence improvement of disosal safety bydevelopement of a safety case for high-level radioactive waste disposal

Energy Technology Data Exchange (ETDEWEB)

Baik, Min Hoon; Ko, Nak Youl; Jeong, Jong Tae; Kim, Kyung Su [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-12-15

Many countries have developed a safety case suitable to their own countries in order to improve the confidence of disposal safety in deep geological disposal of high-level radioactive waste as well as to develop a disposal program and obtain its license. This study introduces and summarizes the meaning, necessity, and development process of the safety case for radioactive waste disposal. The disposal safety is also discussed in various aspects of the safety case. In addition, the status of safety case development in the foreign countries is briefly introduced for Switzerland, Japan, the United States of America, Sweden, and Finland. The strategy for the safety case development that is being developed by KAERI is also briefly introduced. Based on the safety case, we analyze the efforts necessary to improve confidence in disposal safety for high-level radioactive waste. Considering domestic situations, we propose and discuss some implementing methods for the improvement of disposal safety, such as construction of a reliable information database, understanding of processes related to safety, reduction of uncertainties in safety assessment, communication with stakeholders, and ensuring justice and transparency. This study will contribute to the understanding of the safety case for deep geological disposal and to improving confidence in disposal safety through the development of the safety case in Korea for the disposal of high-level radioactive waste.

Safety analysis of the transportation of high-level radioactive waste

International Nuclear Information System (INIS)

Murphy, E.S.; Winegardner, W.K.

1975-01-01

An analysis of the risk from transportation of solidified high-level waste is being performed at Battelle-Northwest as part of a comprehensive study of the management of high-level waste. The risk analysis study makes use of fault trees to identify failure events and to specify combinations of events which could result in breach of containment and a release of radioactive material to the environment. Contributions to risk analysis methodology which have been made in connection with this study include procedures for identification of dominant failure sequences, methods for quantifying the effects of probabilistic failure events, and computer code development. Preliminary analysis based on evaluation of the rail transportation fault tree indicates that the dominant failure sequences for transportation of solidified high-level waste will be those related to railroad accidents. Detailed evaluation of rail accident failure sequences is proceeding and is making use of the limited frequency-severity data which is available in the literature. (U.S.)

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

Using level-I PRA for enhanced safety of the advanced neutron source reactor

International Nuclear Information System (INIS)

Ramsey, C.T.; Linn, M.A.

1995-01-01

The phase-1, level-I probabilistic risk assessment (PRA) of the Advanced Neutron Source (ANS) reactor has been completed as part of the conceptual design phase of this proposed research facility. Since project inception, PRA and reliability concepts have been an integral part of the design evolutions contributing to many of the safety features in the current design. The level-I PRA has been used to evaluate the internal events core damage frequency against project goals and to identify systems important to safety and availability, and it will continue to guide and provide support to accident analysis, both severe and nonsevere. The results also reflect the risk value of defense-in-depth safety features in reducing the likelihood of core damage

Standard model for the safety analysis report of nuclear fuel reprocessing plants

International Nuclear Information System (INIS)

1980-02-01

This norm establishes the Standard Model for the Safety Analysis Report of Nuclear Fuel Reprocessing Plants, comprehending the presentation format, the detailing level of the minimum information required by the CNEN for evaluation the requests of Construction License or Operation Authorization, in accordance with the legislation in force. This regulation applies to the following basic reports: Preliminary Safety Analysis Report - PSAR, integrating part of the requirement of Construction License; and Final Safety Analysis Report (FSAR) which is the integrating part of the requirement for Operation Authorization

Ignalina Safety Analysis Group

International Nuclear Information System (INIS)

Ushpuras, E.

1995-01-01

The article describes the fields of activities of Ignalina NPP Safety Analysis Group (ISAG) in the Lithuanian Energy Institute and overview the main achievements gained since the group establishment in 1992. The group is working under the following guidelines: in-depth analysis of the fundamental physical processes of RBMK-1500 reactors; collection, systematization and verification of the design and operational data; simulation and analysis of potential accident consequences; analysis of thermohydraulic and neutronic characteristics of the plant; provision of technical and scientific consultations to VATESI, Governmental authorities, and also international institutions, participating in various projects aiming at Ignalina NPP safety enhancement. The ISAG is performing broad scientific co-operation programs with both Eastern and Western scientific groups, supplying engineering assistance for Ignalina NPP. ISAG is also participating in the joint Lithuanian - Swedish - Russian project - Barselina, the first Probabilistic Safety Assessment (PSA) study of Ignalina NPP. The work is underway together with Maryland University (USA) for assessment of the accident confinement system for a range of breaks in the primary circuit. At present the ISAG personnel is also involved in the project under the grant from the Nuclear Safety Account, administered by the European Bank for reconstruction and development for the preparation and review of an in-depth safety assessment of the Ignalina plant

Determining supply chain safety stock level and location

Directory of Open Access Journals (Sweden)

Bahareh Amirjabbari

2014-01-01

Full Text Available Purpose: The lean methodology and its principles have widely been applied in supply chain management in recent decades. Manufacturers are one of the most important contributors in a supply chain and inventory plays a paramount role for them to become lean. Therefore, there should be appropriate management of inventory and all of its drivers in accordance with a lean strategy. Safety stock is one of the main drivers of inventory; it protects against increasing the stretch in the breaking points of the supply chain, which in turn can result in possible reduction of inventory. In this paper an optimization model and a simulation model are developed and applied in a real case to optimize the safety stock level with the objective of logistics cost minimization.Design/methodology/approach: In order to optimize the safety stock level while minimizing logistics costs, a nonlinear cost minimization safety stock model is developed in this paper and then it is applied in a real world manufacturing case company. A safety stock simulation model based on appropriate metrics in the case company’s supply chain performance is also provided.Findings: These models result in not only the optimum levels but also locations of safety stock within the supply chain.Originality/value: In this research, two models of cost minimization and simulation have been developed and also applied in a real case company to result in not only optimized levels but also optimized locations of safety stock across the whole supply chain. In addition, the appropriate supply chain performance measurement metrics have been introduced in this paper and the simulation model is developed based on those.

Preliminary safety analysis methodology for the SMART

Energy Technology Data Exchange (ETDEWEB)

Bae, Kyoo Hwan; Chung, Y. J.; Kim, H. C.; Sim, S. K.; Lee, W. J.; Chung, B. D.; Song, J. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-03-01

This technical report was prepared for a preliminary safety analysis methodology of the 330MWt SMART (System-integrated Modular Advanced ReacTor) which has been developed by Korea Atomic Energy Research Institute (KAERI) and funded by the Ministry of Science and Technology (MOST) since July 1996. This preliminary safety analysis methodology has been used to identify an envelope for the safety of the SMART conceptual design. As the SMART design evolves, further validated final safety analysis methodology will be developed. Current licensing safety analysis methodology of the Westinghouse and KSNPP PWRs operating and under development in Korea as well as the Russian licensing safety analysis methodology for the integral reactors have been reviewed and compared to develop the preliminary SMART safety analysis methodology. SMART design characteristics and safety systems have been reviewed against licensing practices of the PWRs operating or KNGR (Korean Next Generation Reactor) under construction in Korea. Detailed safety analysis methodology has been developed for the potential SMART limiting events of main steam line break, main feedwater pipe break, loss of reactor coolant flow, CEA withdrawal, primary to secondary pipe break and the small break loss of coolant accident. SMART preliminary safety analysis methodology will be further developed and validated in parallel with the safety analysis codes as the SMART design further evolves. Validated safety analysis methodology will be submitted to MOST as a Topical Report for a review of the SMART licensing safety analysis methodology. Thus, it is recommended for the nuclear regulatory authority to establish regulatory guides and criteria for the integral reactor. 22 refs., 18 figs., 16 tabs. (Author)

Transient analysis for resolving safety issues

International Nuclear Information System (INIS)

Chao, J.; Layman, W.

1987-01-01

The Nuclear Safety Analysis Center (NSAC) has a Generic Safety Analysis Program to help resolve high priority generic safety issues. This paper describes several high priority safety issues considered at NSAC and how they were resolved by transient analysis using thermal hydraulics and neutronics codes. These issues are pressurized thermal shock (PTS), anticipated transients without scram (ATWS), steam generator tube rupture (SGTR), and reactivity transients in light of the Chernobyl accident

Results of the Safety probabilistic analysis of Level 2 of the CNSNS

International Nuclear Information System (INIS)

Lopez M, R.; Godinez S, V.

2004-01-01

The National Commission of Nuclear Safety and Safeguards (CNSNS) it has concluded the one develop of their Probabilistic Analysis of Safety (APS) of Level 2. The reach of the study it considers internal events to full power and it was developed on the base of the methodology of the NUREG-1150, for what you it was built an Event Tree of the Progression of the Accident (APET) to analyze the 25 States of Damage to the Plant (PDS) obtained of the APS Nl of the CNSNS. In the APET are considered the phenomenology of severe accidents, the performance of mitigation systems and actions of the operator that could modify the evolution of a severe accident in the CNLV, as well as the diverse modes of failure of the primary container and it identifies the trajectories of liberation of radioactive material to the exterior. The conditional probabilities of failure of the primary container were obtained and it was characterized the time so much to which happens the liberation of radioactive material as the quantity of the term liberated source. Also, to establish the times and parameters of the evolution of accidents were selected representative accident sequences of the diverse accident types and their conditions were simulated by means of the MELCOR computer code. Also it was developed a code of parametric compute type XSOR, specific for Laguna Verde, with which it was carried out the estimate of the term source in each one of the release trajectories. In this work the main characteristic ones are presented and results of the APS N2 developed in the CNSNS and they are compared against the model and results of the EIP of the CNLV. (Author)

Derivation methods for clearance levels and safety assessments for very low-level radioactive waste disposal

International Nuclear Information System (INIS)

Okoshi, Minoru

2001-01-01

The clearance level was evaluated by the dose of concrete and metal when they would be recycled and reused from shallow land burial of radioactive facilities. The state of waste after clearance is not specified, so that we studied large scale of exposure pathways. The parameter values used for safety assessment were determined as the average values under the consideration of natural and social environment in Japan. Propriety of these values was confirmed by a probability analysis. On the safety assessment of very low-level waste disposal facility, the disposer pathway and parameters were determined under the consideration of special site conditions (natural and social environment) and properties of waste. However, the same exposure pathway of them used the same model for external (exposure by sky shine' s ray) and internal exposure. The calculation results of estimated pathway showed 1.2x10 -5 mSv/y the largest dose for the external exposure pathway by sky shine's ray. (S.Y.)

Atlantic Richfield Hanford Company californium multiplier/delayed neutron counter safety analysis

International Nuclear Information System (INIS)

Zimmer, W.H.

1976-08-01

The Californium Multiplier (CFX) is a subcritical assembly of uranium surrounding 252 Cf spontaneously fissioning neutron sources; its function is to multiply the neutron flux to a level useful for activation analysis. This document summarizes the safety analysis aspects of the CFX, DNC, pneumatic transfer system, and instrumentation and to detail all the aspects of the total facility as a starting point for the ARHCO Safety Analysis Review. Recognized hazards and steps already taken to neutralize them are itemized

Safety evaluation review of the prototype license application safety analysis report

International Nuclear Information System (INIS)

1991-08-01

The US Nuclear Regulatory Commission (NRC) staff and consultants reviewed a Prototype License Application Safety Analysis Report (PLASAR) submitted by the US Department of Energy (DOE) for the belowground vault (BGV) alternative method of low-level radioactive waste disposal. In Volume 1 of NUREG-1375, the NRC staff provided the safety review results for an earth-mounded concrete bunker PLASAR. In the current report, the staff focused its review on the design, construction, and operational aspects of the BGV PLASAR. The staff developed review comments and questions using the Standard Review Plan (SRP), Rev. 1 (NUREG-1200) as the basis for evaluating the acceptability of the information provided in the BGV PLASAR. The detailed review comments provided in this report are intended to be useful guidance to facility developers and State regulators in addressing issues likely to be encountered in the review of a license application for a low-level-waste disposal facility. 44 refs

Disposal of flow-level radioactive waste in Belgium: A safety analysis for inorganic chemotoxic elements

International Nuclear Information System (INIS)

Mallants, D.; Volckaert, G.; Marivoet, J.; Neerdael, B.

2000-01-01

Low-level radioactive waste often contains large quantities of inorganic chemical substances. Due attention should therefore be given to the safety implications of both the radiological and chemical substances in the waste. Our study develops the safety assessment methodology for surface disposal with emphasis on the potential effects of inorganic nonradiological elements on human health. Contamination of groundwater was considered as the major exposure pathway. The applied methodology first screens all elements on the basis of five criteria. Conservative screening calculations were used to screen out the elements that do not pose danger to humans, and to select those that could have a negative impact and thus require further analysis. The latter was done by first calculating the elemental mass fluxes out of the repository and into the aquifer followed by the calculation of groundwater concentrations. The results showed that on the basis of the screening calculations, 75% of all elements could be classified as non-hazardous. The detailed calculations showed that the majority of the remaining elements had groundwater concentrations below the drinking water or groundwater standards. The results further showed that for a few elements the maximum groundwater concentration was above the standard, but below the background concentrations. (author)

A study of software safety analysis system for safety-critical software

International Nuclear Information System (INIS)

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

2004-01-01

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

Project Guarantee 1985. Final repository for high-level radioactive wastes: The system of safety barriers

International Nuclear Information System (INIS)

Anon.

1985-01-01

Final disposal of radioactive waste involves preventing the waste from returning from the repository location into the biosphere by means of successively arranged containment measures known as safety barriers. In the present volume NGB 85-04 of the series of reports for Project 'Guarantee' 1985, the safety barrier system for the type C repository for high-level waste is described. The barrier parameters which are relevant for safety analysis are quantified and associated error limits and data scatter are given. The aim of the report is to give a summary documentation of the safety analysis input data and their scientific background. For secure containment of radioactive waste safety barriers are used which effectively limit the release of radioactive material from the repository (release barriers) and effectively retard the entry of the original radioactive material into the biosphere (time barriers). Safety barriers take the form of both technically constructed containment measures and the siting of the repository in suitable geological formations. The technical safety barrier system in the case of high-level waste comprises: the waste solidification matrix (borosilicate glass), massive steel canisters, encasement of the waste canisters, encasement of the waste canisters in highly compacted bentonite, sealing of vacant storage space and access routes on repository closure. The natural geological safety barriers - the host rock and overlying formations provide sufficiently long deep groundwater flow times from the repository location to the earth's surface and for additional lengthening of radionuclide migration times by means of various chemical and physical retardation mechanisms. The stability of the geological formations is so great that hydrogeological system is protected for a sufficient length of time from deterioration caused, in particular, by erosion. Observations in the final section of the report indicate that input data for the type C repository safety

Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji; Brayon, Fedrick Charlie Matthew; Mohamed, Faizal

2014-01-01

As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia

Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Mazleha Maskin; Phongsakorn, P.T.; Tonny, A.L.; Fedrick, C.M.B.; Faizal Mohamed; Mohamad Fauzi Saad; Ahmad Razali Ismail; Mohamad Puad Haji Abu

2013-01-01

Full-text: As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia. (author)

Safety analysis of autonomous excavator functionality

International Nuclear Information System (INIS)

Seward, D.; Pace, C.; Morrey, R.; Sommerville, I.

2000-01-01

This paper presents an account of carrying out a hazard analysis to define the safety requirements for an autonomous robotic excavator. The work is also relevant to the growing generic class of heavy automated mobile machinery. An overview of the excavator design is provided and the concept of a safety manager is introduced. The safety manager is an autonomous module responsible for all aspects of system operational safety, and is central to the control system's architecture. Each stage of the hazard analysis is described, i.e. system model creation, hazard definition and hazard analysis. Analysis at an early stage of the design process, and on a system that interfaces directly to an unstructured environment, exposes certain issues relevant to the application of current hazard analysis methods. The approach taken in the analysis is described. Finally, it is explained how the results of the hazard analysis have influenced system design, in particular, safety manager specifications. Conclusions are then drawn about the applicability of hazard analysis of requirements in general, and suggestions are made as to how the approach can be taken further

Solid waste burial grounds interim safety analysis

International Nuclear Information System (INIS)

Saito, G.H.

1994-01-01

This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment

Solid waste burial grounds interim safety analysis

Energy Technology Data Exchange (ETDEWEB)

Saito, G.H.

1994-10-01

This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment.

Safety analysis of the proposed Canadian geologic nuclear waste repository

International Nuclear Information System (INIS)

Prowse, D.R.

1977-01-01

The Canadian program for development and qualification of a geologic repository for emplacement of high-level and long-lived, alpha-emitting waste from irradiated nuclear fuel has been inititiated and is in its initial development stage. Fieldwork programs to locate candidate sites with suitable geological characteristics have begun. Laboratory studies and development of models for use in safety analysis of the emplaced nuclear waste have been initiated. The immediate objective is to complete a simplified safety analysis of a model geologic repository by mid-1978. This analysis will be progressively updated and will form part of an environmental Assessment Report of a Model Fuel Center which will be issued in mid-1979. The long-term objectives are to develop advanced safety assessment models of a geologic repository which will be available by 1980

Safety analysis and synthesis using fuzzy sets and evidential reasoning

International Nuclear Information System (INIS)

Wang, J.; Yang, J.B.; Sen, P.

1995-01-01

This paper presents a new methodology for safety analysis and synthesis of a complex engineering system with a structure that is capable of being decomposed into a hierarchy of levels. In this methodology, fuzzy set theory is used to describe each failure event and an evidential reasoning approach is then employed to synthesise the information thus produced to assess the safety of the whole system. Three basic parameters--failure likelihood, consequence severity and failure consequence probability, are used to analyse a failure event. These three parameters are described by linguistic variables which are characterised by a membership function to the defined categories. As safety can also be clearly described by linguistic variables referred to as the safety expressions, the obtained fuzzy safety score can be mapped back to the safety expressions which are characterised by membership functions over the same categories. This mapping results in the identification of the safety of each failure event in terms of the degree to which the fuzzy safety score belongs to each of the safety expressions. Such degrees represent the uncertainty in safety evaluations and can be synthesised using an evidential reasoning approach so that the safety of the whole system can be evaluated in terms of these safety expressions. Finally, a practical engineering example is presented to demonstrate the proposed safety analysis and synthesis methodology

Safety analysis procedures for PHWR

International Nuclear Information System (INIS)

Min, Byung Joo; Kim, Hyoung Tae; Yoo, Kun Joong

2004-03-01

The methodology of safety analyses for CANDU reactors in Canada, a vendor country, uses a combination of best-estimate physical models and conservative input parameters so as to minimize the uncertainty of the plant behavior predictions. As using the conservative input parameters, the results of the safety analyses are assured the regulatory requirements such as the public dose, the integrity of fuel and fuel channel, the integrity of containment and reactor structures, etc. However, there is not the comprehensive and systematic procedures for safety analyses for CANDU reactors in Korea. In this regard, the development of the safety analyses procedures for CANDU reactors is being conducted not only to establish the safety analyses system, but also to enhance the quality assurance of the safety assessment. In the first phase of this study, the general procedures of the deterministic safety analyses are developed. The general safety procedures are covered the specification of the initial event, selection of the methodology and accident sequences, computer codes, safety analysis procedures, verification of errors and uncertainties, etc. Finally, These general procedures of the safety analyses are applied to the Large Break Loss Of Coolant Accident (LBLOCA) in Final Safety Analysis Report (FSAR) for Wolsong units 2, 3, 4

Safety evaluation status report for the prototype license application safety analysis report

International Nuclear Information System (INIS)

1989-07-01

The US Nuclear Regulatory Commission (NRC) staff and consultants reviewed a Prototype License Application Safety Analysis Report (PLASAR) submitted by the US Department of Energy (DOE) for the earth-mounded concrete bunker (EMCB) alternative method of low-level radioactive waste disposal. The NRC reviewers relied extensively on the Standard Review Plan (SRP), Rev.1 (NUREG-1200), to evaluate the acceptability of the information provided in the EMCB PLASAR. The NRC staff selected certain review areas in the PLASAR for development of safety evaluation report input to provide examples of safety assessments that are necessary as part of a licensing review. Because of the fictitious nature of the assumed disposal site, and the decision to limit the review to essentially first-round review status, the NRC staff report is labeled a ''Safety Evaluation Status Report'' (SESR). Appendix A comprises the NRC review comments and questions on the information that DOE submitted in the PLASAR. The NRC concentrated its review on the design and operations-related portions of the EMCB PLASAR

Main factors determining the KNP units 5 and 6 safety level according to the PSA level 1 result

International Nuclear Information System (INIS)

Manchev, B.; Marinova, B.; Nenkova, B.

2004-01-01

The Probabilistic Safety Analysis (PSA) is a powerful tool for ascertainment of the safety level reached at nuclear power plants operation. The results of PSA determine very clearly the functions, systems, equipment or operator actions that have to be improved in order to increase the plant safety level as a whole. The present report presents the main results of the last upgraded revision of PSA level 1 of units 5 and 6 of KNPP. The objective of the report is to lay emphasis on the factors determining the result obtained, i.e. to demonstrate the scopes whose improvement leads to an increase of the safety level reached at the units power operation. In the frame of the study presented the following categories of initiating events are included: Internal initiating events; Initiating events result of internal fires; Initiating events result of seismic action; Floods. Only the reactor core is considered as a source of radioactive contamination. Only initiating events related to the reactor work on power are analyzed. Unit 5 of KNPP is accepted as a basic unit for the study. All modifications and design changes implemented up to year 2000 are taken into account. The results of PSA level 1 for units 5 and 6 of KNPP covering the risk of internal initiators are presented. The assessment of the core damage due to internal initiators is based on the analysis of 18 groups of initiating events. 932 consequences and two groups of initial events are identified, leading to core damage. As a result of the quantitative calculation, over 15000 minimal cuts for the core damage are obtained. The first 80 cuts bear over 75% of the frequency obtained, and the first 700 cuts bear over 90%. Distribution of the core damage frequency by different groups of initiators is presented in tables and diagrams. A comparison of the result obtained for the reactor core damage of KNPP units 5 and 6 with assessment obtained for similar power plants is presented. The data for different NPPs are taken

Methodology of safety assessment and sensitivity analysis for geologic disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Kimura, Hideo; Takahashi, Tomoyuki; Shima, Shigeki; Matsuzuru, Hideo

1995-01-01

A deterministic safety assessment methodology has been developed to evaluate long-term radiological consequences associated with geologic disposal of high-level radioactive waste, and to demonstrate a generic feasibility of geologic disposal. An exposure scenario considered here is based on a normal evolution scenario which excludes events attributable to probabilistic alterations in the environment. A computer code system GSRW thus developed is based on a non site-specific model, and consists of a set of sub-modules for calculating the release of radionuclides from engineered barriers, the transport of radionuclides in and through the geosphere, the behavior of radionuclides in the biosphere, and radiation exposures of the public. In order to identify the important parameters of the assessment models, an automated procedure for sensitivity analysis based on the Differential Algebra method has been developed to apply to the GSRW. (author)

Operator reliability study for Probabilistic Safety Analysis of an operating research reactor

International Nuclear Information System (INIS)

Mohamed, F.; Hassan, A.; Yahaya, R.; Rahman, I.; Maskin, M.; Praktom, P.; Charlie, F.

2015-01-01

Highlights: • Human Reliability Analysis (HRA) for Level 1 Probabilistic Safety Analysis (PSA) is performed on research nuclear reactor. • Implemented qualitative HRA framework is addressed. • Human Failure Events of significant impact to the reactor safety are derived. - Abstract: A Level 1 Probabilistic Safety Analysis (PSA) for the TRIGA Mark II research reactor of Malaysian Nuclear Agency has been developed to evaluate the potential risk in its operation. In conjunction to this PSA development, Human Reliability Analysis (HRA) is performed in order to determine human contribution to the risk. The aim of this study is to qualitatively analyze human actions (HAs) involved in the operation of this reactor according to the qualitative part of the HRA framework for PSA which is namely the identification, qualitative screening and modeling of HAs. By performing this framework, Human Failure Events (HFEs) of significant impact to the reactor safety are systematically analyzed and incorporated into the PSA structure. A part of the findings in this study will become the input for the subsequent quantitative part of the HRA framework, i.e. the Human Error Probability (HEP) quantification

Standard format and content of a license application for a low-level radioactive waste disposal facility: Safety analysis report

International Nuclear Information System (INIS)

1988-01-01

This document discusses the information that should be provided in the Safety Analysis Report and establishes a uniform format for presenting the information necessary to fulfill the licensing requirements for land disposal of radioactive waste called for in 10 CFR 61. The uniform format will (1) help ensure that the Safety Analysis Report contains the information required by 10 CFR 61, (2) aid the applicant and NRC staff in ensuring that the information is complete, (3) help persons reading the Safety Analysis Report to locate information, and (4) contribute to shortening the time needed for the review process

Standard format and content of a license application for a low-level radioactive waste disposal facility: Safety analysis report

International Nuclear Information System (INIS)

1987-01-01

This document discusses the information that should be provided in the Safety Analysis Report and establishes a uniform format for presenting the information necessary to fulfill the licensing requirements for land disposal of radioactive waste called for in 10 CFR 61. The uniform format will (1) help ensure that the Safety Analysis Report contains the information required by 10 CFR 61, (2) aid the applicant and NRC staff in ensuring that the information is complete, (3) help persons reading the Safety Analysis Report to locate information, and (4) contribute to shortening the time needed for the review process

Reliability Analysis and Calibration of Partial Safety Factors for Redundant Structures

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard

1998-01-01

Redundancy is important to include in the design and analysis of structural systems. In most codes of practice redundancy is not directly taken into account. In the paper various definitions of a deterministic and reliability based redundancy measure are reviewed. It is described how reundancy can...... be included in the safety system and how partial safety factors can be calibrated. An example is presented illustrating how redundancy is taken into account in the safety system in e.g. the Danish codes. The example shows how partial safety factors can be calibrated to comply with the safety level...

АSSESSMENT AND FORECASTING OF FLIGHT SAFETY LEVEL OF AIRLINE

Directory of Open Access Journals (Sweden)

E. S. Prozorov

2015-01-01

Full Text Available The article presents methods based on probability theory and mathematical statistics for solving a number of basic problems: formation and evaluation of the current flight safety level; forecasting the level of flight safety; ranking the objects (planes, pilots in terms of flight safety; evaluation of the presence (or absence of control actions arising in the context of the organization of corporate safety management system. At the same time as the main source of information are considered forward-looking events received from flight data.

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.

Application of a structural model for advanced analysis in the evaluation of nuclear safety

International Nuclear Information System (INIS)

Landesmann, Alexandre; Barros, Francisco Claudio Pereira de; Batista, Eduardo de Miranda

2003-01-01

The Advanced Analysis concept, which means the direct consideration of both physical and geometric nonlinear effects in the analysis and design of steel buildings structures, represents the state-of-art in the field of structural analysis by this beginning of the 21 st century. In this context, the present paper presents an Advanced Analysis methodology applied to the Safety Evaluation of high hazardous civil structures. This Safety Evaluation plays an important part in the regulators position as a step in the licensing process performed by CNEN - Brazilian Nuclear Energy Commission. The proposed Advance Analysis procedure is implemented by a refined second-order plastic hinge model. The application of this model allows to carry out: the description of the inelastic structural behavior; the identification of the collapse mechanism; the ultimate load level; structural safety's level and the service ability limit. (author)

An intelligent hybrid system for surface coal mine safety analysis

Energy Technology Data Exchange (ETDEWEB)

Lilic, N.; Obradovic, I.; Cvjetic, A. [University of Belgrade, Belgrade (Serbia)

2010-06-15

Analysis of safety in surface coal mines represents a very complex process. Published studies on mine safety analysis are usually based on research related to accidents statistics and hazard identification with risk assessment within the mining industry. Discussion in this paper is focused on the application of AI methods in the analysis of safety in mining environment. Complexity of the subject matter requires a high level of expert knowledge and great experience. The solution was found in the creation of a hybrid system PROTECTOR, whose knowledge base represents a formalization of the expert knowledge in the mine safety field. The main goal of the system is the estimation of mining environment as one of the significant components of general safety state in a mine. This global goal is subdivided into a hierarchical structure of subgoals where each subgoal can be viewed as the estimation of a set of parameters (gas, dust, climate, noise, vibration, illumination, geotechnical hazard) which determine the general mine safety state and category of hazard in mining environment. Both the hybrid nature of the system and the possibilities it offers are illustrated through a case study using field data related to an existing Serbian surface coal mine.

Proposal for the improvement of IRD safety culture based on risk analysis

International Nuclear Information System (INIS)

Aguiar, L.A.; Ferreira, P.R.R.; Silveira, C.S.

2017-01-01

The Safety Culture (SC) is a concept about the relationship of individuals and organizations towards the safety in a specific activity. Any organization that carries out activities with risks has a SC, even at minimum levels. People perceive different types of radiation risks in very different ways, therefore, to identify and to analysis of the possible radiation risks resulting from normal operation or accident conditions is an important issue in order to improve the SC in organization. The main is to present guidelines for the improvement of the safety culture in the Institute of Radiation Protection and Dosimetry - IRD through on risk-based approach. The methodology proposed here is: A) select a division of the IRD for case study; B) assess the level of the 10 culture safety basic elements of the IRD division selected; C) conduct a survey of the hazards and risks associated with the various activities developed by the division; D) reassess the level of the 10 basic elements of CS; And E) analyze the results and correlate the impact of risk knowledge on safety culture improvement. The expected result is improvement the safety and of safety culture by understanding of radiation risks and hazards relating to work and to the working environment; and thus enforce a collective commitment to safety by teams and individuals and raise the safety culture to higher levels. (author)

Proposal for the improvement of IRD safety culture based on risk analysis

Energy Technology Data Exchange (ETDEWEB)

Aguiar, L.A.; Ferreira, P.R.R. [Instituto de Radioproteção e Dosimetria (DIRAD/IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Silveira, C.S., E-mail: laguiar@ird.gov.br [Comissão Nacional de Energia Nuclear (DRS/CGMI/CNEN), Rio de Janeiro, RJ (Brazil)

2017-07-01

The Safety Culture (SC) is a concept about the relationship of individuals and organizations towards the safety in a specific activity. Any organization that carries out activities with risks has a SC, even at minimum levels. People perceive different types of radiation risks in very different ways, therefore, to identify and to analysis of the possible radiation risks resulting from normal operation or accident conditions is an important issue in order to improve the SC in organization. The main is to present guidelines for the improvement of the safety culture in the Institute of Radiation Protection and Dosimetry - IRD through on risk-based approach. The methodology proposed here is: A) select a division of the IRD for case study; B) assess the level of the 10 culture safety basic elements of the IRD division selected; C) conduct a survey of the hazards and risks associated with the various activities developed by the division; D) reassess the level of the 10 basic elements of CS; And E) analyze the results and correlate the impact of risk knowledge on safety culture improvement. The expected result is improvement the safety and of safety culture by understanding of radiation risks and hazards relating to work and to the working environment; and thus enforce a collective commitment to safety by teams and individuals and raise the safety culture to higher levels. (author)

Hazard Analysis and Safety Requirements for Small Drone Operations: To What Extent Do Popular Drones Embed Safety?

Science.gov (United States)

Plioutsias, Anastasios; Karanikas, Nektarios; Chatzimihailidou, Maria Mikela

2018-03-01

Currently, published risk analyses for drones refer mainly to commercial systems, use data from civil aviation, and are based on probabilistic approaches without suggesting an inclusive list of hazards and respective requirements. Within this context, this article presents: (1) a set of safety requirements generated from the application of the systems theoretic process analysis (STPA) technique on a generic small drone system; (2) a gap analysis between the set of safety requirements and the ones met by 19 popular drone models; (3) the extent of the differences between those models, their manufacturers, and the countries of origin; and (4) the association of drone prices with the extent they meet the requirements derived by STPA. The application of STPA resulted in 70 safety requirements distributed across the authority, manufacturer, end user, or drone automation levels. A gap analysis showed high dissimilarities regarding the extent to which the 19 drones meet the same safety requirements. Statistical results suggested a positive correlation between drone prices and the extent that the 19 drones studied herein met the safety requirements generated by STPA, and significant differences were identified among the manufacturers. This work complements the existing risk assessment frameworks for small drones, and contributes to the establishment of a commonly endorsed international risk analysis framework. Such a framework will support the development of a holistic and methodologically justified standardization scheme for small drone flights. © 2017 Society for Risk Analysis.

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.

Project Guarantee 1985. Final repository for low- and intermediate-level radioactive wastes: The system of safety barriers

International Nuclear Information System (INIS)

Anon.

1985-01-01

The safety barrier system for the type B repository for low- and intermediate-level waste is described. The barrier parameters which are relevant for safety analysis are quantified and associated error limits and data scatter are given. The aim of the report is to give a summary documentation of the safety analysis input data and their scientific background. For secure containment of radioactive waste safety barriers are used which effectively limit the release of radioactive material from the repository (release barriers) and effectively retard the entry of the original radioactive material into the biosphere (time barriers). In the case of low- and intermediate-level waste the technical safety barrier system comprises: waste solidification matrix (cement, bitumen and resin), immobilisation of the waste packages in containers using liquid cement, concrete repository containers, backfilling of remaining vacant storage space with special concrete, concrete lining of the repository caverns, sealing of access tunnels on final closure of the repository. Natural geological safety barriers - host rock and overlying formations - have the following important functions. Because of its stability, the host rock in the repository zone protects the technical safety barrier system from destruction caused by climatic effects and erosion for a sufficient length of time. It also provides for low water flow and favourable chemistry (reducing conditions)

SIMMER as a safety analysis tool

International Nuclear Information System (INIS)

Smith, L.L.; Bell, C.R.; Bohl, W.R.; Bott, T.F.; Dearing, J.F.; Luck, L.B.

1982-01-01

SIMMER has been used for numerous applications in fast reactor safety, encompassing both accident and experiment analysis. Recent analyses of transition-phase behavior in potential core disruptive accidents have integrated SIMMER testing with the accident analysis. Results of both the accident analysis and the verification effort are presented as a comprehensive safety analysis program

Boundary conditions for pathways, safety analysis and basic criteria for low-level radiation waste site selection

International Nuclear Information System (INIS)

Saverot, P.

1994-01-01

There are three successive periods in the life of a disposal facility: the operating period, the institutional control period and the unrestricted site access period. The purpose of safety analysis of the disposal facility is to ensure that the radiological impacts for each period in the life of the facility are acceptable under all circumstances. Founded on a deterministic approach, this analysis leads to a determination of the maximum quantity of each radionuclide present in the facility at the beginning of the institutional control period in order for the impacts to be considered acceptable. Safety analysis involves the calculation of the radiological impacts of a given radiological inventory under a selected scenario, from all plausible scenarios of radionuclide migration to the environment in both normal and accident conditions, and taking into account other specified variables. The calculation itself involves an assessment of the quantities of radionuclides that could be released to the environment under the specific scenario selected and following identified pathways, and a determination of the resultant exposure, both internal and external, to the public. An iterative approach is used in the performance of pathways analyses. If the pathways analyses result in unacceptable radiological impacts, either the radiological inventory of the site is reduced or barrier characteristics not previously factored into the analysis are taken into account. New pathways analyses are then performed until the results are within the acceptable range. Once accepted by the safety authorities, the radiological inventory becomes the radiological capacity, which is the approved quantities of specific radionuclides that may be disposed of at the site. The following elaborates on the boundary conditions used in safety analyses and describes the types of pathways analyses performed for a LLW disposal facility

Performance and safety analysis of WP-cave concept

International Nuclear Information System (INIS)

Skagius, K.; Svemar, C.

1989-08-01

The report presents a performance safety, and cost analysis of the WP-cave, WPC, concept. In the performance analysis, questions specific to the WPC have been addressed which have been identified to require more detailed studies. Based on the outcome of this analysis, a safety analysis has been made which comprises of the modeling and calculation of radionuclide transport from the repository to the biosphere and the resulting dose exposure to man. The result of the safety analysis indicates that the present design of a WPC repository may give unacceptably high doses. By improving the properties of the bentonite/sand barrier such that the hydraulic conductivity is reduced, or by changing the short-lived steel canisters to more long-lived canisters, e.g. copper canisters, it is judged possible to achieve a sufficiently low level of dose exposure rates to man. The cost for a WPC repository of the studied design is significantly higher than for a KBS-3 repository considering the Swedish conditions and the Swedish amount of spent fuel. The major costs are connected to the excavation and backfilling of the bentonite/sand barrier. The potential for cost savings is high but it is not judged possible to account for savings in such a way that the WPC concept shows lower cost than the KBS-3 concept. (34 figs., 33 tabs., 29 refs.)

Safety analysis of Ignalina NPP during shutdown conditions

International Nuclear Information System (INIS)

Kaliatka, A.; Uspuras, E.

2000-01-01

The accident analysis for the Ignalina NPP with RBMK-1500 reactors at normal operating conditions and at minimum controlled power level (during startup of the reactor) has been performed in the frame of the project I n-Depth Safety Assessment of the Ignalina NPP , which was completed in 1996. However, the plant conditions during the reactor shutdown differ from conditions during reactor operation at full power (equipment status in protection systems, set points for actuation of safety and protection systems, etc.). Results of RELAP5 simulation of two worst initiating events during reactor shutdown - Pressure Header rupture in case of steam reactor cooldown as well as Pressure Header rupture in case of water reactor cooldown are discussed in the paper. Results of analysis shown that reactor are reliably cooled in both cases. Further analysis for all range of initial events during reactor shutdown and at shutdown conditions is recommended. (author)

Crime, perceived safety, and physical activity: A meta-analysis.

Science.gov (United States)

Rees-Punia, Erika; Hathaway, Elizabeth D; Gay, Jennifer L

2018-06-01

Perceived safety from crime and objectively-measured crime rates may be associated with physical inactivity. The purpose of this meta-analysis is to estimate the odds of accumulating high levels of physical activity (PA) when the perception of safety from crime is high and when objectively-measured crime is high. Peer-reviewed studies were identified through PubMed, Web of Science, ProQuest Criminal Justice, and ScienceDirect from earliest record through 2016. Included studies measured total PA, leisure-time PA, or walking in addition to perceived safety from crime or objective measures of crime. Mean odds ratios were aggregated with random effects models, and meta-regression was used to examine effects of potential moderators: country, age, and crime/PA measure. Sixteen cross-sectional studies yielded sixteen effects for perceived safety from crime and four effects for objective crime. Those reporting feeling safe from crime had a 27% greater odds of achieving higher levels of physical activity (OR=1.27 [1.08, 1.49]), and those living in areas with higher objectively-measured crime had a 28% reduced odds of achieving higher levels of physical activity (OR=0.72 [0.61, 0.83]). Effects of perceived safety were highly heterogeneous (I 2 =94.09%), but explored moderators were not statistically significant, likely because of the small sample size. Despite the limited number of effects suitable for aggregation, the mean association between perceived safety and PA was significant. As it seems likely that perceived lack of safety from crime constrains PA behaviors, future research exploring moderators of this association may help guide public health recommendations and interventions. Copyright © 2017 Elsevier Inc. All rights reserved.

77 FR 33777 - General Aviation Safety Forum: Climbing to the Next Level

Science.gov (United States)

2012-06-07

... NATIONAL TRANSPORTATION SAFETY BOARD General Aviation Safety Forum: Climbing to the Next Level The National Transportation Safety Board (NTSB) will convene a 2- day forum focused on safety issues related to... the Next Level,'' will be chaired by NTSB Chairman Deborah A. P. Hersman and all five Board Members...

Use of safety analysis to site comfirmation procedure in case of hard rock repository

International Nuclear Information System (INIS)

Peltonen, E.K.

1984-02-01

The role of safety analysis in a confirmation procedure of a candidate disposal site of radioactive wastes is discussed. Items dealt with include principle reasons and practical goals of the use of safety analysis, methodology of safety analysis and assessment, as well as usefulness and adequacy of the present safety analysis. Safety analysis is a tool, which enables one to estimate quantitatively the possible radiological impacts from the disposal. The results can be compared with the criteria and the suitability conclusions drawn. Because of its systems analytical nature safety analysis is an effective method to reveal, what are the most important factors of the disposal system and the most critical site characteristics inside the lumped parameters often provided by the experimental site investigation methods. Furthermore it gives information on the accuracy needs of different site properties. This can be utilized to judge whether the quality and quantity of the measurements for the characterization are sufficient as well as to guide the further site investigations. A more practical discussion regarding the applicability of the use of safety analysis is presented by an example concerning the assessment of a Finnish candidate site for low- and intermediate-level radioactive waste repository. (author)

Development of safety analysis technology for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Sim, Suk K.; Song, J. H.; Chung, Y. J. and others

1999-03-01

Inherent safety features and safety system characteristics of the SMART integral reactor are investigated in this study. Performance and safety of the SMART conceptual design have been evaluated and confirmed through the performance and safety analyses using safety analysis system codes as well as a preliminary performance and safety analysis methodology. SMART design base events and their acceptance criteria are identified to develop a preliminary PIRT for the SMART integral reactor. Using the preliminary PIRT, a set of experimental program for the thermal hydraulic separate effect tests and the integral effect tests was developed for the thermal hydraulic model development and the system code validation. Safety characteristics as well as the safety issues of the integral reactor has been identified during the study, which will be used to resolve the safety issues and guide the regulatory criteria for the integral reactor. The results of the performance and safety analyses performed during the study were used to feedback for the SMART conceptual design. The performance and safety analysis code systems as well as the preliminary safety analysis methodology developed in this study will be validated as the SMART design evolves. The performance and safety analysis technology developed during the study will be utilized for the SMART basic design development. (author)

ACRR fuel storage racks criticality safety analysis

International Nuclear Information System (INIS)

Bodette, D.E.; Naegeli, R.E.

1997-10-01

This document presents the criticality safety analysis for a new fuel storage rack to support modification of the Annular Core Research Reactor for production of molybdenum-99 at Sandia National Laboratories, Technical Area V facilities. Criticality calculations with the MCNP code investigated various contingencies for the criticality control parameters. Important contingencies included mix of fuel element types stored, water density due to air bubbles or water level for the over-moderated racks, interaction with existing fuel storage racks and fuel storage holsters in the fuel storage pool, neutron absorption of planned rack design and materials, and criticality changes due to manufacturing tolerances or damage. Some limitations or restrictions on use of the new fuel storage rack for storage operations were developed through the criticality analysis and are required to meet the double contingency requirements of criticality safety. As shown in the analysis, this system will remain subcritical under all credible upset conditions. Administrative controls are necessary for loading, moving, and handling the storage rack as well as for control of operations around it. 21 refs., 16 figs., 4 tabs

Computer aided safety analysis 1989

International Nuclear Information System (INIS)

1990-04-01

The meeting was conducted in a workshop style, to encourage involvement of all participants during the discussions. Forty-five (45) experts from 19 countries, plus 22 experts from the GDR participated in the meeting. A list of participants can be found at the end of this volume. Forty-two (42) papers were presented and discussed during the meeting. Additionally an open discussion was held on the possible directions of the IAEA programme on Computer Aided Safety Analysis. A summary of the conclusions of these discussions is presented in the publication. The remainder of this proceedings volume comprises the transcript of selected technical papers (22) presented in the meeting. It is the intention of the IAEA that the publication of these proceedings will extend the benefits of the discussions held during the meeting to a larger audience throughout the world. The Technical Committee/Workshop on Computer Aided Safety Analysis was organized by the IAEA in cooperation with the National Board for Safety and Radiological Protection (SAAS) of the German Democratic Republic in Berlin. The purpose of the meeting was to provide an opportunity for discussions on experiences in the use of computer codes used for safety analysis of nuclear power plants. In particular it was intended to provide a forum for exchange of information among experts using computer codes for safety analysis under the Technical Cooperation Programme on Safety of WWER Type Reactors (RER/9/004) and other experts throughout the world. A separate abstract was prepared for each of the 22 selected papers. Refs, figs tabs and pictures

Estimation of average hazardous-event-frequency for allocation of safety-integrity levels

International Nuclear Information System (INIS)

Misumi, Y.; Sato, Y.

1999-01-01

One of the fundamental concepts of the draft international standard, IEC 61508, is target failure measures to be allocated to Electric/Electronic/Programmable Electronic Safety-Related Systems, i.e. Safety Integrity Levels. The Safety Integrity Levels consist of four discrete probabilistic levels for specifying the safety integrity requirements or the safety functions to be allocated to Electric/Electronic/Programmable Electronic Safety-Related Systems. In order to select the Safety Integrity Levels the draft standard classifies Electric/Electronic/Programmable Electronic Safety-Related Systems into two modes of operation using demand frequencies only. It is not clear which modes of operation should be applied to Electric/Electronic/Programmable Electronic Safety-Related Systems taking into account the demand-state probability and the spurious demand frequency. It is essential for the allocation of Safety Integrity Levels that generic algorithms be derived by involving possible parameters, which make it possible to model the actuality of real systems. The present paper addresses this issue. First of all, the overall system including Electric/Electronic/programmable Electronic Safety-Related Systems is described using a simplified fault-tree. Then, the relationships among demands, demand-states and proof-tests are studied. Overall systems are classified into two groups: a non-demand-state-at-proof-test system which includes both repairable and non-repairable demand states and a constant-demand-frequency system. The new ideas such as a demand-state, spurious demand-state, mean time between detections, rates of d-failure and h-failure, and an h/d ratio are introduced in order to make the Safety Integrity Levels and modes of operation generic and comprehensive. Finally, the overall system is simplified and modeled by fault-trees using Priority-AND gates. At the same time the assumptions for modeling are described. Generic algorithms to estimate hazardous

Subseabed disposal safety analysis

International Nuclear Information System (INIS)

Koplick, C.M.; Kabele, T.J.

1982-01-01

This report summarizes the status of work performed by Analytic Sciences Corporation (TASC) in FY'81 on subseabed disposal safety analysis. Safety analysis for subseabed disposal is divided into two phases: pre-emplacement which includes all transportation, handling, and emplacement activities; and long-term (post-emplacement), which is concerned with the potential hazard after waste is safely emplaced. Details of TASC work in these two areas are provided in two technical reports. The work to date, while preliminary, supports the technical and environmental feasibility of subseabed disposal of HLW

14 CFR 33.75 - Safety analysis.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Safety analysis. 33.75 Section 33.75... STANDARDS: AIRCRAFT ENGINES Design and Construction; Turbine Aircraft Engines § 33.75 Safety analysis. (a... consequences of all failures that can reasonably be expected to occur. This analysis will take into account, if...

14 CFR 35.15 - Safety analysis.

Science.gov (United States)

2010-01-01

... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Safety analysis. 35.15 Section 35.15... STANDARDS: PROPELLERS Design and Construction § 35.15 Safety analysis. (a)(1) The applicant must analyze the.... This analysis will take into account, if applicable: (i) The propeller system in a typical installation...

Safety analysis of spent fuel packaging

International Nuclear Information System (INIS)

Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

1987-01-01

Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)

Compendium of computer codes for the safety analysis of LMFBR's

International Nuclear Information System (INIS)

1975-06-01

A high level of mathematical sophistication is required in the safety analysis of LMFBR's to adequately meet the demands for realism and confidence in all areas of accident consequence evaluation. The numerical solution procedures associated with these analyses are generally so complex and time consuming as to necessitate their programming into computer codes. These computer codes have become extremely powerful tools for safety analysis, combining unique advantages in accuracy, speed and cost. The number, diversity and complexity of LMFBR safety codes in the U. S. has grown rapidly in recent years. It is estimated that over 100 such codes exist in various stages of development throughout the country. It is inevitable that such a large assortment of codes will require rigorous cataloguing and abstracting to aid individuals in identifying what is available. It is the purpose of this compendium to provide such a service through the compilation of code summaries which describe and clarify the status of domestic LMFBR safety codes. (U.S.)

Development of safety analysis technology for integral reactor

International Nuclear Information System (INIS)

Kim, Hee Cheol; Kim, K. K.; Kim, S. H.

2002-04-01

The state-of-the-arts for the integral reactor was performed to investigate the safety features. The safety and performance of SMART were assessed using the technologies developed during the study. For this purpose, the computer code system and the analysis methodology were developed and the safety and performance analyses on SMART basic design were carried out for the design basis event and accident. The experimental facilities were designed for the core flow distribution test and the self-pressurizing pressurizer performance test. The tests on the 2-phase critical flow with non-condensable gas were completed and the results were used to assess the critical flow model. Probabilistic Safety Assessment(PSA) was carried out to evaluate the safety level and to optimize the design by identifying and remedying any weakness in the design. A joint study with KINS was carried out to promote licensing environment. The generic safety issues of integral reactors were identified and the solutions were formulated. The economic evaluation of the SMART desalination plant and the activities related to the process control were carried out in the scope of the study

Auditable safety analysis: High Radiation Level Chemical Development Facility (Buildings 4507 and 4556), Oak Ridge National Laboratory, Oak Ridge, Tennessee

International Nuclear Information System (INIS)

Platfoot, J.H.

1998-07-01

The High-Radiation-Level Chemical Development Facility includes Buildings 4507 and 4556. Building 4507, located immediately to the west of Building 4500N and to the south of Building 4505, is a doubly contained three-level structure constructed in 1957. The most recent use of the facility was for recovery of multi-gram quantities of 244 Cm during the early 1970s and for Liquid Metal Fast Breeder Reactor (LMFBR) fuel studies in the late 1970s. It has remained in safe standby since 1980. Building 4556 is a below-grade filter pit located to the southwest of Building 4507 and was constructed in 1972. Ventilation from the cells in Building 4507 is passed through high-efficiency particulate air (HEPA) filtration in this building prior to being exhausted to the Building 3039 stack system. This building remains in operation to support ventilation requirements for Building 4507. This Auditable Safety Analysis (ASA) was developed in accordance with the requirements in Energy Systems Program Description FS-103PD, Safety Documentation, Revision 1. This ASA identifies and screens all hazards associated with Buildings 4507 and 4556. The only hazard not screened out and requiring further analysis following the initial screening process is radioactive material in the form of surface contamination. The results of this ASA indicate that the hazards associated with Buildings 4507 and 4556 do not pose a significant threat to workers, the public, or the environment

The dependence level analysis between the human actions in NPP Operation

International Nuclear Information System (INIS)

Farcasiu, M.; Nitoi, M.; Apostol, M.; Florescu, G.; Prisecaru, Ilie

2009-01-01

The Human Reliability Analysis (HRA) is an important method in Probabilistic Safety Assessment (PSA) studies and offers desirability for concrete improvement of the man - machine - organization interfaces, reliability and safety. An important step in HRA is the dependence level analysis between the human actions performed by the same person or between the actions performed by different persons, step in quantitative analysis of the human errors probabilities. The purpose of this paper is to develop a model to analyze the dependence level between human actions for Nuclear Power Plant (NPP) operation. The model estimates the conditional human error probabilities (CHEP) and joint human error probabilities (JHEP). The achieved sensitivity analyses determine human performance sensibility to systematic variations for dependence level between human actions. The human error probabilities estimated in this paper are adequate values for integration both in HRA and in PSA realized for NPP. This type of analysis helps in finding and analyzing the ways of reducing the likelihood of human errors, so that the impact of human factor to systems availability, reliability and safety can be realistically estimated. In order to demonstrate the usability of this model an analysis is performed upon the dependences between the necessary human actions in mitigating the consequences of LOCA events, particularly for the case of Cernavoda NPP. (authors)

Development of System Model for Level 1 Probabilistic Safety Assessment of TRIGA PUSPATI Reactor

International Nuclear Information System (INIS)

Tom, P.P; Mazleha Maskin; Ahmad Hassan Sallehudin Mohd Sarif; Faizal Mohamed; Mohd Fazli Zakaria; Shaharum Ramli; Muhamad Puad Abu

2014-01-01

Nuclear safety is a very big issue in the world. As a consequence of the accident at Fukushima, Japan, most of the reactors in the world have been reviewed their safety of the reactors including also research reactors. To develop Level 1 Probabilistic Safety Assessment (PSA) of TRIGA PUSPATI Reactor (RTP), three organizations are involved; Nuclear Malaysia, AELB and UKM. PSA methodology is a logical, deductive technique which specifies an undesired top event and uses fault trees and event trees to model the various parallel and sequential combinations of failures that might lead to an undesired event. Fault Trees (FT) methodology is use in developing of system models. At the lowest level, the Basic Events (BE) of the fault trees (components failure and human errors) are assigned probability distributions. In this study, Risk Spectrum software used to construct the fault trees and analyze the system models. The results of system models analysis such as core damage frequency (CDF), minimum cut set (MCS) and common cause failure (CCF) uses to support decision making for upgrading or modification of the RTP?s safety system. (author)

Gas-cooled reactor safety and accident analysis

International Nuclear Information System (INIS)

1985-12-01

The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers

Partial Safety Factors and Target Reliability Level in Danish Structural Codes

DEFF Research Database (Denmark)

Sørensen, John Dalsgaard; Hansen, J. O.; Nielsen, T. A.

2001-01-01

The partial safety factors in the newly revised Danish structural codes have been derived using a reliability-based calibration. The calibrated partial safety factors result in the same average reliability level as in the previous codes, but a much more uniform reliability level has been obtained....... The paper describes the code format, the stochastic models and the resulting optimised partial safety factors....

Safety analysis - current and future regulatory challenges

Energy Technology Data Exchange (ETDEWEB)

Jamieson, T., E-mail: Terry.Jamieson@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

2015-07-01

'Full text:' The current and future regulatory challenges associated with deterministic safety analysis are reviewed, including: 1. The CNSC's and safety control areas. 2. Traditional safety analysis approach. 3. Experience gained and impact. 4. Current analysis and regulatory approaches. 5. Current status. 6. Complexity and challenges In particular, the technical, regulatory and strategic aspects of these challenges are discussed. (author)

Safety analysis - current and future regulatory challenges

International Nuclear Information System (INIS)

Jamieson, T.

2015-01-01

'Full text:' The current and future regulatory challenges associated with deterministic safety analysis are reviewed, including: 1. The CNSC's and safety control areas. 2. Traditional safety analysis approach. 3. Experience gained and impact. 4. Current analysis and regulatory approaches. 5. Current status. 6. Complexity and challenges In particular, the technical, regulatory and strategic aspects of these challenges are discussed. (author)

Development of safety analysis technology for LMR

International Nuclear Information System (INIS)

Hahn, Do Hee; Kwon, Y. M.; Kim, K. D.

2000-05-01

The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well

Development of safety analysis technology for LMR

Energy Technology Data Exchange (ETDEWEB)

Hahn, Do Hee; Kwon, Y. M.; Kim, K. D. [and others

2000-05-01

The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well.

Management system of health and safety work (SMK3) with job safety analysis (JSA) in PT. Nira Murni construction

Science.gov (United States)

Melliana, Armen, Yusrizal, Akmal, Syarifah

2017-11-01

PT Nira Murni construction is a contractor of PT Chevron Pacific Indonesia which engaged in contractor, fabrication, maintenance construction suppliers, and labor services. The high of accident rate in this company is caused the lack of awareness of workplace safety. Therefore, it requires an effort to reduce the accident rate on the company so that the financial losses can be minimized. In this study, Safe T-Score method is used to analyze the accident rate by measuring the level of frequency. Analysis is continued using risk management methods which identify hazards, risk measurement and risk management. The last analysis uses Job safety analysis (JSA) which will identify the effect of accidents. From the result of this study can be concluded that Job Safety Analysis (JSA) methods has not been implemented properly. Therefore, JSA method needs to follow-up in the next study, so that can be well applied as prevention of occupational accidents.

Safety analysis reports - new strategies

International Nuclear Information System (INIS)

Booth, J.A.

1994-01-01

Within the past year there have been many external changes in the requirements of safety analysis reports. Now there is emphasis on open-quotes graded approachesclose quotes depending on the Hazard Classification of the project. The Energy Facility Contractors Group (EFCOG) has a Safety Analysis Working Group. The results of this group for the past year are discussed as well as the implications for EG ampersand G. New strategies include ideas for incorporating the graded approach, auditable safety documents, additional guidance for Hazard Classification per DOE-STD-1027-92. The emphasis in the paper is on those projects whose hazard classification is category three or less

Development of evaluation method for software safety analysis techniques

International Nuclear Information System (INIS)

Huang, H.; Tu, W.; Shih, C.; Chen, C.; Yang, W.; Yih, S.; Kuo, C.; Chen, M.

2006-01-01

Full text: Full text: Following the massive adoption of digital Instrumentation and Control (I and C) system for nuclear power plant (NPP), various Software Safety Analysis (SSA) techniques are used to evaluate the NPP safety for adopting appropriate digital I and C system, and then to reduce risk to acceptable level. However, each technique has its specific advantage and disadvantage. If the two or more techniques can be complementarily incorporated, the SSA combination would be more acceptable. As a result, if proper evaluation criteria are available, the analyst can then choose appropriate technique combination to perform analysis on the basis of resources. This research evaluated the applicable software safety analysis techniques nowadays, such as, Preliminary Hazard Analysis (PHA), Failure Modes and Effects Analysis (FMEA), Fault Tree Analysis (FTA), Markov chain modeling, Dynamic Flowgraph Methodology (DFM), and simulation-based model analysis; and then determined indexes in view of their characteristics, which include dynamic capability, completeness, achievability, detail, signal/ noise ratio, complexity, and implementation cost. These indexes may help the decision makers and the software safety analysts to choose the best SSA combination arrange their own software safety plan. By this proposed method, the analysts can evaluate various SSA combinations for specific purpose. According to the case study results, the traditional PHA + FMEA + FTA (with failure rate) + Markov chain modeling (without transfer rate) combination is not competitive due to the dilemma for obtaining acceptable software failure rates. However, the systematic architecture of FTA and Markov chain modeling is still valuable for realizing the software fault structure. The system centric techniques, such as DFM and Simulation-based model analysis, show the advantage on dynamic capability, achievability, detail, signal/noise ratio. However, their disadvantage are the completeness complexity

Occupational Health and Safety. Level 1. Level 2. Level 3. Support Materials for Agricultural Training.

Science.gov (United States)

Batman, Kangan; Gadd, Nick; Lucas, Michele

This publication contains the three communication skills units of the three levels of Support Materials for Agricultural Training (SMAT) in the area of occupational health and safety: Level 1 (starting), 2 (continuing), and 3 (completing). The units are designed to help the learner improve his or her written and spoken communication skills needed…

Hazard Identification and Risk Assessment of Health and Safety Approach JSA (Job Safety Analysis) in Plantation Company

Science.gov (United States)

Sugarindra, Muchamad; Ragil Suryoputro, Muhammad; Tiya Novitasari, Adi

2017-06-01

Plantation company needed to identify hazard and perform risk assessment as an Identification of Hazard and Risk Assessment Crime and Safety which was approached by using JSA (Job Safety Analysis). The identification was aimed to identify the potential hazards that might be the risk of workplace accidents so that preventive action could be taken to minimize the accidents. The data was collected by direct observation to the workers concerned and the results were recorded on a Job Safety Analysis form. The data were as forklift operator, macerator worker, worker’s creeper, shredder worker, workers’ workshop, mechanical line worker, trolley cleaning workers and workers’ crepe decline. The result showed that shredder worker value was 30 and had the working level with extreme risk with the risk value range was above 20. So to minimize the accidents could provide Personal Protective Equipment (PPE) which were appropriate, information about health and safety, the company should have watched the activities of workers, and rewards for the workers who obey the rules that applied in the plantation.

Fuel reprocessing: safety analysis of extraction cycles

International Nuclear Information System (INIS)

Dinh, B.; Mauborgne, B.; Baron, P.; Mercier, J.P.

1991-01-01

An essential part of the safety analysis related to the extraction cycles of reprocessing plants, is the analysis of their behaviour during steady-state and transient operations, by means of simulation codes. These codes are based on the chemical properties of the main species involved (distribution coefficient and kinetics) and the hydrodynamics inside the contactors (mixer-settlers and pulsed columns). These codes have been consolidated by comparison of calculations with experimental results. The safety analysis is essentially performed in two steps. The first step is a parametric sensitivity analysis of the chemical flowsheet operated: the effect of a misadjustment (flowrate of feed, solvent, etc) is evaluated by successive steady-state calculations. These calculations help the identification of the sensitive parameters for the risk of plutonium accumulation, while indicating the permissible level of misadjustment. These calculations also serve to identify the parameters which should be measured during plant operation. The second step is the study of transient regimes, for the most sensitive parameters related to plutonium accumulation risk. The aim is to confirm the conclusions of the first step and to check that the characteristic process parameters chosen effectively allow, the early and reliable detection of any drift towards a plutonium accumulating regime. The procedures to drive the process backwards to a specified convenient steady-state regime from a drifting-state are also verified. The identification of the sensitive parameters, the process status parameters and the process transient analysis, allow a good control of process operation. This procedure, applied to the first purification cycle of COGEMA's UP3-A La Hague plant has demonstrated the total safety of facility operations

System and software safety analysis for the ERA control computer

International Nuclear Information System (INIS)

Beerthuizen, P.G.; Kruidhof, W.

2001-01-01

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

System analysis of vehicle active safety problem

Science.gov (United States)

Buznikov, S. E.

2018-02-01

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

AST-500 safety analysis experience

Energy Technology Data Exchange (ETDEWEB)

Falikov, A A; Bakhmetiev, A M; Kuul, V S; Samoilov, O B [OKBM, Nizhny Novgorod (Russian Federation)

1997-09-01

Characteristic AST-type NHR safety features and requirements are described briefly. The main approaches and results of design and beyond-design accidents analyses for the AST-500 NHR, and the results of probabilistic safety assessments are considered. It is concluded that the AST-500 possesses a high safety level in virtue of the development and realization in the design of self-protection, passivity and defence-in-depth principles. (author). 9 refs, 2 figs.

Applying importance-performance analysis to patient safety culture.

Science.gov (United States)

Lee, Yii-Ching; Wu, Hsin-Hung; Hsieh, Wan-Lin; Weng, Shao-Jen; Hsieh, Liang-Po; Huang, Chih-Hsuan

2015-01-01

The Sexton et al.'s (2006) safety attitudes questionnaire (SAQ) has been widely used to assess staff's attitudes towards patient safety in healthcare organizations. However, to date there have been few studies that discuss the perceptions of patient safety both from hospital staff and upper management. The purpose of this paper is to improve and to develop better strategies regarding patient safety in healthcare organizations. The Chinese version of SAQ based on the Taiwan Joint Commission on Hospital Accreditation is used to evaluate the perceptions of hospital staff. The current study then lies in applying importance-performance analysis technique to identify the major strengths and weaknesses of the safety culture. The results show that teamwork climate, safety climate, job satisfaction, stress recognition and working conditions are major strengths and should be maintained in order to provide a better patient safety culture. On the contrary, perceptions of management and hospital handoffs and transitions are important weaknesses and should be improved immediately. Research limitations/implications - The research is restricted in generalizability. The assessment of hospital staff in patient safety culture is physicians and registered nurses. It would be interesting to further evaluate other staff's (e.g. technicians, pharmacists and others) opinions regarding patient safety culture in the hospital. Few studies have clearly evaluated the perceptions of healthcare organization management regarding patient safety culture. Healthcare managers enable to take more effective actions to improve the level of patient safety by investigating key characteristics (either strengths or weaknesses) that healthcare organizations should focus on.

Emergency concepts for the safety level four; Notfallkonzepte der Sicherheitsebene Vier

Energy Technology Data Exchange (ETDEWEB)

Richner, Martin [Axpo Power AG, Doettingen (Switzerland). Kernkraftwerk Beznau

2016-04-15

According to the IAEA Guidelines and the Swiss Safety Guidelines the defence-in depth safety concept for a nuclear power plant consists of four safety levels. Emergency measures for the limitation of beyond design basis accidents are of safety level four. They are referred to as incident management. After the Chernobyl accident in 1986, in Switzerland the former regulatory body HSK (today ENSI) requested several retrofit measures in the field of accident management. The importance of accident management was visible again in Fukushima and demands for preventive measures grew.

Radiation Safety Analysis In The NFEC For Assessing Possible Implementation Of The ICRP-60 Standard

International Nuclear Information System (INIS)

Yowono, I.

1998-01-01

Radiation safety analysis of the 3 facilities in the nuclear fuel element center (NFEC) for assessing possible implementation of the ICRP-60 standard has been done. The analysis has covered the radiation dose received by workers, dose rate in the working area, surface contamination level, air contamination level and the level of radioactive gas release to the environment. The analysis has been based on BATAN regulation and ICRP-60 standard. The result of the analysis has showed that the highest radiation dose received has been found to be only around 15% of the set value in the ICRP-60 standard and only 6% of the set value in the BATAN regulation. Thus the ICRP-60 as radiation safety standard could be implemented without changing the laboratory design

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.

Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

International Nuclear Information System (INIS)

Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook

2007-08-01

This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the modeling

Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

2007-08-15

This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

Current status of safety analysis report for ANPP

International Nuclear Information System (INIS)

Amirjanyan, A.

1999-01-01

Current situation concerning Armenian NPP safety analysis report is considered within the frame of accepted safety practice. Licensing procedure is being developed. Technical support group was established in the Armenian Nuclear Regulatory Authority (ANRA). The task of the group is to study modern methods of NPP in depth safety analysis for technical assistance for the ANRA, and perform independent safety assessments. ANRA will be obliged to demand assistance from various foreign organisations for preparation of different parts of the Safety Analysis Report like determination though certain parts can be prepared in Armenia

Computer codes for level 1 probabilistic safety assessment

International Nuclear Information System (INIS)

1990-06-01

Probabilistic Safety Assessment (PSA) entails several laborious tasks suitable for computer codes assistance. This guide identifies these tasks, presents guidelines for selecting and utilizing computer codes in the conduct of the PSA tasks and for the use of PSA results in safety management and provides information on available codes suggested or applied in performing PSA in nuclear power plants. The guidance is intended for use by nuclear power plant system engineers, safety and operating personnel, and regulators. Large efforts are made today to provide PC-based software systems and PSA processed information in a way to enable their use as a safety management tool by the nuclear power plant overall management. Guidelines on the characteristics of software needed for management to prepare a software that meets their specific needs are also provided. Most of these computer codes are also applicable for PSA of other industrial facilities. The scope of this document is limited to computer codes used for the treatment of internal events. It does not address other codes available mainly for the analysis of external events (e.g. seismic analysis) flood and fire analysis. Codes discussed in the document are those used for probabilistic rather than for phenomenological modelling. It should be also appreciated that these guidelines are not intended to lead the user to selection of one specific code. They provide simply criteria for the selection. Refs and tabs

Regulatory review of probabilistic safety assessment (PSA) Level 2

International Nuclear Information System (INIS)

2001-07-01

Probabilistic safety assessment (PSA) is increasingly being used as part of the decision making process to assess the level of safety of nuclear power plants. The methodologies in use are maturing and the insights gained from the PSAs are being used along with those from deterministic analysis. Many regulatory authorities consider the current state of the art in PSA to be sufficiently well developed for results to be used centrally in the regulatory decision making process-referred to as risk informed regulation. For these applications to be successful, it will be necessary for the regulatory authority to have a high degree of confidence in the PSA. However, at the 1994 IAEA Technical Committee Meeting on Use of PSA in the Regulatory Process and at the OECD Nuclear Energy Agency Committee for Nuclear Regulatory Activities (CNRA) 'Special Issues' meeting in 1997 on Review Procedures and Criteria for Different Regulatory Applications of PSA, it was recognized that formal regulatory review guidance for PSA did not exist. The senior regulators noted that there was a need to produce some international guidance for reviewing PSAs to establish an agreed basis for assessing whether important technological and methodological issues in PSAs are treated adequately and to verify that conclusions reached are appropriate. In 1997, the IAEA and OECD Nuclear Energy Agency agreed to produce, in cooperation, guidance on Regulatory Review of PSA. This led to the publication of IAEA-TECDOC-1135 on the Regulatory Review of Probabilistic Safety Assessment (PSA) Level 1, which gives advice for the review of Level 1 PSA for initiating events occurring at power plants. This TECDOC extends the coverage to address the regulatory review of Level 2 PSA.These publications are intended to provide guidance to regulatory authorities on how to review the PSA for a nuclear power plant to gain confidence that it has been carried out to an acceptable level of quality so that it can be used as the

NKS/SOS-1 seminar on safety analysis

Energy Technology Data Exchange (ETDEWEB)

Lauridsen, K. [Risoe National Lab., Roskilde (Denmark); Anderson, K. [Karinta-Konsult (Sweden); Pulkkinen, U. [VTT Automation (Finland)

2001-05-01

The report describes presentations and discussions at a seminar held at Risoe on March 22-23, 2000. The title of the seminar was NKS/SOS-1 - Safety Analysis. It dealt with issues of relevance for the safety analysis for the entire nuclear safety field (notably reactors and nuclear waste repositories). Such issues were: objectives of safety analysis, risk criteria, decision analysis, expert judgement and risk communication. In addition, one talk dealt with criteria for chemical industries in Europe. The seminar clearly showed that the concept of risk is multidimensional, which makes clarity and transparency essential elements in risk communication, and that there are issues of common concern between different applications, such as how to deal with different kinds of uncertainty and expert judgement. (au)

Patient safety climate profiles across time: Strength and level of safety climate associated with a quality improvement program in Switzerland—A cross-sectional survey study

Science.gov (United States)

Mascherek, Anna C.

2017-01-01

Safety Climate has been acknowledged as an unspecific factor influencing patient safety. However, studies rarely provide in-depth analysis of climate data. As a helpful approach, the concept of “climate strength” has been proposed. In the present study we tested the hypotheses that even if safety climate remains stable on mean-level across time, differences might be evident in strength or shape. The data of two hospitals participating in a large national quality improvement program were analysed for differences in climate profiles at two measurement occasions. We analysed differences on mean-level, differences in percent problematic response, agreement within groups, and frequency histograms in two large hospitals in Switzerland at two measurement occasions (2013 and 2015) applying the Safety Climate Survey. In total, survey responses of 1193 individuals were included in the analyses. Overall, small but significant differences on mean-level of safety climate emerged for some subgroups. Also, although agreement was strong at both time-points within groups, tendencies of divergence or consensus were present in both hospitals. Depending on subgroup and analyses chosen, differences were more or less pronounced. The present study illustrated that taking several measures into account and describing safety climate from different perspectives is necessary in order to fully understand differences and trends within groups and to develop interventions addressing the needs of different groups more precisely. PMID:28753633

Safety analysis reports. Current status (third key report)

International Nuclear Information System (INIS)

1999-01-01

A review of Ukrainian regulations and laws concerned with Nuclear power and radiation safety is presented with an overview of the requirements for the Safety Analysis Report Contents. Status of Safety Analysis Reports (SAR) is listed for each particular Ukrainian NPP including SAR development schedules. Organisational scheme of SAR development works includes: general technical co-ordination on Safety Analysis Report development; list of leading organisations and utilization of technical support within international projects

Measuring patient safety culture: an assessment of the clustering of responses at unit level and hospital level

NARCIS (Netherlands)

Smits, M.; Wagner, C.; Spreeuwenberg, P.; Wal, van der G.

2009-01-01

OBJECTIVES: To test the claim that the Hospital Survey on Patient Safety Culture (HSOPS) measures patient safety culture instead of mere individual attitudes and to determine the most appropriate level (individual, unit or hospital level) for interventions aimed at improving the culture of patient

Measuring patient safety culture : an assessment of the clustering of responses at unit level and hospital level

NARCIS (Netherlands)

Smits, M.; Wagner, C.; Spreeuwenberg, P.; Wal, G. van der; Groenewegen, P.P.

2009-01-01

Objectives: To test the claim that the Hospital Survey on Patient Safety Culture (HSOPS) measures patient safety culture instead of mere individual attitudes and to determine the most appropriate level (individual, unit or hospital level) for interventions aimed at improving the culture of patient

An Analysis of Trainers' Perspectives within an Ecological Framework: Factors that Influence Mine Safety Training Processes.

Science.gov (United States)

Haas, Emily J; Hoebbel, Cassandra L; Rost, Kristen A

2014-09-01

Satisfactory completion of mine safety training is a prerequisite for being hired and for continued employment in the coal industry. Although training includes content to develop skills in a variety of mineworker competencies, research and recommendations continue to specify that specific limitations in the self-escape portion of training still exist and that mineworkers need to be better prepared to respond to emergencies that could occur in their mine. Ecological models are often used to inform the development of health promotion programs but have not been widely applied to occupational health and safety training programs. Nine mine safety trainers participated in in-depth semi-structured interviews. A theoretical analysis of the interviews was completed via an ecological lens. Each level of the social ecological model was used to examine factors that could be addressed both during and after mine safety training. The analysis suggests that problems surrounding communication and collaboration, leadership development, and responsibility and accountability at different levels within the mining industry contribute to deficiencies in mineworkers' mastery and maintenance of skills. This study offers a new technique to identify limitations in safety training systems and processes. The analysis suggests that training should be developed and disseminated with consideration of various levels-individual, interpersonal, organizational, and community-to promote skills. If factors identified within and between levels are addressed, it may be easier to sustain mineworker competencies that are established during safety training.

Work-related driver safety: A multi-level investigation

OpenAIRE

AMANDA ROSE WARMERDAM

2017-01-01

This program of research explored the organisational determinants of work-related road traffic injury in light vehicle fleets. The landscape of risk management in workplace road safety in Australia and organisational practices that influence safe driver behaviour were investigated. Key findings included that safe driving is influenced by factors at multiple levels, including senior managers, supervisors and individual fleet drivers and workplace road safety is not well integrated within curre...

Safety indicators: an efficient tool for a better safety

International Nuclear Information System (INIS)

Aufort, P.; Lars, R.

1993-01-01

Safety indicators based on the examination of the Operating Technical Specifications have been defined with the aim of following the in-operation safety level of French nuclear power plants. These safety indicators are operation feedback tools which permit the a posteriori justification and the adjustment of actual procedures. They would allow detection of an abnormal unavailability occurrence rate or a situation revealing a potential safety problem. So, data acquisition, processing, analysis and display software allowing trend analysis of these indicators has been developed so far as: a reflexion tool for the power plant operators about the safety instructions and the adjustment of preventive maintenance, and a help for decision making at a national level for the examination and the improvement of Operating Technical Specifications. This paper presents the objectives of these safety indicators, the processing tool associated, the preliminary results obtained and more elaborate processing of these indicators. These safety indicators may be very useful in framing probabilistic safety assessments. (author)

Hot Cell Facility (HCF) Safety Analysis Report

Energy Technology Data Exchange (ETDEWEB)

MITCHELL,GERRY W.; LONGLEY,SUSAN W.; PHILBIN,JEFFREY S.; MAHN,JEFFREY A.; BERRY,DONALD T.; SCHWERS,NORMAN F.; VANDERBEEK,THOMAS E.; NAEGELI,ROBERT E.

2000-11-01

This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR.

Approach to uncertainty evaluation for safety analysis

International Nuclear Information System (INIS)

Ogura, Katsunori

2005-01-01

Nuclear power plant safety used to be verified and confirmed through accident simulations using computer codes generally because it is very difficult to perform integrated experiments or tests for the verification and validation of the plant safety due to radioactive consequence, cost, and scaling to the actual plant. Traditionally the plant safety had been secured owing to the sufficient safety margin through the conservative assumptions and models to be applied to those simulations. Meanwhile the best-estimate analysis based on the realistic assumptions and models in support of the accumulated insights could be performed recently, inducing the reduction of safety margin in the analysis results and the increase of necessity to evaluate the reliability or uncertainty of the analysis results. This paper introduces an approach to evaluate the uncertainty of accident simulation and its results. (Note: This research had been done not in the Japan Nuclear Energy Safety Organization but in the Tokyo Institute of Technology.) (author)

Hot Cell Facility (HCF) Safety Analysis Report

International Nuclear Information System (INIS)

MITCHELL, GERRY W.; LONGLEY, SUSAN W.; PHILBIN, JEFFREY S.; MAHN, JEFFREY A.; BERRY, DONALD T.; SCHWERS, NORMAN F.; VANDERBEEK, THOMAS E.; NAEGELI, ROBERT E.

2000-01-01

This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR

Study on a quantitative evaluation method of equipment maintenance level and plant safety level for giant complex plant system

International Nuclear Information System (INIS)

Aoki, Takayuki

2010-01-01

In this study, a quantitative method on maintenance level which is determined by the two factors, maintenance plan and field work implementation ability by maintenance crew is discussed. And also a quantitative evaluation method on safety level for giant complex plant system is discussed. As a result of consideration, the following results were obtained. (1) It was considered that equipment condition after maintenance work was determined by the two factors, maintenance plan and field work implementation ability possessed by maintenance crew. The equipment condition determined by the two factors was named as 'equipment maintenance level' and its quantitative evaluation method was clarified. (2) It was considered that CDF in a nuclear power plant, evaluated by using a failure rate counting the above maintenance level was quite different from CDF evaluated by using existing failure rates including a safety margin. Then, the former CDF was named as 'plant safety level' of plant system and its quantitative evaluation method was clarified. (3) Enhancing equipment maintenance level means an improvement of maintenance quality. That results in the enhancement of plant safety level. Therefore, plant safety level should be always watched as a plant performance indicator. (author)

Development of High-Level Safety Requirements for a Pyroprocessing Facility

Energy Technology Data Exchange (ETDEWEB)

Seo, Seok Jun; Jo, Woo Jin; You, Gil Sung; Choung, Won Myung; Lee, Ho Hee; Kim, Hyun Min; Jeon, Hong Rae; Ku, Jeong Hoe; Lee, Hyo Jik [KAERI, Daejeon (Korea, Republic of)

2016-05-15

Korea Atomic Energy Research Institute (KAERI) has been developing a pyroproceesing technology to reduce the waste volume and recycle some elements. The pyroprocessing includes several treatment processes which are related with not only radiological and physical but also chemical and electrochemical properties. Thus, it is of importance to establish safety design requirements considering all the aspects of those properties for a reliable pyroprocessing facility. In this study, high-level requirements are presented in terms of not only radiation protection, nuclear criticality, fire protection, and seismic safety but also confinement and chemical safety for the unique characteristics of a pyroprocessing facility. Several high-level safety design requirements such as radiation protection, nuclear criticality, fire protection, seismic, confinement, and chemical processing were presented for a pyroprocessing facility. The requirements must fulfill domestic and international safety technology standards for a nuclear facility. Furthermore, additional requirements should be considered for the unique electrochemical treatments in a pyroprocessing facility.

Safety analysis of a high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shimazu, Akira; Morimoto, Toshio

1975-01-01

In recent years, in order to satisfy the social requirements of environment and safety and also to cope with the current energy stringency, the installation of safe nuclear power plants is indispensable. Herein, safety analysis and evaluation to confirm quantitatively the safety design of a nuclear power plant become more and more important. The safety analysis and its methods for a high temperature gas-cooled reactor are described, with emphasis placed on the practices by Fuji Electric Manufacturing Co. Fundamental rule of securing plant safety ; safety analysis in normal operation regarding plant dynamic characteristics and radioactivity evaluation ; and safety analysis at the time of accidents regarding plant response to the accidents and radioactivity evaluation are explained. (Mori, K.)

The Structure and Application of High Level Safety Goals. A Review by the MDEP Sub-committee on Safety Goals

International Nuclear Information System (INIS)

2011-01-01

One of the aims of MDEP is to work towards greater harmonisation of regulatory requirements. To achieve this aim, it is necessary that there is a degree of convergence on the safety goals that are required to be met by designers and operators. The term 'safety goals' is defined to cover all health and safety requirements which must be met: these may be deterministic rules and/or probabilistic targets. They should cover the safety of workers, public and the environment in line with the IAEA's Basic Safety Objective; encompassing safety in normal operation through to severe accidents. All regulators have safety goals, but these are expressed in many different ways and exercises in comparing them frequently are done at a very low level eg specific temperatures in the reactor vessel. The differences in the requirements from different regulators are difficult to resolve as the goals are derived using different principles and assumptions and are for a specific technology. Therefore MDEP set up a sub-committee to investigate a different approach. This approach was to start with the top level goals and to derive a structure and means of deriving lower tier goals that can be seen to be clearly related to the higher level ones. This approach has the potential to greatly assist in the process of harmonisation of regulatory requirements. The paper reviews the high level goals used in MDEP countries and the relevant work of international groups. From these it draws broad conclusions that the form of the framework should be an Hierarchical Structure of Safety Goals, incorporating an extended Defense-in-Depth approach. The basis concept is that the higher level safety goals can then developed, in a coherent and consistent manner, into lower level safety goals and targets that can be applied within the design and operation of reactors, with a clear connection between the different levels. This structured approach is technology-neutral and is sufficiently flexible that it can be

Improving patient safety in radiotherapy through error reporting and analysis

International Nuclear Information System (INIS)

Findlay, Ú.; Best, H.; Ottrey, M.

2016-01-01

Aim: To improve patient safety in radiotherapy (RT) through the analysis and publication of radiotherapy errors and near misses (RTE). Materials and methods: RTE are submitted on a voluntary basis by NHS RT departments throughout the UK to the National Reporting and Learning System (NRLS) or directly to Public Health England (PHE). RTE are analysed by PHE staff using frequency trend analysis based on the classification and pathway coding from Towards Safer Radiotherapy (TSRT). PHE in conjunction with the Patient Safety in Radiotherapy Steering Group publish learning from these events, on a triannual and summarised on a biennial basis, so their occurrence might be mitigated. Results: Since the introduction of this initiative in 2010, over 30,000 (RTE) reports have been submitted. The number of RTE reported in each biennial cycle has grown, ranging from 680 (2010) to 12,691 (2016) RTE. The vast majority of the RTE reported are lower level events, thus not affecting the outcome of patient care. Of the level 1 and 2 incidents reported, it is known the majority of them affected only one fraction of a course of treatment. This means that corrective action could be taken over the remaining treatment fractions so the incident did not have a significant impact on the patient or the outcome of their treatment. Analysis of the RTE reports demonstrates that generation of error is not confined to one professional group or to any particular point in the pathway. It also indicates that the pattern of errors is replicated across service providers in the UK. Conclusion: Use of the terminology, classification and coding of TSRT, together with implementation of the national voluntary reporting system described within this report, allows clinical departments to compare their local analysis to the national picture. Further opportunities to improve learning from this dataset must be exploited through development of the analysis and development of proactive risk management strategies

Human performance analysis in the frame of probabilistic safety assessment of research reactors

International Nuclear Information System (INIS)

Farcasiu, Mita; Nitoi, Mirela; Apostol, Minodora; Turcu, I.; Florescu, Gh.

2005-01-01

Full text: The analysis of operating experience has identified the importance of human performance in reliability and safety of research reactors. In Probabilistic Safety Assessment (PSA) of nuclear facilities, human performance analysis (HPA) is used in order to estimate human error contribution to the failure of system components or functions. HPA is a qualitative and quantitative analysis of human actions identified for error-likely situations or accident-prone situations. Qualitative analysis is used to identify all man-machine interfaces that can lead to an accident, types of human interactions which may mitigate or exacerbate the accident, types of human errors and performance shaping factors. Quantitative analysis is used to develop estimates of human error probability as effects of human performance in reliability and safety. The goal of this paper is to accomplish a HPA in the PSA frame for research reactors. Human error probabilities estimated as results of human actions analysis could be included in system event tree and/or system fault tree. The achieved sensitivity analyses determine human performance sensibility at systematically variations both for dependencies level between human actions and for operator stress level. The necessary information was obtained from operating experience of research reactor TRIGA from INR Pitesti. The required data were obtained from generic data bases. (authors)

Levels of safety satisfactory for commercialization of the breeder

International Nuclear Information System (INIS)

Ferguson, R.L.

1979-01-01

A brief discussion is presented of the Department of Energy's LMFBR safety program and the safety levels which DOE believes would be satisfactory for the commercialization of the breeder are indicated. Some observations are offered on the Three Mile Island accident and some of its implications are discussed for the LMFBR program

Applications of probabilistic risk analysis in nuclear criticality safety design

International Nuclear Information System (INIS)

Chang, J.K.

1992-01-01

Many documents have been prepared that try to define the scope of the criticality analysis and that suggest adding probabilistic risk analysis (PRA) to the deterministic safety analysis. The report of the US Department of Energy (DOE) AL 5481.1B suggested that an accident is credible if the occurrence probability is >1 x 10 -6 /yr. The draft DOE 5480 safety analysis report suggested that safety analyses should include the application of methods such as deterministic safety analysis, risk assessment, reliability engineering, common-cause failure analysis, human reliability analysis, and human factor safety analysis techniques. The US Nuclear Regulatory Commission (NRC) report NRC SG830.110 suggested that major safety analysis methods should include but not be limited to risk assessment, reliability engineering, and human factor safety analysis. All of these suggestions have recommended including PRA in the traditional criticality analysis

Software safety hazard analysis

International Nuclear Information System (INIS)

Lawrence, J.D.

1996-02-01

Techniques for analyzing the safety and reliability of analog-based electronic protection systems that serve to mitigate hazards in process control systems have been developed over many years, and are reasonably well understood. An example is the protection system in a nuclear power plant. The extension of these techniques to systems which include digital computers is not well developed, and there is little consensus among software engineering experts and safety experts on how to analyze such systems. One possible technique is to extend hazard analysis to include digital computer-based systems. Software is frequently overlooked during system hazard analyses, but this is unacceptable when the software is in control of a potentially hazardous operation. In such cases, hazard analysis should be extended to fully cover the software. A method for performing software hazard analysis is proposed in this paper

Improving the safety of Ukrainian NPP to reach an internationally accepted level

International Nuclear Information System (INIS)

Bozhko, S.; Helske, J.; Janke, R.; Mayoral, C.

2013-01-01

This paper summarizes the safety status and the modernization progress of Ukrainian NPPs towards an internationally accepted level of safety. After a brief discussion of the concept of what is called an 'international accepted level' for new and operating NPPs, the status of Russian type WWER and in particular the Ukrainian NPPs is presented. Then, the performed investigations of the gaps between international accepted level and the original status of Ukrainian NPPs are presented. The safety objectives of the modernization programs, some examples of defence in depth improvements, and an overall view of the modernization programs of Ukrainian NPPs are produced. Then, few important safety improvements implemented at the oldest Ukrainian WWER-1000 South Ukraine-1 are given in more detail. Finally, a conclusion presents the current status on the way to fulfill the national safety targets and to reach an internationally accepted level for all the Ukrainian NPPs. The paper is followed by the slides of the presentation. (authors)

Infusing Reliability Techniques into Software Safety Analysis

Science.gov (United States)

Shi, Ying

2015-01-01

Software safety analysis for a large software intensive system is always a challenge. Software safety practitioners need to ensure that software related hazards are completely identified, controlled, and tracked. This paper discusses in detail how to incorporate the traditional reliability techniques into the entire software safety analysis process. In addition, this paper addresses how information can be effectively shared between the various practitioners involved in the software safety analyses. The author has successfully applied the approach to several aerospace applications. Examples are provided to illustrate the key steps of the proposed approach.

Incorporation of advanced accident analysis methodology into safety analysis reports

International Nuclear Information System (INIS)

2003-05-01

The IAEA Safety Guide on Safety Assessment and Verification defines that the aim of the safety analysis should be by means of appropriate analytical tools to establish and confirm the design basis for the items important to safety, and to ensure that the overall plant design is capable of meeting the prescribed and acceptable limits for radiation doses and releases for each plant condition category. Practical guidance on how to perform accident analyses of nuclear power plants (NPPs) is provided by the IAEA Safety Report on Accident Analysis for Nuclear Power Plants. The safety analyses are performed both in the form of deterministic and probabilistic analyses for NPPs. It is customary to refer to deterministic safety analyses as accident analyses. This report discusses the aspects of using the advanced accident analysis methods to carry out accident analyses in order to introduce them into the Safety Analysis Reports (SARs). In relation to the SAR, purposes of deterministic safety analysis can be further specified as (1) to demonstrate compliance with specific regulatory acceptance criteria; (2) to complement other analyses and evaluations in defining a complete set of design and operating requirements; (3) to identify and quantify limiting safety system set points and limiting conditions for operation to be used in the NPP limits and conditions; (4) to justify appropriateness of the technical solutions employed in the fulfillment of predetermined safety requirements. The essential parts of accident analyses are performed by applying sophisticated computer code packages, which have been specifically developed for this purpose. These code packages include mainly thermal-hydraulic system codes and reactor dynamics codes meant for the transient and accident analyses. There are also specific codes such as those for the containment thermal-hydraulics, for the radiological consequences and for severe accident analyses. In some cases, codes of a more general nature such

ARIES-RS safety design and analysis

International Nuclear Information System (INIS)

Steiner, D.; El-Guebaly, L.; Herring, S.; Khater, H.; Mogahed, E.; Thayer, R.; Tillack, M.S.

1997-01-01

The ARIES-RS safety design and analysis focused on achieving two objectives: (1) The avoidance of sheltering or evacuation in the event of an accident; and (2) the generation of only low-level waste, no greater than Class C. The ARIES-RS baseline design employs V-4Cr-4Ti as the blanket structural material and a low activation ferritic steel in the reflector and shield. In the event of a LOCA, the baseline design first wall maximum temperature falls in the range of 1100-1200 C. For this temperature range, the hazard assessment indicates that the dose at the site boundary will be less than 1 rem per year. Thus, no sheltering or evacuation would be required in the event of a LOCA. Although the baseline design satisfies the first safety objective noted above, a first wall maximum temperature of ∝1100-1200 C would likely compromise the integrity of the vanadium blanket structure and would require blanket replacement following such a temperature excursion. To avoid this situation, a modified blanket design incorporating supplemental heat removal is also proposed. Preliminary analysis of this modified design suggests that the first wall maximum temperature can be kept below the temperature range of concern, ∝1000-1100 C, in the event of a LOCA. When the ferritic steel used in the reflector and shield is one reduced in Ir and Ag impurities, all in-vessel components qualify for near-surface shallow land burial as Class C low-level waste. (orig.)

Software safety analysis practice in installation phase

Energy Technology Data Exchange (ETDEWEB)

Huang, H. W.; Chen, M. H.; Shyu, S. S., E-mail: hwhwang@iner.gov.t [Institute of Nuclear Energy Research, No. 1000 Wenhua Road, Chiaan Village, Longtan Township, 32546 Taoyuan County, Taiwan (China)

2010-10-15

This work performed a software safety analysis in the installation phase of the Lung men nuclear power plant in Taiwan, under the cooperation of Institute of Nuclear Energy Research and Tpc. The US Nuclear Regulatory Commission requests licensee to perform software safety analysis and software verification and validation in each phase of software development life cycle with Branch Technical Position 7-14. In this work, 37 safety grade digital instrumentation and control systems were analyzed by failure mode and effects analysis, which is suggested by IEEE standard 7-4.3.2-2003. During the installation phase, skew tests for safety grade network and point to point tests were performed. The failure mode and effects analysis showed all the single failure modes can be resolved by the redundant means. Most of the common mode failures can be resolved by operator manual actions. (Author)

Software safety analysis practice in installation phase

International Nuclear Information System (INIS)

Huang, H. W.; Chen, M. H.; Shyu, S. S.

2010-10-01

This work performed a software safety analysis in the installation phase of the Lung men nuclear power plant in Taiwan, under the cooperation of Institute of Nuclear Energy Research and Tpc. The US Nuclear Regulatory Commission requests licensee to perform software safety analysis and software verification and validation in each phase of software development life cycle with Branch Technical Position 7-14. In this work, 37 safety grade digital instrumentation and control systems were analyzed by failure mode and effects analysis, which is suggested by IEEE standard 7-4.3.2-2003. During the installation phase, skew tests for safety grade network and point to point tests were performed. The failure mode and effects analysis showed all the single failure modes can be resolved by the redundant means. Most of the common mode failures can be resolved by operator manual actions. (Author)

Procedures for conducting probabilistic safety assessments of nuclear power plants (level 2). Accident progression, containment analysis and estimation of accident source terms

International Nuclear Information System (INIS)

1995-01-01

The present publication on Level 2 PSA is based on a compilation and review of practices in various Member States. It complements Safety Series No. 50-P-4, issued in 1992, on Procedures for Conducting Probabilistic Safety Assessments of Nuclear Power Plants (Level 1). Refs, figs and tabs

A Systematic Analysis of Functional Safety Certification Practices in Industrial Robot Software Development

Directory of Open Access Journals (Sweden)

Tong Xie

2017-01-01

Full Text Available For decades, industry robotics have delivered on the promise of speed, efficiency and productivity. The last several years have seen a sharp resurgence in the orders of industrial robots in China, and the areas addressed within industrial robotics has extended into safety-critical domains. However, safety standards have not yet been implemented widely in academia and engineering applications, particularly in robot software development. This paper presents a systematic analysis of functional safety certification practices in software development for the safety-critical software of industrial robots, to identify the safety certification practices used for the development of industrial robots in China and how these practices comply with the safety standard requirements. Reviewing from Chinese academic papers, our research shows that safety standards are barely used in software development of industrial robot. The majority of the papers propose various solutions to achieve safety, but only about two thirds of the papers refer to non-standardized approaches that mainly address the systematic level rather than the software development level. In addition, our research shows that with the development of artificial intelligent, an emerging field is still on the quest for standardized and suitable approaches to develop safety-critical software.

Guidelines for nuclear reactor equipments safety-analysis

International Nuclear Information System (INIS)

1978-01-01

The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

Preliminary Integrated Safety Analysis Status Report

International Nuclear Information System (INIS)

Gwyn, D.

2001-01-01

This report provides the status of the potential Monitored Geologic Repository (MGR) Integrated Safety Analysis (EA) by identifying the initial work scope scheduled for completion during the ISA development period, the schedules associated with the tasks identified, safety analysis issues encountered, and a summary of accomplishments during the reporting period. This status covers the period from October 1, 2000 through March 30, 2001

Analysis of Critical Characteristics for Safety Graded Personnel Computers in the KNICS Architecture

International Nuclear Information System (INIS)

Lee, Hyun Chul; Lee, Dong Young

2009-01-01

Critical characteristics analysis of a safety related item is to identify characteristics to be verified to replace an original item with the dedicated item. It is sure that the dedicated item meeting critical characteristics would perform its intended safety function instead of the specified item. KNICS project developed two safety systems: IDiPS RPS (Reactor Protection System) and IDiPS ESF-CCS (Engineered Safety Features-Component Control System). Two safety systems of IDiPS are equipped with personnel computers, so-called COMs (Cabinet Operator Modules), in their cabinets. The personnel computers, COMs, are responsible for safety system monitoring, testing, and maintaining. Even though two safety systems are safety critical system, the personnel computers of two systems, i.e. COMs, are not graded as safety-graded items. Regulation requirements are expected to be strengthened, and the functions of the personnel computer may be enhanced to include safety-related functions and safety functions, it would be necessary that the grade of the personnel computers is adjusted to a higher level, the safety grade. To try to upgrade a non safety system, i.e. COMs, to a safety system, its safety functions and requirements, i.e. critical characteristics, must be identified and verified. This paper describes the process of the identification of critical characteristics and the results of analysis

TU-EF-BRD-01: Topics in Quality and Safety Research and Level of Evidence

Energy Technology Data Exchange (ETDEWEB)

Pawlicki, T. [UCSD Medical Center (United States)

2015-06-15

Research related to quality and safety has been a staple of medical physics academic activities for a long time. From very early on, medical physicists have developed new radiation measurement equipment and analysis techniques, created ever increasingly accurate dose calculation models, and have vastly improved imaging, planning, and delivery techniques. These and other areas of interest have improved the quality and safety of radiotherapy for our patients. With the advent of TG-100, quality and safety is an area that will garner even more research interest in the future. As medical physicists pursue quality and safety research in greater numbers, it is worthwhile to consider what actually constitutes research on quality and safety. For example, should the development of algorithms for real-time EPID-based in-vivo dosimetry be defined as “quality and safety” research? How about the clinical implementation of such as system? Surely the application of failure modes and effects analysis to a clinical process would be considered quality and safety research, but is this type of research that should be included in the medical physics peer-reviewed literature? The answers to such questions are of critical importance to set researchers in a direction that will provide the greatest benefit to our field and the patients we serve. The purpose of this symposium is to consider what constitutes research in the arena of quality and safety and differentiate it from other research directions. The key distinction here is developing the tool itself (e.g. algorithms for EPID dosimetry) vs. studying the impact of the tool with some quantitative metric. Only the latter would I call quality and safety research. Issues of ‘basic’ versus ‘applied’ quality and safety research will be covered as well as how the research results should be structured to provide increasing levels of support that a quality and safety intervention is effective and sustainable. Examples from existing

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

System-level hazard analysis using the sequence-tree method

International Nuclear Information System (INIS)

Huang, H.-W.; Shih Chunkuan; Yih Swu; Chen, M.-H.

2008-01-01

A system-level PHA using the sequence-tree method is presented to perform safety-related digital I and C system SSA. The conventional PHA involves brainstorming among experts on various portions of the system to identify hazards through discussions. However, since the conventional PHA is not a systematic technique, the analysis results depend strongly on the experts' subjective opinions. The quality of analysis cannot be appropriately controlled. Therefore, this study presents a system-level sequence tree based PHA, which can clarify the relationship among the major digital I and C systems. This sequence-tree-based technique has two major phases. The first phase adopts a table to analyze each event in SAR Chapter 15 for a specific safety-related I and C system, such as RPS. The second phase adopts a sequence tree to recognize the I and C systems involved in the event, the working of the safety-related systems and how the backup systems can be activated to mitigate the consequence if the primary safety systems fail. The defense-in-depth echelons, namely the Control echelon, Reactor trip echelon, ESFAS echelon and Monitoring and indicator echelon, are arranged to build the sequence-tree structure. All the related I and C systems, including the digital systems and the analog back-up systems, are allocated in their specific echelons. This system-centric sequence-tree analysis not only systematically identifies preliminary hazards, but also vulnerabilities in a nuclear power plant. Hence, an effective simplified D3 evaluation can also be conducted

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

An overview-probabilistic safety analysis for research reactors

International Nuclear Information System (INIS)

Liu Jinlin; Peng Changhong

2015-01-01

For long-term application, Probabilistic Safety Analysis (PSA) has proved to be a valuable tool for improving the safety and reliability of power reactors. In China, 'Nuclear safety and radioactive pollution prevention 'Twelfth Five Year Plan' and the 2020 vision' raises clearly that: to develop probabilistic safety analysis and aging evaluation for research reactors. Comparing with the power reactors, it reveals some specific features in research reactors: lower operating power, lower coolant temperature and pressure, etc. However, the core configurations may be changed very often and human actions play an important safety role in research reactors due to its specific experimental requirement. As a result, there is a necessary to conduct the PSA analysis of research reactors. This paper discusses the special characteristics related to the structure and operation and the methods to develop the PSA of research reactors, including initiating event analysis, event tree analysis, fault tree analysis, dependent failure analysis, human reliability analysis and quantification as well as the experimental and external event evaluation through the investigation of various research reactors and their PSAs home and abroad, to provide the current situation and features of research reactors PSAs. (author)

RISMC Advanced Safety Analysis Project Plan – FY 2015 - FY 2019

Energy Technology Data Exchange (ETDEWEB)

Szilard, Ronaldo H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Youngblood, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-09-01

In this report, a project plan is developed, focused on industry applications, using Risk-Informed Safety Margin Characterization (RISMC) tools and methods applied to realistic, relevant, and current interest issues to the operating nuclear fleet. RISMC focuses on modernization of nuclear power safety analysis (tools, methods and data); implementing state-of-the-art modeling techniques (which include, for example, enabling incorporation of more detailed physics as they become available); taking advantage of modern computing hardware; and combining probabilistic and mechanistic analyses to enable a risk informed safety analysis process. The modernized tools will maintain the current high level of safety in our nuclear power plant fleet, while providing an improved understanding of safety margins and the critical parameters that affect them. Thus, the set of tools will provide information to inform decisions on plant modifications, refurbishments, and surveillance programs, while improving economics. This set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. The proposed plan will focus on application of the RISMC toolkit, in particular, solving realistic problems of important current issues to the nuclear industry, in collaboration with plant owners and operators to demonstrate the usefulness of these tools in decision making.

From mutual recognition to mutual scientific opinion? Constitutional framework for risk analysis in EU food safety law

NARCIS (Netherlands)

Szajkowska, A.

2009-01-01

Regulation 178/2002 (the so-called General Food Law – GFL) codifies risk analysis as the core principle of the modern food safety policy. This article places the GFL in EU multi-level food safety governance and analyses the impact of risk analysis, the precautionary principle and mechanisms of

Ignalina Safety Analysis Group's report for the year 1998

International Nuclear Information System (INIS)

Uspuras, E.; Augutis, J.; Bubelis, E.; Cesna, B.; Kaliatka, A.

1999-02-01

Results of Ignalina NPP Safety Analysis Group's research are presented. The main fields of group's activities in 1998 were following: safety analysis of reactor's cooling system, safety analysis of accident localization system, investigation of the problem graphite - fuel channel, reactor core modelling, assistance to the regulatory body VATESI in drafting regulations and reviewing safety reports presented by Ignalina NPP during the process of licensing of unit 1

Operating plant safety analysis needs

International Nuclear Information System (INIS)

Young, M.Y.; Love, D.S.

1992-01-01

The primary objective for nuclear power station owners is to operate and manage their plants safely. However, there is also a need to provide economical electric power, which requires that the unit be operated as efficiently as possible, consistent with the safety requirements. The objectives cited above can be achieved through the identification and use of available margins inherent in the plant design. As a result of conservative licensing and analytical approaches taken in the past, many of these margins may be found in the safety analysis limits within which plants currently operate. Improvements in the accuracy of the safety analysis, and a more realistic treatment of plant initial and boundary conditions, can make this margin available for a variety of uses which enhance plant performance, help to reduce O and M costs, and may help to extend licensed operation. Opportunities for improvement exist in several areas in the accident analysis normally performed for Chapter 15 of the FSAR. For example, recent modifications to the ECCS rule, 10CFR50.46 and Appendix K, allow use of margins previously unavailable in the analysis of the Loss of Coolant Accident (LOCA). To take advantage of this regulatory change, new methods are being developed to analyze both the large and small break loss of coolant accident (LOCA). As this margin is used, enhancements in the analysis of other transients will become necessary. The paper discusses accident analysis methods, future development needs, and analysis margin utilization in specific accident scenarios

Computer codes for safety analysis

International Nuclear Information System (INIS)

Holland, D.F.

1986-11-01

Computer codes for fusion safety analysis have been under development in the United States for about a decade. This paper will discuss five codes that are currently under development by the Fusion Safety Program. The purpose and capability of each code will be presented, a sample given, followed by a discussion of the present status and future development plans

Analysis of the survey results for the safety culture attitudes of HANARO management division staff

Energy Technology Data Exchange (ETDEWEB)

Lee, M.; Hwang, S. Y.; Lim, I. C.; Kang, T. J.; Choi, H. Y

2003-12-01

This report provides the that analysis results of the surveys on the safety culture attitude and related topics for the HANARO Management Division staff. The survey form consisted of the questionnaire on the safety culture attitude, happiness and life style, drinking habit and traffic safety attitude. The key information drawn from the analysis are as follow; First, the level of safety culture attitude of the staff was confirmed and it was found that the attitude improved when the present analysis results were compared with them for the survey in 1998. Second, the degree of happiness that the staff feel is higher than the average Korean and they enjoy more leisure time than the average Korean. Third, the staff drink more alcohol than the average Korean. Thus, the drinking habit must be improved for the health of the staff. Fourth, the questionnaire of the happiness and the traffic safety attitude should be improved to verify the hypotheses on the relation between them and the safety culture attitude.

Analysis of the survey results for the safety culture attitudes of HANARO management division staff

International Nuclear Information System (INIS)

Lee, M.; Hwang, S. Y.; Lim, I. C.; Kang, T. J.; Choi, H. Y.

2003-12-01

This report provides the that analysis results of the surveys on the safety culture attitude and related topics for the HANARO Management Division staff. The survey form consisted of the questionnaire on the safety culture attitude, happiness and life style, drinking habit and traffic safety attitude. The key information drawn from the analysis are as follow; First, the level of safety culture attitude of the staff was confirmed and it was found that the attitude improved when the present analysis results were compared with them for the survey in 1998. Second, the degree of happiness that the staff feel is higher than the average Korean and they enjoy more leisure time than the average Korean. Third, the staff drink more alcohol than the average Korean. Thus, the drinking habit must be improved for the health of the staff. Fourth, the questionnaire of the happiness and the traffic safety attitude should be improved to verify the hypotheses on the relation between them and the safety culture attitude

Removing unreasonable conservatisms in DOE safety analysis

International Nuclear Information System (INIS)

BISHOP, G.E.

1999-01-01

While nuclear safety analyses must always be conservative, invoking excessive conservatisms does not provide additional margins of safety. Rather, beyond a fairly narrow point, conservatisms skew a facility's true safety envelope by exaggerating risks and creating unreasonable bounds on what is required for safety. The conservatism has itself become unreasonable. A thorough review of the assumptions and methodologies contained in a facility's safety analysis can provide substantial reward, reducing both construction and operational costs without compromising actual safety

Safety assessment of research reactors and preparation of the safety analysis report

International Nuclear Information System (INIS)

1994-01-01

This Safety Guide presents guidelines, approved by international consensus, for the preparation, review and assessment of safety documentation for research reactors such as the Safety Analysis Report. While the Guide is most applicable to research reactors in the design and construction stage, it is also recommended for use during relicensing or reassessment of existing reactors

Application of Software Safety Analysis Methods

International Nuclear Information System (INIS)

Park, G. Y.; Hur, S.; Cheon, S. W.; Kim, D. H.; Lee, D. Y.; Kwon, K. C.; Lee, S. J.; Koo, Y. H.

2009-01-01

A fully digitalized reactor protection system, which is called the IDiPS-RPS, was developed through the KNICS project. The IDiPS-RPS has four redundant and separated channels. Each channel is mainly composed of a group of bistable processors which redundantly compare process variables with their corresponding setpoints and a group of coincidence processors that generate a final trip signal when a trip condition is satisfied. Each channel also contains a test processor called the ATIP and a display and command processor called the COM. All the functions were implemented in software. During the development of the safety software, various software safety analysis methods were applied, in parallel to the verification and validation (V and V) activities, along the software development life cycle. The software safety analysis methods employed were the software hazard and operability (Software HAZOP) study, the software fault tree analysis (Software FTA), and the software failure modes and effects analysis (Software FMEA)

Nuclear safety in Slovak Republic. Safety analysis reports for WWER 440 reactors

International Nuclear Information System (INIS)

Rohar, S.

1999-01-01

Implementation of nuclear power program is connected to establishment of regulatory body for safe regulation of siting, construction, operation and decommissioning of nuclear installations. Licensing being one of the most important regulatory surveillance activity is based on independent regulatory review and assessment of information on nuclear safety for particular nuclear facility. Documents required to be submitted to the regulatory body by the licensee in Slovakia for the review and assessment usually named Safety Analysis Report (SAR) are presented in detail in this paper. Current status of Safety Analysis Reports for Bohunice V-1, Bohunice V-2 and Mochovce NPP is shown

From Safety Analysis to Formal Specification

DEFF Research Database (Denmark)

Hansen, Kirsten Mark; Ravn, Anders P.; Stavridou, Victoria

1998-01-01

Software for safety critical systems must deal with the hazards identified bysafety analysis. This paper investigates, how the results of onesafety analysis technique, fault trees, are interpreted as software safetyrequirements to be used in the program design process. We propose thatfault tree...... analysis and program development use the samesystem model. This model is formalized in areal-time, interval logic, based on a conventional dynamic systems modelwith state evolving over time. Fault trees are interpreted astemporal formulas, and it is shown how such formulas can be usedfor deriving safety...

Business of Nuclear Safety Analysis Office, Nuclear Technology Test Center

International Nuclear Information System (INIS)

Hayakawa, Masahiko

1981-01-01

The Nuclear Technology Test Center established the Nuclear Safety Analysis Office to execute newly the works concerning nuclear safety analysis in addition to the works related to the proving tests of nuclear machinery and equipments. The regulations for the Nuclear Safety Analysis Office concerning its organization, business and others were specially decided, and it started the business formally in August, 1980. It is a most important subject to secure the safety of nuclear facilities in nuclear fuel cycle as the premise of developing atomic energy. In Japan, the strict regulation of safety is executed by the government at each stage of the installation, construction, operation and maintenance of nuclear facilities, based on the responsibility for the security of installers themselves. The Nuclear Safety Analysis Office was established as the special organ to help the safety examination related to the installation of nuclear power stations and others by the government. It improves and puts in order the safety analysis codes required for the cross checking in the safety examination, and carries out safety analysis calculation. It is operated by the cooperation of the Science and Technology Agency and the Agency of Natural Resources and Energy. The purpose of establishment, the operation and the business of the Nuclear Safety Analysis Office, the plan of improving and putting in order of analysis codes, and the state of the similar organs in foreign countries are described. (Kako, I.)

Establishment of Safety Analysis System and Technology for CANDU Reactors

International Nuclear Information System (INIS)

Park, Joo Hwan; Rhee, B. W.; Min, B. J.; Kim, H. T.; Kim, W. Y.; Yoon, C.; Chun, J. S.; Cho, M. S.; Jeong, J. Y.; Kang, H. S.

2007-06-01

The following 4 research items have been studied to establish a CANDU safety analysis system and to develop the relevant elementary technology for CANDU reactors. First, to improve and validate the CANDU design and operational safety analysis codes, the CANDU physics cell code WIMS-CANDU was improved, and validated, and an analysis of the moderator subcooling and pressure tube integrity has been performed for the large break LOCAs without ECCS. Also a CATHENA model and a CFD model for a post-blowdown fuel channel analysis have been developed and validated against two high temperature thermal-chemical experiments, CS28-1 and 2. Second, to improve the integrated operating system of the CANDU safety analysis codes, an extension has been made to them to include the core and fuel accident analyses, and a web-based CANDU database, CANTHIS version 2.0 was completed. Third, to assess the applicability of the ACR-7 safety analysis methodology to CANDU-6 the ACR-7 safety analysis methods were reviewed and the safety analysis methods of ACR-7 applicable to CANDU-6 were recommended. Last, to supplement and improve the existing CANDU safety analysis procedures, detailed analysis procedures have been prepared for individual accident scenarios. The results of this study can be used to resolve the CANDU safety issues, to improve the current design and operational safety analysis codes, and to technically support the Wolsong site to resolve their problems

FLIGHT SAFETY MANAGEMENT PROBLEMS AND EVALUATION OF FLIGHT SAFETY LEVEL OF AN AVIATION ENTERPRISE

OpenAIRE

B. V. Zubkov; H. E. Fourar

2017-01-01

This article is devoted to studying the problem of safety management system (SMS) and evaluating safety level of an aviation enterprise.This article discusses the problems of SMS, presented at the 41st meeting of the Russian Aviation Production Commanders Club in June 2014 in St. Petersburg in connection with the verification of the status of the CA of the Russian Federation by the International Civil Aviation Organization (ICAO) in the same year, a set of urgent measures to eliminate the def...

AGNES - safety reassessment of Paks NPP

International Nuclear Information System (INIS)

Gado, J.

1995-01-01

The main goal of the AGNES (Advanced General and New Evaluation of Safety) project for the reassessment of the safety of Paks Nuclear Power Plant, Hungary, was to improve the safety culture of the technology at Paks. A report was prepared on the reassessment of the Paks NPP safety. The analysis was divided into four groups: systems analysis, analysis of design basis accidents, severe accident analysis, and level 1 probabilistic safety analysis. Proposed safety enhancement measures are discussed. (N.T.)

Nuclear safety culture and nuclear safety supervision

International Nuclear Information System (INIS)

Chai Jianshe

2013-01-01

In this paper, the author reviews systematically and summarizes up the development process and stage characteristics of nuclear safety culture, analysis the connotation and characteristics of nuclear safety culture, sums up the achievements of our country's nuclear safety supervision, dissects the challenges and problems of nuclear safety supervision. This thesis focused on the relationship between nuclear safety culture and nuclear safety supervision, they are essential differences, but there is a close relationship. Nuclear safety supervision needs to introduce some concepts of nuclear safety culture, lays emphasis on humanistic care and improves its level and efficiency. Nuclear safety supervision authorities must strengthen nuclear safety culture training, conduct the development of nuclear safety culture, make sure that nuclear safety culture can play significant roles. (author)

Establishment of Safety Analysis System and Technology for CANDU Reactors

International Nuclear Information System (INIS)

Min, Byung Joo; Kim, W. Y.; Kim, H. T.; Rhee, B. W.; Yoon, C.; Kang, H. S.; Yoo, K. J.

2005-03-01

To improve the CANDU design/operation safety analysis codes and the CANDU safety analysis methodology, the following works have been done. From the development of the lattice codes (WIMS/CANDU), the lattice model simulates the real core lattice geometry and the effect of the pressure tube creep to the core lattice parameter has been evaluated. From the development of the 3-dimensional thermal-hydraulic analysis model of the moderator behavior (CFX4-CAMO), validation of the model against STERN Lab experiment has been executed. The butterfly-shaped grid structure and the 3-dimensional flow resistance model for porous media were developed and applied to the moderator analysis for Wolsong units 2/3/4. The single fuel channel analysis codes for blowdown and post-blowdown were unified by CATHENA. The 3-dimensional fuel channel analysis model (CFX-CACH) has been developed for validation of CATHENA fuel channel analysis model. The interlinking analysis system (CANVAS) of the thermal-hydraulic safety analysis codes for the primary heat transport system and containment system has been executed. The database system of core physics and thermal-hydraulics experimental data for safety analysis has been established on the URL: http://CANTHIS.kaeri.re.kr. For documentation and Standardization of the general safety analysis procedure, the general safety analysis procedure is developed and applied to a large break LOCA. The present research results can be utilized for establishment of the independent safety analysis technology and acquisition of the optimal safety analysis technology

Ignalina NPP Safety Analysis: Models and Results

International Nuclear Information System (INIS)

Uspuras, E.

1999-01-01

Research directions, linked to safety assessment of the Ignalina NPP, of the scientific safety analysis group are presented: Thermal-hydraulic analysis of accidents and operational transients; Thermal-hydraulic assessment of Ignalina NPP Accident Localization System and other compartments; Structural analysis of plant components, piping and other parts of Main Circulation Circuit; Assessment of RBMK-1500 reactor core and other. Models and main works carried out last year are described. (author)

A meta-analysis of personality and workplace safety: addressing unanswered questions.

Science.gov (United States)

Beus, Jeremy M; Dhanani, Lindsay Y; McCord, Mallory A

2015-03-01

[Correction Notice: An Erratum for this article was reported in Vol 100(2) of Journal of Applied Psychology (see record 2015-08139-001). Table 3 contained formatting errors. Minus signs used to indicate negative statistical estimates within the table were inadvertently changed to m-dashes. All versions of this article have been corrected.] The purpose of this meta-analysis was to address unanswered questions regarding the associations between personality and workplace safety by (a) clarifying the magnitude and meaning of these associations with both broad and facet-level personality traits, (b) delineating how personality is associated with workplace safety, and (c) testing the relative importance of personality in comparison to perceptions of the social context of safety (i.e., safety climate) in predicting safety-related behavior. Our results revealed that whereas agreeableness and conscientiousness were negatively associated with unsafe behaviors, extraversion and neuroticism were positively associated with them. Of these traits, agreeableness accounted for the largest proportion of explained variance in safety-related behavior and openness to experience was unrelated. At the facet level, sensation seeking, altruism, anger, and impulsiveness were all meaningfully associated with safety-related behavior, though sensation seeking was the only facet that demonstrated a stronger relationship than its parent trait (i.e., extraversion). In addition, meta-analytic path modeling supported the theoretical expectation that personality's associations with accidents are mediated by safety-related behavior. Finally, although safety climate perceptions accounted for the majority of explained variance in safety-related behavior, personality traits (i.e., agreeableness, conscientiousness, neuroticism) still accounted for a unique and substantive proportion of the explained variance. Taken together, these results substantiate the value of considering personality traits as key

Reload safety analysis automation tools

International Nuclear Information System (INIS)

Havlůj, F.; Hejzlar, J.; Vočka, R.

2013-01-01

Performing core physics calculations for the sake of reload safety analysis is a very demanding and time consuming process. This process generally begins with the preparation of libraries for the core physics code using a lattice code. The next step involves creating a very large set of calculations with the core physics code. Lastly, the results of the calculations must be interpreted, correctly applying uncertainties and checking whether applicable limits are satisfied. Such a procedure requires three specialized experts. One must understand the lattice code in order to correctly calculate and interpret its results. The next expert must have a good understanding of the physics code in order to create libraries from the lattice code results and to correctly define all the calculations involved. The third expert must have a deep knowledge of the power plant and the reload safety analysis procedure in order to verify, that all the necessary calculations were performed. Such a procedure involves many steps and is very time consuming. At ÚJV Řež, a.s., we have developed a set of tools which can be used to automate and simplify the whole process of performing reload safety analysis. Our application QUADRIGA automates lattice code calculations for library preparation. It removes user interaction with the lattice code and reduces his task to defining fuel pin types, enrichments, assembly maps and operational parameters all through a very nice and user-friendly GUI. The second part in reload safety analysis calculations is done by CycleKit, a code which is linked with our core physics code ANDREA. Through CycleKit large sets of calculations with complicated interdependencies can be performed using simple and convenient notation. CycleKit automates the interaction with ANDREA, organizes all the calculations, collects the results, performs limit verification and displays the output in clickable html format. Using this set of tools for reload safety analysis simplifies

Software safety analysis application in installation phase

International Nuclear Information System (INIS)

Huang, H. W.; Yih, S.; Wang, L. H.; Liao, B. C.; Lin, J. M.; Kao, T. M.

2010-01-01

This work performed a software safety analysis (SSA) in the installation phase of the Lungmen nuclear power plant (LMNPP) in Taiwan, under the cooperation of INER and TPC. The US Nuclear Regulatory Commission (USNRC) requests licensee to perform software safety analysis (SSA) and software verification and validation (SV and V) in each phase of software development life cycle with Branch Technical Position (BTP) 7-14. In this work, 37 safety grade digital instrumentation and control (I and C) systems were analyzed by Failure Mode and Effects Analysis (FMEA), which is suggested by IEEE Standard 7-4.3.2-2003. During the installation phase, skew tests for safety grade network and point to point tests were performed. The FMEA showed all the single failure modes can be resolved by the redundant means. Most of the common mode failures can be resolved by operator manual actions. (authors)

Safety analysis, risk assessment, and risk acceptance criteria

International Nuclear Information System (INIS)

Jamali, K.

1997-01-01

This paper discusses a number of topics that relate safety analysis as documented in the Department of Energy (DOE) safety analysis reports (SARs), probabilistic risk assessments (PRA) as characterized primarily in the context of the techniques that have assumed some level of formality in commercial nuclear power plant applications, and risk acceptance criteria as an outgrowth of PRA applications. DOE SARs of interest are those that are prepared for DOE facilities under DOE Order 5480.23 and the implementing guidance in DOE STD-3009-94. It must be noted that the primary area of application for DOE STD-3009 is existing DOE facilities and that certain modifications of the STD-3009 approach are necessary in SARs for new facilities. Moreover, it is the hazard analysis (HA) and accident analysis (AA) portions of these SARs that are relevant to the present discussions. Although PRAs can be qualitative in nature, PRA as used in this paper refers more generally to all quantitative risk assessments and their underlying methods. HA as used in this paper refers more generally to all qualitative risk assessments and their underlying methods that have been in use in hazardous facilities other than nuclear power plants. This discussion includes both quantitative and qualitative risk assessment methods. PRA has been used, improved, developed, and refined since the Reactor Safety Study (WASH-1400) was published in 1975 by the Nuclear Regulatory Commission (NRC). Much debate has ensued since WASH-1400 on exactly what the role of PRA should be in plant design, reactor licensing, 'ensuring' plant and process safety, and a large number of other decisions that must be made for potentially hazardous activities. Of particular interest in this area is whether the risks quantified using PRA should be compared with numerical risk acceptance criteria (RACs) to determine whether a facility is 'safe.' Use of RACs requires quantitative estimates of consequence frequency and magnitude

Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

Energy Technology Data Exchange (ETDEWEB)

Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

2001-02-01

This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines.

Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

International Nuclear Information System (INIS)

Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

2001-02-01

This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines

Status of SPACE Safety Analysis Code Development

International Nuclear Information System (INIS)

Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

2009-01-01

In 2006, the Korean the Korean nuclear industry started developing a thermal-hydraulic analysis code for safety analysis of PWR(Pressurized Water Reactor). The new code is named as SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). The SPACE code can solve two-fluid, three-field governing equations in one dimensional or three dimensional geometry. The SPACE code has many component models required for modeling a PWR, such as reactor coolant pump, safety injection tank, etc. The programming language used in the new code is C++, for new generation of engineers who are more comfortable with C/C++ than old FORTRAN language. This paper describes general characteristics of SPACE code and current status of SPACE code development

New enhancements to SCALE for criticality safety analysis

International Nuclear Information System (INIS)

Hollenbach, D.F.; Bowman, S.M.; Petrie, L.M.; Parks, C.V.

1995-01-01

As the speed, available memory, and reliability of computer hardware increases and the cost decreases, the complexity and usability of computer software will increase, taking advantage of the new hardware capabilities. Computer programs today must be more flexible and user friendly than those of the past. Within available resources, the SCALE staff at Oak Ridge National Laboratory (ORNL) is committed to upgrading its computer codes to keep pace with the current level of technology. This paper examines recent additions and enhancements to the criticality safety analysis sections of the SCALE code package. These recent additions and enhancements made to SCALE can be divided into nine categories: (1) new analytical computer codes, (2) new cross-section libraries, (3) new criticality search sequences, (4) enhanced graphical capabilities, (5) additional KENO enhancements, (6) enhanced resonance processing capabilities, (7) enhanced material information processing capabilities, (8) portability of the SCALE code package, and (9) other minor enhancements, modifications, and corrections to SCALE. Each of these additions and enhancements to the criticality safety analysis capabilities of the SCALE code system are discussed below

Status of generic actions items and safety analysis system of PHWR

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Min, Byung Joo

2001-05-01

This report described the review results of a GAIs(Generic Action Item) currently issued on safety analysis of PHWR(Pressurized Heavy Water Reactor) and the research activities and positions to solve the GAIs in each country which possess PHWRs. eviewing the Final Safety Analysis Report for Wolsong-2/3/4 Units, the safety analysis methodology, classification for accident scenarios, safety analysis codes, their interface, etc.. were described. From the present review report, it is intended to establish the CANDU safety analysis system by providing the better understandings and development plans for the safety analysis of PHWR. esults.

Role of in-house safety analysis and research activities in regulatory decision making

International Nuclear Information System (INIS)

Pradhan, Santosh K.; Nagrale, Dhanesh B.; Gaikwad, Avinash J.

2015-01-01

Achievement of an acceptable level of nuclear safety is an essential requirement for the peaceful utilization of nuclear energy. The success of Global Nuclear Safety Regime is built upon a foundation of research. Such research has been sponsored by Governments and industry and has led to improved designs, safer and more reliable plant operation, and improvements in operating plant efficiency. A key element of this research has been the nuclear safety research performed or sponsored by regulatory organizations. In part, it has been the safety research performed or sponsored by regulatory organizations that has contributed to improved safety and has laid the foundation for activities such as risk-informed regulation, plant life extension, improved plant performance (e.g. power uprates) and new plant designs. The regulatory research program is meant to improve the regulatory authority’s knowledge where uncertainty exists, where safety margins are not well-characterized, and where regulatory decisions need to be confirmed in existing or new designs and technologies. The regulatory body get research initiated either in-house or by the licensee or through technical support organizations (TSOs). Research and analysis carried out within the regulatory body is of immense value in this context. This could be in the form of analysis of safety significant events, analysis of severe accidents, review of operating experience, independent checks of critical designs and even review of operator responses under different situations towards arriving at modifications to training programmes and licensing procedures for operating personnel. A latent benefit of regulatory research carried out by the regulators themselves is that it improves their technical competence considerably which in turn leads to high quality safety reviews and improved regulation in general. The aim of the present paper is to provide an overview of role of regulatory research and the in-house regulatory safety

Safety strategy and safety analysis of nuclear power plants

International Nuclear Information System (INIS)

Franzen, L.F.

1976-01-01

The safety strategy for nuclear power plants is characterized by the fact that the high level of safety was attained not as a result of experience, but on the basis of preventive accident analyses and the finding derived from such analyses. Although, in these accident analyses, the deterministic approach is predominant, it is supplemented by reliability analyses. The accidents analyzed in nuclear licensing procedures cover a wide spectrum from minor incidents to the design basis accidents which determine the design of the safety devices. The initial and boundary conditions, which are essentail for accident analyses, and the determination of the loads occurring in various states during regular operation and in accidents flow into the design of the individual systems and components. The inevitable residual risk and its origins are discussed. (orig.) [de

Technical safety Organisations (TSO) contribute to European Nuclear Safety

International Nuclear Information System (INIS)

Repussard, J.

2010-01-01

Nuclear safety and radiation protection rely on science to achieve high level prevention objectives, through the analysis of safety files proposed by the licensees. The necessary expertise needs to be exercised so as to ensure adequate independence from nuclear operators, appropriate implementation of state of the art knowledge, and a broad spectrum of analysis, adequately ranking the positive and negative points of the safety files. The absence of a Europe-wide nuclear safety regime is extremely costly for an industry which has to cope with a highly competitive and open international environment, but has to comply with fragmented national regulatory systems. Harmonization is therefore critical, but such a goal is difficult to achieve. Only a gradual policy, made up of planned steps in each of the three key dimensions of the problem (energy policy at EU level, regulatory harmonization, consolidation of Europe-wide technical expertise capability) can be successful to achieve the required integration on the basis of the highest safety levels. TSO's contribute to this consolidation, with the support of the EC, in the fields of research (EURATOM-Programmes), of experience feedback analysis (European Clearinghouse), of training and knowledge management (European Training and Tutoring Institute, EUROSAFE). The TSO's network, ETSON, is becoming a formal organisation, able to enter into formal dialogue with EU institutions. However, nuclear safety nevertheless remains a world wide issue, requiring intensive international cooperation, including on TSO issues. (author)

Impact of ITER liquid metal design options on safety level and licensing - Sweden

International Nuclear Information System (INIS)

Harfors, C.; Devell, L.; Johansson, Kjell; Lundell, B.; Rolandsson, S.

1993-01-01

The safety level and licensability of five design options for ITER coolant, breeding material and structural material are assessed, with emphasis on some specified accident scenarios. The safety level is assessed in terms of barrier requirements and the feasibility to construct and qualify such a barrier. The licensability in Sweden of each design option is assessed based on the indicated safety level and on a judgement of the technical feasibility to construct and qualify the ITER tokamak itself, based on the selected design option. 20 refs

Development of safety analysis and constraint detection techniques for process interaction errors

Energy Technology Data Exchange (ETDEWEB)

Fan, Chin-Feng, E-mail: csfanc@saturn.yzu.edu.tw [Computer Science and Engineering Dept., Yuan-Ze University, Taiwan (China); Tsai, Shang-Lin; Tseng, Wan-Hui [Computer Science and Engineering Dept., Yuan-Ze University, Taiwan (China)

2011-02-15

Among the new failure modes introduced by computer into safety systems, the process interaction error is the most unpredictable and complicated failure mode, which may cause disastrous consequences. This paper presents safety analysis and constraint detection techniques for process interaction errors among hardware, software, and human processes. Among interaction errors, the most dreadful ones are those that involve run-time misinterpretation from a logic process. We call them the 'semantic interaction errors'. Such abnormal interaction is not adequately emphasized in current research. In our static analysis, we provide a fault tree template focusing on semantic interaction errors by checking conflicting pre-conditions and post-conditions among interacting processes. Thus, far-fetched, but highly risky, interaction scenarios involve interpretation errors can be identified. For run-time monitoring, a range of constraint types is proposed for checking abnormal signs at run time. We extend current constraints to a broader relational level and a global level, considering process/device dependencies and physical conservation rules in order to detect process interaction errors. The proposed techniques can reduce abnormal interactions; they can also be used to assist in safety-case construction.

Development of safety analysis and constraint detection techniques for process interaction errors

International Nuclear Information System (INIS)

Fan, Chin-Feng; Tsai, Shang-Lin; Tseng, Wan-Hui

2011-01-01

Among the new failure modes introduced by computer into safety systems, the process interaction error is the most unpredictable and complicated failure mode, which may cause disastrous consequences. This paper presents safety analysis and constraint detection techniques for process interaction errors among hardware, software, and human processes. Among interaction errors, the most dreadful ones are those that involve run-time misinterpretation from a logic process. We call them the 'semantic interaction errors'. Such abnormal interaction is not adequately emphasized in current research. In our static analysis, we provide a fault tree template focusing on semantic interaction errors by checking conflicting pre-conditions and post-conditions among interacting processes. Thus, far-fetched, but highly risky, interaction scenarios involve interpretation errors can be identified. For run-time monitoring, a range of constraint types is proposed for checking abnormal signs at run time. We extend current constraints to a broader relational level and a global level, considering process/device dependencies and physical conservation rules in order to detect process interaction errors. The proposed techniques can reduce abnormal interactions; they can also be used to assist in safety-case construction.

Integrated framework for dynamic safety analysis

International Nuclear Information System (INIS)

Kim, Tae Wan; Karanki, Durga R.

2012-01-01

In the conventional PSA (Probabilistic Safety Assessment), detailed plant simulations by independent thermal hydraulic (TH) codes are used in the development of accident sequence models. Typical accidents in a NPP involve complex interactions among process, safety systems, and operator actions. As independent TH codes do not have the models of operator actions and full safety systems, they cannot literally simulate the integrated and dynamic interactions of process, safety systems, and operator responses. Offline simulation with pre decided states and time delays may not model the accident sequences properly. Moreover, when stochastic variability in responses of accident models is considered, defining all the combinations for simulations will be cumbersome task. To overcome some of these limitations of conventional safety analysis approach, TH models are coupled with the stochastic models in the dynamic event tree (DET) framework, which provides flexibility to model the integrated response due to better communication as all the accident elements are in the same model. The advantages of this framework also include: Realistic modeling in dynamic scenarios, comprehensive results, integrated approach (both deterministic and probabilistic models), and support for HRA (Human Reliability Analysis)

OASIS: An automotive analysis and safety engineering instrument

International Nuclear Information System (INIS)

Mader, Roland; Armengaud, Eric; Grießnig, Gerhard; Kreiner, Christian; Steger, Christian; Weiß, Reinhold

2013-01-01

In this paper, we describe a novel software tool named OASIS (AutOmotive Analysis and Safety EngIneering InStrument). OASIS supports automotive safety engineering with features allowing the creation of consistent and complete work products and to simplify and automate workflow steps from early analysis through system development to software development. More precisely, it provides support for (a) model creation and reuse, (b) analysis and documentation and (c) configuration and code generation. We present OASIS as a part of a tool chain supporting the application of a safety engineering workflow aligned with the automotive safety standard ISO 26262. In particular, we focus on OASIS' (1) support for property checking and model correction as well as its (2) support for fault tree generation and FMEA (Failure Modes and Effects Analysis) table generation. Finally, based on the case study of hybrid electric vehicle development, we demonstrate that (1) and (2) are able to strongly support FTA (Fault Tree Analysis) and FMEA

Safety analysis of the UTSI-CFFF superconducting magnet

International Nuclear Information System (INIS)

Turner, L.R.; Wang, S.T.; Smith, R.P.; VanderArend, P.C.; Hsu, Y.H.

1979-01-01

In designing a large superconducting magnet such as the UTSI-CFFF dipole, great attention must be devoted to the safety of the magnet and personnel. The conductor for the UTSI-CFFF magnet incorporates much copper stabilizer, which both insures its cryostability, and contributes to the magnet safety. The quench analysis and the cryostat fault condition analysis are presented. Two analyses of exposed turns follow; the first shows that gas cooling protects uncovered turns; the second, that the cryostat pressure relief system protects them. Finally the failure mode and safety analysis is presented

Road Safety Risk Assessment: An Analysis of Transport Policy and Management for Low-, Middle-, and High-Income Asian Countries

Directory of Open Access Journals (Sweden)

Syyed Adnan Raheel Shah

2018-02-01

Full Text Available Road safety assessment has played a crucial role in the theory and practice of transport management systems. This paper focuses on risk evaluation in the Asian region by exploring the interaction between road safety risk and influencing factors. In the first stage, a data envelopment analysis (DEA method is applied to calculate and rank the road safety risk levels of Asian countries. In the second stage, a structural equation model (SEM with latent variables is applied to analyze the interaction between the road safety risk level and the latent variables, measured by six observed performance indicators, i.e., financial impact, institutional framework, infrastructure and mobility, legislation and policy, vehicular road users, and trauma management. Finally, this paper illustrates the applicability of this DEA-SEM approach for road safety performance analysis.

Updated safety analysis of ITER

Energy Technology Data Exchange (ETDEWEB)

Taylor, Neill, E-mail: neill.taylor@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

2011-10-15

An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

Updated safety analysis of ITER

International Nuclear Information System (INIS)

Taylor, Neill; Baker, Dennis; Ciattaglia, Sergio; Cortes, Pierre; Elbez-Uzan, Joelle; Iseli, Markus; Reyes, Susana; Rodriguez-Rodrigo, Lina; Rosanvallon, Sandrine; Topilski, Leonid

2011-01-01

An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.

Reliability Analysis for Safety Grade PLC(POSAFE-Q)

International Nuclear Information System (INIS)

Choi, Kyung Chul; Song, Seung Whan; Park, Gang Min; Hwang, Sung Jae

2012-01-01

Safety Grade PLC(Programmable Logic Controller), POSAFE-Q, was developed recently in accordance with nuclear regulatory and requirements. In this paper, describe reliability analysis for digital safety grade PLC (especially POSAFE-Q). Reliability analysis scope is Prediction, Calculation of MTBF (Mean Time Between Failure), FMEA (Failure Mode Effect Analysis), PFD (Probability of Failure on Demand). (author)

Status of safety analysis reports

Energy Technology Data Exchange (ETDEWEB)

Cserhati, A

1999-06-01

The safety regulation connected to both of the Atomic Acts from 1980 and 1996 requires preparation of the Preliminary Safety Analysis Report (PSAR) as well as Final SAR (FSAR). In this respect the licensing procedure for the construction and commissioning of Paks NPP did not formally deviate from the standards applied in developed countries; this is particularly true if comparison is made with the standards applied for commissioning NPPs in the second half of the seventies. By the time the overall development of internationally accepted safety standards and some existing deficiencies of earlier SAR made necessary a general reassessment of the plant safety (AGNES project). The carried out PSR for Paks-1 and 2 also added a valuable contribution to the SAR content, however a formal update of SAR is not made yet. A Hungarian nuclear authority decree from 1997 obligates the licensee to prepare and submit a major upgrade of FSAR until the mid of 2000, after finishing the PSR for Paks-3 and 4. From this date a periodic update of FSAR is required every year. The operational license renewal affects only the PSR but not the FSAR updating. The new Nuclear Safety Code outlines the contents of PSAR and FSAR, based on US NRC Reg. Guide 1. 70. Rev. 3. Hungary by now can fulfill the upgrading of SAR without major external technical or financial help. The AGNES project covered the safety analysis chapters of SAR. It was financed mainly by the country. In the project there have been involved in limited cases as performers the VTT (Finland), Belgatom (Belgium), GRS (Germany), etc., the IVO (Finland) fulfilled tasks of an independent reviewer for safety analysis. The AGNES had certain interconnection with the similar IAEA RER safety reassessment project for WWER-440/213. The PSR for Paks-1 and 2 have been carried out by the Paks staff from the resources of the plant. During the evaluation of several parts of Paks-3 and 4 PSR documentation the authority intends to use certain

Status of safety analysis reports

International Nuclear Information System (INIS)

Cserhati, A.

1999-01-01

The safety regulation connected to both of the Atomic Acts from 1980 and 1996 requires preparation of the Preliminary Safety Analysis Report (PSAR) as well as Final SAR (FSAR). In this respect the licensing procedure for the construction and commissioning of Paks NPP did not formally deviate from the standards applied in developed countries; this is particularly true if comparison is made with the standards applied for commissioning NPPs in the second half of the seventies. By the time the overall development of internationally accepted safety standards and some existing deficiencies of earlier SAR made necessary a general reassessment of the plant safety (AGNES project). The carried out PSR for Paks-1 and 2 also added a valuable contribution to the SAR content, however a formal update of SAR is not made yet. A Hungarian nuclear authority decree from 1997 obligates the licensee to prepare and submit a major upgrade of FSAR until the mid of 2000, after finishing the PSR for Paks-3 and 4. From this date a periodic update of FSAR is required every year. The operational license renewal affects only the PSR but not the FSAR updating. The new Nuclear Safety Code outlines the contents of PSAR and FSAR, based on US NRC Reg. Guide 1. 70. Rev. 3. Hungary by now can fulfill the upgrading of SAR without major external technical or financial help. The AGNES project covered the safety analysis chapters of SAR. It was financed mainly by the country. In the project there have been involved in limited cases as performers the VTT (Finland), Belgatom (Belgium), GRS (Germany), etc., the IVO (Finland) fulfilled tasks of an independent reviewer for safety analysis. The AGNES had certain interconnection with the similar IAEA RER safety reassessment project for WWER-440/213. The PSR for Paks-1 and 2 have been carried out by the Paks staff from the resources of the plant. During the evaluation of several parts of Paks-3 and 4 PSR documentation the authority intends to use certain

Safety analysis of the nuclear chemistry Building 151

International Nuclear Information System (INIS)

Kvam, D.

1984-01-01

This report summarizes the results of a safety analysis that was done on Building 151. The report outlines the methodology, the analysis, and the findings that led to the low hazard classification. No further safety evaluation is indicated at this time. 5 tables

DESIGN PACKAGE 1E SYSTEM SAFETY ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

M. Salem

1995-06-23

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

The influence of sodium fires on LMFBRs safety analysis

Energy Technology Data Exchange (ETDEWEB)

Justin, F [DSN/Centre de Fontenay-aux-Roses, Fontenay-aux-Roses (France)

1979-03-01

In a sodium cooled reactor, sodium fires are accidental conditions to be taken into account in safety analysis. For the various sodium categories, fire conditions, associated risks, safety analysis objectives and detailed corresponding issues are indicated, An experimental research program can be deduced from these considerations. This report covers the following: safety analysis methodology; primary sodium fires; secondary sodium fires; auxiliary sodium fires, and related experimental research programs.

The influence of sodium fires on LMFBRs safety analysis

International Nuclear Information System (INIS)

Justin, F.

1979-01-01

In a sodium cooled reactor, sodium fires are accidental conditions to be taken into account in safety analysis. For the various sodium categories, fire conditions, associated risks, safety analysis objectives and detailed corresponding issues are indicated, An experimental research program can be deduced from these considerations. This report covers the following: safety analysis methodology; primary sodium fires; secondary sodium fires; auxiliary sodium fires, and related experimental research programs

Meta-analysis of surgical safety checklist effects on teamwork, communication, morbidity, mortality, and safety.

Science.gov (United States)

Lyons, Vanessa E; Popejoy, Lori L

2014-02-01

The purpose of this study is to examine the effectiveness of surgical safety checklists on teamwork, communication, morbidity, mortality, and compliance with safety measures through meta-analysis. Four meta-analyses were conducted on 19 studies that met the inclusion criteria. The effect size of checklists on teamwork and communication was 1.180 (p = .003), on morbidity and mortality was 0.123 (p = .003) and 0.088 (p = .001), respectively, and on compliance with safety measures was 0.268 (p teamwork and communication, reduce morbidity and mortality, and improve compliance with safety measures. This meta-analysis is limited in its generalizability based on the limited number of studies and the inclusion of only published research. Future research is needed to examine possible moderating variables for the effects of surgical safety checklists.

Reliability analysis of software based safety functions

International Nuclear Information System (INIS)

Pulkkinen, U.

1993-05-01

The methods applicable in the reliability analysis of software based safety functions are described in the report. Although the safety functions also include other components, the main emphasis in the report is on the reliability analysis of software. The check list type qualitative reliability analysis methods, such as failure mode and effects analysis (FMEA), are described, as well as the software fault tree analysis. The safety analysis based on the Petri nets is discussed. The most essential concepts and models of quantitative software reliability analysis are described. The most common software metrics and their combined use with software reliability models are discussed. The application of software reliability models in PSA is evaluated; it is observed that the recent software reliability models do not produce the estimates needed in PSA directly. As a result from the study some recommendations and conclusions are drawn. The need of formal methods in the analysis and development of software based systems, the applicability of qualitative reliability engineering methods in connection to PSA and the need to make more precise the requirements for software based systems and their analyses in the regulatory guides should be mentioned. (orig.). (46 refs., 13 figs., 1 tab.)

Issues affecting advanced passive light-water reactor safety analysis

International Nuclear Information System (INIS)

Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

1992-01-01

Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented

STARS software tool for analysis of reliability and safety

International Nuclear Information System (INIS)

Poucet, A.; Guagnini, E.

1989-01-01

This paper reports on the STARS (Software Tool for the Analysis of Reliability and Safety) project aims at developing an integrated set of Computer Aided Reliability Analysis tools for the various tasks involved in systems safety and reliability analysis including hazard identification, qualitative analysis, logic model construction and evaluation. The expert system technology offers the most promising perspective for developing a Computer Aided Reliability Analysis tool. Combined with graphics and analysis capabilities, it can provide a natural engineering oriented environment for computer assisted reliability and safety modelling and analysis. For hazard identification and fault tree construction, a frame/rule based expert system is used, in which the deductive (goal driven) reasoning and the heuristic, applied during manual fault tree construction, is modelled. Expert system can explain their reasoning so that the analyst can become aware of the why and the how results are being obtained. Hence, the learning aspect involved in manual reliability and safety analysis can be maintained and improved

Strain-Level Metagenomic Analysis of the Fermented Dairy Beverage Nunu Highlights Potential Food Safety Risks.

Science.gov (United States)

Walsh, Aaron M; Crispie, Fiona; Daari, Kareem; O'Sullivan, Orla; Martin, Jennifer C; Arthur, Cornelius T; Claesson, Marcus J; Scott, Karen P; Cotter, Paul D

2017-08-15

The rapid detection of pathogenic strains in food products is essential for the prevention of disease outbreaks. It has already been demonstrated that whole-metagenome shotgun sequencing can be used to detect pathogens in food but, until recently, strain-level detection of pathogens has relied on whole-metagenome assembly, which is a computationally demanding process. Here we demonstrated that three short-read-alignment-based methods, i.e., MetaMLST, PanPhlAn, and StrainPhlAn, could accurately and rapidly identify pathogenic strains in spinach metagenomes that had been intentionally spiked with Shiga toxin-producing Escherichia coli in a previous study. Subsequently, we employed the methods, in combination with other metagenomics approaches, to assess the safety of nunu, a traditional Ghanaian fermented milk product that is produced by the spontaneous fermentation of raw cow milk. We showed that nunu samples were frequently contaminated with bacteria associated with the bovine gut and, worryingly, we detected putatively pathogenic E. coli and Klebsiella pneumoniae strains in a subset of nunu samples. Ultimately, our work establishes that short-read-alignment-based bioinformatics approaches are suitable food safety tools, and we describe a real-life example of their utilization. IMPORTANCE Foodborne pathogens are responsible for millions of illnesses each year. Here we demonstrate that short-read-alignment-based bioinformatics tools can accurately and rapidly detect pathogenic strains in food products by using shotgun metagenomics data. The methods used here are considerably faster than both traditional culturing methods and alternative bioinformatics approaches that rely on metagenome assembly; therefore, they can potentially be used for more high-throughput food safety testing. Overall, our results suggest that whole-metagenome sequencing can be used as a practical food safety tool to prevent diseases or to link outbreaks to specific food products. Copyright �

Probabilistic analysis of safety in industrial irradiation plants

International Nuclear Information System (INIS)

Alderete, F.; Elechosa, C.

2006-01-01

of the components and related systems with the radiological safety. The advantage of the application of the APS is the high grade of knowledge of the installation that is acquired. For that even with a qualitative analysis it is possible to verify the safety level of the same ones and to identify the areas more advisable and susceptible of improvement, being able to apply that learned to the licensing of new facilities. (Author)

Technical Issues and Proposes on the Legislation of Probabilistic Safety Assessment in Periodic Safety Review

International Nuclear Information System (INIS)

Hwang, Seok-Won; Jeon, Ho-Jun; Na, Jang-Hwan

2015-01-01

Korean Nuclear Power Plants have performed a comprehensive safety assessment reflecting design and procedure changes and using the latest technology every 10 years. In Korea, safety factors of PSR are revised to 14 by revision of IAEA Safety Guidelines in 2003. In the revised safety guidelines, safety analysis field was subdivided into deterministic safety analysis, PSA (Probabilistic safety analysis), and hazard analysis. The purpose to examine PSA as a safety factor on PSR is to make sure that PSA results and assumptions reflect the latest state of NPPs, validate the level of computer codes and analytical models, and evaluate the adequacy of PSA instructions. In addition, its purpose is to derive the plant design change, operating experience of other plants and safety enhancement items as well. In Korea, PSA is introduced as a new factor. Thus, the overall guideline development and long-term implementation strategy are needed. Today in Korea, full-power PSA model revision and low-power and shutdown (LPSD) PSA model development is being performed as a part of the post Fukushima action items for operating plants. The scope of the full-power PSA is internal/external level 1, 2 PSA. But in case of fire PSA, the scope is level 1 PSA using new method, NUREG/CR-6850. In case of LPSD PSA, level 1 PSA for all operating plants, and level 2 PSA for 2 demonstration plants are under development. The result of the LPSD PSA will be used as major input data for plant specific SAMG (Severe Accident Management Guideline). The scope of PSA currently being developed in Korea cannot fulfill 'All Mode, All Scope' requirements recommended in the IAEA Safety Guidelines. Besides the legislation of PSA, step-by-step development strategy for non-performed scopes such as level 3 PSA and new fire PSA is one of the urgent issues in Korea. This paper suggests technical issues and development strategies for each PSA technical elements.

Construction Safety And Health Factors At The Industry Level: The Case Of Singapore

Directory of Open Access Journals (Sweden)

Charles Y.J. Cheah

2007-12-01

Full Text Available The construction industry is one of the most hazardous industries due to the unique nature of its products and the processes involved. Recent occurrences of highly publicized construction site accidents in Singapore have highlighted the immediate needs for the local industry to address safety and attention at the industry level. The objective of this paper is to examine issues and critical factors affecting S&H standards in Singapore. Clearly, collective efforts should be pursued at the industry level as the country moves towards the ultimate safety management strategy of self-regulation. The findings also indicate that the challenge of making worksites safe should not be placed solely on the contractors but should be shared by all parties affecting the value chain of construction, including the developers, the consultants and the government. The factors identified through factor analysis may inform legislators and industry practitioners in terms of the sources of problems and help develop effective strategies for improvement. Some of the experiences mentioned in the paper could also be relevant to other countries facing similar circumstances.

Probabilistic safety assessment of Tehran Research Reactor using systems analysis programs for hands-on integrated reliability evaluations

International Nuclear Information System (INIS)

Hosseini, M.H.; Nematollahi, M.R.; Sepanloo, K.

2004-01-01

Probabilistic safety assessment application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this document the application of the probabilistic safety assessment to the Tehran Research Reactor is presented. The level 1 practicabilities safety assessment application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantifications, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using systems analysis programs for hands-on integrated reliability evaluations software

Safety analysis methodology for OPR 1000

International Nuclear Information System (INIS)

Hwang-Yong, Jun

2005-01-01

Full text: Korea Electric Power Research Institute (KEPRI) has been developing inhouse safety analysis methodology based on the delicate codes available to KEPRI to overcome the problems arising from currently used vendor oriented methodologies. For the Loss of Coolant Accident (LOCA) analysis, the KREM (KEPRI Realistic Evaluation Methodology) has been developed based on the RELAP-5 code. The methodology was approved for the Westinghouse 3-loop plants by the Korean regulatory organization and the project to extent the methodology to the Optimized Power Reactor 1000 (OPR1000) has been ongoing since 2001. Also, for the Non-LOCA analysis, the KNAP (Korea Non-LOCA Analysis Package) has been developed using the UNICORN-TM code system. To demonstrate the feasibility of these codes systems and methodologies, some typical cases of the design basis accidents mentioned in the final safety analysis report (FSAR) were analyzed. (author)

Project Guarantee 1985. Final repository for low- and intermediate level radioactive wastes: Safety report

International Nuclear Information System (INIS)

Anon.

1985-01-01

Storage of radioactive waste must delay the return of radionuclides to the biosphere for a long period of time and must maintain the release rates at a sufficiently low level for all time. This is achieved with the aid of a series of safety barriers which consist, on the one hand, of technical barriers in the repository and, on the other hand , of natural geological barriers as they occur at the repository location. In order to assess the efficiency of the barriers, the working methods of the technical barriers and the host rock must be understood. This understanding is transferred into quantitative models in order to calculate the safety of the repository. The individual barriers and the methods used to modelling their functions were described in volume NGB 85-07 of the Project Guarantee 1985 report series and the data necessary for modelling were given. The models and data are used in the safety analysis, the results of which are contained in the present report. Safety considerations show that models are available in Switzerland which allow, in principle, an assessment of the long-term behaviour of a repository for low- and intermediate-level waste. The evaluation of earlier studies and experimental work, suitable laboratory measurements and results from field research enable compilation of a representative data-set so that the requirements for quantitative statements on safety of final disposal are met from this side also. The safety calculations show that the radiation doses calculated for a base case scenario with realistic/conservative parameter values are negligibly low. Also, radiation doses which are clearly under the protection standard of 10 mrem per year result for conservative values and the cumulation of several conservative assumptions. Even assuming exposure of the repository by erosion, a radiotoxicity of the soil formed results which is under natural values

Swimming pool reactor reliability and safety analysis

International Nuclear Information System (INIS)

Li Zhaohuan

1997-01-01

A reliability and safety analysis of Swimming Pool Reactor in China Institute of Atomic Energy is done by use of event/fault tree technique. The paper briefly describes the analysis model, analysis code and main results. Meanwhile it also describes the impact of unassigned operation status on safety, the estimation of effectiveness of defense tactics in maintenance against common cause failure, the effectiveness of recovering actions on the system reliability, the comparison of occurrence frequencies of the core damage by use of generic and specific data

Safety analysis of the present status of the research reactor 'RA' at 'Vinca' Institute

International Nuclear Information System (INIS)

Jovic, V.; Jovic, L.; Zivotic, Z.; Milovanovic, Dj.

1995-01-01

Safety analysis of the nuclear facility which has been out of work for a long time and whose future is not defined at the present moment, can not be connected to the usual, normatively regulated system analysis procedure in both operational and accidental regimes. Therefore, the safety analysis of the present status of the present status of the reactor RA is related to system and components analysis which, in present conditions maintain their nuclear functions operational. In the first place, it refers to components and equipment in which radioactive radiation generation still exists and to installations and equipment maintaining radiation level below permitted limit. in the context of the analysis the following areas are being covered: present status characteristics, accidental events while operating period from 1959. to 1984., nuclear fuels and radioactive waste inventory, basic characteristics and status of safety-related systems and equipment, radiation protection, potential accident analysis at present status of the reactor RA, potential accidental situations due to natural events (earthquakes, water flood) or man-induced events and security. 8 refs

Preliminary safety analysis of unscrammed events for KLFR

International Nuclear Information System (INIS)

Kim, S.J.; Ha, G.S.

2005-01-01

The report presents the design features of KLFR; Safety Analysis Code; steady-state calculation results and analysis results of unscrammed events. The calculations of the steady-state and unscrammed events have been performed for the conceptual design of KLFR using SSC-K code. UTOP event results in no fuel damage and no centre-line melting. The inherent safety features are demonstrated through the analysis of ULOHS event. Although the analysis of ULOF has much uncertainties in the pump design, the analysis results show the inherent safety characteristics. 6% flow of rated flow of natural circulation is formed in the case of ULOF. In the metallic fuel rod, the cladding temperature is somewhat high due to the low heat transfer coefficient of lead. ULOHS event should be considered in design of RVACS for long-term cooling

SACS2: Dynamic and Formal Safety Analysis Method for Complex Safety Critical System

International Nuclear Information System (INIS)

Koh, Kwang Yong; Seong, Poong Hyun

2009-01-01

Fault tree analysis (FTA) is one of the most widely used safety analysis technique in the development of safety critical systems. However, over the years, several drawbacks of the conventional FTA have become apparent. One major drawback is that conventional FTA uses only static gates and hence can not capture dynamic behaviors of the complex system precisely. Although several attempts such as dynamic fault tree (DFT), PANDORA, formal fault tree (FFT) and so on, have been made to overcome this problem, they can not still do absolute or actual time modeling because they adapt relative time concept and can capture only sequential behaviors of the system. Second drawback of conventional FTA is its lack of rigorous semantics. Because it is informal in nature, safety analysis results heavily depend on an analyst's ability and are error-prone. Finally reasoning process which is to check whether basic events really cause top events is done manually and hence very labor-intensive and timeconsuming for the complex systems. In this paper, we propose a new safety analysis method for complex safety critical system in qualitative manner. We introduce several temporal gates based on timed computational tree logic (TCTL) which can represent quantitative notion of time. Then, we translate the information of the fault trees into UPPAAL query language and the reasoning process is automatically done by UPPAAL which is the model checker for time critical system

MSSV Modeling for Wolsong-1 Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Moon, Bok Ja; Choi, Chul Jin; Kim, Seoung Rae [KEPCO EandC, Daejeon (Korea, Republic of)

2010-10-15

The main steam safety valves (MSSVs) are installed on the main steam line to prevent the overpressurization of the system. MSSVs are held in closed position by spring force and the valves pop open by internal force when the main steam pressure increases to open set pressure. If the overpressure condition is relieved, the valves begin to close. For the safety analysis of anticipated accident condition, the safety systems are modeled conservatively to simulate the accident condition more severe. MSSVs are also modeled conservatively for the analysis of over-pressurization accidents. In this paper, the pressure transient is analyzed at over-pressurization condition to evaluate the conservatism for MSSV models

Level of knowledge among the population of radiation safety basic issues

Directory of Open Access Journals (Sweden)

S. A. Zelencova

2015-01-01

Full Text Available The goal of research was to determine the level of knowledge among the population on issues like sources of ionising radiation, methods of ionising radiation measurement, measures of self-protection in case of threating or actual radioactive pollution in the district, and to study self-estimation by the population of their knowledge of radiation safety issues. Research was carried out using the method of questioning of population groups in three regions close to the places of previous peaceful nuclear explosions (Arkhangelsk, Murmansk and Tyumen regions, and in five Far East regions of the Russian Federation (Kamchatka, Khabarovsk, Primorsky, Magadan and South-Sakhalin regions after radiation accident in Japan at "Fukushima-1" NPP. This research included processing of 243 questionnaires from the regions close to places of previous peaceful nuclear explosions and 216 questionnaires from the Far East regions.The analysis of obtained questioning results enabled to make the following conclusions: the level of knowledge among the population about the basic concepts of radiation safety appeared to be generally low among respondents of all eight territories. Considerable number of respondents in seven groups correctly mentioned the x-ray device as a source of ionising radiation (from 71 to 88 % of answers. In Murmansk region – only 52 % of the answers. Respondents of the same seven groups often correctly answered the question on how to detect ionising radiation (only with devices – from 68 to 98 % in different groups. The smallest number of correct answers to this question (42 % is also noted among respondents from the Murmansk region.Level of knowledge on self-protection measures at threating or actual radioactive pollution of the places of residence appeared a little higher among the Far East region population, who had actual concerns regarding the threat of radioactive pollution at the present time. However, in all eight investigated groups

2005 dossier: granite. Tome: safety analysis of the geologic disposal; Dossier 2005: granite. Tome analyse de surete du stockage geologique

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-07-01

This document makes a status of the researches carried out by the French national agency of radioactive wastes (ANDRA) about the safety aspects of the geologic disposal of high-level and long-lived (HLLL) radioactive wastes in granite formations. Content: 1 - safety approach: context and general goal, references, design approach by safety functions, safety approach during the construction-exploitation-observation-closure phase, safety analysis during the post-closure phase; 2 - general description: HLLL wastes, granitic environment, general structure of the architecture of a disposal facility; 3 - safety functions and disposal design: general context, safety functions of the long-term disposal, design dispositions retained to answer the functions; 4 - operational safety: people's protection, radiological risks during exploitation, risk analysis in accident situation; 5 - qualitative safety analysis: methodology, main results of the analysis of the features, events and processes (FEP) database; 6 - disposal efficiency evaluation during post-closure phase: calculation models, calculation tools used for the modeling of radionuclides transport, calculation results and main lessons. (J.S.)

2005 dossier: granite. Tome: safety analysis of the geologic disposal; Dossier 2005: granite. Tome analyse de surete du stockage geologique

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-07-01

This document makes a status of the researches carried out by the French national agency of radioactive wastes (ANDRA) about the safety aspects of the geologic disposal of high-level and long-lived (HLLL) radioactive wastes in granite formations. Content: 1 - safety approach: context and general goal, references, design approach by safety functions, safety approach during the construction-exploitation-observation-closure phase, safety analysis during the post-closure phase; 2 - general description: HLLL wastes, granitic environment, general structure of the architecture of a disposal facility; 3 - safety functions and disposal design: general context, safety functions of the long-term disposal, design dispositions retained to answer the functions; 4 - operational safety: people's protection, radiological risks during exploitation, risk analysis in accident situation; 5 - qualitative safety analysis: methodology, main results of the analysis of the features, events and processes (FEP) database; 6 - disposal efficiency evaluation during post-closure phase: calculation models, calculation tools used for the modeling of radionuclides transport, calculation results and main lessons. (J.S.)

Documented Safety Analysis for the B695 Segment

Energy Technology Data Exchange (ETDEWEB)

Laycak, D

2008-09-11

This Documented Safety Analysis (DSA) was prepared for the Lawrence Livermore National Laboratory (LLNL) Building 695 (B695) Segment of the Decontamination and Waste Treatment Facility (DWTF). The report provides comprehensive information on design and operations, including safety programs and safety structures, systems and components to address the potential process-related hazards, natural phenomena, and external hazards that can affect the public, facility workers, and the environment. Consideration is given to all modes of operation, including the potential for both equipment failure and human error. The facilities known collectively as the DWTF are used by LLNL's Radioactive and Hazardous Waste Management (RHWM) Division to store and treat regulated wastes generated at LLNL. RHWM generally processes low-level radioactive waste with no, or extremely low, concentrations of transuranics (e.g., much less than 100 nCi/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., {sup 90}Sr, {sup 137}Cs, or {sup 3}H. The mission of the B695 Segment centers on container storage, lab-packing, repacking, overpacking, bulking, sampling, waste transfer, and waste treatment. The B695 Segment is used for storage of radioactive waste (including transuranic and low-level), hazardous, nonhazardous, mixed, and other waste. Storage of hazardous and mixed waste in B695 Segment facilities is in compliance with the Resource Conservation and Recovery Act (RCRA). LLNL is operated by the Lawrence Livermore National Security, LLC, for the Department of Energy (DOE). The B695 Segment is operated by the RHWM Division of LLNL. Many operations in the B695 Segment are performed under a Resource Conservation and Recovery Act (RCRA) operation plan, similar to commercial treatment operations with best demonstrated available technologies. The buildings of the B695 Segment were designed and built considering such operations, using proven building

Documented Safety Analysis for the B695 Segment

International Nuclear Information System (INIS)

Laycak, D.

2008-01-01

This Documented Safety Analysis (DSA) was prepared for the Lawrence Livermore National Laboratory (LLNL) Building 695 (B695) Segment of the Decontamination and Waste Treatment Facility (DWTF). The report provides comprehensive information on design and operations, including safety programs and safety structures, systems and components to address the potential process-related hazards, natural phenomena, and external hazards that can affect the public, facility workers, and the environment. Consideration is given to all modes of operation, including the potential for both equipment failure and human error. The facilities known collectively as the DWTF are used by LLNL's Radioactive and Hazardous Waste Management (RHWM) Division to store and treat regulated wastes generated at LLNL. RHWM generally processes low-level radioactive waste with no, or extremely low, concentrations of transuranics (e.g., much less than 100 nCi/g). Wastes processed often contain only depleted uranium and beta- and gamma-emitting nuclides, e.g., 90 Sr, 137 Cs, or 3 H. The mission of the B695 Segment centers on container storage, lab-packing, repacking, overpacking, bulking, sampling, waste transfer, and waste treatment. The B695 Segment is used for storage of radioactive waste (including transuranic and low-level), hazardous, nonhazardous, mixed, and other waste. Storage of hazardous and mixed waste in B695 Segment facilities is in compliance with the Resource Conservation and Recovery Act (RCRA). LLNL is operated by the Lawrence Livermore National Security, LLC, for the Department of Energy (DOE). The B695 Segment is operated by the RHWM Division of LLNL. Many operations in the B695 Segment are performed under a Resource Conservation and Recovery Act (RCRA) operation plan, similar to commercial treatment operations with best demonstrated available technologies. The buildings of the B695 Segment were designed and built considering such operations, using proven building systems, and keeping

Economic consideration of nuclear safety and cost benefit analysis in nuclear safety regulation

International Nuclear Information System (INIS)

Choi, Y. S.; Choi, K. S.; Choi, K. W.; Song, I. J.; Park, D. K.

2001-01-01

For the optimization of nuclear safety regulation, understanding of economic aspects of it becomes increasingly important together with the technical approach used so far to secure nuclear safety. Relevant economic theories on private and public goods were reviewed to re-illuminate nuclear safety from the economic perspective. The characteristics of nuclear safety as a public good was reviewed and discussed in comparison with the car safety as a private safety good. It was shown that the change of social welfare resulted from the policy change induced can be calculated by the summation of compensating variation(CV) of individuals. It was shown that the value of nuclear safety could be determined in monetary term by this approach. The theoretical background and history of cost benefit analysis of nuclear safety regulation were presented and topics for future study were suggested

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.

Safety culture and accident analysis-A socio-management approach based on organizational safety social capital

International Nuclear Information System (INIS)

Rao, Suman

2007-01-01

One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, Key drivers of safety culture in many organizations are regulation, audits, safety training, various types of employee exhortations to comply with safety norms, etc. However, less evident factors like networking relationships and social trust amongst employees, as also extended networking relationships and social trust of organizations with external stakeholders like government, suppliers, regulators, etc., which constitute the safety social capital in the Organization-seem to also influence the sustenance of organizational safety culture. Can erosion in safety social capital cause deterioration in safety culture and contribute to accidents? If so, how does it contribute? As existing accident analysis models do not provide answers to these questions, CAMSoC (Curtailing Accidents by Managing Social Capital), an accident analysis model, is proposed. As an illustration, five accidents: Bhopal (India), Hyatt Regency (USA), Tenerife (Canary Islands), Westray (Canada) and Exxon Valdez (USA) have been analyzed using CAMSoC. This limited cross-industry analysis provides two key socio-management insights: the biggest source of motivation that causes deviant behavior leading to accidents is 'Faulty Value Systems'. The second biggest source is 'Enforceable Trust'. From a management control perspective, deterioration in safety culture and resultant accidents is more due to the 'action controls' rather than explicit 'cultural controls'. Future research directions to enhance the model's utility through layering are addressed briefly

The PEC reactor. Safety analysis: Detailed reports

Energy Technology Data Exchange (ETDEWEB)

1988-01-01

In the safety-analysis of the PEC Brasimone reactor (Italy), attention was focused on the role of plant-incident analysis during the design stage and the conclusions reached. The analysis regarded the following: thermohydraulic incidents at full power; incidents with the reactor shut down; reactivity incidents; core local faults; analysis of fuel-handling incidents; engineered safeguards and passive safety features; coolant leakage and sodium fires; research and development studies on the seismic behaviour of the PEC fast reactor; generalized sodium fire; severe accidents, accident sequences with shudown; reference accident. Both the theoretical and experimental analyses demonstrated the adequacy of the design of the PEC fast reactor, aimed at minimizing the consequences of a hypothetical disruptive core accident with mechanical energy release. It was shown that the containment barriers were sized correctly and that the residual heat from a disassembled core would be removed. The re-evaluation of the source term emphasized the conservative nature of the hypotheses assumed in the preliminary safety analysis for calculating the risk to the public.

Safety of geologic disposal of high level radioactive waste

International Nuclear Information System (INIS)

Zaitsu, Tomohisa; Ishiguro, Katsuhiko; Masuda, Sumio

1992-01-01

This article introduces current concepts of geologic disposal of high level radioactive waste and its safety. High level radioactive waste is physically stabilized by solidifying it in a glass form. Characteristics of deep geologic layer are presented from the viewpoint of geologic disposal. Reconstruction of multi-barrier system receives much attention to secure the safety of geologic disposal. It is important to research performance assessment of multi-barrier system for preventing dissolution or transfer of radionuclides into the ground water. Physical and chemical modeling for the performance assessment is outlined in the following terms: (1) chemical property of deep ground water, (2) geochemical modeling of artificial barrier spatial water, (3) hydrology of deep ground water, (4) hydrology of the inside of artificial barrier, and (5) modeling of radionuclide transfer from artificial barrier. (N.K.)

Low- and Intermediate Level Radioactive Waste Disposal Environmental and Safety Assessment Activities in Slovenia

International Nuclear Information System (INIS)

Marc, D.; Loose, A.; Urbanc, J.

1998-01-01

The protection of the environment is one of the main concerns in the management of radioactive waste, especially in repository planning. In different stages of repository lifetime the environmental assessment has different functions: it can be used as a decision making process and as a planning, communication and management tool. Safety assessment as a procedure for evaluating the performance of a disposal system, and its potential radiological impact on human health and environment, is also required. Following the international recommendations and Slovene legislation, a presentation is given of the role and importance of the environmental and safety assessment activities in the early stages following concept development and site selection for a low- and intermediate level radioactive waste (LILW) repository in Slovenia. As a case study, a short overview is also given of the preliminary safety assessment that has been carried out in the analysis of possibilities for long-lived LILW disposal in Slovenia. (author)

DOE high-level waste tank safety program Final report, Task 002

International Nuclear Information System (INIS)

1998-01-01

The overall objective of the work on Task 002 was to provide LANL with support to the DOE High-Level Waste Tank Safety program. The objective of the work was to develop safety documentation in support of the unsafe tank mitigation activities at Hanford. The work includes the development of safety assessment and an environmental assessment. All tasks which were assigned under this Task Order were completed. Descriptions of the objectives of each task and effort performed to complete each objective are provided. The two tasks were: Task 2.1--safety assessment for instrumentation insertion; and Task 2.2--environmental assessment

Optimum Safety Levels and Design Rules for the Icelandic Type Berm Breakwater

DEFF Research Database (Denmark)

Sigurdarson, Sigurdur; van der Meer, Jentsje W.; Burcharth, Hans F.

2007-01-01

Guidance on selection of breakwater types and related design safety levels for breakwaters are almost non-existent, which is the reason that PIANC has initiated working group 47 on this subject. This paper presents ongoing work particulary on the Icelandic type berm breakwater within the PIANC...... working group. It will concentrate on design guidance and on the optimum safety levels for this type of structure....

Safety analysis of disposal of decommissioning waste from the Olkiluoto nuclear power plant - PURKU-93

International Nuclear Information System (INIS)

Vieno, T.; Meszaros, F.; Nordman, H.; Taivassalo, V.

1993-12-01

Decommissioning waste from the Olkiluoto nuclear power plant will be disposed of at the depth between 60 and 100 meters in the bedrock at the power plant site. The existing VLJ repository for low and medium level operating waste will be extended with three new silos for the decommissioning waste of the TVO I and II reactors and the spent fuel interim store at the Olkiluoto site. Besides dismantling waste also used fuel boxes, control rods and other activated metal components accumulated during the operation of the reactors will be disposed of in the repository. The safety analysis is based on the detailed decommissioning plan of the Olkiluoto power plants and the comprehensive safety analysis carried out for the Final Safety Analysis Report of the VLJ repository. (58 refs., 31 figs., 38 tabs.)

Evaluation of health and safety impacts of defense high-level waste in geologic repositories

International Nuclear Information System (INIS)

Smith, E.D.; Kocher, D.C.; Witherspoon, J.P.

1985-02-01

Pursuant to the requirement of the Nuclear Waste Policy Act of 1982 that the President evaluate the use of commercial high-level waste repositories for the disposal of defense high-level wastes, a comparative assessment has been performed of the potential health and safety impacts of disposal of defense wastes in commercial or defense-only repositories. Simplified models were used to make quantitative estimates of both long- and short-term health and safety impacts of several options for defense high-level waste disposal. The results indicate that potential health and safety impacts are not likely to vary significantly among the different disposal options for defense wastes. Estimated long-term health and safety impacts from all defense-waste disposal options are somewhat less than those from commercial waste disposal, while short-term health and safety impacts appear to be insensitive to the differences between defense and commercial wastes. In all cases, potential health and safety impacts are small because of the need to meet stringent standards promulgated by the US Environmental Protection Agency and the US Nuclear Regulatory Commission. We conclude that health and safety impacts should not be a significant factor in the choice of a disposal option for defense high-level wastes. 20 references, 14 tables

Predisposal Management of Low and Intermediate Level Radioactive Waste. Safety Guide

International Nuclear Information System (INIS)

2009-01-01

The objective of this Safety Guide is to provide regulatory bodies and the operators that generate and manage radioactive waste with recommendations on how to meet the principles and requirements established for the predisposal management of low and intermediate level waste. Contents: 1. Introduction; 2. Protection of human health and the environment; 3. Roles and responsibilities; 4. General safety considerations; 5. Safety features for the predisposal management of LILW; 6. Record keeping and reporting; 7. Safety assessment; 8. Quality assurance; Annex I: Nature and sources of LILW from nuclear facilities; Annex II: Development of specifications for waste packages; Annex III: Site conditions, processes and events for consideration in a safety assessment (external natural phenomena); Annex IV: Site conditions, processes and events for consideration in a safety assessment (external human induced phenomena); Annex V: Postulated initiating events for consideration in a safety assessment (internal phenomena).

Safety during sea transport of radioactive materials. Probabilistic safety analysis of package fro sea surface fire accident

International Nuclear Information System (INIS)

Matsuoka, Takeshi; Obara, Isonori; Akutsu, Yukio; Aritomi, Masanori

2000-01-01

The ships carrying irradiated nuclear fuel, plutonium and high level radioactive wastes(INF materials) are designed to keep integrity of packaging based on the various safety and fireproof measures, even if the ship encounters a maritime fire accident. However, granted that the frequency is very low, realistic severe accidents should be evaluated. In this paper, probabilistic safety assessment method is applied to evaluate safety margin for severe sea fire accidents using event tree analysis. Based on our separate studies, the severest scenario was estimated as follows; an INF transport ship collides with oil tanker and induces a sea surface fire. Probability data such as ship's collision, oil leakage, ignition, escape from fire region, operations of cask cooling system and water flooding systems were also introduced from above mentioned studies. The results indicate that the probability of which packages cannot keep their integrity during the sea surface fire accident is very low and sea transport of INF materials is carried out very safely. (author)

Analysis on relation between safety input and accidents

Institute of Scientific and Technical Information of China (English)

YAO Qing-guo; ZHANG Xue-mu; LI Chun-hui

2007-01-01

The number of safety input directly determines the level of safety, and there exists dialectical and unified relations between safety input and accidents. Based on the field investigation and reliable data, this paper deeply studied the dialectical relationship between safety input and accidents, and acquired the conclusions. The security situation of the coal enterprises was related to the security input rate, being effected little by the security input scale, and build the relationship model between safety input and accidents on this basis, that is the accident model.

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．

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.)

Optimization of safety equipment outages improves safety

International Nuclear Information System (INIS)

Cepin, Marko

2002-01-01

Testing and maintenance activities of safety equipment in nuclear power plants are an important potential for risk and cost reduction. An optimization method is presented based on the simulated annealing algorithm. The method determines the optimal schedule of safety equipment outages due to testing and maintenance based on minimization of selected risk measure. The mean value of the selected time dependent risk measure represents the objective function of the optimization. The time dependent function of the selected risk measure is obtained from probabilistic safety assessment, i.e. the fault tree analysis at the system level and the fault tree/event tree analysis at the plant level, both extended with inclusion of time requirements. Results of several examples showed that it is possible to reduce risk by application of the proposed method. Because of large uncertainties in the probabilistic safety assessment, the most important result of the method may not be a selection of the most suitable schedule of safety equipment outages among those, which results in similarly low risk. But, it may be a prevention of such schedules of safety equipment outages, which result in high risk. Such finding increases the importance of evaluation speed versus the requirement of getting always the global optimum no matter if it is only slightly better that certain local one

Predisposal management of low and intermediate level radioactive waste. Safety guide

International Nuclear Information System (INIS)

2003-01-01

considered in this publication begins with the refining and conversion of uranium concentrates. Recommendations on the management of radioactive waste from the mining and milling of uranium and thorium ores are provided. Some parts of the nuclear fuel cycle generate both high level waste and LLW. The management of high level waste itself generates LLW. The predisposal management of this LLW is included in the scope of this Safety Guide. Recommendations on the predisposal management of high level waste are provided. The recommendations in this Safety Guide primarily concern complex management activities for LLW. The regulatory body should decide which parts of this Safety Guide are relevant and appropriate for particular circumstances, and the extent to which the recommendations and guidance apply. This Safety Guide provides only introductory material on the transport and storage of LLW. Requirements and recommendations are provided in Refs. There may be non-radiological hazards associated with the predisposal management of LLW. Some guidance is given on the safety measures to be taken against non-radiological hazards if they have potential consequences for radiation safety. However, detailed recommendations are beyond the scope of this Safety Guide. The user should seek guidance from the regulatory body in the areas of health and safety and environmental protection

Safety systems and safety analysis of the Qinshan phase III CANDU nuclear power plant

International Nuclear Information System (INIS)

Cai Jianping; Shen Sen; Barkman, N.

1999-01-01

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

Do we need an integrative approach to food safety at the country level?

International Nuclear Information System (INIS)

Ristic, G.

2002-01-01

Scientific data show increasing evidence of relationship between food safety and food standards on one hand and public health concern on the other hand. In FR Yugoslavia in 1989 the system of reporting on food safety issues on federal and republic level was established. The system provides data on laboratory analysis of 22 food items (bread, milk, meat and meat products, vegetables, processed vegetables etc). Those items were and still are tested on food quality and safety parameters such as microbiological, chemical and radio nuclides. Seldom all required testing on chemical and radio nuclides are performed, so we lack exact risk assessment for those contaminants. Further, during war conflict in FR Yugoslavia and also due to industrial hazards in neighbouring countries (Rumania, Hungary) high quantities of PCBs, dioxins, heavy metals, arsenic compounds and other toxic compounds contaminated the environment. In the soil and in some food products (animal fats predominantly) radionuclides originating from Chernobyl hazard can still be detected. In order to identify the level of exposure to chemical and radio nuclide contaminants in the food chain it is essential to test intensively and systematically food from animal and from plant origin. In order to prevent entering the contaminants to the food chain new recommendations from WHO, FAO and EU suggest implementation of integrative approach to food safety and control over the whole chain of food production from 'farm to table'. This approach provides control of the contaminants in soil, water, air, control over primary food production (covering animal feed too), intensive control over processing with implementation of HACCP system, but also, over transportation, retail trade, street food and home made food too. In our country creation of the map of the polluted areas, and actions in order to treat the pollution should accompany implementation of this new food safety system. The need for assessment of the level of

Do we need an integrative approach to food safety at the country level?

Energy Technology Data Exchange (ETDEWEB)

Ristic, G E-mail:risticg@eunet yu [Department of Nutrition, Medical Faculty, Belgrade (Yugoslavia)

2002-05-01

Scientific data show increasing evidence of relationship between food safety and food standards on one hand and public health concern on the other hand. In FR Yugoslavia in 1989 the system of reporting on food safety issues on federal and republic level was established. The system provides data on laboratory analysis of 22 food items (bread, milk, meat and meat products, vegetables, processed vegetables etc). Those items were and still are tested on food quality and safety parameters such as microbiological, chemical and radio nuclides. Seldom all required testing on chemical and radio nuclides are performed, so we lack exact risk assessment for those contaminants. Further, during war conflict in FR Yugoslavia and also due to industrial hazards in neighbouring countries (Rumania, Hungary) high quantities of PCBs, dioxins, heavy metals, arsenic compounds and other toxic compounds contaminated the environment. In the soil and in some food products (animal fats predominantly) radionuclides originating from Chernobyl hazard can still be detected. In order to identify the level of exposure to chemical and radio nuclide contaminants in the food chain it is essential to test intensively and systematically food from animal and from plant origin. In order to prevent entering the contaminants to the food chain new recommendations from WHO, FAO and EU suggest implementation of integrative approach to food safety and control over the whole chain of food production from 'farm to table'. This approach provides control of the contaminants in soil, water, air, control over primary food production (covering animal feed too), intensive control over processing with implementation of HACCP system, but also, over transportation, retail trade, street food and home made food too. In our country creation of the map of the polluted areas, and actions in order to treat the pollution should accompany implementation of this new food safety system. The need for assessment of the level of

Challenges on innovations of newly-developed safety analysis codes

International Nuclear Information System (INIS)

Yang, Yanhua; Zhang, Hao

2016-01-01

With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

Challenges on innovations of newly-developed safety analysis codes

Energy Technology Data Exchange (ETDEWEB)

Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City

2016-05-15

With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

Safety culture and accident analysis-A socio-management approach based on organizational safety social capital

Energy Technology Data Exchange (ETDEWEB)

Rao, Suman [Risk Analyst (India)]. E-mail: sumanashokrao@yahoo.co.in

2007-04-11

One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, Key drivers of safety culture in many organizations are regulation, audits, safety training, various types of employee exhortations to comply with safety norms, etc. However, less evident factors like networking relationships and social trust amongst employees, as also extended networking relationships and social trust of organizations with external stakeholders like government, suppliers, regulators, etc., which constitute the safety social capital in the Organization-seem to also influence the sustenance of organizational safety culture. Can erosion in safety social capital cause deterioration in safety culture and contribute to accidents? If so, how does it contribute? As existing accident analysis models do not provide answers to these questions, CAMSoC (Curtailing Accidents by Managing Social Capital), an accident analysis model, is proposed. As an illustration, five accidents: Bhopal (India), Hyatt Regency (USA), Tenerife (Canary Islands), Westray (Canada) and Exxon Valdez (USA) have been analyzed using CAMSoC. This limited cross-industry analysis provides two key socio-management insights: the biggest source of motivation that causes deviant behavior leading to accidents is 'Faulty Value Systems'. The second biggest source is 'Enforceable Trust'. From a management control perspective, deterioration in safety culture and resultant accidents is more due to the 'action controls' rather than explicit 'cultural controls'. Future research directions to enhance the model's utility through layering are addressed briefly.

Thoughts of the nuclear safety culture and 'star-level' management

International Nuclear Information System (INIS)

Wang Sen

2004-01-01

From the point of view that enterprise management has come into the stage of cultural management, this article divides the contents of nuclear safety culture into target management, safety management, quality management, site management, cost management, authority management, teamwork, information communication and continuous improvement. Each aspect win be classified by five 'star- level's according to the appearance, and the present situation of the company should be assessed with those star-level indices so as to find out the disadvantages. Improvement will follow with the promotion of company management level. (author)

Construction safety and waste management an economic analysis

CERN Document Server

Li, Rita Yi Man

2015-01-01

This monograph presents an analysis of construction safety problems and on-site safety measures from an economist’s point of view. The book includes examples from both emerging countries, e.g. China and India, and developed countries, e.g. Australia and Hong Kong. Moreover, the author covers an analysis on construction safety knowledge sharing by means of updatable mobile technology such as apps in Androids and iOS platform mobile devices. The target audience comprises primarily researchers and experts in the field but the book may also be beneficial for graduate students.

Development of 3D CFD simulation method in nuclear reactor safety analysis

International Nuclear Information System (INIS)

Rosli Darmawan; Mariah Adam

2012-01-01

One of the most prevailing issues in the operation of nuclear reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing have always brought about public fear on anything related to nuclear. Most findings on the nuclear reactor accidents are closely related to the reactor cooling system. Thus, the understanding of the behaviour of reactor cooling system is very important to ensure the development and improvement on safety can be continuously done. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last three decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. This paper discusses the development of 3D CFD usage in nuclear reactor safety analysis worldwide. A brief review on the usage of CFD at Malaysia's Reactor TRIGA PUSPATI is also presented. (author)

Using Addenda in Documented Safety Analysis Reports

International Nuclear Information System (INIS)

Swanson, D.S.; Thieme, M.A.

2003-01-01

This paper discusses the use of addenda to the Radioactive Waste Management Complex (RWMC) Documented Safety Analysis (DSA) located at the Idaho National Engineering and Environmental Laboratory (INEEL). Addenda were prepared for several systems and processes at the facility that lacked adequate descriptive information and hazard analysis in the DSA. They were also prepared for several new activities involving unreviewed safety questions (USQs). Ten addenda to the RWMC DSA have been prepared since the last annual update

System safety analysis of an autonomous mobile robot

International Nuclear Information System (INIS)

Bartos, R.J.

1994-01-01

Analysis of the safety of operating and maintaining the Stored Waste Autonomous Mobile Inspector (SWAMI) II in a hazardous environment at the Fernald Environmental Management Project (FEMP) was completed. The SWAMI II is a version of a commercial robot, the HelpMate trademark robot produced by the Transitions Research Corporation, which is being updated to incorporate the systems required for inspecting mixed toxic chemical and radioactive waste drums at the FEMP. It also has modified obstacle detection and collision avoidance subsystems. The robot will autonomously travel down the aisles in storage warehouses to record images of containers and collect other data which are transmitted to an inspector at a remote computer terminal. A previous study showed the SWAMI II has economic feasibility. The SWAMI II will more accurately locate radioactive contamination than human inspectors. This thesis includes a System Safety Hazard Analysis and a quantitative Fault Tree Analysis (FTA). The objectives of the analyses are to prevent potentially serious events and to derive a comprehensive set of safety requirements from which the safety of the SWAMI II and other autonomous mobile robots can be evaluated. The Computer-Aided Fault Tree Analysis (CAFTA copyright) software is utilized for the FTA. The FTA shows that more than 99% of the safety risk occurs during maintenance, and that when the derived safety requirements are implemented the rate of serious events is reduced to below one event per million operating hours. Training and procedures in SWAMI II operation and maintenance provide an added safety margin. This study will promote the safe use of the SWAMI II and other autonomous mobile robots in the emerging technology of mobile robotic inspection

Selecting of key safety parameters in reactor nuclear safety supervision

International Nuclear Information System (INIS)

He Fan; Yu Hong

2014-01-01

The safety parameters indicate the operational states and safety of research reactor are the basis of nuclear safety supervision institution to carry out effective supervision to nuclear facilities. In this paper, the selecting of key safety parameters presented by the research reactor operating unit to National Nuclear Safety Administration that can express the research reactor operational states and safety when operational occurrence or nuclear accident happens, and the interrelationship between them are discussed. Analysis shows that, the key parameters to nuclear safety supervision of research reactor including design limits, operational limits and conditions, safety system settings, safety limits, acceptable limits and emergency action level etc. (authors)

Preliminary Safety Analysis Report for the Tokamak Physics Experiment

International Nuclear Information System (INIS)

Motloch, C.G.; Bonney, R.F.; Levine, J.D.; Masson, L.S.; Commander, J.C.

1995-04-01

This Preliminary Safety Analysis Report (PSAR), includes an indication of the magnitude of facility hazards, complexity of facility operations, and the stage of the facility life-cycle. It presents the results of safety analyses, safety assurance programs, identified vulnerabilities, compensatory measures, and, in general, the rationale describing why the Tokamak Physics Experiment (TPX) can be safely operated. It discusses application of the graded approach to the TPX safety analysis, including the basis for using Department of Energy (DOE) Order 5480.23 and DOE-STD-3009-94 in the development of the PSAR

Methodology of safety evaluation about land disposal of low level radioactive wastes

International Nuclear Information System (INIS)

Suzuki, Atsuyuki

1986-01-01

Accompanying the progress of the construction project of low level radioactive waste storage facilities in Aomori Prefecture, the full scale land disposal of low level radioactive wastes shows its symptom also in Japan. In this report, the scientific methodology to explain the safety about the land disposal of low level radioactive wastes is discussed. The land disposal of general wastes by shallow burying has already had sufficient results. In the case of low level radioactive wastes, also the land disposal by shallow burying is considered. Low level radioactive wastes can be regarded as one form of industrial wastes, as there are many common parts in the scientific and theoretical base of the safety. Attention is paid most to the contamination of ground water. Low level radioactive wastes are solid wastes, accordingly the degree of contamination should be less. The space in which ground water existes, the phenomena of ground water movement, the phenomena of ground water dispersion and Fick's law, the adsorption effect of strata, and the evaluation of source term are explained. These are the method to analyze the degree of contamination from safety evaluation viewpoint. (Kako, I.)

Model extension and improvement for simulator-based software safety analysis

Energy Technology Data Exchange (ETDEWEB)

Huang, H.-W. [Department of Engineering and System Science, National Tsing Hua University (NTHU), 101 Section 2 Kuang Fu Road, Hsinchu, Taiwan (China) and Institute of Nuclear Energy Research (INER), No. 1000 Wenhua Road, Chiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)]. E-mail: hwhwang@iner.gov.tw; Shih Chunkuan [Department of Engineering and System Science, National Tsing Hua University (NTHU), 101 Section 2 Kuang Fu Road, Hsinchu, Taiwan (China); Yih Swu [Department of Computer Science and Information Engineering, Ching Yun University, 229 Chien-Hsin Road, Jung-Li, Taoyuan County 320, Taiwan (China); Chen, M.-H. [Institute of Nuclear Energy Research (INER), No. 1000Wenhua Road, Chiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Lin, J.-M. [Taiwan Power Company (TPC), 242 Roosevelt Road, Section 3, Taipei 100, Taiwan (China)

2007-05-15

One of the major concerns when employing digital I and C system in nuclear power plant is digital system may introduce new failure mode, which differs with previous analog I and C system. Various techniques are under developing to analyze the hazard originated from software faults in digital systems. Preliminary hazard analysis, failure modes and effects analysis, and fault tree analysis are the most extensive used techniques. However, these techniques are static analysis methods, cannot perform dynamic analysis and the interactions among systems. This research utilizes 'simulator/plant model testing' technique classified in (IEEE Std 7-4.3.2-2003, 2003. IEEE Standard for Digital Computers in Safety Systems of Nuclear Power Generating Stations) to identify hazards which might be induced by nuclear I and C software defects. The recirculation flow system, control rod system, feedwater system, steam line model, dynamic power-core flow map, and related control systems of PCTran-ABWR model were successfully extended and improved. The benchmark against ABWR SAR proves this modified model is capable to accomplish dynamic system level software safety analysis and better than the static methods. This improved plant simulation can then further be applied to hazard analysis for operator/digital I and C interface interaction failure study, and the hardware-in-the-loop fault injection study.

Requirement analysis of the safety-critical software implementation for the nuclear power plant

International Nuclear Information System (INIS)

Chang, Hoon Seon; Jung, Jae Cheon; Kim, Jae Hack; Nam, Sang Ku; Kim, Hang Bae

2005-01-01

The safety critical software shall be implemented under the strict regulation and standards along with hardware qualification. In general, the safety critical software has been implemented using functional block language (FBL) and structured language like C in the real project. Software design shall comply with such characteristics as; modularity, simplicity, minimizing the use of sub-routine, and excluding the interrupt logic. To meet these prerequisites, we used the computer-aided software engineering (CASE) tool to substantiate the requirements traceability matrix that were manually developed using Word processors or Spreadsheets. And the coding standard and manual have been developed to confirm the quality of software development process, such as; readability, consistency, and maintainability in compliance with NUREG/CR-6463. System level preliminary hazard analysis (PHA) is performed by analyzing preliminary safety analysis report (PSAR) and FMEA document. The modularity concept is effectively implemented for the overall module configurations and functions using RTP software development tool. The response time imposed on the basis of the deterministic structure of the safety-critical software was measured

Reliability analysis of diverse safety logic systems of fast breeder reactor

International Nuclear Information System (INIS)

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

2006-01-01

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

Regulatory review of probabilistic safety assessment (PSA) level 1

International Nuclear Information System (INIS)

2000-02-01

Probabilistic safety assessment (PSA) is increasingly being used as part of the decision making process to assess the level of safety of nuclear power plants. The methodologies in use are maturing and the insights gained from the PSAs are being used along with those from the deterministic analysis. Many regulatory authorities consider that the current state of the art in PSA (especially Level 1 PSA) is sufficiently well developed that it can be used centrally in the regulatory decision making process - referred to as 'risk informed regulation'. For these applications to be successful, it will be necessary for regulatory authorities to have a high degree of confidence in PSA. However, at the IAEA Technical Committee Meeting on Use of PSA in the Regulatory Process in 1994 and at the OECD Nuclear Energy Agency Committee for Nuclear Regulatory Activities (CNRA) 'Special Issues' Meeting in 1997 on Review Procedures and Criteria for Different Regulatory Applications of PSA, it was recognized that formal regulatory review guidance for PSA did not exist. The senior regulators noted that there was a need to produce some international guidance for reviewing PSAs to establish an agreed basis for assessing whether important technological and methodological issues in PSAs are treated adequately and to verify that conclusions reached are appropriate. In 1997 the IAEA and OECD Nuclear Energy Agency agreed to produce in co-operation a technical document on the regulatory review of PSA. This publication is intended to provide guidance to regulatory authorities on how to review the PSA for a nuclear power plant to gain confidence that it has been carried out to an acceptable standard so that it can be used as the basis for taking risk informed decisions within a regulatory decision making process. The document gives guidance on how to set about reviewing a PSA and on the technical issues that need to be addressed. This publication gives guidance for the review of Level 1 PSA for

Uncertainty analysis for Ulysses safety evaluation report

International Nuclear Information System (INIS)

Frank, M.V.

1991-01-01

As part of the effort to review the Ulysses Final Safety Analysis Report and to understand the risk of plutonium release from the Ulysses spacecraft General Purpose Heat Source---Radioisotope Thermal Generator (GPHS-RTG), the Interagency Nuclear Safety Review Panel (INSRP) and the author performed an integrated, quantitative analysis of the uncertainties of the calculated risk of plutonium release from Ulysses. Using state-of-art probabilistic risk assessment technology, the uncertainty analysis accounted for both variability and uncertainty of the key parameters of the risk analysis. The results show that INSRP had high confidence that risk of fatal cancers from potential plutonium release associated with calculated launch and deployment accident scenarios is low

Nuclear safety in France after Fukushima - Critical analysis of complementary safety assessments (CSA) carried out on French nuclear installations after Fukushima

International Nuclear Information System (INIS)

Makhijani, Arjun; Marignac, Yves

2012-02-01

This report proposes a critical analysis of the approach carried out on the basis of the CSA (complementary safety assessment), from their specifications to the IRSN conclusions. It is notably based on the analysis performed by EDF on three nuclear sites (Gravelines, Civaux and Flamanville) which encompass the different levels of the nuclear power plants in France and the EPR project under construction, and on the analysis performed by Areva for La Hague reprocessing plants. Due to the short delay, only some sites and some problems have been considered. The CSA methodology is described. The EDF approach is discussed as well as the IRSN analysis of reports made by EDF, and then the different case studies. Beyond the conclusions of these reports, the authors highlight several major possible accidents which must be taken into account. They also outline that this CSA approach is a good starting point for the strengthening of nuclear safety

COLD-SAT feasibility study safety analysis

Science.gov (United States)

Mchenry, Steven T.; Yost, James M.

1991-01-01

The Cryogenic On-orbit Liquid Depot-Storage, Acquisition, and Transfer (COLD-SAT) satellite presents some unique safety issues. The feasibility study conducted at NASA-Lewis desired a systems safety program that would be involved from the initial design in order to eliminate and/or control the inherent hazards. Because of this, a hazards analysis method was needed that: (1) identified issues that needed to be addressed for a feasibility assessment; and (2) identified all potential hazards that would need to be controlled and/or eliminated during the detailed design phases. The developed analysis method is presented as well as the results generated for the COLD-SAT system.

Guidance for preparation of safety analysis reports for nonreactor facilities and operations

International Nuclear Information System (INIS)

1992-01-01

Department of Energy (DOE) Orders 5480.23, ''Nuclear Safety Analysis Reports,'' and 5481.1B, ''Safety Analysis and Review System'' require the preparation of appropriate safety analyses for each DOE operation and subsequent significant modifications including decommissioning, and independent review of each safety analysis. The purpose of this guide is to assist in the preparation and review of safety documentation for Oak Ridge Field Office (OR) nonreactor facilities and operation. Appendix A lists DOE Orders, NRC Regulatory Guides and other documents applicable to the preparation of safety analysis reports

Safety analysis and risk assessment handbook

International Nuclear Information System (INIS)

Peterson, V.L.; Colwell, R.G.; Dickey, R.L.

1997-01-01

This Safety Analysis and Risk Assessment Handbook (SARAH) provides guidance to the safety analyst at the Rocky Flats Environmental Technology Site (RFETS) in the preparation of safety analyses and risk assessments. Although the older guidance (the Rocky Flats Risk Assessment Guide) continues to be used for updating the Final Safety Analysis Reports developed in the mid-1980s, this new guidance is used with all new authorization basis documents. With the mission change at RFETS came the need to establish new authorization basis documents for its facilities, whose functions had changed. The methodology and databases for performing the evaluations that support the new authorization basis documents had to be standardized, to avoid the use of different approaches and/or databases for similar accidents in different facilities. This handbook presents this new standardized approach. The handbook begins with a discussion of the requirements of the different types of authorization basis documents and how to choose the one appropriate for the facility to be evaluated. It then walks the analyst through the process of identifying all the potential hazards in the facility, classifying them, and choosing the ones that need to be analyzed further. It then discusses the methods for evaluating accident initiation and progression and covers the basic steps in a safety analysis, including consequence and frequency binning and risk ranking. The handbook lays out standardized approaches for determining the source terms of the various accidents (including airborne release fractions, leakpath factors, etc.), the atmospheric dispersion factors appropriate for Rocky Flats, and the methods for radiological and chemical consequence assessments. The radiological assessments use a radiological open-quotes templateclose quotes, a spreadsheet that incorporates the standard values of parameters, whereas the chemical assessments use the standard codes ARCHIE and ALOHA

Multivariate time series analysis of SafetyNet data. SafetyNet, Building the European Road Safety Observatory, Workpackage 7, Deliverable 7.7.

NARCIS (Netherlands)

Commandeur, J.J.F. Bijleveld, F.D. & Bergel, R.

2009-01-01

This deliverable provides an application of theories and methods documented in Deliverables 7.4 and 7.5 of work package 7 of the SafetyNet project. In this deliverable, use of select analysis techniques is demonstrated through real world road safety analysis problems, using aggregate data which may

Current activities and future trends in reliability analysis and probabilistic safety assessment in Hungary

International Nuclear Information System (INIS)

Hollo, E.; Toth, J.

1986-01-01

In Hungary reliability analysis (RA) and probabilistic safety assessment (PSA) of nuclear power plants was initiated 3 years ago. First, computer codes for automatic fault tree analysis (CAT, PREP) and numerical evaluation (REMO, KITT1,2) were adapted. Two main case studies - detailed availability/reliability calculation of diesel sets and analysis of safety systems influencing event sequences induced by large LOCA - were performed. Input failure data were taken from publications, a need for failure and reliability data bank was revealed. Current and future activities involves: setup of national data bank for WWER-440 units; full-scope level-I PSA of PAKS NPP in Hungary; operational safety assessment of particular problems at PAKS NPP. In the present article the state of RA and PSA activities in Hungary, as well as the main objectives of ongoing work are described. A need for international cooperation (for unified data collection of WWER-440 units) and for IAEA support (within Interregional Program INT/9/063) is emphasized. (author)

Associations between safety culture and employee engagement over time: a retrospective analysis.

Science.gov (United States)

Daugherty Biddison, Elizabeth Lee; Paine, Lori; Murakami, Peter; Herzke, Carrie; Weaver, Sallie J

2016-01-01

With the growth of the patient safety movement and development of methods to measure workforce health and success have come multiple modes of assessing healthcare worker opinions and attitudes about work and the workplace. Safety culture, a group-level measure of patient safety-related norms and behaviours, has been proposed to influence a variety of patient safety outcomes. Employee engagement, conceptualised as a positive, work-related mindset including feelings of vigour, dedication and absorption in one's work, has also demonstrated an association with a number of important worker outcomes in healthcare. To date, the relationship between responses to these two commonly used measures has been poorly characterised. Our study used secondary data analysis to assess the relationship between safety culture and employee engagement over time in a sample of >50 inpatient hospital units in a large US academic health system. With >2000 respondents in each of three time periods assessed, we found moderate to strong positive correlations (r=0.43-0.69) between employee engagement and four Safety Attitudes Questionnaire domains. Independent collection of these two assessments may have limited our analysis in that minimally different inclusion criteria resulted in some differences in the total respondents to the two instruments. Our findings, nevertheless, suggest a key area in which healthcare quality improvement efforts might be streamlined. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/

[Economic analysis versus the principle of guaranteed safety in blood transfusion].

Science.gov (United States)

Moatti, J P; Loubière, S; Rotily, M

2000-06-01

This article shows that policies aimed at reducing risks of infectious agents transmissible through blood unfortunately follow a law of 'diminishing returns': increasing marginal costs have to be devoted for limited reductions in the risks of contamination through blood donations. Therefore, the economic cost-effectiveness analysis is appropriate to identify screening strategies which may minimize costs to reach a certain level of safety. Moreover, economic analysis can contribute to public debates about the level of residual risk that society is willing to accept. Empirical results from French studies about screening for hepatitis C virus (HCV) in individuals who have received blood transfusions and in blood donations are presented to illustrate these points.

Engineered safeguards and passive safety features (safety analysis detailed report no. 6)

Energy Technology Data Exchange (ETDEWEB)

1988-01-15

The Safety-Analysis Summary lists the reactor's safety aspects for passive and active prevention of severe accidents and mitigation of accident consequences, i.e., intrinsic and passive protections of the plant; intrinsic and passive protections of the core; inherent decay-heat removal systems; rapid-shutdown systems; four physical containment barriers. This report goes into further details regarding some of this aspects.

Computer aided safety analysis

International Nuclear Information System (INIS)

1988-05-01

The document reproduces 20 selected papers from the 38 papers presented at the Technical Committee/Workshop on Computer Aided Safety Analysis organized by the IAEA in co-operation with the Institute of Atomic Energy in Otwock-Swierk, Poland on 25-29 May 1987. A separate abstract was prepared for each of these 20 technical papers. Refs, figs and tabs

Holistic safety analysis for advanced nuclear power plants

International Nuclear Information System (INIS)

Alvarenga, M.A.B.; Guimaraes, A.C.F.

1992-01-01

This paper reviews the basic methodology of safety analysis used in the ANGRA-I and ANGRA-II nuclear power plants, its weaknesses, the problems with public acceptance of the risks, the future of the nuclear energy in Brazil, as well as recommends a new methodology, HOLISTIC SAFETY ANALYSIS, to be used both in the design and licensing phases, for advanced reactors. (author)

EUMENES, a computer software for managing the radiation safety program information at an institutional level

International Nuclear Information System (INIS)

Hernandez Saiz, Alejandro; Cornejo Diaz, Nestor; Valdes Ramos, Maryzury; Martinez Gonzalez, Alina; Gonzalez Rodriguez, Niurka; Vergara Gil, Alex

2008-01-01

The correct application of national and international regulations in the field of Radiological Safety requires the implementation of Radiation Safety Programs appropriate to the developed practice. These Programs demand the preparation and keeping of an important number of records and data, the compliance with working schedules, systematic quality controls, audits, delivery of information to the Regulatory Authority, the execution of radiological assessments, etc. Therefore, it is unquestionable the necessity and importance of having a computer tool to support the management of the information related to the Radiation Safety Program in any institution. The present work describes a computer program that allows the efficient management of these data. Its design was based on the IAEA International Basic Safety Standards recommendations and on the requirements of the Cuban national standards, with the objective of being flexible enough to be applied in most of the institutions using ionizing radiations. The most important records of Radiation Safety Programs were incorporated and reports can be generated by the users. An additional tools-module allows the user to access to a radionuclide data library, and to carry out different calculations of interest in radiological protection. The program has been developed in Borland Delphi and manages Microsoft Access databases. It is a user friendly code that aims to support the optimization of Radiation Safety Programs. The program contributes to save resources and time, as the generated information is electronically kept and transmitted. The code has different security access levels according to the user responsibility at the institution and also provides for the analysis of the introduced data, in a quick and efficient way, as well as to notice deadlines, the exceeding of reference levels and situations that require attention. (author)

Special characteristics of the safety analysis of HWRs

International Nuclear Information System (INIS)

Kugler, G.

1980-01-01

Two lectures are presented in this report. The CANDU-PHW reactor is used as a model for discussion. The first lecture describes the distinctive features of the CANDU reactor, and how they impact on reactor safety. In the second lecture the Canadian safety philosophy, the safety design objective, and other selected topics on reactor safety analysis are discussed. The material in this report was selected with a view to assisting those not familiar with the CANDU heavy water reactor design in evaluating the distinctive safety aspects of these reactors. (orig./RW)

SCALE 5: Powerful new criticality safety analysis tools

International Nuclear Information System (INIS)

Bowman, Stephen M.; Hollenbach, Daniel F.; Dehart, Mark D.; Rearden, Bradley T.; Gauld, Ian C.; Goluoglu, Sedat

2003-01-01

Version 5 of the SCALE computer software system developed at Oak Ridge National Laboratory, scheduled for release in December 2003, contains several significant new modules and sequences for criticality safety analysis and marks the most important update to SCALE in more than a decade. This paper highlights the capabilities of these new modules and sequences, including continuous energy flux spectra for processing multigroup problem-dependent cross sections; one- and three-dimensional sensitivity and uncertainty analyses for criticality safety evaluations; two-dimensional flexible mesh discrete ordinates code; automated burnup-credit analysis sequence; and one-dimensional material distribution optimization for criticality safety. (author)

Preparation of safety analysis reports (SARs) for near surface radioactive waste disposal facilities. Format and content of SARs

International Nuclear Information System (INIS)

1995-02-01

All facilities at which radioactive wastes are processed, stored and disposed of have the potential for causing hazards to humans and to the environment. Precautions must be taken in the siting, design and operation of the facilities to ensure that an adequate level of safety is achieved. The processes by which this is evaluated is called safety assessment. An important part of safety assessment is the documentation of the process. A well prepared safety analysis report (SAR) is essential if approval of the facility is to be obtained from the regulatory authorities. This TECDOC describes the format and content of a safety analysis report for a near surface radioactive waste disposal facility and will serve essentially as a checklist in this respect

Research on the improvement of nuclear safety

International Nuclear Information System (INIS)

Yoo, Keon Joong; Kim, Dong Soo; Kim, Hui Dong; Park, Chang Kyu

1993-06-01

To improve the nuclear safety, this project is divided into three areas which are the development of safety analysis technology, the development of severe accident analysis technology and the development of integrated safety assessment technology. 1. The development of safety analysis technology. The present research aims at the development of necessary technologies for nuclear safety analysis in Korea. Establishment of the safety analysis technologies enables to reduce the expenditure both by eliminating excessive conservatisms incorporated in nuclear reactor design and by increasing safety margins in operation. It also contributes to improving plant safety through realistic analyses of the Emergency Operating Procedures (EOP). 2. The development of severe accident analysis technology. By the computer codes (MELCOR and CONTAIN), the in-vessel and the ex-vessel severe accident phenomena are simulated. 3. The development of integrated safety assessment technology. In the development of integrated safety assessment techniques, the included research areas are the improvement of PSA computer codes, the basic study on the methodology for human reliability analysis (HRA) and common cause failure (CCF). For the development of the level 2 PSA computer code, the basic research for the interface between level 1 and 2 PSA, the methodology for the treatment of containment event tree are performed. Also the new technologies such as artificial intelligence, object-oriented programming techniques are used for the improvement of computer code and the assessment techniques

SAFETY ANALYSIS APPROACH TO TANK 241-SY-101 REMEDIATION ACTIVITIES

International Nuclear Information System (INIS)

RYAN, G.W.

2000-01-01

An Unreviewed Safety Question was declared related to the unexplained waste surface level growth in high-level radioactive waste storage Tank 241-SY-101 at the Hanford Site in Richland, Washington. Because the waste surface level in Tank 241-SY-101 was growing in a manner inconsistent with previous behavior, the following issues of concern were recognized: (1) The continually rising surface level had the potential to reach physical encumbrances or limits within the tank (e.g., instrumentation, cameras, established Authorization Basis limits, and the double containment boundary) and the potential to significantly change the consequences of previously analyzed accidents (e.g., flammable gas deflagrations). (2) The presence of new hazards because of significant quantities of flammable gas retained in the crust (e.g., crust collapse gas-release events). (3) The potential to inhibit information gathering related to the existing hazards in the tank (e.g., unable to determine surface level to assess the potential for large gas releases). In response to this situation, a Contractor Project Team, which included Department of Energy representation, was formed to constructively address the issue. The team was responsible for developing and evaluating remediation options and executing the chosen option for remediating the surface level rise issue for Tank 241-SY-101. From an Authorization Basis perspective, the following important aspects will be discussed in this paper: (1) The integrated nature of the Project Team. The team consisted of all the organizations necessary to ensure that the time available to remediate Tank 241-SY-101 was effectively used. Most notable is the connectivity of the Nuclear Safety and Licensing organization with the Engineering, Design, and Operations organizations. (2) The ability of the safety analysis support to adjust to and address evolving Project Team goals and dynamic tank conditions. (3) Due to the urgency to mitigate this developing issue

Basic concept on safety regulation for land disposal of low level radioactive solid wastes

International Nuclear Information System (INIS)

1985-01-01

As to the land disposal of low level radioactive solid wastes, to which the countermeasures have become the urgent problem at present, it is considered to be a realistic method to finally store the solid wastes concentratedly outside the sites of nuclear power stations and others, and effort has been exerted by those concerned to realize it. Besides, as for extremely low level radioactive solid wastes, the measures of disposing them corresponding to the radioactivity level are necessary, and the concrete method has been examined. The Committee on Safety Regulation for Radioactive Wastes has discussed the safety regulation for those since April, 1984, and the basic concept on the safety regulation was worked up. It is expected that the safety of the land disposal of low level radioactive solid wastes can be ensured when the safety regulation is carried out in conformity with this basic concept. The present status of the countermeasures to the land disposal of low level radioactive solid wastes is shown. As the concrete method, the disposal in shallow strate has been generally adopted. At present, the plan for the final storage in Aomori Prefecture is considered, and it will be started with the first stage of four-stage control. (Kako, I.)

Level II Probabilistic Safety Analysis Methodology for the Application to GEN-IV Sodium-cooled Fast Reactor

International Nuclear Information System (INIS)

Park, S. Y.; Kim, T. W.; Han, S. H.; Jeong, H. Y.

2010-03-01

The Korea Atomic Energy Research Institute (KAERI) has been developing liquid metal reactor (LMR) design technologies under a National Nuclear R and D Program. Nevertheless, there is no experience of the probabilistic safety assessment (PSA) domestically for a fast reactor with the metal fuel. Therefore, the objective of this study is to establish the methodologies of risk assessment for the reference design of GEN-IV sodium fast reactor (SFR). An applicability of the PSA methodology of U. S. NRC and PRISM plant to the domestic GEN-IV SFR has been studied. The study contains a plant damage state analysis, a containment event tree analysis, and a source-term release category binning process

Behavioral integrity for safety, priority of safety, psychological safety, and patient safety : a team-level study

NARCIS (Netherlands)

Leroy, H.; Dierynck, B.; Anseel, F.; Simons, T.; Halbesleben, J.R.; McCaughey, D.; Savage, G.T.; Sels, L.

2012-01-01

This article clarifies how leader behavioral integrity for safety helps solve follower's double bind between adhering to safety protocols and speaking up about mistakes against protocols. Path modeling of survey data in 54 nursing teams showed that head nurse behavioral integrity for safety

Comparing performance level estimation of safety functions in three distributed structures

International Nuclear Information System (INIS)

Hietikko, Marita; Malm, Timo; Saha, Heikki

2015-01-01

The capability of a machine control system to perform a safety function is expressed using performance levels (PL). This paper presents the results of a study where PL estimation was carried out for a safety function implemented using three different distributed control system structures. Challenges relating to the process of estimating PLs for safety related distributed machine control functions are highlighted. One of these examines the use of different cabling schemes in the implementation of a safety function and its effect on the PL evaluation. The safety function used as a generic example in PL calculations relates to a mobile work machine. It is a safety stop function where different technologies (electrical, hydraulic and pneumatic) can be utilized. It was detected that by replacing analogue cables with digital communication the system structure becomes simpler with less number of failing components, which can better the PL of the safety function. - Highlights: • Integration in distributed systems enables systems with less components. • It offers high reliability and diagnostic properties. • Analogue signals create uncertainty in signal reliability and difficult diagnostics

An Analysis of Trainers' Perspectives within an Ecological Framework: Factors that Influence Mine Safety Training Processes

Directory of Open Access Journals (Sweden)

Emily J. Haas

2014-09-01

Conclusion: This study offers a new technique to identify limitations in safety training systems and processes. The analysis suggests that training should be developed and disseminated with consideration of various levels—individual, interpersonal, organizational, and community—to promote skills. If factors identified within and between levels are addressed, it may be easier to sustain mineworker competencies that are established during safety training.

Utilization of the MCNP-3A code for criticality safety analysis

International Nuclear Information System (INIS)

Maragni, M.G.; Moreira, J.M.L.

1996-01-01

In the last decade, Brazil started to operate facilities for processing and storing uranium in different forms. The necessity of criticality safety analysis appeared in the design phase of the uranium pilot process plants and also in the licensing of transportation and storage of fissile materials. The 2-MW research reactor and the Angra I power plant also required criticality safety assessments because their spent-fuel storage was approaching full-capacity utilization. The criticality safety analysis in Brazil has been based on KENO IV code calculations, which present some difficulties for correct geometry representation. The MCNP-3A code is not reported to be used frequently for criticality safety analysis in Brazil, but its good geometry representation makes it a possible tool for treating problems of complex geometry. A set of benchmark tests was performed to verify its applicability for criticality safety analysis in Brazil. This paper presents several benchmark tests aimed at selecting a set of options available in the MCNP-3A code that would be adequate for criticality safety analysis. The MCNP-3A code is also compared with the KENO-IV code regarding its performance for criticality safety analysis

Safety analysis of tritium processing system based on PHA

International Nuclear Information System (INIS)

Fu Wanfa; Luo Deli; Tang Tao

2012-01-01

Safety analysis on primary confinement of tritium processing system for TBM was carried out with Preliminary Hazard Analysis. Firstly, the basic PHA process was given. Then the function and safe measures with multiple confinements about tritium system were described and analyzed briefly, dividing the two kinds of boundaries of tritium transferring through, that are multiple confinement systems division and fluid loops division. Analysis on tritium releasing is the key of PHA. Besides, PHA table about tritium releasing was put forward, the causes and harmful results being analyzed, and the safety measures were put forward also. On the basis of PHA, several kinds of typical accidents were supposed to be further analyzed. And 8 factors influencing the tritium safety were analyzed, laying the foundation of evaluating quantitatively the safety grade of various nuclear facilities. (authors)

Deterministic Safety Analysis for Nuclear Power Plants. Specific Safety Guide (Spanish Edition)

International Nuclear Information System (INIS)

2012-01-01

. The IAEA takes seriously the enduring challenge for users and regulators everywhere: that of ensuring a high level of safety in the use of nuclear materials and radiation sources around the world. Their continuing utilization for the benefit of humankind must be managed in a safe manner, and the IAEA safety standards are designed to facilitate the achievement of that goal.

Safety Assessment for Research Reactors and Preparation of the Safety Analysis Report. Specific Safety Guide

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-11-15

The IAEA's Statute authorizes the Agency to 'establish or adopt' standards of safety for protection of health and minimization of danger to life and property' - standards that the IAEA must use in its own operations, and which States can apply by means of their regulatory provisions for nuclear and radiation safety. The IAEA does this in consultation with the competent organs of the United Nations and with the specialized agencies concerned. A comprehensive set of high quality standards under regular review is a key element of a stable and sustainable global safety regime, as is the IAEA's assistance in their application. The IAEA commenced its safety standards programme in 1958. The emphasis placed on quality, fitness for purpose and continuous improvement has led to the widespread use of the IAEA standards throughout the world. The Safety Standards Series now includes unified Fundamental Safety Principles, which represent an international consensus on what must constitute a high level of protection and safety. With the strong support of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its standards. Standards are only effective if they are properly applied in practice. The IAEA's safety services encompass design, siting and engineering safety, operational safety, radiation safety, safe transport of radioactive material and safe management of radioactive waste, as well as governmental organization, regulatory matters and safety culture in organizations. These safety services assist Member States in the application of the standards and enable valuable experience and insights to be shared. Regulating safety is a national responsibility, and many States have decided to adopt the IAEA's standards for use in their national regulations. For parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions

Safety Assessment for Research Reactors and Preparation of the Safety Analysis Report. Specific Safety Guide

International Nuclear Information System (INIS)

2011-01-01

The IAEA's Statute authorizes the Agency to 'establish or adopt' standards of safety for protection of health and minimization of danger to life and property' - standards that the IAEA must use in its own operations, and which States can apply by means of their regulatory provisions for nuclear and radiation safety. The IAEA does this in consultation with the competent organs of the United Nations and with the specialized agencies concerned. A comprehensive set of high quality standards under regular review is a key element of a stable and sustainable global safety regime, as is the IAEA's assistance in their application. The IAEA commenced its safety standards programme in 1958. The emphasis placed on quality, fitness for purpose and continuous improvement has led to the widespread use of the IAEA standards throughout the world. The Safety Standards Series now includes unified Fundamental Safety Principles, which represent an international consensus on what must constitute a high level of protection and safety. With the strong support of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its standards. Standards are only effective if they are properly applied in practice. The IAEA's safety services encompass design, siting and engineering safety, operational safety, radiation safety, safe transport of radioactive material and safe management of radioactive waste, as well as governmental organization, regulatory matters and safety culture in organizations. These safety services assist Member States in the application of the standards and enable valuable experience and insights to be shared. Regulating safety is a national responsibility, and many States have decided to adopt the IAEA's standards for use in their national regulations. For parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions

Improvement of Managers’ Safety Knowledge through Scientifically Reasonable Interviews

Directory of Open Access Journals (Sweden)

Paas Õnnela

2015-11-01

Full Text Available The safety management system has been analysed in 16 Estonian enterprises using the MISHA method (Method for Industrial Safety and Health Activity Assessment. The factor analysis (principal component analysis and varimax with Kaiser analysis has been implemented for the interpretation of the results on safety performance at the enterprises implementing OHSAS 18001 and the ones that do not implement OHSAS 18001. The division of the safety areas into four parts for a better understanding of the safety level and its improvement possibilities has been proven through the statistical analysis. The connections between the questions aimed to clarify the safety level and performance at the enterprises have been set based on the statistics. New learning package “training through the questionnaires” has been worked out in the current paper for the top and middle-level managers to improve their safety knowledge, where the MISHA questionnaire has been taken as the basis.

Human factors, system safety, and systems engineering in the transportation of U.S. high-level waste

International Nuclear Information System (INIS)

Price, D.L.; Chu, S.C.

1993-01-01

The U.S. Nuclear Waste Technical Review Board is an independent agency charged with evaluating the technical and scientific validity of the U.S. Department of Energy's program to manage the disposal of spent fuel and defense high-level waste. The Board has continued to emphasize the importance of using a true system approach in designing the waste management system. The Board has recommended the application of basic design disciplines such as human factors, system safety, and systems engineering. A top-level system study needs to be undertaken that focuses on minimizing handling. The analysis must be well done, in a timely manner, and without the inclusion in the analysis of arbitrary and artificial constraints. (author)

Study on the development of safety regulations for geological disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Wei Fangxin

2012-01-01

The development of regulations under Regulations on Safety Management of Radioactive Waste has become necessary as the issuance of it. The regulations related to geological disposal of high-level radioactive waste can promote the progress of research and development on geological disposal of high-level radioactive waste in China. This paper has present suggestions on development of regulations on geological disposal of high-level radioactive waste by analyzing development of safety regulations on geological disposal of high-level radioactive waste in foreign countries and problems occurred in China and discussed important issues related to the development of safety regulations on geological disposal of high-level radioactive waste. (author)

Analysis of Paks NPP Personnel Activity during Safety Related Event Sequences

International Nuclear Information System (INIS)

Bareith, A.; Hollo, Elod; Karsa, Z.; Nagy, S.

1998-01-01

Within the AGNES Project (Advanced Generic and New Evaluation of Safety) the Level-1 PSA model of the Paks NPP Unit 3 was developed in form of a detailed event tree/fault tree structure (53 initiating events, 580 event sequences, 6300 basic events are involved). This model gives a good basis for quantitative evaluation of potential consequences of actually occurred safety-related events, i.e. for precursor event studies. To make these studies possible and efficient, the current qualitative event analysis practice should be reviewed and a new additional quantitative analysis procedure and system should be developed and applied. The present paper gives an overview of the method outlined for both qualitative and quantitative analyses of the operator crew activity during off-normal situations. First, the operator performance experienced during past operational events is discussed. Sources of raw information, the qualitative evaluation process, the follow-up actions, as well as the documentation requirements are described. Second, the general concept of the proposed precursor event analysis is described. Types of modeled interactions and the considered performance influences are presented. The quantification of the potential consequences of the identified precursor events is based on the task-oriented, Level-1 PSA model of the plant unit. A precursor analysis system covering the evaluation of operator activities is now under development. Preliminary results gained during a case study evaluation of a past historical event are presented. (authors)

A Study of Time Response for Safety-Related Operator Actions in Non-LOCA Safety Analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, Min Seok; Lee, Sang Seob; Park, Min Soo; Lee, Gyu Cheon; Kim, Shin Whan [KEPCO E and C Company, Daejeon (Korea, Republic of)

2014-10-15

The classification of initiating events for safety analysis report (SAR) chapter 15 is categorized into moderate frequency events (MF), infrequent events (IF), and limiting faults (LF) depending on the frequency of its occurrence. For the non-LOCA safety analysis with the purpose to get construction or operation license, however, it is assumed that the operator response action to mitigate the events starts at 30 minutes after the initiation of the transient regardless of the event categorization. Such an assumption of corresponding operator response time may have over conservatism with the MF and IF events and results in a decrease in the safety margin compared to its acceptance criteria. In this paper, the plant conditions (PC) are categorized with the definitions in SAR 15 and ANS 51.1. Then, the consequence of response for safety-related operator action time is determined based on the PC in ANSI 58.8. The operator response time for safety analysis regarding PC are reviewed and suggested. The clarifying alarm response procedure would be required for the guideline to reduce the operator response time when the alarms indicate the occurrence of the transient.

Safety sans Frontières: An International Safety Culture Model.

Science.gov (United States)

Reader, Tom W; Noort, Mark C; Shorrock, Steven; Kirwan, Barry

2015-05-01

The management of safety culture in international and culturally diverse organizations is a concern for many high-risk industries. Yet, research has primarily developed models of safety culture within Western countries, and there is a need to extend investigations of safety culture to global environments. We examined (i) whether safety culture can be reliably measured within a single industry operating across different cultural environments, and (ii) if there is an association between safety culture and national culture. The psychometric properties of a safety culture model developed for the air traffic management (ATM) industry were examined in 17 European countries from four culturally distinct regions of Europe (North, East, South, West). Participants were ATM operational staff (n = 5,176) and management staff (n = 1,230). Through employing multigroup confirmatory factor analysis, good psychometric properties of the model were established. This demonstrates, for the first time, that when safety culture models are tailored to a specific industry, they can operate consistently across national boundaries and occupational groups. Additionally, safety culture scores at both regional and national levels were associated with country-level data on Hofstede's five national culture dimensions (collectivism, power distance, uncertainty avoidance, masculinity, and long-term orientation). MANOVAs indicated safety culture to be most positive in Northern Europe, less so in Western and Eastern Europe, and least positive in Southern Europe. This indicates that national cultural traits may influence the development of organizational safety culture, with significant implications for safety culture theory and practice. © 2015 Society for Risk Analysis.

Development of safety analysis technology for LMR

International Nuclear Information System (INIS)

Hahn, Do Hee; Kwon, Y. M.; Suk, S. D.

2002-05-01

In the present study, the KALIMER safety analysis has been made for the transients considered in the design concept, hypothetical core disruptive accident (HCDA), and containment performance with the establishment of the design basis. Such analyses have not been possible without the computer code improvement, and the experience attained during this research period must have greatly contributed to the achievement of the self reliance in the domestic technology establishment on the safety analysis areas of the conceptual design. The safety analysis codes have been improved to extend their applicable ranges for detailed conceptual design, and a basic computer code system has been established for HCDA analysis. A code-to-code comparison analysis has been performed as a part of code verification attempt, and the leading edge technology of JNC also has been brought for the technology upgrade. In addition, the research and development on the area of the database establishment has been made for the efficient and systematic project implementation of the conceptual design, through performances on the development of a project scheduling management, integration of the individually developed technology, establishment of the product database, and so on, taking into account coupling of the activities conducted in each specific area

Towards a Fuzzy Bayesian Network Based Approach for Safety Risk Analysis of Tunnel-Induced Pipeline Damage.

Science.gov (United States)

Zhang, Limao; Wu, Xianguo; Qin, Yawei; Skibniewski, Miroslaw J; Liu, Wenli

2016-02-01

Tunneling excavation is bound to produce significant disturbances to surrounding environments, and the tunnel-induced damage to adjacent underground buried pipelines is of considerable importance for geotechnical practice. A fuzzy Bayesian networks (FBNs) based approach for safety risk analysis is developed in this article with detailed step-by-step procedures, consisting of risk mechanism analysis, the FBN model establishment, fuzzification, FBN-based inference, defuzzification, and decision making. In accordance with the failure mechanism analysis, a tunnel-induced pipeline damage model is proposed to reveal the cause-effect relationships between the pipeline damage and its influential variables. In terms of the fuzzification process, an expert confidence indicator is proposed to reveal the reliability of the data when determining the fuzzy probability of occurrence of basic events, with both the judgment ability level and the subjectivity reliability level taken into account. By means of the fuzzy Bayesian inference, the approach proposed in this article is capable of calculating the probability distribution of potential safety risks and identifying the most likely potential causes of accidents under both prior knowledge and given evidence circumstances. A case concerning the safety analysis of underground buried pipelines adjacent to the construction of the Wuhan Yangtze River Tunnel is presented. The results demonstrate the feasibility of the proposed FBN approach and its application potential. The proposed approach can be used as a decision tool to provide support for safety assurance and management in tunnel construction, and thus increase the likelihood of a successful project in a complex project environment. © 2015 Society for Risk Analysis.

Analysis of high burnup fuel safety issues

Energy Technology Data Exchange (ETDEWEB)

Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

2000-12-01

Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.

Analysis of high burnup fuel safety issues

International Nuclear Information System (INIS)

Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

2000-12-01

Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development

Criticality Safety Evaluation of Hanford Site High Level Waste Storage Tanks

Energy Technology Data Exchange (ETDEWEB)

ROGERS, C.A.

2000-02-17

This criticality safety evaluation covers operations for waste in underground storage tanks at the high-level waste tank farms on the Hanford site. This evaluation provides the bases for criticality safety limits and controls to govern receipt, transfer, and long-term storage of tank waste. Justification is provided that a nuclear criticality accident cannot occur for tank farms operations, based on current fissile material and operating conditions.

Criticality Safety Evaluation of Hanford Site High-Level Waste Storage Tanks

International Nuclear Information System (INIS)

ROGERS, C.A.

2000-01-01

This criticality safety evaluation covers operations for waste in underground storage tanks at the high-level waste tank farms on the Hanford site. This evaluation provides the bases for criticality safety limits and controls to govern receipt, transfer, and long-term storage of tank waste. Justification is provided that a nuclear criticality accident cannot occur for tank farms operations, based on current fissile material and operating conditions

LOCA analysis of SCWR-M with passive safety system

Energy Technology Data Exchange (ETDEWEB)

Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Fu, S.W. [Navy University of Engineering, Wuhan, Hubei (China); Xu, Z.H. [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China); Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Nuclear Technology, Karlsruhe Institute of Technology (KIT), Kaiserstr. 12, 76131 Karlsruhe (Germany)

2013-06-15

Highlights: • Application of the ATHLET-SC code to the trans-critical analysis for SCWR. • Development of a passive safety system for SCWR-M. • Analysis of hot/cold leg LOCA behaviour with different break size. • Introduction of some mitigation measures for SCWR-M -- Abstract: A new SCWR conceptual design (mixed spectrum supercritical water cooled reactor: SCWR-M) is proposed by Shanghai Jiao Tong University (SJTU). R and D activities covering core design, safety system design and code development of SCWR-M are launched at SJTU. Safety system design and analysis is one of the key tasks during the development of SCWR-M. Considering the current advanced reactor design, a new passive safety system for SCWR-M including isolation cooling system (ICS), accumulator injection system (ACC), gravity driven cooling system (GDCS) and automatic depressurization system (ADS) is proposed. Based on the modified and preliminarily assessed system code ATHLET-SC, loss of coolant accident (LOCA) analysis for hot and cold leg is performed in this paper. Three different break sizes are analyzed to clarify the hot and cold LOCA characteristics of the SCWR-M. The influence of the break location and break size on the safety performance of SCWR-M is also concluded. Several measures to induce the core coolant flow and to mitigate core heating up are also discussed. The results achieved so far demonstrate the feasibility of the proposed passive safety system to keep the SCWR-M core at safety condition during loss of coolant accident.

PA activity by using nuclear power plant safety demonstration and analysis

International Nuclear Information System (INIS)

Tsuchiya, Mitsuo; Kamimae, Rie

1999-01-01

INS/NUPEC presents one of Public acceptance (PA) methods for nuclear power in Japan, 'PA activity by using Nuclear Power Plant Safety Demonstration and Analysis', by using one of videos which is explained and analyzed accident events (Loss of Coolant Accident). Safety regulations of The National Government are strictly implemented in licensing at each of basic design and detailed design. To support safety regulation activities conducted by the National Government, INS/NLTPEC continuously implement Safety demonstration and analysis. With safety demonstration and analysis, made by assuming some abnormal conditions, what impacts could be produced by the assumed conditions are forecast based on specific design data on a given nuclear power plants. When analysis results compared with relevant decision criteria, the safety of nuclear power plants is confirmed. The decision criteria are designed to help judge if or not safety design of nuclear power plants is properly made. The decision criteria are set in the safety examination guidelines by taking sufficient safety allowance based on the latest technical knowledge obtained from a wide range of tests and safety studies. Safety demonstration and analysis is made by taking the procedure which are summarized in this presentation. In Japan, various PA (Public Acceptance) pamphlets and videos on nuclear energy have been published. But many of them focused on such topics as necessity or importance of nuclear energy, basic principles of nuclear power generation, etc., and a few described safety evaluation particularly of abnormal and accident events in accordance with the regulatory requirements. In this background, INS/NUPEC has been making efforts to prepare PA pamphlets and videos to explain the safety of nuclear power plants, to be simple and concrete enough, using various analytical computations for abnormal and accident events. In results, PA activity of INS/NUPEC is evaluated highly by the people

Investigation of Relationship between Level of Awareness around Health, Safety and Environment Management System and Its Effects on Safety Climate and Risk Perception by Employees in an Iran Oil Refinery, 2015

Directory of Open Access Journals (Sweden)

Milad Ahmadi Marzaleh

2017-02-01

Full Text Available Special attention of and oil, gas and petrochemical large corporation in the world to the HSE management system is due to its importance in the designing and development of products, services and processes by considering its health, safety and environment requirements. Staff's perception of the existing job risks has a significant impact on their safe behavior at work. This study was conducted to determine the relationship between safety climate and staff's perception of risk with an awareness level among employees of the HSE management system in an oil refinery in Kermanshah. The study population was employees in one of the oil refineries in Iran. After designing and questionnaire psychometric assessment of staff knowledge of HSE management system (Cronbach's alpha was 0.9 and its validity was assessed by certified professionals, Loughborough safety climate questionnaire and Flin risk perception questionnaire were used. Data analysis was performed using SPSS V22 software. Results showed that the relationship between safety climate and awareness level of the HSE management system; also the relationship between safety climate and perception of risk was also getting significant. However, the relationship between perception of risk and awareness level of the HSE management system was not significant. The results of this study showed a moderate awareness of HSE Management System in refinery workers. In this regard, appropriate and proper management policy should be committed to improving the situation. The results of this study is a profile the situation of safety climate in the refinery, which can be used as an indicator for the development of preventive policies and evaluate the performance of the organization's safety and the results of the safety improvement organization.