340 waste handling facility interim safety basis

Energy Technology Data Exchange (ETDEWEB)

VAIL, T.S.

1999-04-01

This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people.

340 waste handling facility interim safety basis

International Nuclear Information System (INIS)

VAIL, T.S.

1999-01-01

This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people

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.

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

International Nuclear Information System (INIS)

SCHWEHR, B.A.

1999-01-01

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

Interim safety basis compliance matrix for Trenches 31 and 34

International Nuclear Information System (INIS)

Ames, R.R.

1994-01-01

The tables provided in this document identify the specific requirements and basis for the administrative controls established in the Westinghouse Hanford Company (WHC) Solid Waste Burial Ground (SWBG) Interim Safety Basis (ISB) for operation of the Project W-025, Mixed Waste Lined Landfill (Trenches 31 and 34). The tables document the necessary controls and implementing procedures to ensure compliance with the requirements of the ISB. These requirements provide a basis for future Unreviewed Safety Questions (USQ) screening of applicable procedure changes, proposed physical modifications, tests, experiments, and occurrences. Table 1 provides the SWBG interim Operational Safety Requirements administrative controls matrix. The specific assumptions and commitments used in the safety analysis documents applicable to disposal of mixed wastes in Trenches 31 and 34 are provided in Table 2. Table 3 is provided to document the potential engineered and administrative mitigating features identified in the Preliminary Hazard Analysis (PHA) for disposal of mixed waste

Discussion on the Criterion for the Safety Certification Basis Compilation - Brazilian Space Program Case

Science.gov (United States)

Niwa, M.; Alves, N. C.; Caetano, A. O.; Andrade, N. S. O.

2012-01-01

The recent advent of the commercial launch and re- entry activities, for promoting the expansion of human access to space for tourism and hypersonic travel, in the already complex ambience of the global space activities, brought additional difficulties over the development of a harmonized framework of international safety rules. In the present work, with the purpose of providing some complementary elements for global safety rule development, the certification-related activities conducted in the Brazilian space program are depicted and discussed, focusing mainly on the criterion for certification basis compilation. The results suggest that the composition of a certification basis with the preferential use of internationally-recognized standards, as is the case of ISO standards, can be a first step toward the development of an international safety regulation for commercial space activities.

A new approach to determine the environmental qualification requirements for the safety related equipment

International Nuclear Information System (INIS)

Hasnaoui, C.; Parent, G.

2000-01-01

The objective of the environmental qualification of safety related equipment is to ensure that the plant defense-in-depth is not compromised by common mode failures following design basis accidents with a harsh environment. A new approach based on safety functions has been developed to determine what safety-related equipment is required to function during and after a design basis accident, as well as their environmental qualification requirements. The main feature of this approach is to use auxiliary safety functions established from safety requirements as credited in the safety analyses. This approach is undertaken in three steps: identification of the auxiliary safety functions of each main safety function; determination of the main equipment groups required for each auxiliary safety function; and review of the safety analyses for design basis accidents in order to determine the credited auxiliary safety functions and their mission times for each accident scenario. Some of the benefits of the proposed approach for the determination of the safety environmental qualification requirements are: a systematic approach for the review of safety analyses based on a safety function check list, and the insurance, with the availability of the safety functions, that Gentilly-2 defense-in-depth would not be compromised by design basis accidents with a harsh environment. (author)

Comparative analysis of safety related site characteristics

International Nuclear Information System (INIS)

Andersson, Johan

2010-12-01

This document presents a comparative analysis of site characteristics related to long-term safety for the two candidate sites for a final repository for spent nuclear fuel in Forsmark (municipality of Oesthammar) and in Laxemar (municipality of Oskarshamn) from the point of view of site selection. The analyses are based on the updated site descriptions of Forsmark /SKB 2008a/ and Laxemar /SKB 2009a/, together with associated updated repository layouts and designs /SKB 2008b and SKB 2009b/. The basis for the comparison is thus two equally and thoroughly assessed sites. However, the analyses presented here are focussed on differences between the sites rather than evaluating them in absolute terms. The document serves as a basis for the site selection, from the perspective of long-term safety, in SKB's application for a final repository. A full evaluation of safety is made for a repository at the selected site in the safety assessment SR-Site /SKB 2011/, referred to as SR-Site main report in the following

Comparative analysis of safety related site characteristics

Energy Technology Data Exchange (ETDEWEB)

Andersson, Johan (ed.)

2010-12-15

This document presents a comparative analysis of site characteristics related to long-term safety for the two candidate sites for a final repository for spent nuclear fuel in Forsmark (municipality of Oesthammar) and in Laxemar (municipality of Oskarshamn) from the point of view of site selection. The analyses are based on the updated site descriptions of Forsmark /SKB 2008a/ and Laxemar /SKB 2009a/, together with associated updated repository layouts and designs /SKB 2008b and SKB 2009b/. The basis for the comparison is thus two equally and thoroughly assessed sites. However, the analyses presented here are focussed on differences between the sites rather than evaluating them in absolute terms. The document serves as a basis for the site selection, from the perspective of long-term safety, in SKB's application for a final repository. A full evaluation of safety is made for a repository at the selected site in the safety assessment SR-Site /SKB 2011/, referred to as SR-Site main report in the following

Just in Time DSA-The Hanford Nuclear Safety Basis Strategy

Energy Technology Data Exchange (ETDEWEB)

Olinger, S. J.; Buhl, A. R.

2002-02-26

The U.S. Department of Energy, Richland Operations Office (RL) is responsible for 30 hazard category 2 and 3 nuclear facilities that are operated by its prime contractors, Fluor Hanford Incorporated (FHI), Bechtel Hanford, Incorporated (BHI) and Pacific Northwest National Laboratory (PNNL). The publication of Title 10, Code of Federal Regulations, Part 830, Subpart B, Safety Basis Requirements (the Rule) in January 2001 imposed the requirement that the Documented Safety Analyses (DSA) for these facilities be reviewed against the requirements of the Rule. Those DSA that do not meet the requirements must either be upgraded to satisfy the Rule, or an exemption must be obtained. RL and its prime contractors have developed a Nuclear Safety Strategy that provides a comprehensive approach for supporting RL's efforts to meet its long term objectives for hazard category 2 and 3 facilities while also meeting the requirements of the Rule. This approach will result in a reduction of the total number of safety basis documents that must be developed and maintained to support the remaining mission and closure of the Hanford Site and ensure that the documentation that must be developed will support: compliance with the Rule; a ''Just-In-Time'' approach to development of Rule-compliant safety bases supported by temporary exemptions; and consolidation of safety basis documents that support multiple facilities with a common mission (e.g. decontamination, decommissioning and demolition [DD&D], waste management, surveillance and maintenance). This strategy provides a clear path to transition the safety bases for the various Hanford facilities from support of operation and stabilization missions through DD&D to accelerate closure. This ''Just-In-Time'' Strategy can also be tailored for other DOE Sites, creating the potential for large cost savings and schedule reductions throughout the DOE complex.

Just in Time DSA-The Hanford Nuclear Safety Basis Strategy

International Nuclear Information System (INIS)

Olinger, S. J.; Buhl, A. R.

2002-01-01

The U.S. Department of Energy, Richland Operations Office (RL) is responsible for 30 hazard category 2 and 3 nuclear facilities that are operated by its prime contractors, Fluor Hanford Incorporated (FHI), Bechtel Hanford, Incorporated (BHI) and Pacific Northwest National Laboratory (PNNL). The publication of Title 10, Code of Federal Regulations, Part 830, Subpart B, Safety Basis Requirements (the Rule) in January 2001 imposed the requirement that the Documented Safety Analyses (DSA) for these facilities be reviewed against the requirements of the Rule. Those DSA that do not meet the requirements must either be upgraded to satisfy the Rule, or an exemption must be obtained. RL and its prime contractors have developed a Nuclear Safety Strategy that provides a comprehensive approach for supporting RL's efforts to meet its long term objectives for hazard category 2 and 3 facilities while also meeting the requirements of the Rule. This approach will result in a reduction of the total number of safety basis documents that must be developed and maintained to support the remaining mission and closure of the Hanford Site and ensure that the documentation that must be developed will support: compliance with the Rule; a ''Just-In-Time'' approach to development of Rule-compliant safety bases supported by temporary exemptions; and consolidation of safety basis documents that support multiple facilities with a common mission (e.g. decontamination, decommissioning and demolition [DD and D], waste management, surveillance and maintenance). This strategy provides a clear path to transition the safety bases for the various Hanford facilities from support of operation and stabilization missions through DD and D to accelerate closure. This ''Just-In-Time'' Strategy can also be tailored for other DOE Sites, creating the potential for large cost savings and schedule reductions throughout the DOE complex

Just in Time DSA the Hanford Nuclear Safety Basis Strategy

Energy Technology Data Exchange (ETDEWEB)

JACKSON, M.W.

2002-06-01

The U.S. Department of Energy, Richland Operations Office (RL) is responsible for 30 hazard category 2 and 3 nuclear facilities that are operated by its prime contractors, Fluor Hanford, Incorporated (FHI), Bechtel Hanford, Incorporated (BHI) and Pacific Northwest National Laboratory (PNNL). The publication of Title 10, Code of Federal Regulations, Part 830, Subpart B, Safely Basis Requirements (the Rule) in January 2001 requires that the Documented Safety Analyses (DSA) for these facilities be reviewed against the requirements of the Rule. Those DSAs that do not meet the requirements must either be upgraded to satisfy the Rule, or an exemption must be obtained. RL and its prime contractors have developed a Nuclear Safety Strategy that provides a comprehensive approach for supporting RL's efforts to meet its long-term objectives for hazard category 2 and 3 facilities while also meeting the requirements of the Rule. This approach will result in a reduction of the total number of safety basis documents that must be developed and maintained to support the remaining mission and closure of the Hanford Site and ensure that the documentation that must be developed will support: Compliance with the Rule; A ''Just-In-Time'' approach to development of Rule-compliant safety bases supported by temporary exemptions; and Consolidation of safety basis documents that support multiple facilities with a common mission (e.g. decontamination, decommissioning and demolition [DD&D], waste management, surveillance and maintenance). This strategy provides a clear path to transition the safety bases for the various Hanford facilities from support of operation and stabilization missions through DD&D to accelerate closure. This ''Just-In-Time'' Strategy can also be tailored for other DOE Sites, creating the potential for large cost savings and schedule reductions throughout the DOE complex.

Just in Time DSA the Hanford Nuclear Safety Basis Strategy

International Nuclear Information System (INIS)

JACKSON, M.W.

2002-01-01

The U.S. Department of Energy, Richland Operations Office (RL) is responsible for 30 hazard category 2 and 3 nuclear facilities that are operated by its prime contractors, Fluor Hanford, Incorporated (FHI), Bechtel Hanford, Incorporated (BHI) and Pacific Northwest National Laboratory (PNNL). The publication of Title 10, Code of Federal Regulations, Part 830, Subpart B, Safely Basis Requirements (the Rule) in January 2001 requires that the Documented Safety Analyses (DSA) for these facilities be reviewed against the requirements of the Rule. Those DSAs that do not meet the requirements must either be upgraded to satisfy the Rule, or an exemption must be obtained. RL and its prime contractors have developed a Nuclear Safety Strategy that provides a comprehensive approach for supporting RL's efforts to meet its long-term objectives for hazard category 2 and 3 facilities while also meeting the requirements of the Rule. This approach will result in a reduction of the total number of safety basis documents that must be developed and maintained to support the remaining mission and closure of the Hanford Site and ensure that the documentation that must be developed will support: Compliance with the Rule; A ''Just-In-Time'' approach to development of Rule-compliant safety bases supported by temporary exemptions; and Consolidation of safety basis documents that support multiple facilities with a common mission (e.g. decontamination, decommissioning and demolition [DD and D], waste management, surveillance and maintenance). This strategy provides a clear path to transition the safety bases for the various Hanford facilities from support of operation and stabilization missions through DD and D to accelerate closure. This ''Just-In-Time'' Strategy can also be tailored for other DOE Sites, creating the potential for large cost savings and schedule reductions throughout the DOE complex

Safety basis for the 241-AN-107 mixer pump installation and caustic addition

International Nuclear Information System (INIS)

Van Vleet, R.J.

1994-01-01

This safety Basis was prepared to determine whether or not the proposed activities of installing a 76 HP jet mixer pump and the addition of approximately 50,000 gallons of 19 M (50:50 wt %) aqueous caustic are within the safety envelope as described by Tank Farms (chapter six of WHC-SD-WM-ISB-001, Rev. 0). The safety basis covers the components, structures and systems for the caustic addition and mixer pump installation. These include: installation of the mixer pump and monitoring equipment; operation of the mixer pump, process monitoring equipment and caustic addition; the pump stand, caustic addition skid, the electrical skid, the video camera system and the two densitometers. Also covered is the removal and decontamination of the mixer pump and process monitoring system. Authority for this safety basis is WHC-IP-0842 (Waste Tank Administration). Section 15.9, Rev. 2 (Unreviewed Safety Questions) of WHC-IP-0842 requires that an evaluation be performed for all physical modifications

Transuranic waste storage and assay facility (TRUSAF) interim safety basis

International Nuclear Information System (INIS)

Gibson, K.D.

1995-09-01

The TRUSAF ISB is based upon current facility configuration and procedures. The purpose of the document is to provide the basis for interim operation or restrictions on interim operations and the authorization basis for the TRUSAF at the Hanford Site. The previous safety analysis document TRUSAF hazards Identification and Evaluation (WHC 1977) is superseded by this document

Central waste complex interim safety basis

International Nuclear Information System (INIS)

Cain, F.G.

1995-01-01

This interim safety basis provides the necessary information to conclude that hazards at the Central Waste Complex are controlled and that current and planned activities at the CWC can be conducted safely. CWC is a multi-facility complex within the Solid Waste Management Complex that receives and stores most of the solid wastes generated and received at the Hanford Site. The solid wastes that will be handled at CWC include both currently stored and newly generated low-level waste, low-level mixed waste, contact-handled transuranic, and contact-handled TRU mixed waste

Safety-related site characteristics - a relative comparison of the Forsmark reference areas

International Nuclear Information System (INIS)

Winberg, Anders

2010-12-01

SKB has over the years from 2002 to 2008 conducted site investigations in Forsmark and Laxemar, with associated site modeling, design and safety analysis. In mid-2009 Forsmark was selected on the basis of analysis made as site for a future repository for spent nuclear fuel. Based on defined safety-related geoscientific location factors data from Forsmark are compared in relative terms with data from a number of locations in Sweden, previously studied by SKB. The factors compared include: the rock's composition and structures, future climate evolution, rock mechanical conditions, earthquakes, groundwater flow, groundwater composition, delay of solutes, and the ability to characterize and describe the location. Past comparisons of these properties for the selected sites show that none of these sites collectively show any significant benefit over Forsmark site for a repository. This does not preclude that there may be places on the basis of an overall assessment of geoscientific location factors could be equivalent to Forsmark

Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

Energy Technology Data Exchange (ETDEWEB)

David Petti; Jim Kinsey; Dave Alberstein

2014-01-01

Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

DEVELOPING SAFETY INDICATORS ON THE BASIS OF THE ICAO RECOMMENDATIONS

Directory of Open Access Journals (Sweden)

V. D. Sharov

2014-01-01

Full Text Available The article offers direct use of the recommendations of SMM ICAO Doc.9859, 3rd ed. 2013, for calculation the target and alert levels of safety indicators. Examples of calculation based on data of 2011 and monitoring of the current indicators during 2012 are presented. Safety indicators for airlines in terms of “numbers of incidents per 1000 flight hours” could be calculated on the basis of the state values through the «coefficient of conformity».

Authorization Basis Safety Classification of Transfer Bay Bridge Crane at the 105-K Basins

International Nuclear Information System (INIS)

CHAFFEE, G.A.

2000-01-01

This supporting document provides the bases for the safety classification for the K Basin transfer bay bridge crane and the bases for the Structures, Systems, and Components (SSC) safety classification. A table is presented that delineates the safety significant components. This safety classification is based on a review of the Authorization Basis (AB). This Authorization Basis review was performed regarding AB and design baseline issues. The primary issues are: (1) What is the AB for the safety classification of the transfer bay bridge crane? (2) What does the SSC safety classification ''Safety Significant'' or ''Safety Significant for Design Only'' mean for design requirements and quality requirements for procurement, installation and maintenance (including replacement of parts) activities for the crane during its expected life time? The AB information on the crane was identified based on review of Department of Energy--Richland Office (RL) and Spent Nuclear Fuel (SNF) Project correspondence, K Basin Safety Analysis Report (SAR) and RL Safety Evaluation Reports (SERs) of SNF Project SAR submittals. The relevant correspondence, actions and activities taken and substantive directions or conclusions of these documents are provided in Appendix A

Fundamentals of a graded approach to safety-related equipment setpoints

International Nuclear Information System (INIS)

Woodruff, B.A.; Cash, J.S. Jr.; Bockhorst, R.M.

1993-01-01

The concept of using a graded approach to reconstitute instrument setpoints associated with safety-related equipment was first presented to the industry by the U.S. Nuclear Regulatory Commission during the 1992 ISA/POWID Symposium in Kansas City, Missouri. The graded approach establishes that the manner in which a utility analyzes and documents setpoints is related to each setpoint's relative importance to safety. This allows a utility to develop separate requirements for setpoints of varying levels of safety significance. A graded approach to setpoints is a viable strategy that minimizes extraneous effort expended in resolving difficult issues that arise when formal setpoint methodology is applied blindly to all setpoints. Close examination of setpoint methodology reveals that the application of a graded approach is fundamentally dependent on the analytical basis of each individual setpoint

DARHT: INTEGRATION OF AUTHORIZATION BASIS REQUIREMENTS AND WORKER SAFETY

International Nuclear Information System (INIS)

MC CLURE, D. A.; NELSON, C. A.; BOUDRIE, R. L.

2001-01-01

This document describes the results of consensus agreements reached by the DARHT Safety Planning Team during the development of the update of the DARHT Safety Analysis Document (SAD). The SAD is one of the Authorization Basis (AB) Documents required by the Department prior to granting approval to operate the DARHT Facility. The DARHT Safety Planning Team is lead by Mr. Joel A. Baca of the Department of Energy Albuquerque Operations Office (DOE/AL). Team membership is drawn from the Department of Energy Albuquerque Operations Office, the Department of Energy Los Alamos Area Office (DOE/LAAO), and several divisions of the Los Alamos National Laboratory. Revision 1 of the DARHT SAD had been written as part of the process for gaining approval to operate the Phase 1 (First Axis) Accelerator. Early in the planning stage for the required update of the SAD for the approval to operate both Phase 1 and Phase 2 (First Axis and Second Axis) DARHT Accelerator, it was discovered that a conflict existed between the Laboratory approach to describing the management of facility and worker safety

10 CFR Appendix A to Subpart B of... - General Statement of Safety Basis Policy

Science.gov (United States)

2010-01-01

... at all levels. Performing work in accordance with the safety basis for a nuclear facility is the..., safety, and health into work planning and execution (48 CFR 970.5223-1, Integration of Environment, Safety and Health into Work Planning and Execution) and the DEAR clause on laws, regulations, and DOE...

Recent findings relating to firefighter safety zones

Science.gov (United States)

Bret Butler; Russ Parsons; William Mell

2015-01-01

Designation of safety zones is a primary duty of all wildland firefighters. Unfortunately, information regarding what constitutes an adequate safety zone is inadequately defined. Measurements of energy release from wildland fires have been used to develop an empirically based safety zone guideline. The basis for this work is described here.

Working group 1A - basis for the standard-safety

International Nuclear Information System (INIS)

Whipple, C.

1993-01-01

This paper presents a summary of the progress made by working group 1A (Basis for the Safety Standard) during the Electric Power Research Institute's EPRI Workshop on the technical basis of EPA HLW Disposal Criteria, March 1993. This group discussed the semantics of terms within the standard 40 CFR Part 191, the implementation of this standard, the advanced notice of rulemaking, the issue of emitting carbon-14 through a gaseous pathway, the strategy of dealing with standards for contamination of drinking water and groundwater, the 100,000 year time frame, and the analysis of specific comments. The specific comments dealt with the cost effectiveness of the standard, the dose histogram for populations and individuals, groundwater definition and the underlying technology driver for this standard

Scientific basis for a safety case of deep geological repositories

Energy Technology Data Exchange (ETDEWEB)

Noseck, Ulrich; Becker, Dirk-Alexander; Brasser, Thomas [and others

2012-11-15

Within this project strategies and methods to build a safety case for deep geological repositories are further developed. This includes also the scientific fundamentals as a basis of the safety case. In the international framework the methodology of the Safety Case is frequently applied and continuously improved. According to definitions from IAEA and NEA the Safety Case is a compilation of arguments and facts, which describe, quantify and support the safety and the degree of confidence in the safety of the geological repository. The safety of the geological repository should be demonstrated by the safety case. The safety case is the basis for essential decisions during a repository programme. It comprises the results of safety assessments in combination with additional information like multiple lines of evidence and a discussion of robustness and quality of the repository, its design and the quality of all safety assessments including the basic assumptions. A crucial element of the Safety Case is the long-term safety analysis, i.e. the systematic analysis of the hazards connected with the facility and the capability of site and repository design to ensure the required safety functions and to fulfill the technical claims. Long-term safety analysis requires a powerful and qualified programme package, which contains appropriate hardware and software as well as well trained and experienced modellers performing the model calculations. The calculation tools used within safety cases need to be checked and verified and continuously adapted to the state-of-the-art science and technology. Especially it needs to be applicable to a real repository system. For the assessment of safety a deep process understanding is necessary. The R and D work performed within this project will contribute to the improvement of process and system understanding as well as to the further development of methods and strategies applied in the safety case. Emphasis was put on the following aspects

Scientific basis for a safety case of deep geological repositories

International Nuclear Information System (INIS)

Noseck, Ulrich; Becker, Dirk-Alexander; Brasser, Thomas

2012-11-01

Within this project strategies and methods to build a safety case for deep geological repositories are further developed. This includes also the scientific fundamentals as a basis of the safety case. In the international framework the methodology of the Safety Case is frequently applied and continuously improved. According to definitions from IAEA and NEA the Safety Case is a compilation of arguments and facts, which describe, quantify and support the safety and the degree of confidence in the safety of the geological repository. The safety of the geological repository should be demonstrated by the safety case. The safety case is the basis for essential decisions during a repository programme. It comprises the results of safety assessments in combination with additional information like multiple lines of evidence and a discussion of robustness and quality of the repository, its design and the quality of all safety assessments including the basic assumptions. A crucial element of the Safety Case is the long-term safety analysis, i.e. the systematic analysis of the hazards connected with the facility and the capability of site and repository design to ensure the required safety functions and to fulfill the technical claims. Long-term safety analysis requires a powerful and qualified programme package, which contains appropriate hardware and software as well as well trained and experienced modellers performing the model calculations. The calculation tools used within safety cases need to be checked and verified and continuously adapted to the state-of-the-art science and technology. Especially it needs to be applicable to a real repository system. For the assessment of safety a deep process understanding is necessary. The R and D work performed within this project will contribute to the improvement of process and system understanding as well as to the further development of methods and strategies applied in the safety case. Emphasis was put on the following aspects

Safety of Research Reactors. Safety Requirements

International Nuclear Information System (INIS)

2010-01-01

The main objective of this Safety Requirements publication is to provide a basis for safety and a basis for safety assessment for all stages in the lifetime of a research reactor. Another objective is to establish requirements on aspects relating to regulatory control, the management of safety, site evaluation, design, operation and decommissioning. Technical and administrative requirements for the safety of research reactors are established in accordance with these objectives. This Safety Requirements publication is intended for use by organizations engaged in the site evaluation, design, manufacturing, construction, operation and decommissioning of research reactors as well as by regulatory bodies

Climate and climate-related issues for the safety assessment SR-Site

International Nuclear Information System (INIS)

2010-12-01

The purpose of this report is to document current scientific knowledge on climate and climate-related conditions, relevant to the long-term safety of a KBS-3 repository, to a level required for an adequate treatment in the safety assessment SR-Site. The report also presents a number of dedicated studies on climate and selected climate-related processes of relevance for the assessment of long term repository safety. Based on this information, the report presents a number of possible future climate developments for Forsmark, the site selected for building a repository for spent nuclear fuel in Sweden (Figure 1-1). The presented climate developments are used as basis for the selection and analysis of SR-Site safety assessment scenarios in the SR-Site main report /SKB 2011/. The present report is based on research conducted and published by SKB as well as on research reported in the general scientific literature

Climate and climate-related issues for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

2010-12-15

The purpose of this report is to document current scientific knowledge on climate and climate-related conditions, relevant to the long-term safety of a KBS-3 repository, to a level required for an adequate treatment in the safety assessment SR-Site. The report also presents a number of dedicated studies on climate and selected climate-related processes of relevance for the assessment of long term repository safety. Based on this information, the report presents a number of possible future climate developments for Forsmark, the site selected for building a repository for spent nuclear fuel in Sweden (Figure 1-1). The presented climate developments are used as basis for the selection and analysis of SR-Site safety assessment scenarios in the SR-Site main report /SKB 2011/. The present report is based on research conducted and published by SKB as well as on research reported in the general scientific literature

Safety in relation to risk and benefit

International Nuclear Information System (INIS)

Siddall, E.

1985-01-01

The proper definition and quantification of human safety is discussed and from this basis the historical development of our present very high standard of safety is traced. It is shown that increased safety is closely associated with increased wealth, and the quantitative relationship between then is derived from different sources of evidence. When this factor is applied to the production of wealth by industry, a safety benefit is indicated which exceeds the asserted risks by orders of magnitude. It is concluded that present policies and attitudes in respect to the safety of industry may be diametrically wrong. (orig.) [de

Relative hazard potential: the basis for definition of safety criteria for fast reactors

International Nuclear Information System (INIS)

Cave, L.; Ilberg, D.

1977-02-01

One of the main safety criteria to be met for larger thermal reactors is that the probability of exceeding the dose limits imposed by 10 CRF 100 should not be greater than 10 per reactor year. The potential hazard presented by a fast reactor could be substantially greater than that due to an LWR. The potential for harm of a reactor system may be judged by the effects which would arise from a severe accident. Several different types of effects may be considered: number of latent fatal cancers; number of deaths due to acute effects; number of thyroid tumors or nodules; extent of property damage; and genetic effects. Analytical methods for comparison are employed in this paper. A second important parameter reviewed in this report is the radio-toxicity attributed to the various isotopes. It was found that the worst conceivable accident to a 1000 MW(e) fast reactor would lead to effects on health greater by an order of magnitude than the worst accident usually considered for an LWR. Therefore, some reconsideration of the need for additional safety criteria for LMFBRs, as a guide to designers in relation to the control of the effects of very severe accidents, is desirable

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.

PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

Energy Technology Data Exchange (ETDEWEB)

Weber, P.; Umminger, K.J.; Schoen, B. [Siemens AG Power Generation Group (KWU), Erlangen (France)

1995-09-01

The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where the decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).

Materials Safety Data Sheets: the basis for control of toxic chemicals

Energy Technology Data Exchange (ETDEWEB)

Ketchen, E.E.; Porter, W.E.

1979-09-01

The Material Safety Data Sheets contained in this volume are the basis for the Toxic Chemical Control Program developed by the Industrial Hygiene Department, Health Division, ORNL. The three volumes are the update and expansion of ORNL/TM-5721 and ORNL/TM-5722 Material Safety Data Sheets: The Basis for Control of Toxic Chemicals, Volume I and Volume II. As such, they are a valuable adjunct to the data cards issued with specific chemicals. The chemicals are identified by name, stores catalog number where appropriate, and sequence numbers from the NIOSH Registry of Toxic Effects of Chemical Substances, 1977 Edition, if available. The data sheets were developed and compiled to aid in apprising the employees of hazards peculiar to the handling and/or use of specific toxic chemicals. Space limitation necessitate the use of descriptive medical terms and toxicological abbreviations. A glossary and an abbreviation list were developed to define some of those sometimes unfamiliar terms and abbreviations. The page numbers are keyed to the catalog number in the chemical stores at ORNL.

IAEA activity related to safety of nuclear desalination

International Nuclear Information System (INIS)

Gasparini, M.

2000-01-01

The nuclear plants for desalination to be built in the future will have to meet the standards of safety required for the best nuclear power plants currently in operation or being designed. The current safety approach, based on the achievement of the fundamental safety functions and defence in depth strategy, has been shown to be a sound foundation for the safety and protection of public health, and gives the plant the capability of dealing with a large variety of sequences, even beyond the design basis. The Department of Nuclear Safety of the IAEA is involved in many activities, the most important of which are to establish safety standards, and to provide various safety services and technical knowledge in many Technical Co-operation assistance projects. The department is also involved in other safety areas, notably in the field of future reactors. The IAEA is carrying out a project on the safety of new generation reactors, including those used for desalination, with the objective of fostering an exchange of information on safety approaches, promoting harmonization among Member States and contributing towards the development and revision of safety standards and guidelines for nuclear power plant design. The safety, regulatory and environmental concerns in nuclear powered desalination are those related directly to nuclear power plants, with due consideration given to the coupling process. The protection of product water against radioactive contamination must be ensured. An effective infrastructure, including appropriate training, a legal framework and regulatory regime, is a prerequisite to considering use of nuclear power for desalination plants, also in those countries with limited industrial infrastructures and little experience in nuclear technology or safety. (author)

Safety basis for selected activities in single-shell tanks with flammable gas concerns. Revision 1

International Nuclear Information System (INIS)

Schlosser, R.L.

1996-01-01

This is full revision to Revision 0 of this report. The purpose of this report is to provide a summary of analyses done to support activities performed for single-shell tanks. These activities are encompassed by the flammable gas Unreviewed Safety Question (USQ). The basic controls required to perform these activities involve the identification, elimination and/or control of ignition sources and monitoring for flammable gases. Controls are implemented through the Interim Safety Basis (ISB), IOSRs, and OSDs. Since this report only provides a historical compendium of issues and activities, it is not to be used as a basis to perform USQ screenings and evaluations. Furthermore, these analyses and others in process will be used as the basis for developing the Flammable Gas Topical Report for the ISB Upgrade

Value-impact assessment of safety-related modifications

International Nuclear Information System (INIS)

Knowles, W.M.C.; Dinnie, K.S.; Gordon, C.W.

1992-01-01

Like other nuclear utilities, Ontario Hydro, as part of its risk management activities, continually assesses the safety of its nuclear operations. In addition, new regulatory requirements are being applied to the older nuclear power plants. Both of these result in proposed plant modifications designed to reduce the risk to the public. However, modifications to an operating plant can have serious economic effects, and the resources, both financial and personnel, required for the implementation of these modifications are limited. Thus, all potential benefits and effects of a proposed modification must be thoroughly investigated to judge whether the modification is beneficial. Ontario Hydro has begun to use comprehensive value-impact assessments, utilizing plant-specific probabilistic risk assessments (PRAs), as tools to provide an informed basis for judgments on the benefit of safety-related modifications. The results from value-impact assessments can also be used to prioritize the implementation of these modifications

Ferrocyanide Safety Program: Safety criteria for ferrocyanide watch list tanks

International Nuclear Information System (INIS)

Postma, A.K.; Meacham, J.E.; Barney, G.S.

1994-01-01

This report provides a technical basis for closing the ferrocyanide Unreviewed Safety Question (USQ) at the Hanford Site. Three work efforts were performed in developing this technical basis. The efforts described herein are: 1. The formulation of criteria for ranking the relative safety of waste in each ferrocyanide tank. 2. The current classification of tanks into safety categories by comparing available information on tank contents with the safety criteria; 3. The identification of additional information required to resolve the ferrocyanide safety issue

29 CFR 1975.2 - Basis of authority.

Science.gov (United States)

2010-07-01

... Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR (CONTINUED) COVERAGE OF EMPLOYERS UNDER THE WILLIAMS-STEIGER OCCUPATIONAL SAFETY AND HEALTH ACT OF 1970 § 1975.2 Basis... Occupational Safety and Health Act of 1970, is derived mainly from the Commerce Clause of the Constitution...

Reactor safety under design basis flood condition for inland sites

International Nuclear Information System (INIS)

Hajela, S.; Bajaj, S.S.; Samota, A.; Verma, U.S.P.; Warudkar, A.S.

2002-01-01

Full text: In June 1994, there was an incident of flooding at Kakrapar Atomic Power Station (KAPS) due to combination of heavy rains and mechanical failure in the operation of gates at the adjoining weir. An indepth review of the incident was carried out and a number of flood protection measures were recommended and were implemented at site. As part of this review, a safety analysis was also done to demonstrate reactor safety with a series of failures considered in the flood protection features. For each inland NPP site, as part of design, different flood scenarios are analysed to arrive at design basis flood (DBF) level. This level is estimated based on worst combination of heavy local precipitation, flooding in river, failure of upstream/downstream water control structures

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Licensing decisions and safety research related to LMFBR accidents

International Nuclear Information System (INIS)

Denise, R.P.; Speis, T.P.; Kelber, C.N.; Curtis, R.T.

1977-01-01

The licensing approach which ensures adequate protection of the public health and safety against serious accidents is described. This paper describes the role of core melt and core disruptive accidents in the design, safety research, and licensing processes, using the Clinch River Breeder Reactor (CRBR) as a focal point. Major design attention is placed on the prevention of these accidents so that the probability of core melt accidents is reduced to a sufficiently low level that they are not treated as design basis accidents. Additional requirements are placed upon the design to further reduce residual risk. This licensing process is supported by a confirmatory research program designed to provide an independent basis for licensing judgements. It has as a goal the resolution of generic safety issues prior to the establishment of a commercial LMFBR industry. The program includes accident analysis, experiments in materials interactions, aerosol transport and system integrity and planning for new safety test facilities. The problems are approached in a multi-disciplinary functional manner that identifies key safety issues and centralizes efforts to resolve them. The near term objectives of the program support the licensing of the Clinch River Breeder Reactor (CRBR) and the proposed Prototype Large Breeder Reactor (PLBR). The long term objectives of the program support the licensing of commercial LMFBRs during the late 1980's and beyond. This safety research is designed to provide an independent basis for the licensing judgements which must be made by the Nuclear Regulatory Commission

Formal safety assessment based on relative risks model in ship navigation

Energy Technology Data Exchange (ETDEWEB)

Hu Shenping [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: sphu@mmc.shmtu.edu.cn; Fang Quangen [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: qgfang@mmc.shmtu.edu.cn; Xia Haibo [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: hbxia@mmc.shmtu.edu.cn; Xi Yongtao [Merchant Marine College, Shanghai Maritime University, 1550, Pudong Dadao, Shanghai 200135 (China)]. E-mail: xiyt@mmc.shmtu.edu.cn

2007-03-15

Formal safety assessment (FSA) is a structured and systematic methodology aiming at enhancing maritime safety. It has been gradually and broadly used in the shipping industry nowadays around the world. On the basis of analysis and conclusion of FSA approach, this paper discusses quantitative risk assessment and generic risk model in FSA, especially frequency and severity criteria in ship navigation. Then it puts forward a new model based on relative risk assessment (MRRA). The model presents a risk-assessment approach based on fuzzy functions and takes five factors into account, including detailed information about accident characteristics. It has already been used for the assessment of pilotage safety in Shanghai harbor, China. Consequently, it can be proved that MRRA is a useful method to solve the problems in the risk assessment of ship navigation safety in practice.

Formal safety assessment based on relative risks model in ship navigation

International Nuclear Information System (INIS)

Hu Shenping; Fang Quangen; Xia Haibo; Xi Yongtao

2007-01-01

Formal safety assessment (FSA) is a structured and systematic methodology aiming at enhancing maritime safety. It has been gradually and broadly used in the shipping industry nowadays around the world. On the basis of analysis and conclusion of FSA approach, this paper discusses quantitative risk assessment and generic risk model in FSA, especially frequency and severity criteria in ship navigation. Then it puts forward a new model based on relative risk assessment (MRRA). The model presents a risk-assessment approach based on fuzzy functions and takes five factors into account, including detailed information about accident characteristics. It has already been used for the assessment of pilotage safety in Shanghai harbor, China. Consequently, it can be proved that MRRA is a useful method to solve the problems in the risk assessment of ship navigation safety in practice

14 CFR 60.37 - FSTD qualification on the basis of a Bilateral Aviation Safety Agreement (BASA).

Science.gov (United States)

2010-01-01

... Bilateral Aviation Safety Agreement (BASA). 60.37 Section 60.37 Aeronautics and Space FEDERAL AVIATION... CONTINUING QUALIFICATION AND USE § 60.37 FSTD qualification on the basis of a Bilateral Aviation Safety... on International Civil Aviation for the sponsor of an FSTD located in that contracting State may be...

The deep geologic repository technology programme: toward a geoscience basis for understanding repository safety

International Nuclear Information System (INIS)

Jensen, M.R.

2007-01-01

Within the Deep Geologic Repository Technology Programme (DGRTP) several Geoscience activities are focused on advancing the understanding of groundwater flow system evolution and geochemical stability in a Canadian Shield setting as affected by long-term climate change. A key aspect is developing confidence in predictions of groundwater flow patterns and residence times as they relate to the safety of a deep geologic repository for used nuclear fuel waste. This is being achieved through a coordinated multi-disciplinary approach intent on: i) demonstrating coincidence between independent geo-scientific data; ii) improving the traceability of geo-scientific data and its interpretation within a conceptual descriptive model(s); iii) improving upon methods to assess and demonstrate robustness in flow domain prediction(s) given inherent flow domain uncertainties (i.e. spatial chemical/physical property distributions, boundary conditions) in time and space; and iv) improving awareness amongst geo-scientists as to the utility of various geo-scientific data in supporting a safety case for a deep geologic repository. This multi-disciplinary DGRTP approach is yielding an improved understanding of groundwater flow system evolution and stability in Canadian Shield settings that is further contributing to the geo-scientific basis for understanding and communicating aspects of DGR safety. (author)

Organic reactivity analysis in Hanford single-shell tanks: Experimental and modeling basis for an expanded safety criterion

International Nuclear Information System (INIS)

Fauske, H.; Grigsby, J.M.; Turner, D.A.; Babad, H.; Meacham, J.E.

1996-01-01

De-spite demonstrated safe storage in terms of chemical stability of the Hanford high level waste for many decades, including decreasing waste temperatures and continuing aging of chemicals to less energetic states, concerns continue relative to assurance of long-term safe storage. Review of potential chemical safety hazards has been of particular recent interest in response to serious incidents within the Nuclear Weapons Complexes in the former Soviet Union (the 1957 Kyshtym and the 1993 Tomsk-7 incidents). Based upon an evaluation of the extensive new information and understanding that have developed over the last few years, it is concluded that the Hanford waste is stored safely and that concerns related to potential chemical safety hazards are not warranted. Spontaneous bulk runaway reactions of the Kyshtym incident type and other potential condensed-phase propagating reactions can be ruled out by assuring appropriate tank operating controls are in place and by limiting tank intrusive activities. This paper summarizes the technical basis for this position

The biological basis of plutonium safety standards

International Nuclear Information System (INIS)

Mole, R.H.

1976-01-01

Since no radiation injury or cancer in man can, as yet, be directly attributed to Pu, all safety standards for Pu must be determined by reference to other safety standards, development of which is discussed. A system of safety standards must be based on links with real damage, such as the requirement for 226 Ra in bone. The type of biological information required for making standards realistic is considered in relation to Pu and Ra in bone. Also considered are the possible effects of Pu in soft tissue such as bone marrow. Not only dose, but also the number of cells exposed to the dose are important biologically and cellular aspects are examined. Since there is no positive evidence of Pu toxicity relevant information on other α emitters must be examined. The observed effectiveness of Ra, daughters of 222 Ra and 232 Th in causing mutations and cancer, is surveyed. Reference is made to the necessity of improving the ICRP system, currently based on the critical organ concept, by recognising the need for summation of risks in other organs where exposure to Pu is concerned. Improved biological understanding particularly that of hereditary damage, in recent years leads to less pessimistic thinking on the effects of ionizing radiations. The immediate need appears to be for consistency in safety standards. (U.K.)

General safety basis development guidance for environmental restoration decontamination and decommissioning

International Nuclear Information System (INIS)

Ellingson, D.R.; Kerr, N.; Bohlander, K.; Hansen, J.; Crowley, W.

1994-02-01

Safety analyses have the objective of contributing to two essential ingredients of a successful operation. The first is promoting the safety of the operation through worker involvement in information development (safety basis). The second is obtaining approval to conduct the operation (authorization). Typically these ingredients are assembled under separate programs covered by separate DOE requirements. DOE authorization relies on successful development of a document containing up to 21 topics written in terms and language suited to reviewers and approvers. Safety relies on successful training and procedures that convert the technical documented information into terms and language understandable to the worker. This separation can lead to successful incorporation of one ingredient independent of the other. At best, this separation may result in a safe but unauthorized operation; at worst, the separation may result in an unsafe operation authorized to proceed. This guide is based on experiences gained by contractors who have integrated rather than separated the safety and authorization. The short duration of ER/D ampersand D activities, the uncertainties of hazards, and the publicly expressed desire for demonstrable progress in cleanup activities add emphasis to the need to integrate rather than separate and develop new programs. Experience-based information has been useful to workers, safety analysis practitioners, and reviewers in the following ways: (1) Acquiring or developing the needed information in a useful form; (2) Managing the uncertainties during activity development and operation; (3) Identifying the subset of applicable requirements for an activity; (4) Developing the appropriate level of documentation detail for a specific activity; and (5) Increasing the usefulness and use of safety analysis (ownership)

Common basis of establishing safety standards and other safety decision-making levels for different sources of health risk

International Nuclear Information System (INIS)

Demin, V.F.

2002-01-01

Current approaches in establishing safety standards and other decision-making levels for different sources of health risk are critically analysed. To have a common basis for this decision-making a specific risk index R is recommended. In the common sense R is quantitatively defined as LLE caused by the annual exposure to the risk source considered: R = annual exposure, damage (LLE) from the exposure unit. This common definition is also rewritten in specific forms for a set of different risk sources (ionising radiation, chemical pollutants, etc): for different risk sources the exposure can be measured with different quantities (the probability of death, the exposure dose, etc.). R is relative LLE: LLE in years referred to 1 year under the risk. The dimension of this value is [year/year]. In the statistical sense R is conditionally the share of the year, which is lost due to exposure to a risk source during this year. In this sense R can be called as the relative damage. Really lifetime years are lost after the exposure. R can be in some conditional sense considered as a dimensionless quantity. General safety standards R n for the public and occupational workers have been suggested in terms of this index: R n = 0.0007 and 0.01 accordingly. Secondary safety standards are derived for a number of risk sources (ionising radiation, environmental chemical pollutants, etc). Values of R n are chosen in such a way that to have the secondary radiation BSS being equivalent to the current one's. Other general and derived levels for safety decision-making are also proposed including the de-minimus levels. Their possible dependence on the national or regional health-demographic data (HDD) is considered. Such issues as the ways of the integration and averaging of risk indices considered through the national or regional HDD for different risk sources and the use of non-threshold linear exposure - response relationships for ionising radiation and chemical pollutants are analysed

Safety-related site characteristics - a relative comparison of the Forsmark reference areas; Saekerhetsrelaterade platsegenskaper - en relativ jaemfoerelse av Forsmark med referensomraaden

Energy Technology Data Exchange (ETDEWEB)

Winberg, Anders (Conterra AB, Uppsala (Sweden))

2010-12-15

SKB has over the years from 2002 to 2008 conducted site investigations in Forsmark and Laxemar, with associated site modeling, design and safety analysis. In mid-2009 Forsmark was selected on the basis of analysis made as site for a future repository for spent nuclear fuel. Based on defined safety-related geoscientific location factors data from Forsmark are compared in relative terms with data from a number of locations in Sweden, previously studied by SKB. The factors compared include: the rock's composition and structures, future climate evolution, rock mechanical conditions, earthquakes, groundwater flow, groundwater composition, delay of solutes, and the ability to characterize and describe the location. Past comparisons of these properties for the selected sites show that none of these sites collectively show any significant benefit over Forsmark site for a repository. This does not preclude that there may be places on the basis of an overall assessment of geoscientific location factors could be equivalent to Forsmark

Flammable gas tank safety program: Technical basis for gas analysis and monitoring

International Nuclear Information System (INIS)

Estey, S.D.

1998-01-01

Several Hanford waste tanks have been observed to exhibit periodic releases of significant quantities of flammable gases. Because potential safety issues have been identified with this type of waste behavior, applicable tanks were equipped with instrumentation offering the capability to continuously monitor gases released from them. This document was written to cover three primary areas: (1) describe the current technical basis for requiring flammable gas monitoring, (2) update the technical basis to include knowledge gained from monitoring the tanks over the last three years, (3) provide the criteria for removal of Standard Hydrogen Monitoring System(s) (SHMS) from a waste tank or termination of other flammable gas monitoring activities in the Hanford Tank farms

Safety-related control air systems

International Nuclear Information System (INIS)

Anon.

1977-01-01

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

Optimization of the nuclear power engineering safety on the basis of social and economic parameters

International Nuclear Information System (INIS)

Kozlov, V.F.; Kuz'min, I.I.; Lystsov, V.N.; Amosova, T.V.; Makhutov, N.A.; Men'shikov, V.F.

1995-01-01

Principle of optimization of nuclear power engineering safety is presented on the basis of estimating the risks to the man's health with an account of peculiarities of socio-economic system and other types of economic activities in the region. Average expected duration of forthcoming life and costs of its prolongation serve as a unit for measuring the man's safety. It is shown that if the expenditures on NPP technical safety exceed the scientifically substantiated costs for this region with application of the above principle, than the risk for population will exceed the minimum achievable level. 8 refs., 2 figs., 1 tab

To the problem of the statistical basis of evaluation of the mechanical safety factor

International Nuclear Information System (INIS)

Tsyganov, S.V.

2009-01-01

The methodology applied for the safety factor assessment of the WWER fuel cycles uses methods and terms of statistics. Value of the factor is calculated on the basis of estimation of probability to meet predefined limits. Such approach demands the special attention to the statistical properties of parameters of interest. Considering the mechanical constituents of the engineering factor it is assumed uncertainty factors of safety parameters are stochastic values. It characterized by probabilistic distributions that can be unknown. Traditionally in the safety factor assessment process the unknown parameters are estimated from the conservative points of view. This paper analyses how the refinement of the factors distribution parameters is important for the assessment of the mechanical safety factor. For the analysis the statistical approach is applied for modelling of different type of factor probabilistic distributions. It is shown the significant influence of the shape and parameters of distributions for some factors on the value of mechanical safety factor. (Authors)

To the problem of the statistical basis of evaluation of the mechanical safety factor

International Nuclear Information System (INIS)

Tsyganov, S.

2009-01-01

The methodology applied for the safety factor assessment of the VVER fuel cycles uses methods and terms of statistics. Value of the factor is calculated on the basis of estimation of probability to meet predefined limits. Such approach demands the special attention to the statistical properties of parameters of interest. Considering the mechanical constituents of the engineering factor it is assumed uncertainty factors of safety parameters are stochastic values. It characterized by probabilistic distributions that can be unknown. Traditionally in the safety factor assessment process the unknown parameters are estimated from the conservative points of view. This paper analyses how the refinement of the factors distribution parameters is important for the assessment of the mechanical safety factor. For the analysis the statistical approach is applied for modelling of different type of factor probabilistic distributions. It is shown the significant influence of the shape and parameters of distributions for some factors on the value of mechanical safety factor. (author)

ITER safety task NID-5a: ITER tritium environmental source terms - safety analysis basis

International Nuclear Information System (INIS)

Natalizio, A.; Kalyanam, K.M.

1994-09-01

The Canadian Fusion Fuels Technology Project's (CFFTP) is part of the contribution to ITER task NID-5a, Initial Tritium Source Term. This safety analysis basis constitutes the first part of the work for establishing tritium source terms and is intended to solicit comments and obtain agreement. The analysis objective is to provide an early estimate of tritium environmental source terms for the events to be analyzed. Events that would result in the loss of tritium are: a Loss of Coolant Accident (LOCA), a vacuum vessel boundary breach. a torus exhaust line failure, a fuelling machine process boundary failure, a fuel processing system process boundary failure, a water detritiation system process boundary failure and an isotope separation system process boundary failure. 9 figs

The Argentine Approach to Radiation Safety: Its Ethical Basis

International Nuclear Information System (INIS)

Gonzalez, A.J.

2011-01-01

The ethical bases of Argentina's radiation safety approach are reviewed. The applied principles are those recommended and established internationally, namely: the principle of justification of decisions that alters the radiation exposure situation; the principle of optimization of protection and safety; the principle of individual protection for restricting possible inequitable outcomes of optimized safety; and the implicit principle of inter generational prudence for protection future generations and the habitat. The principles are compared vis-a-vis the prevalent ethical doctrines: justification vis-a-vis teleology; optimization vis-a-vis utilitarianism; individual protection vis-a-vis de ontology; and, inter generational prudence vis-a-vis aretaicism (or virtuosity). The application of the principles and their ethics in Argentina is analysed. These principles are applied to All exposure to radiation harm; namely, to exposures to actual doses and to exposures to actual risk and potential doses, including those related to the safety of nuclear installations, and they are harmonized and applied in conjunction. It is concluded that building a bridge among all available ethical doctrines and applying it to radiation safety against actual doses and actual risk and potential doses is at the roots of the successful nuclear regulatory experience in Argentina.

Setting clear expectations for safety basis development

International Nuclear Information System (INIS)

MORENO, M.R.

2003-01-01

DOE-RL has set clear expectations for a cost-effective approach for achieving compliance with the Nuclear Safety Management requirements (10 CFR 830, Nuclear Safety Rule) which will ensure long-term benefit to Hanford. To facilitate implementation of these expectations, tools were developed to streamline and standardize safety analysis and safety document development resulting in a shorter and more predictable DOE approval cycle. A Hanford Safety Analysis and Risk Assessment Handbook (SARAH) was issued to standardized methodologies for development of safety analyses. A Microsoft Excel spreadsheet (RADIDOSE) was issued for the evaluation of radiological consequences for accident scenarios often postulated for Hanford. A standard Site Documented Safety Analysis (DSA) detailing the safety management programs was issued for use as a means of compliance with a majority of 3009 Standard chapters. An in-process review was developed between DOE and the Contractor to facilitate DOE approval and provide early course correction. As a result of setting expectations and providing safety analysis tools, the four Hanford Site waste management nuclear facilities were able to integrate into one Master Waste Management Documented Safety Analysis (WM-DSA)

Configuration control during maintenance of safety related equipment

International Nuclear Information System (INIS)

Irish, C.S.

2001-01-01

Possibly the most important aspect of performing maintenance of safety related equipment is maintaining the component's original design basis. Assuring that the repaired item will perform the same safety function within the original performance and equipment qualification parameters is commonly referred to as configuration control. Maintaining configuration control of a technologically current well documented item is easy. Unfortunately, this does not describe most safety related items requiring maintenance within the global nuclear industry. Items such as motors, transformers, metal clad switchgear (low and medium voltage circuit breakers), refrigeration compressors, and electronic components (i.e. circuit boards, power supplies, regulators, etc.) which routinely require repair have been in service for twenty plus years. As a result, finding replacement parts and or material to repair the items to the original condition is becoming more and more difficult. An added difficulty is the lack of original technical documentation available on the item which is being repaired. The lack of technical documentation makes it difficult to identify replacement material and parts when the original part or material is not available. The lack of documentation also makes it difficult to test the repaired item to make sure that the original configuration has been maintained after the repair. The presentation will discuss the details of repairing various items including motors, metal clad switchgear, refrigeration compressors and power supplies and the controls which are necessary to maintain the configuration of the original item. The discussion will include the Quality Assurance and engineering necessary to identify and evaluate replacement material and parts necessary to perform repairs on safety related equipment when the original material or part is not available. Examples of repairs which required different parts or materials than the original to be used in the repair will be

Configuration control during maintenance of safety related equipment

International Nuclear Information System (INIS)

Irish, C.S.

2003-01-01

Possibly the most important aspect of performing maintenance of safety related equipment is maintaining the component's original design basis. Assuring that the repaired item will perform the same safety function within the original performance and equipment qualification parameters is commonly referred to as configuration control. Maintaining configuration control of a technologically current well documented item is easy. Unfortunately, this does not describe most safety related items requiring maintenance within the global nuclear industry. Items such as motors, transformers, metal clad switchgear (low and medium voltage circuit breakers), refrigeration compressors, and electronic components (i.e. circuit boards, power supplies, regulators, etc.) which routinely require repair have been in service for twenty plus years. As a result, finding replacement parts and or material to repair the items to the original condition is becoming more and more difficult. An added difficulty is the lack of original technical documentation available on the item which is being repaired. The lack of technical documentation makes it difficult to identify replacement material and parts when the original part or material is not available. The lack of documentation also makes it difficult to test the repaired item to make sure that the original configuration has been maintained after the repair. The presentation will discuss the details of repairing various items including motors, metal clad switchgear, refrigeration compressors and power supplies and the controls which are necessary to maintain the configuration of the original item. The discussion will include the Quality Assurance and engineering necessary to identify and evaluate replacement material and parts necessary to perform repairs on safety related equipment when the original material or part is not available. Examples of repairs which required different parts or materials than the original to be used in the repair will be

Seismic methodology in determining basis earthquake for nuclear installation

International Nuclear Information System (INIS)

Ameli Zamani, Sh.

2008-01-01

Design basis earthquake ground motions for nuclear installations should be determined to assure the design purpose of reactor safety: that reactors should be built and operated to pose no undue risk to public health and safety from earthquake and other hazards. Regarding the influence of seismic hazard to a site, large numbers of earthquake ground motions can be predicted considering possible variability among the source, path, and site parameters. However, seismic safety design using all predicted ground motions is practically impossible. In the determination of design basis earthquake ground motions it is therefore important to represent the influences of the large numbers of earthquake ground motions derived from the seismic ground motion prediction methods for the surrounding seismic sources. Viewing the relations between current design basis earthquake ground motion determination and modem earthquake ground motion estimation, a development of risk-informed design basis earthquake ground motion methodology is discussed for insight into the on going modernization of the Examination Guide for Seismic Design on NPP

The cohort of the atomic bomb survivors major basis of radiation safety regulations

CERN Document Server

Rühm, W; Nekolla, E A

2006-01-01

Since 1950 about 87 000 A-bomb survivors from Hiroshima and Nagasaki have been monitored within the framework of the Life Span Study, to quantify radiation-induced late effects. In terms of incidence and mortality, a statistically significant excess was found for leukemia and solid tumors. In another major international effort, neutron and gamma radiation doses were estimated, for those survivors (Dosimetry System DS02). Both studies combined allow the deduction of risk coefficients that serve as a basis for international safety regulations. As an example, current results on all solid tumors combined suggest an excess relative risk of 0.47 per Sievert for an attained age of 70 years, for those who were exposed at an age of 30 years. After exposure to an effective dose of one Sievert the solid tumor mortality would thus be about 50% larger than that expected for a similar cohort not exposed to any ionizing radiation from the bombs.

Assessment of the food safety issues related to genetically modified foods

NARCIS (Netherlands)

Kuiper, H.A.; Kleter, G.A.; Noteborn, H.P.J.M.; Kok, E.J.

2001-01-01

International consensus has been reached on the principles regarding evaluation of the food safety of genetically modified plants. The concept of substantial equivalence has been developed as part of a safety evaluation framework, based on the idea that existing foods can serve as a basis for

Safety design guide for safety related systems for CANDU 9

International Nuclear Information System (INIS)

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

1996-03-01

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

Safety design guide for safety related systems for CANDU 9

Energy Technology Data Exchange (ETDEWEB)

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

1996-03-01

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

Criteria for safety-related operator actions

International Nuclear Information System (INIS)

Gray, L.H.; Haas, P.M.

1983-01-01

The Safety-Related Operator Actions (SROA) Program was designed to provide information and data for use by NRC in assessing the performance of nuclear power plant (NPP) control room operators in responding to abnormal/emergency events. The primary effort involved collection and assessment of data from simulator training exercises and from historical records of abnormal/emergency events that have occurred in operating plants (field data). These data can be used to develop criteria for acceptability of the use of manual operator action for safety-related functions. Development of criteria for safety-related operator actions are considered

Safety assessment as basis for the decision making process

International Nuclear Information System (INIS)

Ilie, P.; Didita, L.; Danchiv, A.

2005-01-01

. The results of sensitivity and uncertainty analyses related to the input parameters will be presented. A practical application of decision making process in context of post-closure safety assessment will be presented, where decision framework means demonstration of compliance with radiological criteria. The analysis is focused on assessment of ground water pathway in the site selection phase of repository development and the ISAM methodology will be used as a decision tool to identify if a candidate site meets safety requirements for construction of disposal facility. If a decision is made that the results of the safety assessment are inadequate the following step is the identification and prioritisation of activities that will make the safety assessment acceptable. Even if the results are considered acceptable, the assessment results will be used to help prioritise the future activities at the site. (authors)

Structural Aging Program approach to providing an improved basis for aging management of safety-related concrete structures

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.; Ellingwood, B.

1993-01-01

The Structural Aging (SAG) Program is being conducted at the Oak Ridge National Laboratory (ORNL) for the United States Nuclear Regulatory Commission (USNRC). The SAG Program is addressing the aging management of safety-related concrete structures in nuclear power plants for the purpose of providing improved technical bases for their continued service. The program is organized into four tasks: Program Management, Materials Property Data Base, Structural Component Assessment/Repair Technologies, and Quantitative Methodology for Continued Service Determinations. Objectives and a summary of recent accomplishments under each of these tasks are presented

Development of safety related technology and infrastructure for safety assessment

International Nuclear Information System (INIS)

Venkat Raj, V.

1997-01-01

Development and optimum utilisation of any technology calls for the building up of the necessary infrastructure and backup facilities. This is particularly true for a developing country like India and more so for an advanced technology like nuclear technology. Right from the inception of its nuclear power programme, the Indian approach has been to develop adequate infrastructure in various areas such as design, construction, manufacture, installation, commissioning and safety assessment of nuclear plants. This paper deals with the development of safety related technology and the relevant infrastructure for safety assessment. A number of computer codes for safety assessment have been developed or adapted in the areas of thermal hydraulics, structural dynamics etc. These codes have undergone extensive validation through data generated in the experimental facilities set up in India as well as participation in international standard problem exercises. Side by side with the development of the tools for safety assessment, the development of safety related technology was also given equal importance. Many of the technologies required for the inspection, ageing assessment and estimation of the residual life of various components and equipment, particularly those having a bearing on safety, were developed. This paper highlights, briefly, the work carried out in some of the areas mentioned above. (author)

Reliability Analysis on NPP's Safety-Related Control Module with Field Data

International Nuclear Information System (INIS)

Lee, Sang Yong; Jung, Jae Hyun; Kim, Seong Hun

2006-01-01

The automatic control systems used in nuclear power plant (NPP) consists of numerous control modules that can be considered to be a network of components various complex ways. The control modules require relatively high reliability than industrial electronic products. Reliability prediction provides the rational basis of system designs and also provides the safety significance of system operations. The aim of this paper is to minimize the deficiencies of the traditional reliability prediction method calculation using the available field return data. This way is possible to do more realistic reliability assessment. SAMCHANG Enterprise Company (SEC) has established database containing high quality data at the module and component level from module maintenance in NPP. On the basis of these, this paper compares results that add failure record (field data) to Telcordia-SR-332 reliability prediction model with MIL-HDBK-217F prediction results

Concept for creating program-technical complex of safety monitoring with system of safety parameters presentation functions on the basis of routine WWER-1000 systems

International Nuclear Information System (INIS)

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

2005-01-01

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

Westinghouse Hanford Company safety analysis reports and technical safety requirements upgrade program

International Nuclear Information System (INIS)

Busche, D.M.

1995-09-01

During Fiscal Year 1992, the US Department of Energy, Richland Operations Office (RL) separately transmitted the following US Department of Energy (DOE) Orders to Westinghouse Hanford Company (WHC) for compliance: DOE 5480.21, ''Unreviewed Safety Questions,'' DOE 5480.22, ''Technical Safety Requirements,'' and DOE 5480.23, ''Nuclear Safety Analysis Reports.'' WHC has proceeded with its impact assessment and implementation process for the Orders. The Orders are closely-related and contain some requirements that are either identical, similar, or logically-related. Consequently, WHC has developed a strategy calling for an integrated implementation of the three Orders. The strategy is comprised of three primary objectives, namely: Obtain DOE approval of a single list of DOE-owned and WHC-managed Nuclear Facilities, Establish and/or upgrade the ''Safety Basis'' for each Nuclear Facility, and Establish a functional Unreviewed Safety Question (USQ) process to govern the management and preservation of the Safety Basis for each Nuclear Facility. WHC has developed policy-revision and facility-specific implementation plans to accomplish near-term tasks associated with the above strategic objectives. This plan, which as originally submitted in August 1993 and approved, provided an interpretation of the new DOE Nuclear Facility definition and an initial list of WHC-managed Nuclear Facilities. For each current existing Nuclear Facility, existing Safety Basis documents are identified and the plan/status is provided for the ISB. Plans for upgrading SARs and developing TSRs will be provided after issuance of the corresponding Rules

Safety Review related to Commercial Grade Digital Equipment in Safety System

International Nuclear Information System (INIS)

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

2013-01-01

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

Optimal replacement policy for safety-related multi-component multi-state systems

International Nuclear Information System (INIS)

Xu Ming; Chen Tao; Yang Xianhui

2012-01-01

This paper investigates replacement scheduling for non-repairable safety-related systems (SRS) with multiple components and states. The aim is to determine the cost-minimizing time for replacing SRS while meeting the required safety. Traditionally, such scheduling decisions are made without considering the interaction between the SRS and the production system under protection, the interaction being essential to formulate the expected cost to be minimized. In this paper, the SRS is represented by a non-homogeneous continuous time Markov model, and its state distribution is evaluated with the aid of the universal generating function. Moreover, a structure function of SRS with recursive property is developed to evaluate the state distribution efficiently. These methods form the basis to derive an explicit expression of the expected system cost per unit time, and to determine the optimal time to replace the SRS. The proposed methodology is demonstrated through an illustrative example.

Qualification of safety-related valve actuators

International Nuclear Information System (INIS)

Anon.

1981-01-01

This Standard describes the qualification of all types of power-driven valve actuators, including damper actuators, for safety-related functions in nuclear power generating stations. It may also be used to separately qualify actuator components. This Standard establishes the minimum requirements for, and guidance regarding, the methods and procedures for qualification of all safety-related functions of power-driven valve actuators

Highway Safety Program Manual: Volume 8: Alcohol in Relation to Highway Safety.

Science.gov (United States)

National Highway Traffic Safety Administration (DOT), Washington, DC.

Volume 8 of the 19-volume Highway Safety Program Manual (which provides guidance to State and local governments on preferred highway safety practices) concentrates on alcohol in relation to highway safety. The purpose and objectives of the alcohol program are outlined. Federal authority in the area of highway safety and general policies regarding…

IAEA activities related to safety indicators, time frames and reference scenarios

International Nuclear Information System (INIS)

Batandjieva, B.; Hioki, K.; Metcalf, P.

2002-01-01

The fundamental principles for the safe management of radioactive waste have been agreed internationally and form the basis for the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management that entered into force in June 2001. Protection of human health and the environment and safety of facilities (including radioactive waste disposal facilities) are widely recognised principles to be followed and demonstrated in post-closure safety assessment of waste repositories. Dose and risk are at present internationally agreed safety criteria, used for judging the acceptability of such facilities. However, there have been a number of activities initiated and co-ordinated by the International Atomic Energy Agency (IAEA) which have provided an international forum for discussion and consensus building on the use safety indicators which are complementary to dose and risk. The Agency has been working on the definition of other safety indicators, such as flux, time, environmental concentration, etc.; the desired characteristics, and use of these indicators in different time frames. The IAEA has focused on safety indicators related to geological disposal, exploring their role in the development of a safety case, evaluating the advantages and disadvantages of using other safety indicators and how they complement the dose and risk indicators. The use of these indicators have been discussed also from regulatory perspective, mainly in terms of achieving reasonable assurance and confidence in safety assessments for waste repositories and decision making in the presence of uncertainty in the context of disposal of long-lived waste. Considerable effort has also been expended by the Agency on the development and application of principles for defining critical groups and biospheres for deep geological repositories. One of the important and successful IAEA programmes in this field is the Biosphere Modelling and Assessment (BIOMASS) project

Health risk from radioactive and chemical environmental contamination: common basis for assessment and safety decision making

International Nuclear Information System (INIS)

Demin, V.

2004-01-01

To meet the growing practical need in risk analysis in Russia health risk assessment tools and regulations have been developed in the frame of few federal research programs. RRC Kurchatov Institute is involved in R and D on risk analysis activity in these programs. One of the objectives of this development is to produce a common, unified basis of health risk analysis for different sources of risk. Current specific and different approaches in risk assessment and establishing safety standards developed for chemicals and ionising radiation are analysed. Some recommendations are given to produce the common approach. A specific risk index R has been proposed for safety decision-making (establishing safety standards and other levels of protective actions, comparison of various sources of risk, etc.). The index R is defined as the partial mathematical expectation of lost years of healthy life (LLE) due to exposure during a year to a risk source considered. The more concrete determinations of this index for different risk sources derived from the common definition of R are given. Generic safety standards (GSS) for the public and occupational workers have been suggested in terms of this index. Secondary specific safety standards have been derived from GSS for ionizing radiation and a number of other risk sources including environmental chemical pollutants. Other general and derived levels for decision-making have also been proposed including the e-minimum level. Their possible dependence on the national or regional health-demographic data is shortly considered. Recommendations are given on methods and criteria for comparison of various sources of risk. Some examples of risk comparison are demonstrated in the frame of different comparison tasks. The paper has been prepared on the basis of the research work supported by International Science and Technology Centre, Moscow (project no. 2558). (author)

FLIGHT SAFETY CONTROL OF THE BASIS OF UNCERTAIN RISK EVALUATION WITH NON-ROUTINE FLIGHT CONDITIONS INVOLVED

Directory of Open Access Journals (Sweden)

2016-01-01

Full Text Available The article deals with methods of forecasting the level of aviation safety operation of aircraft systems on the basis of methods of evaluation the risks of negative situations as a consequence of a functional loss of initial properties of the system with critical violations of standard modes of the aircraft. Mathematical Models of Risks as a Danger Measure of Discrete Random Events in Aviation Systems are presented. Technological Schemes and Structure of Risk Control Proce- dures without the Probability are illustrated as Methods of Risk Management System in Civil Aviation. The assessment of the level of safety and quality and management of aircraft, made not only from the standpoint of reliability (quality and consumer properties, but also from the position of ICAO on the basis of a risk-based approach. According to ICAO, the security assessment is performed by comparing the calculated risk with an acceptable level. The approach justifies the use of qualitative evaluation techniques safety in the forms of proactive forecasted and predictive risk management adverse impacts to aviation operations of various kinds, including the space sector and nuclear energy. However, for the events such as accidents and disasters, accidents with the aircraft, fighters in a training flight, during the preparation of the pilots on the training aircraft, etc. there is no required statistics. Density of probability distribution (p. d. f. of these events are only hypothetical, unknown with "hard tails" that completely eliminates the application of methods of confidence intervals in the traditional approaches to the assessment of safety in the form of the probability analysis.

Safety-related control air systems - approved 1977

International Nuclear Information System (INIS)

Anon.

1978-01-01

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

The Labour Contract as a Basis of Appearance of Labour Relations

OpenAIRE

Вишновецька, Світлана Василівна; Національний авіаційний університет; Такаджі, Людмила Геннадіївна; Національний авіаційний університет

2016-01-01

The article is devoted to research of labour contract as a basis of appearance of labour relations. The problems of conclusion of labour contract, its entry into force, the time of appearance of the labour relationship are very topical. The labour contract is the main form of realization of the right to work, a basis of appearance of labour relations that makes legal base for their functioning. Before December 28th, 2014 the actual admission to work was a basis of appearance of labour relatio...

Radiation safety

International Nuclear Information System (INIS)

1996-04-01

Most of the ionizing radiation that people are exposed to in day-to-day activities comes from natural, rather than manmade, sources. The health effects of radiation - both natural and artificial - are relatively well understood and can be effectively minimized through careful safety measures and practices. The IAEA, together with other international and expert organizations, is helping to promote and institute Basic Safety Standards on an international basis to ensure that radiation sources and radioactive materials are managed for both maximum safety and human benefit

Food Safety and Sanitary Practices of Selected Hotels in Batangas Province, Philippines: Basis of Proposed Enhancement Measures

Directory of Open Access Journals (Sweden)

April M. Perez

2017-02-01

Full Text Available This study assessed the extent of food safety and sanitary practices of selected hotels in Batangas province as basis of proposed enhancement measures. The study utilized descriptive method to describe food safety and sanitary practices of selected hotels in Batangas province with a total of 8 hotels (256 respondents. Purposive sampling was used in the study. The questionnaires were designed using the provision of the Sanitation Code of the Philippines, validated and finalized to come up with legitimate results. The study showed that there were eight (8 hotel respondents classified as two, three, four star with considerable years of experience and adequate number of employees. The hotels demonstrated the food safety and sanitary practices always in the areas of restaurant, bar service, catering and banquet and room service. The significant pair-wise comparison for restaurant, bar service, catering and banquet and room service shows that 2 star hotels greatly differs. The researcher recommends that the management should maintain high standard of food safety and sanitary practices among its staff, upgrade the food safety and sanitary practices for food safety accreditation, continuous training of the hotel managers/employees on food safety and sanitary practices.

Final safety analysis report (FSAR) for waste receiving and processing (WRAP) facility

International Nuclear Information System (INIS)

Weidert, J.R.

1997-01-01

This safety analysis report provides a summary description of the WRAP Facility, focusing on significant safety-related characteristics of the location and facility design. This report demonstrates that adherence to the safety basis wi11 ensure necessary operational safety considerations have been addressed sufficiently and justifies the adequacy of the safety basis in protecting the health and safety of the public, workers, and the environment

Electric and mechanical basic parameters to elaborate a process for a technical verification of safety related design modifications

International Nuclear Information System (INIS)

Lamuno Fernandez, Mercedes; La Roca Mallofre, GISEL; Bano Azcon, Alberto

2010-01-01

This paper presents a systematic process to check a design in order to achieve all the requirements that regulations demand. Nuclear engineers must verify that a design is done according to the safety requirements, and this paper presents how we have elaborated a process to improve the technical project verification. For a faster, better and easier verification process, here we summarize how to select the electric and mechanical basic parameters, which ensure the correct project verification of safety related design modifications. This process considers different aspects, which guarantee that the design preserves the availability, reliability and functional capability of the Structures, Systems and Components needed to operate the Nuclear Power Station with security. Electric and mechanical reference parameters are identified and discussed as well as others related ones, which are critical to safety. The implementation procedure to develop tasks performed in any company that has a quality plan is a requirement. On the engineering business, it is important not to use the personal criteria to do a technical analysis of a project; although, many times it is the checker's criteria and knowledge responsibility to ensure the correct development of a design modification. Then, the checker capabilities are the basis of the modification verification. This kind of procedure's development is not easy, because in an engineering project with important technical contents, there are multiple scenarios, but lots of them have a common basis. If we can identify the technical common basis of these projects, we will make good project verification but there are many difficulties we can encounter along this process. (authors)

EPRI program in water reactor safety

International Nuclear Information System (INIS)

Loewenstein, W.B.; Gelhaus, F.; Gopalakrishnan, A.

1975-01-01

The basis for EPRI's water reactor safety program is twofold. First is compilation and development of fundamental background data necessary for quantified light-water reactor (LWR) safety assurance appraisals. Second is development of realistic and experimentally bench-marked analytical procedures. The results are expected either to confirm the safety margins in current operating parameters, and to identify overly conservative controls, or, in some cases, to provide a basis for improvements to further minimize uncertainties in expected performance. Achievement of these objectives requires the synthesis of related current and projected experimental-analytical projects toward establishment of an experimentally-based analysis for the assurance of safety for LWRs

Characterisation of Liquefaction Effects for Beyond-Design Basis Safety Assessment of Nuclear Power Plants

Science.gov (United States)

Bán, Zoltán; Győri, Erzsébet; János Katona, Tamás; Tóth, László

2015-04-01

-tree procedure. Earlier studies have shown that the potentially liquefiable layer at Paks Nuclear Power Plant is situated in relatively large depth. Therefore the applicability and adequacy of the methods at high overburden pressure is important. In case of existing facilities, the geotechnical data gained before construction aren't sufficient for the comprehensive liquefaction analysis. Performance of new geotechnical survey is limited. Consequently, the availability of the data has to be accounted while selection the analysis methods. Considerations have to be made for dealing with aleatory uncertainty related to the knowledge of the soil conditions. It is shown in the paper, a careful comparison and analysis of the results obtained by different methodologies provides the basis of the selection of practicable methods for the safety analysis of nuclear power plant for beyond design basis liquefaction hazard.

Critical review of methodology and application of risk ranking for prioritisation of food and feed related issues, on the basis of the size of anticipated health impact

NARCIS (Netherlands)

Fels-Klerx, van der H.J.; Asselt, van E.D.; Raley, M.; Poulsen, M.; Korsgaard, H.; Bredsdorff, L.; Nauta, M.; Flari, V.; Agostino, D' M.; Coles, D.G.; Frewer, L.J.

2015-01-01

This study aimed to critically review methodologies for ranking of risks related to feed/food safety and nutritional hazards, on the basis of their anticipated human health impact. An extensive systematic literature review was performed to identify and characterize the available methodologies for

Development Trends in Nuclear Technology and Related Safety Aspects

International Nuclear Information System (INIS)

Kuczera, B.; Juhn, P.E.; Fukuda, K.

2002-01-01

The IAEA Safety Standards Series include, in a hierarchical manner, the categories of Safety Fundamentals, Safety Requirements and Safety Guides, which define the elements necessary to ensure the safety of nuclear installations. In the same way as nuclear technology and scientific knowledge advance continuously, also safety requirements may change with these advances. Therefore, in the framework of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) one important aspect among others refers to user requirements on the safety of innovative nuclear installations, which may come into operation within the next fifty years. In this respect, the major objectives of the INPRO sub-task 'User Requirements and Nuclear Energy Development Criteria in the Area of Safety' have been: a. to overview existing national and international requirements in the safety area, b. to define high level user requirements in the area of safety of innovative nuclear technologies, c. to compile and to analyze existing innovative reactor and fuel cycle technology enhancement concepts and approaches intended to achieve a high degree of safety, and d. to identify the general areas of safety R and D needs for the establishment of these technologies. During the discussions it became evident that the application of the defence in depth strategy will continue to be the overriding approach for achieving the general safety objective in nuclear power plants and fuel cycle facilities, where the emphasis will be shifted from mitigation of accident consequences more towards prevention of accidents. In this context, four high level user requirements have been formulated for the safety of innovative nuclear reactors and fuel cycles. On this basis safety strategies for innovative reactor designs are highlighted in each of the five levels of defence in depth and specific requirements are discussed for the individual components of the fuel cycle. (authors)

Qualification of FPGA-Based Safety-Related PRM System

International Nuclear Information System (INIS)

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

2011-01-01

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

Technical basis for environmental qualification of computer-based safety systems in nuclear power plants

International Nuclear Information System (INIS)

Korsah, K.; Wood, R.T.; Tanaka, T.J.; Antonescu, C.E.

1997-01-01

This paper summarizes the results of research sponsored by the US Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. This research was conducted by the Oak Ridge National Laboratory (ORNL) and Sandia National Laboratories (SNL). ORNL investigated potential failure modes and vulnerabilities of microprocessor-based technologies to environmental stressors, including electromagnetic/radio-frequency interference, temperature, humidity, and smoke exposure. An experimental digital safety channel (EDSC) was constructed for the tests. SNL performed smoke exposure tests on digital components and circuit boards to determine failure mechanisms and the effect of different packaging techniques on smoke susceptibility. These studies are expected to provide recommendations for environmental qualification of digital safety systems by addressing the following: (1) adequacy of the present preferred test methods for qualification of digital I and C systems; (2) preferred standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging in qualification testing for equipment that is to be located in mild environments; and (5) determination of an appropriate approach to address smoke in a qualification program

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

Science.gov (United States)

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

2015-01-01

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

Towards a Formal Basis for Modular Safety Cases

Science.gov (United States)

Denney, Ewen; Pai, Ganesh

2015-01-01

Safety assurance using argument-based safety cases is an accepted best-practice in many safety-critical sectors. Goal Structuring Notation (GSN), which is widely used for presenting safety arguments graphically, provides a notion of modular arguments to support the goal of incremental certification. Despite the efforts at standardization, GSN remains an informal notation whereas the GSN standard contains appreciable ambiguity especially concerning modular extensions. This, in turn, presents challenges when developing tools and methods to intelligently manipulate modular GSN arguments. This paper develops the elements of a theory of modular safety cases, leveraging our previous work on formalizing GSN arguments. Using example argument structures we highlight some ambiguities arising through the existing guidance, present the intuition underlying the theory, clarify syntax, and address modular arguments, contracts, well-formedness and well-scopedness of modules. Based on this theory, we have a preliminary implementation of modular arguments in our toolset, AdvoCATE.

Review of safety related control room function research based on experience from nuclear power plants in Finland

International Nuclear Information System (INIS)

Juslin, K.; Wahlstroem, B.; Rinttilae, E.

1985-01-01

A comprehensive human engineering research programme was established in the second half of the 1970's at the Technical Research Centre of Finland (VTT). The research is performed in cooperation with the utility companies Imatran Voima Oy (IVO) and Teollisuuden Voima Oy (TVO) and includes topics such as Handling of alarm information, Disturbance analysis systems, Assessment of control rooms and Validation of safety parameter display systems. Reference is also made to the Finnish contribution to the OECD Halden Reactor Project (Halden) and the Nordic Liaison Committee for Atomic Energy (NKA) research projects. In this paper feasible realization alternatives of safety related control room functions are discussed on the basis of experience from the nuclear power plants in Finland, which at present are equipped with extensive process computer systems. A proposal for future power plant information systems is described. It is intended that this proposal will serve as the basis for future computer systems at nuclear power plants in Finland. (author)

Safety assessment of dangerous goods transport enterprise based on the relative entropy aggregation in group decision making model.

Science.gov (United States)

Wu, Jun; Li, Chengbing; Huo, Yueying

2014-01-01

Safety of dangerous goods transport is directly related to the operation safety of dangerous goods transport enterprise. Aiming at the problem of the high accident rate and large harm in dangerous goods logistics transportation, this paper took the group decision making problem based on integration and coordination thought into a multiagent multiobjective group decision making problem; a secondary decision model was established and applied to the safety assessment of dangerous goods transport enterprise. First of all, we used dynamic multivalue background and entropy theory building the first level multiobjective decision model. Secondly, experts were to empower according to the principle of clustering analysis, and combining with the relative entropy theory to establish a secondary rally optimization model based on relative entropy in group decision making, and discuss the solution of the model. Then, after investigation and analysis, we establish the dangerous goods transport enterprise safety evaluation index system. Finally, case analysis to five dangerous goods transport enterprises in the Inner Mongolia Autonomous Region validates the feasibility and effectiveness of this model for dangerous goods transport enterprise recognition, which provides vital decision making basis for recognizing the dangerous goods transport enterprises.

Preservation of FFTF Data Related to Passive Safety Testing

International Nuclear Information System (INIS)

Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.

2010-01-01

experience to a large-size LMR and obtain data for validating design analysis computer codes, and (3) to develop and test passive safety enhancements that might be used for future LMRs. These tests were designed to provide data sufficient to allow separation of fuel temperature effects from structural temperature effects. The data developed through this testing program were used to verify the predictive capability of passive safety analysis methods as well as provide a data base for calibrating design tools such as the SASSYS/SAS4A codes. These tests were instrumental in improving understanding of reactivity feedback mechanisms in LMRs and demonstrating passive safety margins available in an LMR. Knowledge preservation at the FFTF is focused on the areas of design, construction, startup, and operation of the reactor. This information may be of potential use for international exchanges with other LMR programs around the world. This information provides the basis for creating benchmarks for validating and testing large scale computer programs. All information preserved to date is now being stored and categorized consistent with the IAEA international standardized taxonomy. The test results information exists in several different formats depending upon the final stage of the test evaluation. Over 100 documents relevant to passive safety testing have been identified and are being recovered, scanned, and catalogued. Attempts to recover plant data tapes are also in progress. Documents related to passive safety testing are now being categorized consistent with internationally agreed upon IAEA standards. Documents are being converted to electronic format compatible with a general search engine being developed by INL. The data from the FFTF passive safety tests provides experimental verification of structural reactivity effects that should be very useful to innovative designers seeking to optimize passive safety in the design of new LMRs.

Advances in safety related maintenance

International Nuclear Information System (INIS)

2000-03-01

The maintenance of systems, structures and components in nuclear power plants (NPPs) plays an important role in assuring their safe and reliable operation. Worldwide, NPP maintenance managers are seeking to reduce overall maintenance costs while maintaining or improving the levels of safety and reliability. Thus, the issue of NPP maintenance is one of the most challenging aspects of nuclear power generation. There is a direct relation between safety and maintenance. While maintenance alone (apart from modifications) will not make a plant safer than its original design, deficient maintenance may result in either an increased number of transients and challenges to safety systems or reduced reliability and availability of safety systems. The confidence that NPP structures, systems and components will function as designed is ultimately based on programmes which monitor both their reliability and availability to perform their intended safety function. Because of this, approaches to monitor the effectiveness of maintenance are also necessary. An effective maintenance programme ensures that there is a balance between the improvement in component reliability to be achieved and the loss of component function due to maintenance downtime. This implies that the safety level of an NPP should not be adversely affected by maintenance performed during operation. The nuclear industry widely acknowledges the importance of maintenance in NPP safety and operation and therefore devotes great efforts to develop techniques, methods and tools to aid in maintenance planning, follow-up and optimization, and in assuring the effectiveness of maintenance

Challenges in the management of gas voids in safety related systems

International Nuclear Information System (INIS)

Ezekoye, L.I.; Turkowski, W.M.; Ferraraccio, F.P.; Swartz, M.M.

2009-01-01

of mechanisms, such as maintenance, gas desorption, vortex activities, cavitation, etc. This paper discusses the sources of gas into safety related systems and the challenges associated with management of gas voids in these systems. Various issues such as gas monitoring, gas purging, gas transport characterization in piping, gas ingestion in safety related pumps, water hammer in discharge piping, and gas impact on the core cooling during design basis accidents are discussed. Industry efforts to support the management of gases in these systems are also presented

Challenges in the management of gas voids in safety related systems

Energy Technology Data Exchange (ETDEWEB)

Ezekoye, L.I.; Turkowski, W.M.; Ferraraccio, F.P.; Swartz, M.M. [Westinghouse Electric Company LLC, Pittsburgh (United States)

2009-04-15

a number of mechanisms, such as maintenance, gas desorption, vortex activities, cavitation, etc. This paper discusses the sources of gas into safety related systems and the challenges associated with management of gas voids in these systems. Various issues such as gas monitoring, gas purging, gas transport characterization in piping, gas ingestion in safety related pumps, water hammer in discharge piping, and gas impact on the core cooling during design basis accidents are discussed. Industry efforts to support the management of gases in these systems are also presented.

Current safety issues related to research reactor operation

International Nuclear Information System (INIS)

Alcala-Ruiz, F.

2000-01-01

The Agency has included activities on research reactor safety in its Programme and Budget (P and B) since its inception in 1957. Since then, these activities have traditionally been oriented to fulfil the Agency's functions and obligations. At the end of the decade of the eighties, the Agency's Research Reactor Safety Programme (RRSP) consisted of a limited number of tasks related to the preparation of safety related publications and the conduct of safety missions to research reactor facilities. It was at the beginning of the nineties when the RRSP was upgraded and expanded as a subprogramme of the Agency's P and B. This subprogramme continued including activities related to the above subjects and started addressing an increasing number of issues related to the current situation of research reactors (in operation and shut down) around the world such as reactor ageing, modifications and decommissioning. The present paper discusses some of the above issues as recognised by various external review or advisory groups (e.g., Peer Review Groups under the Agency's Performance Programme Appraisal System (PPAS) or the standing International Nuclear Safety Advisory Group (INSAG)) and the impact of their recommendations on the preparation and implementation of the part of the Agency's P and B relating to the above subject. (author)

Safety-related requirements for photovoltaic modules and arrays

Science.gov (United States)

Levins, A.; Smoot, A.; Wagner, R.

1984-01-01

Safety requirements for photovoltaic module and panel designs and configurations for residential, intermediate, and large scale applications are investigated. Concepts for safety systems, where each system is a collection of subsystems which together address the total anticipated hazard situation, are described. Descriptions of hardware, and system usefulness and viability are included. A comparison of these systems, as against the provisions of the 1984 National Electrical Code covering photovoltaic systems is made. A discussion of the Underwriters Laboratory UL investigation of the photovoltaic module evaluated to the provisions of the proposed UL standard for plat plate photovoltaic modules and panels is included. Grounding systems, their basis and nature, and the advantages and disadvantages of each are described. The meaning of frame grounding, circuit groundings, and the type of circuit ground are covered.

Critical review of methodology and application of risk ranking for prioritisation of food and feed related issues, on the basis of the size of anticipated health impact

DEFF Research Database (Denmark)

van der Fels-Klerx, H. J.; van Asselt, E. D.; Raley, M.

, an overarching framework was developed for selection of the appropriate method(s) that could be used for risk ranking of feed and food related hazards, on the basis of human health impact. This framework has the format of a decision tool, with which – given the characteristics of the risk ranking question...... at hand - the most appropriate method(s) can be selected. Application of this overall framework to several case studies showed it can be a useful tool for risk managers/assessors to select the most suitable method for risk ranking of feed/food and diet related hazards, on the basis of expected human......This study aimed to critically review methodologies for ranking of risks related to feed/food safety and nutritional hazards, on the basis of their anticipated human health impact. An extensive systematic literature review was performed to identify and characterize the available methodologies...

Supporting Fernald Site Closure with Integrated Health and Safety Plans as Documented Safety Analyses

International Nuclear Information System (INIS)

Kohler, S.; Brown, T.; Fisk, P.; Krach, F.; Klein, B.

2004-01-01

At the Fernald Closure Project (FCP) near Cincinnati, Ohio, environmental restoration activities are supported by Documented Safety Analyses (DSAs) that combine the required project-specific Health and Safety Plans, Safety Basis Requirements (SBRs), and Process Requirements (PRs) into single Integrated Health and Safety Plans (I-HASPs). These integrated DSAs employ Integrated Safety Management methodology in support of simplified restoration and remediation activities that, so far, have resulted in the decontamination and demolition (D and D) of over 200 structures, including eight major nuclear production plants. There is one of twelve nuclear facilities still remaining (Silos containing uranium ore residues) with its own safety basis documentation. This paper presents the status of the FCP's safety basis documentation program, illustrating that all of the former nuclear facilities and activities have now replaced. Basis of Interim Operations (BIOs) with I-HASPs as their safety basis during the closure process

Safety of Nuclear Power Plants: Design. Specific Safety Requirements

International Nuclear Information System (INIS)

2012-01-01

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

Preliminary Evaluation Methodology of ECCS Performance for Design Basis LOCA Redefinition

International Nuclear Information System (INIS)

Kang, Dong Gu; Ahn, Seung Hoon; Seul, Kwang Won

2010-01-01

To improve their existing regulations, the USNRC has made efforts to develop the risk-informed and performance-based regulation (RIPBR) approaches. As a part of these efforts, the rule revision of 10CFR50.46 (ECCS Acceptance Criteria) is underway, considering some options for 4 categories of spectrum of break sizes, ECCS functional reliability, ECCS evaluation model, and ECCS acceptance criteria. Since the potential for safety benefits and unnecessary burden reduction from design basis LOCA redefinition is high relative to other options, the USNRC is proceeding with the rulemaking for design basis LOCA redefinition. An instantaneous break with a flow rate equivalent to a double ended guillotine break (DEGB) of the largest primary piping system in the plant is widely recognized as an extremely unlikely event, while redefinition of design basis LOCA can affect the existing regulatory practices and approaches. In this study, the status of the design basis LOCA redefinition and OECD/NEA SMAP (Safety Margin Action Plan) methodology are introduced. Preliminary evaluation methodology of ECCS performance for LOCA is developed and discussed for design basis LOCA redefinition

Seismic qualification of multiple interconnected safety-related cabinets in a high seismic zone

International Nuclear Information System (INIS)

Khan, M.R.; Chen, W.H.W.; Wang, T.Y.

1993-01-01

Certain safety-related multiple, interconnected electrical cabinets and the devices contained therein are required to perform their intended safety functions during and after a design basis seismic event. In general, seismic testing is performed to ensure the structural integrity of the cabinets and the functionality of their associated devices. Constrained by the shake table capacity, seismic testing is usually performed only for a limited number of interconnected cabinets. Also, original shake table tests performed usually did not provide detailed response information at various locations inside the cabinets. For operational and maintenance purposes, doors and panels of some cabinets may need to be opened while the adjacent cabinets are required to remain functional. In addition, in-cabinet response spectra need to be generated for the seismic qualification of new devices and the replacement parts. Consequently, seismic analysis of safety-related multiple, interconnected cabinets is frequently required for configurations which are different from the original tested conditions. This paper presents results of seismic tests of three interconnected safety-related cabinets and finite element analyses performed to compare the analytical results with those obtained from the cabinet seismic tests. Parametric analyses are performed to determine how many panels and doors can be opened while the adjacent cabinets still remain functional. The study indicates that for cabinets located in a high seismic zone, the critical damping of the cabinet is significantly higher than 5% to 7% typically used in qualifying electrical equipment. For devices mounted on the cabinet doors to performed their intended safety function, it requires stiffening of doors and that these doors be properly bolted to the cabinet frame. It also shows that even though doors and panels bolted to the cabinet frame are the primary seismic resistant element of the cabinet, opening of a limited number of them

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

Energy Technology Data Exchange (ETDEWEB)

T. A. Tomberlin; S. B. Grover

2004-11-01

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

International Nuclear Information System (INIS)

T. A. Tomberlin; S. B. Grover

2004-01-01

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment

Selection of design basis event for modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi

2016-06-01

Japan Atomic Energy Agency (JAEA) has been investigating safety requirements and basic approach of safety guidelines for modular High Temperature Gas-cooled Reactor (HTGR) aiming to increase internarial contribution for nuclear safety by developing an international HTGR safety standard under International Atomic Energy Agency. In this study, we investigate a deterministic approach to select design basis events utilizing information obtained from probabilistic approach. In addition, selections of design basis events are conducted for commercial HTGR designed by JAEA. As a result, an approach for selecting design basis event considering multiple failures of safety systems is established which has not been considered as design basis in the safety guideline for existing nuclear facility. Furthermore, selection of design basis events for commercial HTGR has completed. This report provides an approach and procedure for selecting design basis events of modular HTGR as well as selected events for the commercial HTGR, GTHTR300. (author)

Application of life-cycle information for advancement in safety of nuclear fuel cycle facilities. Application of safety information to advanced safety management support system

International Nuclear Information System (INIS)

Suzuki, Kazuhiko; Ishida, Michihiko

2005-08-01

Risk management is major concern to nuclear energy reprocessing plants to improve plant and process reliability and ensure their safety. This is because we are required to predict potential risks before any accident or disaster occurs. The advancement of safety design and safety systems technologies showed large amount of useful safety-related knowledge that can be of great importance to plant operation to reduce operation risks and ensure safety. This research proposes safety knowledge modeling framework on the basis of ontology technologies to systematically construct plant knowledge model, which includes plant structure, operation, and the associated behaviors. In such plant knowledge model safety related information is defined and linked to the different elements of plant knowledge model. Ontology editor is employed to define the basic concepts and their inter-relations, which are used to capture and construct plant safety knowledge. In order to provide detailed safety knowledgebase, HAZOP results are analyzed and structured so that safety-related knowledge are identified and structured within the plant knowledgebase. The target safety knowledgebase includes: failures, deviations, causes, consequences, and fault propagation as mapped to plant knowledge. The proposed ontology-based safety framework is applied on case study nuclear plant to structure failures, causes, consequences, and fault propagation, which are used to support plant operation. (author)

Optimum Safety Levels for Breakwaters

DEFF Research Database (Denmark)

Burcharth, H. F.; Sørensen, John Dalsgaard

2005-01-01

Optimum design safety levels for rock and cube armoured rubble mound breakwaters without superstructure are investigated by numerical simulations on the basis of minimization of the total costs over the service life of the structure, taking into account typical uncertainties related to wave...... statistics and structure response. The study comprises the influence of interest rate, service lifetime, downtime costs and damage accumulation. Design limit states and safety classes for breakwaters are discussed. The results indicate that optimum safety levels are somewhat higher than the safety levels...

Safety related terms for advanced nuclear plants

International Nuclear Information System (INIS)

1995-12-01

The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

Safety related terms for advanced nuclear plants

International Nuclear Information System (INIS)

1991-09-01

The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

Safety management of software-based equipment

CERN Document Server

Boulanger, Jean-Louis

2013-01-01

A review of the principles of the safety of software-based equipment, this book begins by presenting the definition principles of safety objectives. It then moves on to show how it is possible to define a safety architecture (including redundancy, diversification, error-detection techniques) on the basis of safety objectives and how to identify objectives related to software programs. From software objectives, the authors present the different safety techniques (fault detection, redundancy and quality control). "Certifiable system" aspects are taken into account throughout the book. C

TECHNICAL BASIS DOCUMENT FOR NATURAL EVENT HAZARDS

International Nuclear Information System (INIS)

KRIPPS, L.J.

2006-01-01

This technical basis document was developed to support the documented safety analysis (DSA) and describes the risk binning process and the technical basis for assigning risk bins for natural event hazard (NEH)-initiated accidents. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls

The Safety Case and Safety Assessment for the Disposal of Radioactive Waste

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-09-15

This Safety Guide provides guidance and recommendations on meeting the safety requirements in respect of the safety case and supporting safety assessment for the disposal of radioactive waste. The safety case and supporting safety assessment provide the basis for demonstration of safety and for licensing of radioactive waste disposal facilities and assist and guide decisions on siting, design and operations. The safety case is also the main basis on which dialogue with interested parties is conducted and on which confidence in the safety of the disposal facility is developed. This Safety Guide is relevant for operating organizations preparing the safety case as well as for the regulatory body responsible for developing the regulations and regulatory guidance that determine the basis and scope of the safety case. Contents: 1. Introduction; 2. Demonstrating the safety of radioactive waste disposal; 3. Safety principles and safety requirements; 4. The safety case for disposal of radioactive waste; 5. Radiological impact assessment for the period after closure; 6. Specific issues; 7. Documentation and use of the safety case; 8. Regulatory review process.

Safety campaigns. TIS Launches New Safety Information Campaign

CERN Multimedia

2001-01-01

Need to start a new installation and worried about safety aspects? Or are you newly responsible for safety matters in a CERN building? Perhaps you're simply interested in how to make the working environment safer for yourself and your colleagues. Whatever the case, a new information campaign launched by TIS this week can help. The most visible aspects of the new campaign will be posters distributed around the Laboratory treating a different subject each month. The Web site - http://safety.cern.ch/ - which provides all safety related information. But these are not the only aspects of the new campaign. Members of the TIS/GS group, whose contact details can be found on the safety web site, are available to give information and advice on a one-to-one basis at any time. The campaign's launch has been timed to coincide with European Safety Week, organized by the European Agency for Safety and Health at Work and the subject treated in the first posters is safety inspection. This particular topic only concerns thos...

An analysis of electronic health record-related patient safety concerns

Science.gov (United States)

Meeks, Derek W; Smith, Michael W; Taylor, Lesley; Sittig, Dean F; Scott, Jean M; Singh, Hardeep

2014-01-01

Objective A recent Institute of Medicine report called for attention to safety issues related to electronic health records (EHRs). We analyzed EHR-related safety concerns reported within a large, integrated healthcare system. Methods The Informatics Patient Safety Office of the Veterans Health Administration (VA) maintains a non-punitive, voluntary reporting system to collect and investigate EHR-related safety concerns (ie, adverse events, potential events, and near misses). We analyzed completed investigations using an eight-dimension sociotechnical conceptual model that accounted for both technical and non-technical dimensions of safety. Using the framework analysis approach to qualitative data, we identified emergent and recurring safety concerns common to multiple reports. Results We extracted 100 consecutive, unique, closed investigations between August 2009 and May 2013 from 344 reported incidents. Seventy-four involved unsafe technology and 25 involved unsafe use of technology. A majority (70%) involved two or more model dimensions. Most often, non-technical dimensions such as workflow, policies, and personnel interacted in a complex fashion with technical dimensions such as software/hardware, content, and user interface to produce safety concerns. Most (94%) safety concerns related to either unmet data-display needs in the EHR (ie, displayed information available to the end user failed to reduce uncertainty or led to increased potential for patient harm), software upgrades or modifications, data transmission between components of the EHR, or ‘hidden dependencies’ within the EHR. Discussion EHR-related safety concerns involving both unsafe technology and unsafe use of technology persist long after ‘go-live’ and despite the sophisticated EHR infrastructure represented in our data source. Currently, few healthcare institutions have reporting and analysis capabilities similar to the VA. Conclusions Because EHR-related safety concerns have complex

Environmental tests of a digital safety channel: An investigation of stress-related vulnerabilities of computer-based safety system

International Nuclear Information System (INIS)

Korsah, K.; Wilson, T.L.; Wood, R.; Tanaka, T.

1997-01-01

This article presents the results of environmental stress tests performed on an experimental digital safety channel (EDSC) assembled at the Oak Ridge National Laboratory as part of the Qualification of Advanced Instrumentation and Controls Systems Research program, which was sponsored by the US Nuclear Regulatory Commission. The program is expected to provide recommendations for environmental qualification of digital safety systems. The purpose of the study was to investigate potential vulnerabilities of distributed computer systems used in safety applications when subjected to environmental stressors. The EDSC assembled for the tests employs technologies and digital subsystems representative of those proposed for use in advanced light-water reactors or as retrofits in existing plants. Subsystems include computers, electrical and optical serial communication links, fiber-optic network links, analog-to-digital and digital-to-analog converters, and multiplexers. The EDSC was subjected to selected stressors that are a potential risk to digital equipment in a mild environment. The selected stressors were electromagnetic and radiofrequency interferences (EMI-RFI), temperature, humidity, and smoke exposure. The stressors were applied at levels of intensity considerably higher than the safety channel is likely to experience in a normal nuclear power plant environment. Ranges of stress were selected at a sufficiently high level to induce errors so that failure modes that are characteristic of the technologies employed could be identified. On the basis of the incidence of functional errors observed during testing, EMI-RFI, smoke exposure, and high temperature coupled with high relative humidity, in that order, were found to have the greatest impact of the stressors investigated. The most prevalent stressor-induced upsets, as well as the most severe, were found to occur during the EMI-RFI tests

Light Water Reactor Sustainability Program Technical Basis Guide Describing How to Perform Safety Margin Configuration Risk Management

Energy Technology Data Exchange (ETDEWEB)

Curtis Smith; James Knudsen; Bentley Harwood

2013-08-01

The INL has carried out a demonstration of the RISMC approach for the purpose of configuration risk management. We have shown how improved accuracy and realism can be achieved by simulating changes in risk – as a function of different configurations – in order to determine safety margins as the plant is modified. We described the various technical issues that play a role in these configuration-based calculations with the intent that future applications can take advantage of the analysis benefits while avoiding some of the technical pitfalls that are found for these types of calculations. Specific recommendations have been provided on a variety of topics aimed at improving the safety margin analysis and strengthening the technical basis behind the analysis process.

Quality Control Activities Related to Mechanical Maintenance of Safety Related Components at Krsko NPP

International Nuclear Information System (INIS)

Djakovic, D.

2016-01-01

For successful, safe and reliable operation of nuclear power plant, maintenance processes have to be systematically controlled and procedures for quality control of maintenance activities shall be established. This is requested by the quality assurance program, which shall provide control over activities affecting the quality of structures, systems, and components, considering their importance to safety. As a part of Quality and Nuclear Oversight Division (QNOD; SKV), the Quality Control Department (QC) provides quality control activities, which are deeply involved in maintenance processes at Krsko NPP, both on safety related and non-safety related (non-nuclear safety) components. QC activities on safety related components have to fulfil all requirements, which will enable the components to perform their intended safety functions. This paper describes quality control activities related to mechanical maintenance of safety related components at Krsko NPP and significant role of the Krsko plant QC Department in three particular maintenance cases connected with safety related components. In these three specific cases, the QC has confirmed its importance in compliance with quality assurance program and presented its significant added value in providing safe and reliable operation of the plant. The first maintenance activity was installation of nozzle check valves in the scope of a modification for improving regulation of spent fuel pit pumps. The QC Department performed receipt inspection of the valves. Using non-destructive examination methods and X-ray spectrometry, it was found out that the valve diffuser was made of improper material, which could cause progressive corrosion of the valve diffuser in borated water and consequently a loss of safety function of the valves followed by long-term consequences. The second one was the receipt inspection of containment ventilation fan coolers. The coolers were claimed and sent back to the supplier because the QC Department

Categorization of safety related motor operated valve safety significance for Ulchin Unit 3

International Nuclear Information System (INIS)

Kang, D. I.; Kim, K. Y.

2002-03-01

We performed a categorization of safety related Motor Operated Valve (MOV) safety significance for Ulchin Unit 3. The safety evaluation of MOV of domestic nuclear power plants affects the generic data used for the quantification of MOV common cause failure ( CCF) events in Ulchin Units 3 PSA. Therefore, in this study, we re-estimated the MGL(Multiple Greek Letter) parameter used for the evaluation of MOV CCF probabilities in Ulchin Units 3 Probabilistic Safety Assessment (PSA) and performed a classification of the MOV safety significance. The re-estimation results of the MGL parameter show that its value is decreased by 30% compared with the current value in Ulchin Unit 3 PSA. The categorization results of MOV safety significance using the changed value of MGL parameter shows that the number of HSSCs(High Safety Significant Components) is decreased by 54.5% compared with those using the current value of it in Ulchin Units 3 PSA

Aviation and healthcare: a comparative review with implications for patient safety.

Science.gov (United States)

Kapur, Narinder; Parand, Anam; Soukup, Tayana; Reader, Tom; Sevdalis, Nick

2016-01-01

Safety in aviation has often been compared with safety in healthcare. Following a recent article in this journal, the UK government set up an Independent Patient Safety Investigation Service, to emulate a similar well-established body in aviation. On the basis of a detailed review of relevant publications that examine patient safety in the context of aviation practice, we have drawn up a table of comparative features and a conceptual framework for patient safety. Convergence and divergence of safety-related behaviours across aviation and healthcare were derived and documented. Key safety-related domains that emerged included Checklists, Training, Crew Resource Management, Sterile Cockpit, Investigation and Reporting of Incidents and Organisational Culture. We conclude that whilst healthcare has much to learn from aviation in certain key domains, the transfer of lessons from aviation to healthcare needs to be nuanced, with the specific characteristics and needs of healthcare borne in mind. On the basis of this review, it is recommended that healthcare should emulate aviation in its resourcing of staff who specialise in human factors and related psychological aspects of patient safety and staff wellbeing. Professional and post-qualification staff training could specifically include Cognitive Bias Avoidance Training, as this appears to play a key part in many errors relating to patient safety and staff wellbeing.

Technical basis document for natural event hazards

International Nuclear Information System (INIS)

CARSON, D.M.

2003-01-01

This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process and the technical basis for assigning risk bins for natural event hazards (NEH)-initiated representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report

Safety design requirements for safety systems and components of JSFR

International Nuclear Information System (INIS)

Kubo, Shigenobu; Shimakawa, Yoshio; Yamano, Hidemasa; Kotake, Shoji

2011-01-01

Safety design requirements for JSFR were summarized taking the development targets of the FaCT project and design feature of JSFR into account. The related safety principle and requirements for Monju, CRBRP, PRISM, SPX, LWRs, IAEA standards, goals of GIF, basic principle of INPRO etc. were also taken into account so that the safety design requirements can be a next-generation global standard. The development targets for safety and reliability are set based on those of FaCT, namely, ensuring safety and reliability equal to future LWR and related fuel cycle facilities. In order to achieve these targets, the defence-in-depth concept is used as the basic safety design principle. General features of the safety design requirements are 1) Achievement of higher reliability, 2) Achievement of higher inspectability and maintainability, 3) Introduction of passive safety features, 4) Reduction of operator action needs, 5) Design consideration against Beyond Design Basis Events, 6) In-Vessel Retention of degraded core materials, 7) Prevention and mitigation against sodium chemical reactions, and 8) Design against external events. The current specific requirements for each system and component are summarized taking the basic design concept of JSFR into account, which is an advanced loop-type large-output power plant with a mixed-oxide-fuelled core. (author)

Assessment of the long-term safety of repositories. Scientific basis

International Nuclear Information System (INIS)

Noseck, Ulrich; Becker, Dirk; Fahrenholz, Christine

2008-12-01

The project contributed to increase the scientific knowledge on the long-term safety assessment and the safety cases of a radioactive waste repository. International guidelines and more recent safety cases from other countries were evaluated. The feasibility study of the three safety indicators ''individual dose rate'', ''radiotoxicity concentration in the biosphere water'' and ''radiotoxicity flux from the geosphere'' showed that due to the independently derived corresponding reference values these indicators describe three different safety statements. The combination of the three values can give a stronger argument for the safety of the repository system. Another important methodological aspect of the safety cases is the definition and selection of scenarios, one of these the human intrusion scenario. Various human intrusion scenarios are considered in the different nations, which differ significantly with respect to type and time scale, the exposition type and exposition pathway. Further progress has been achieved in how to treat human intrusion scenarios in a German post-closure safety case. Another port of the project dealt with the impact of specific geochemical processes on the long-term safety of the repository. The impact of climate changes on the long-term safety of a radioactive waste repository in rock salt was investigated with respect to processes in the overburden and the biosphere where highest impact is expected. Sofa simplified models and only discrete climate estates have been considered

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

International Nuclear Information System (INIS)

2012-01-01

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

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

International Nuclear Information System (INIS)

2012-01-01

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

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

International Nuclear Information System (INIS)

2012-01-01

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

Radiobiological basis for setting neutron radiation safety standards

International Nuclear Information System (INIS)

Straume, T.

1985-01-01

Present neutron standards, adopted more than 20 yr ago from a weak radiobiological data base, have been in doubt for a number of years and are currently under challenge. Moreover, recent dosimetric re-evaluations indicate that Hiroshima neutron doses may have been much lower than previously thought, suggesting that direct data for neutron-induced cancer in humans may in fact not be available. These recent developments make it urgent to determine the extent to which neutron cancer risk in man can be estimated from data that are available. Two approaches are proposed here that are anchored in particularly robust epidemiological and experimental data and appear most likely to provide reliable estimates of neutron cancer risk in man. The first approach uses gamma-ray dose-response relationships for human carcinogenesis, available from Nagasaki (Hiroshima data are also considered), together with highly characterized neutron and gamma-ray data for human cytogenetics. When tested against relevant experimental data, this approach either adequately predicts or somewhat overestimates neutron tumorigenesis (and mutagenesis) in animals. The second approach also uses the Nagasaki gamma-ray cancer data, but together with neutron RBEs from animal tumorigenesis studies. Both approaches give similar results and provide a basis for setting neutron radiation safety standards. They appear to be an improvement over previous approaches, including those that rely on highly uncertain maximum neutron RBEs and unnecessary extrapolations of gamma-ray data to very low doses. Results suggest that, at the presently accepted neutron dose limit of 0.5 rad/yr, the cancer mortality risk to radiation workers is not very different from accidental mortality risks to workers in various nonradiation occupations

Study of system safety evaluation on LTO of national project. NISA safety research project on system safety of nuclear power plants

International Nuclear Information System (INIS)

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

2012-01-01

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

Analysis of Electrical Safety Conditions Taking into Account Soil Conductivity Determined on the Basis of Fuzzy Logic

OpenAIRE

Manusov, V.Z.; Zaytseva, N.M.

2017-01-01

The goal of this work is to prove a possibility of determining soil parameters that influence its conductivity being the basis of grounding, step voltage and touch voltage calculation. This in its turn increases the safety level of electric equipment operation. The article is devoted to development of new, no conventional models of soil conductivity using the theory of fuzzy sets and fuzzy logic. The description of the solution includes the following sections: fuzzy models of specific electri...

Assessing propensity to learn from safety-related events

NARCIS (Netherlands)

Drupsteen, L.; Wybo, J.L.

2015-01-01

Most organisations aim to use experience from the past to improve safety, for instance through learning from safety-related incidents and accidents. Whether an organisation is able to learn successfully can however only be determined afterwards. So far, there are no proactive measures to assess

OSR encapsulation basis -- 100-KW

International Nuclear Information System (INIS)

Meichle, R.H.

1995-01-01

The purpose of this report is to provide the basis for a change in the Operations Safety Requirement (OSR) encapsulated fuel storage requirements in the 105 KW fuel storage basin which will permit the handling and storing of encapsulated fuel in canisters which no longer have a water-free space in the top of the canister. The scope of this report is limited to providing the change from the perspective of the safety envelope (bases) of the Safety Analysis Report (SAR) and Operations Safety Requirements (OSR). It does not change the encapsulation process itself

Approach to developing a ground-motion design basis for facilities important to safety at Yucca Mountain

International Nuclear Information System (INIS)

King, J.L.

1990-01-01

This paper discusses a methodology for developing a ground-motion design basis for prospective facilities at Yucca Mountain that are important to safety. The methodology utilizes a guasi-deterministic construct called the 10,000-year cumulative-slip earthquake that is designed to provide a conservative, robust, and reproducible estimate of ground motion that has a one-in-ten chance of occurring during the preclosure period. This estimate is intended to define a ground-motion level for which the seismic design would ensure minimal disruption to operations engineering analyses to ensure safe performance are included

Occupational safety in the petroleum industry

Energy Technology Data Exchange (ETDEWEB)

Elsner, W

1987-03-01

The original technique-oriented accident prevention today has grown to a comprehensive occupational workers protection system. Modern occupational safety requires latest strategies. Side by side with technical and organizational measures we see duties for all superiors directed to plant related occupational safety. These new principles of leadership on the basis of occupational safety policies from top management require equivalent tactics to cause change in behaviour of the employees. Such a not only formulated but also accepted safety strategy is extremely clear by its positive results in the petroleum industry.

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.

A risk-informed framework for establishing a beyond design basis safety basis for external hazards

Energy Technology Data Exchange (ETDEWEB)

Amico, P. [Hughes Associates, Inc, Baltimore, MD (United States); Anoba, R. [Hughes Associates, Inc, Raleigh, NC (United States); Najafi, B. [Hughes Associates, Inc., Los Gatos, CA (United States)

2014-07-01

The events at Fukushima Daiichi taught us that meeting a deterministic design basis requirement for external hazards does not assure that the risk is low. As observed at the plant, the two primary reasons for this are failure cliffs above the design basis event and that combined hazard effects are not considered in design. Because the possible combinations of design basis exceedences and external hazard combinations are very large and complex, an approach focusing only on the most important ones is needed. For this reason, a risk informed approach is the most effective approach, which is discussed in this paper. (author)

radiation safety culture for developing country: Basis for s minimum operational radiation protection programme

International Nuclear Information System (INIS)

Rozental, J. J.

1997-01-01

The purpose of this document is to present a methodology for an integrated strategy aiming at establishing an adequate radiation Safety infrastructure for developing countries, non major power reactor programme. Its implementation will allow these countries, about 50% of the IAEA's Member States, to improve marginal radiation safety, specially to those recipients of technical assistance and do not meet the Minimum radiation Safety Requirements of the IAEA's Basic Safety Standards for radiation protection Progress in the implementation of safety regulations depends on the priority of the government and its understanding and conviction about the basic requirements for protection against the risks associated with exposure to ionizing radiation. There is no doubt to conclude that the reasons for the deficiency of sources control and dose limitation are related to the lack of an appropriate legal and regulatory framework, specially considering the establishment of an adequate legislation; A minimum legal infrastructure; A minimum operational radiation safety programme; Alternatives for a Point of Optimum Contact, to avoid overlap and conflict, that is: A 'Memorandum of Understanding' among Regulatory Authorities in the Country, dealing with similar type of licensing and inspection

77 FR 6411 - Training, Qualification, and Oversight for Safety-Related Railroad Employees

Science.gov (United States)

2012-02-07

... Oversight for Safety-Related Railroad Employees AGENCY: Federal Railroad Administration (FRA), Department of... establishing minimum training standards for each category and subcategory of safety-related railroad employee... or contractor that employs one or more safety-related railroad employee to develop and submit a...

Safety-related operator actions: methodology for developing criteria

International Nuclear Information System (INIS)

Kozinsky, E.J.; Gray, L.H.; Beare, A.N.; Barks, D.B.; Gomer, F.E.

1984-03-01

This report presents a methodology for developing criteria for design evaluation of safety-related actions by nuclear power plant reactor operators, and identifies a supporting data base. It is the eleventh and final NUREG/CR Report on the Safety-Related Operator Actions Program, conducted by Oak Ridge National Laboratory for the US Nuclear Regulatory Commission. The operator performance data were developed from training simulator experiments involving operator responses to simulated scenarios of plant disturbances; from field data on events with similar scenarios; and from task analytic data. A conceptual model to integrate the data was developed and a computer simulation of the model was run, using the SAINT modeling language. Proposed is a quantitative predictive model of operator performance, the Operator Personnel Performance Simulation (OPPS) Model, driven by task requirements, information presentation, and system dynamics. The model output, a probability distribution of predicted time to correctly complete safety-related operator actions, provides data for objective evaluation of quantitative design criteria

Operating experience and systems analysis at Trillo NPP: A program intended for systematic review of plant safety systems to assess design basis requirements compliance

International Nuclear Information System (INIS)

Vega, R. de la

1996-01-01

The program was defined to apply to all plant safety systems and/or systems included in plant Technical Specifications. The goal of the program was to ensure, by systematic design, construction, and commissioning review, the adequacy of safety systems, structures and components to fulfill their safety functions. Also, as a result of the program, it was established that a complete, unambiguous, systematic, design basis definition shall take place. And finally, a complete documental review of the plant design shall result from the program execution

Safety and Security Interface Technology Initiative

International Nuclear Information System (INIS)

Dr. Michael A. Lehto; Kevin J. Carroll; Dr. Robert Lowrie

2007-01-01

Earlier this year, the Energy Facility Contractors Group (EFCOG) was asked to assist in developing options related to acceleration deployment of new security-related technologies to assist meeting design base threat (DBT) needs while also addressing the requirements of 10 CFR 830. NNSA NA-70, one of the working group participants, designated this effort the Safety and Security Interface Technology Initiative (SSIT). Relationship to Workshop Theme. ''Supporting Excellence in Operations Through Safety Analysis'', (workshop theme) includes security and safety personnel working together to ensure effective and efficient operations. One of the specific workshop elements listed in the call for papers is ''Safeguards/Security Integration with Safety''. This paper speaks directly to this theme. Description of Work. The EFCOG Safety Analysis Working Group (SAWG) and the EFCOG Security Working Group formed a core team to develop an integrated process involving both safety basis and security needs allowing achievement of the DBT objectives while ensuring safety is appropriately considered. This effort garnered significant interest, starting with a two day breakout session of 30 experts at the 2006 Safety Basis Workshop. A core team was formed, and a series of meetings were held to develop that process, including safety and security professionals, both contractor and federal personnel. A pilot exercise held at Idaho National Laboratory (INL) in mid-July 2006 was conducted as a feasibility of concept review. Work Results. The SSIT efforts resulted in a topical report transmitted from EFCOG to DOE/NNSA in August 2006. Elements of the report included: Drivers and Endstate, Control Selections Alternative Analysis Process, Terminology Crosswalk, Safety Basis/Security Documentation Integration, Configuration Control, and development of a shared ''tool box'' of information/successes. Specific Benefits. The expectation or end state resulting from the topical report and associated

Status and trends in IAEA safety standards

International Nuclear Information System (INIS)

Lipar, M.

2004-01-01

While safety is a national responsibility, international standards and approaches to safety promote consistency and facilitate international technical co-operation and trade, and help to provide assurance that nuclear and radiation related technologies are used safely. The standards also provide support for States in meeting their international obligations. One general international obligation is that a State must not pursue activities that cause damage in another State. More specific obligations on Contracting States are set out in international safety related conventions. The internationally agreed IAEA safety standards provide the basis for States to demonstrate that they are meeting these obligations. These standards are founded in the IAEA's Statute, which authorizes the Agency to establish standards of safety for nuclear and radiation related facilities and activities and to provide for their application. The safety standards reflect an international consensus on what constitutes a high level of safety for protecting people and the environment. (orig.) [de

Safety related events at nuclear installations in 1995

DEFF Research Database (Denmark)

Korsbech, Uffe C C

1996-01-01

Nuclear safety related events of significance at least corresponding to level 2 of the International Nuclear Event Scale are described. In 1995 only two events occured at nuclear power plants, and four events occured at plants using ionizing radiation for processing or research.......Nuclear safety related events of significance at least corresponding to level 2 of the International Nuclear Event Scale are described. In 1995 only two events occured at nuclear power plants, and four events occured at plants using ionizing radiation for processing or research....

The power of simplification: Operator interface with the AP1000R during design-basis and beyond design-basis events

International Nuclear Information System (INIS)

Williams, M. G.; Mouser, M. R.; Simon, J. B.

2012-01-01

to be reliable in these conditions. The primary goal of any such actions is to maintain or refill the passive inventory available to cool the core, containment and spent fuel pool in the safety-related and seismically qualified Passive Containment Cooling Water Storage Tank (PCCWST). The seismically-qualified, ground-mounted Passive Containment Cooling Ancillary Water Storage Tank (PCCAWST) is also available for this function as appropriate. The primary effect of these actions would be to increase the coping time for the AP1000 during design basis events, as well as events such as those described above, from 72 hours without operator intervention to 7 days with minimal operator actions. These Operator actions necessary to protect the health and safety of the public are addressed in the Post-72 Hour procedures, as well as some EOPs, AOPs, ARPs and the Severe Accident Management Guidelines (SAMGs). Should the event continue to become more severe and plant conditions degrade further with indications of inadequate core cooling, the SAMGs provide guidance for strategies to address these hypothetical severe accident conditions. The AP1000 SAMG diagnoses and actions are prioritized to first utilize the AP1000 features that are expected to retain a damaged core inside the reactor vessel. Only one strategy is undertaken at any time. This strategy will be followed and its effectiveness evaluated before other strategies are undertaken. This is a key feature of both the symptom-oriented AP1000 EOPs and the AP1000 SAMGs which maximizes the probability of retaining a damaged core inside the reactor vessel and containment while minimizing the chances for confusion and human errors during implementation. The AP1000 SAMGs are simple and straight-forward and have been developed with considerable input from human factors and plant operations experts. Most importantly, and different from severe accident management strategies for other plants, the AP1000 SAMGs do not require diagnosis of

Some Subjects and Relations According to the Act about Safety at Work

Directory of Open Access Journals (Sweden)

Marino Đ. Učur

2015-01-01

Full Text Available Complex relations in the field of safety at work could not be present without the subjects which have a specific status and specific rights, obligations and responsibilities regulated by the Occupational Health and Safety Act. This paper deals with: employer’s designated employee for the implementation of occupational health and safety activities, employees’ elected representative for health and safety protection at work, occupational medicine specialist, occupational health and safety specialist and the committee for safety at work in the relations of safety at work.

Seismic design and performance of nuclear safety related RC structures based on new seismic design principle

International Nuclear Information System (INIS)

Murugan, R.; Sivathanu Pillai, C.; Chattopadhyaya, S.; Sundaramurthy, C.

2011-01-01

Full text: Seismic design of safety related Reinforced Concrete (RC) structures of Nuclear power plants (NPP) in India as per the present AERB codal procedures tries to ensure predominantly elastic behaviour under OBE so that the features of Nuclear Power Plant (NPP) necessary for continued safe operation are designed to remain functional and prevent accident (collapse) of NPP under SSE for which certain Structures, Systems and Components (SSCs) those are necessary to ensure the capability to shut down the reactor safely, are designed to remain functional. While the seismic design principles of non safety related structures as per Indian code (IS 1893-2002) are ensuring elastic behaviour under DBE and inelastic behaviour under MCE by utilizing ductility and energy dissipation capacity of the structure effectively. The design principle of AERB code is ensuring elastic behaviour under OBE and is not enlightening much inference about the overall structural behaviour under SSE (only ensuring the capability of certain SSCs required for safe shutdown of reactor). Various buildings and structures of Indian Nuclear power plant are classified from the basis of associated safety functions in a descending order in according with their roles in preventions and mitigation of an accident or support functions for prevention. This paper covers a comprehensive seismic analysis and design methodology based on the AERB codal provisions followed for safety related RC structure taking Diesel Generator Building of PFBR as a case study and study and investigates its performance under OBE and SSE by carrying out Non-linear static Pushover analysis. Based on the analysis, observed variations, recommendations are given for getting the desired performance level so as to implement performance based design in the future NPP design

Basis for the safety approach for design and assessment of Generation IV nuclear systems

International Nuclear Information System (INIS)

Fiorini, G.L.; Leahy, T.

2009-01-01

The primary objective of the RSWG is the implementation of a harmonized approach on long-term safety, and to address risk and regulatory issues in development of the next generation of nuclear systems. To this end, the group is proposing safety goals and evaluation methodology applicable for the design and assessment of future systems. The paper resumes the content of the first RSWG report which provides insights for the safety approach and assists the GIF Systems Steering Committee as well as the GIF Experts Group and the GIF Policy Group for the definition of the most adequate safety related Gen IV R and D. The document is also an essential contributor to help identifying the needed supportive crosscut R and D effort (i.e. applicable to all the innovative nuclear technologies). Although the report presents a number of thoughts and recommendations, it really represents only the start of the efforts for the RSWG. (author)

Study on the KALIMER safety approach

International Nuclear Information System (INIS)

Kim, Eui Kwang; Han, Do Hee; Kim, Young Cheol.

1997-01-01

This study describes KALIMER's safety approach, how to establish the safety criteria and temperature limit, how to define safety evaluation events, and some safety research and development needs items. It is recommended that the KALIMER's approach to safety use seven levels of safety design and a defense-in-depth design approach with particular emphasis on inherent passive features. In order to establish as set DBEs for KALIMER safety evaluation, the procedure is explained how to define safety evaluation events. Final selection is to be determined later with the final establishment of design concepts. On the basis of preliminary studies and evaluation of the plant safety related areas, the KALIMER and PRISM have following three main difference that may require special research and development for KALIMER. (author). 7 refs., 6 tabs., 6 figs

Final report on the safety assessment of sodium cetearyl sulfate and related alkyl sulfates as used in cosmetics.

Science.gov (United States)

Fiume, Monice; Bergfeld, Wilma F; Belsito, Donald V; Klaassen, Curtis D; Marks, James G; Shank, Ronald C; Slaga, Thomas J; Snyder, Paul W; Alan Andersen, F

2010-05-01

Sodium cetearyl sulfate is the sodium salt of a mixture of cetyl and stearyl sulfate. The other ingredients in this safety assessment are also alkyl salts, including ammonium coco-sulfate, ammonium myristyl sulfate, magnesium coco-sulfate, sodium cetyl sulfate, sodium coco/hydrogenated tallow sulfate, sodium coco-sulfate, sodium decyl sulfate, sodium ethylhexyl sulfate, sodium myristyl sulfate, sodium oleyl sulfate, sodium stearyl sulfate, sodium tallow sulfate, sodium tridecyl sulfate, and zinc coco-sulfate. These ingredients are surfactants used at concentrations from 0.1% to 29%, primarily in soaps and shampoos. Many of these ingredients are not in current use. The Cosmetic Ingredient Review (CIR) Expert Panel previously completed a safety assessment of sodium and ammonium lauryl sulfate. The data available for sodium lauryl sulfate and ammonium lauryl sulfate provide sufficient basis for concluding that sodium cetearyl sulfate and related alkyl sulfates are safe in the practices of use and concentration described in the safety assessment.

Analysis of regulatory requirement for beyond design basis events of SMART

International Nuclear Information System (INIS)

Kim, W. S.; Seol, K. W.

2000-01-01

To enhance the safety of SMART reactor, safety and regulatory requirements associated with beyond design basis events (beyond BDE), which were developed and applied to advanced light water reactor designs, were analyzed along with a design status of passive reactor. And, based on these requirements, their applicability on the SMART design was evaluated. In the design aspect, severe accident prevention and mitigation features, containment performance, and accident management were analyzed. The evaluation results show that the requirement related to beyond DBE such as ATWS, loss of residual heat removal during shutdown operation, station blackout, fire, inter-system LOCA, and well-known events from severe accident phenomena is applicable to the SMART design. However, comprehensive approach against beyond DBE is not yet provided in the SMART design, and then it is required to designate and analyze the beyond DBE-related features. This study is expected to contribute to efforts to improve plant safety and to establish regulatory requirements for safety review

Key issues on safety design basis selection and safety assessment

International Nuclear Information System (INIS)

An, S.; Togo, Y.

1976-01-01

In current fast reactor design in Japan, four design accident conditions and four design seismic conditions are adopted as the design base classifications. These are classified by the considerations on both likelihood of occurrence and the severeness of the consequences. There are several major problem areas in safety design consideration such as core accident problems which include fuel sodium interaction, fuel failure propagation and residual decay heat removal, and decay heat removal systems problems which is more or less the problem of selection of appropriate system and of assurance of high reliability of the system. In view of licensing, two kinds of accidents are postulated in evaluating the adequacy of a reactor site. The one is the ''major accident'' which is the accident to give most severe radiation hazard to the public from technical point of view. The other is the ''hypothetical accident'', induced public accident of which is severer than that of major accident. While the concept of the former is rather unique to Japanese licensing, the latter is almost equivalent to design base hypothetical accident of the US practice. In this paper, design bases selections, key safety issues and some of the licensing considerations in Japan are described

SOS-1 seminar about safety culture

International Nuclear Information System (INIS)

Wahlstroem, B.; Hammar, L.

2000-01-01

The aim of the seminar was to discuss safety culture in nuclear power utilities, and to exchange experiences about how the term safety culture is accepted by the personnel. The titles of the presentations are: 1) Organisational culture. General ideas as basis for organising; 2) Safety culture - ability and will; 3) View on safety culture at Swedish and Finnish nuclear power plants; 4) Safety culture at Barsebaeck Power Company; 5) Safety culture at Olkiluoto Nuclear Power Plant; 6) How do we improve the safety culture at OKG AB?; 7) Safety culture activities at Ringhals; 8) Aspects in relation to safety culture; 9) Development of regulatory activities/effectiveness of STUK - development as an aspect of culture; 10) Organisational culture research at STUK's Department of Nuclear Reactor Regulation; 11) The IAEA safety culture services; 12) Industrial safety - different perspectives and cultures. (EHS)

78 FR 29392 - Embedded Digital Devices in Safety-Related Systems, Systems Important to Safety, and Items Relied...

Science.gov (United States)

2013-05-20

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0098] Embedded Digital Devices in Safety-Related Systems, Systems Important to Safety, and Items Relied on for Safety AGENCY: Nuclear Regulatory Commission. ACTION... (NRC) is issuing for public comment Draft Regulatory Issue Summary (RIS) 2013-XX, ``Embedded Digital...

Safety-related instrumentation and control systems for nuclear power plants

International Nuclear Information System (INIS)

1984-01-01

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

An analysis of electronic health record-related patient safety incidents.

Science.gov (United States)

Palojoki, Sari; Mäkelä, Matti; Lehtonen, Lasse; Saranto, Kaija

2017-06-01

The aim of this study was to analyse electronic health record-related patient safety incidents in the patient safety incident reporting database in fully digital hospitals in Finland. We compare Finnish data to similar international data and discuss their content with regard to the literature. We analysed the types of electronic health record-related patient safety incidents that occurred at 23 hospitals during a 2-year period. A procedure of taxonomy mapping served to allow comparisons. This study represents a rare examination of patient safety risks in a fully digital environment. The proportion of electronic health record-related incidents was markedly higher in our study than in previous studies with similar data. Human-computer interaction problems were the most frequently reported. The results show the possibility of error arising from the complex interaction between clinicians and computers.

Safety criteria for siting a nuclear power plant

International Nuclear Information System (INIS)

2001-01-01

The guide sets forth requirements for safety of the population and the environment in nuclear power plant siting. It also sets out the general basis for procedures employed by other competent authorities when they issue regulations or grant licences. On request STUK (Radiation and Nuclear Safety Authority of Finland) issues case-specific statements about matters relating to planning and about other matters relating to land use in the environment of nuclear power plants

The power of simplification: Operator interface with the AP1000{sup R} during design-basis and beyond design-basis events

Energy Technology Data Exchange (ETDEWEB)

Williams, M. G.; Mouser, M. R.; Simon, J. B. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

designed to be reliable in these conditions. The primary goal of any such actions is to maintain or refill the passive inventory available to cool the core, containment and spent fuel pool in the safety-related and seismically qualified Passive Containment Cooling Water Storage Tank (PCCWST). The seismically-qualified, ground-mounted Passive Containment Cooling Ancillary Water Storage Tank (PCCAWST) is also available for this function as appropriate. The primary effect of these actions would be to increase the coping time for the AP1000 during design basis events, as well as events such as those described above, from 72 hours without operator intervention to 7 days with minimal operator actions. These Operator actions necessary to protect the health and safety of the public are addressed in the Post-72 Hour procedures, as well as some EOPs, AOPs, ARPs and the Severe Accident Management Guidelines (SAMGs). Should the event continue to become more severe and plant conditions degrade further with indications of inadequate core cooling, the SAMGs provide guidance for strategies to address these hypothetical severe accident conditions. The AP1000 SAMG diagnoses and actions are prioritized to first utilize the AP1000 features that are expected to retain a damaged core inside the reactor vessel. Only one strategy is undertaken at any time. This strategy will be followed and its effectiveness evaluated before other strategies are undertaken. This is a key feature of both the symptom-oriented AP1000 EOPs and the AP1000 SAMGs which maximizes the probability of retaining a damaged core inside the reactor vessel and containment while minimizing the chances for confusion and human errors during implementation. The AP1000 SAMGs are simple and straight-forward and have been developed with considerable input from human factors and plant operations experts. Most importantly, and different from severe accident management strategies for other plants, the AP1000 SAMGs do not require diagnosis

Experience on environmental qualification of safety-related components for Darlington Nuclear Generating Station

International Nuclear Information System (INIS)

Yu, A.S.; Kukreti, B.M.

1987-01-01

The proliferation of Nuclear Power Plant safety concerns has lead to increasing attention over the Environmental Qualification (EQ) of Nuclear Power Plant Safety-Related Components to provide the assurance that the safety related equipment will meet their intended functions during normal operation and postulated accident conditions. The environmental qualification of these components is also a Licensing requirement for Darlington Nuclear Generating Station. This paper provides an overview of EQ and the experience of a pilot project, in the qualification of the Main Moderator System safety-related functions for the Darlington Nuclear Generating Station currently under construction. It addresses the various phases of qualification from the identification of the EQ Safety-Related Components List, definition of location specific service conditions (normal, adbnormal and accident), safety-related functions, Environmental Qualification Assessments and finally, an EQ system summary report for the Main Moderator System. The results of the pilot project are discussed and the methodology reviewed. The paper concludes that the EQ Program developed for Darlington Nuclear Generating Station, as applied to the qualification of the Main Moderator System, contained all the elements necessary in the qualification of safety-related equipment. The approach taken in the qualification of the Moderator safety-related equipment proves to provide a sound framework for the qualification of other safety-related components in the station

Blanket safety by GEMSAFE methodology

International Nuclear Information System (INIS)

Sawada, Tetsuo; Saito, Masaki

2001-01-01

General Methodology of Safety Analysis and Evaluation for Fusion Energy Systems (GEMSAFE) has been applied to a number of fusion system designs, such as R-tokamak, Fusion Experimental Reactor (FER), and the International Thermonuclear Experimental Reactor (ITER) designs in the both stages of Conceptual Design Activities (CDA) and Engineering Design Activities (EDA). Though the major objective of GEMSAFE is to reasonably select design basis events (DBEs) it is also useful to elucidate related safety functions as well as requirements to ensure its safety. In this paper, we apply the methodology to fusion systems with future tritium breeding blankets and make clear which points of the system should be of concern from safety ensuring point of view. In this context, we have obtained five DBEs that are related to the blanket system. We have also clarified the safety functions required to prevent accident propagations initiated by those blanket-specific DBEs. The outline of the methodology is also reviewed. (author)

Problems in Food Safety of Hunan Province and Countermeasures

Institute of Scientific and Technical Information of China (English)

Fanfan; OUYANG; Fangming; DENG

2014-01-01

In recent years,serious food safety accidents are of frequent occurrence. Although government has taken many practical and feasible measures to contain food safety accidents,new food safety accidents still emerge in large numbers. In this situation,food safety control is a long-term and arduous task to be performed jointly by many government departments. Finally,it presents corresponding countermeasures and recommendations on the basis of current situations of food safety in Hunan Province,problem causes,in combination with control measures related to food safety both at home and abroad.

Allowed outage time for test and maintenance - Optimization of safety

International Nuclear Information System (INIS)

Cepin, M.; Mavko, B.

1997-01-01

The main objective of the project is the development and application of methodologies for improvement and optimization of test and maintenance activities for safety related equipment in NPPs on basis of their enhanced safety. The probabilistic safety assessment serves as a base, which does not mean the replacement of the deterministic analyses but the consideration of probabilistic safety assessment results as complement to deterministic results. 15 refs, 2 figs

Is road safety management linked to road safety performance?

Science.gov (United States)

Papadimitriou, Eleonora; Yannis, George

2013-10-01

This research aims to explore the relationship between road safety management and road safety performance at country level. For that purpose, an appropriate theoretical framework is selected, namely the 'SUNflower' pyramid, which describes road safety management systems in terms of a five-level hierarchy: (i) structure and culture, (ii) programmes and measures, (iii) 'intermediate' outcomes'--safety performance indicators (SPIs), (iv) final outcomes--fatalities and injuries, and (v) social costs. For each layer of the pyramid, a composite indicator is implemented, on the basis of data for 30 European countries. Especially as regards road safety management indicators, these are estimated on the basis of Categorical Principal Component Analysis upon the responses of a dedicated road safety management questionnaire, jointly created and dispatched by the ETSC/PIN group and the 'DaCoTA' research project. Then, quasi-Poisson models and Beta regression models are developed for linking road safety management indicators and other indicators (i.e. background characteristics, SPIs) with road safety performance. In this context, different indicators of road safety performance are explored: mortality and fatality rates, percentage reduction in fatalities over a given period, a composite indicator of road safety final outcomes, and a composite indicator of 'intermediate' outcomes (SPIs). The results of the analyses suggest that road safety management can be described on the basis of three composite indicators: "vision and strategy", "budget, evaluation and reporting", and "measurement of road user attitudes and behaviours". Moreover, no direct statistical relationship could be established between road safety management indicators and final outcomes. However, a statistical relationship was found between road safety management and 'intermediate' outcomes, which were in turn found to affect 'final' outcomes, confirming the SUNflower approach on the consecutive effect of each layer

Accelerator production of tritium authorization basis strategy

International Nuclear Information System (INIS)

Miller, L.A.; Edwards, J.; Rose, S.

1996-01-01

The Accelerator Production of Tritium (APT) project has proposed a strategy to develop the APT authorization basis and safety case based on DOE orders and fundamental requirements for safe operation. The strategy is viable regardless of whether the APT is regulated by DOE or by an external regulatory body. Currently the operation of Department of Energy (DOE) facilities is authorized by DOE and regulated by DOE orders and regulations while meeting the environmental protection requirements of the Environmental Protection Agency (EPA) and the states. In the spring of 1994, Congress proposed legislation and held hearings related to requiring all DOE operations to be subject to external regulation. On January 25, 1995, DOE, with the support of the White House Council on Environmental Quality, created the Advisory Committee on External Regulation of Department of Energy Nuclear Safety. This committee divided its recommendations into three areas: (1) facility safety, (2) worker safety, and (3) environmental protection. In the area of facility safety the committee recommended external regulation of DOE nuclear facilities by either the Nuclear Regulatory Commission (NRC) or a restructured Defense Nuclear Facilities Safety Board (DNFSB). In the area of worker safety, the committee recommended that the Occupational Safety and Health Administration (OSHA) regulate DOE nuclear facilities. In the environmental protection area, the committee did not recommend a change in the regulation by the EPA and the states of DOE nuclear facilities. If these recommendations are accepted, all DOE nuclear facilities will be impacted to some extent

Relation between water chemistry and operational safety

International Nuclear Information System (INIS)

Oliveira, M.F. de.

1991-01-01

This report describes the relation between chemistry/radiochemistry and operational safety, the technics bases for chemical and radiochemical parameters and an analysis of the Annual Report of Angra I Operation and OSRAT Mission report to 1989 in this area too. Furthermore it contains the transcription of the technical Specifications related to the chemistry and radiochemistry for Angra I. (author)

Definitions of engineered safety features and related features for nuclear power plants

International Nuclear Information System (INIS)

1986-01-01

In light water moderated, light water cooled nuclear power plants, definitions are given of engineered safety features which are designed to suppress or prevent dispersion of radioactive materials due to damage etc. of fuel at the times of power plant failures, and of related features which are designed to actuate or operate the engineered safety features. Contents are the following: scope of engineered safety features and of related features; classification of engineered safety features (direct systems and indirect systems) and of related features (auxiliaries, emergency power supply, and protective means). (Mori, K.)

PRELIMINARY SELECTION OF MGR DESIGN BASIS EVENTS

International Nuclear Information System (INIS)

Kappes, J.A.

1999-01-01

The purpose of this analysis is to identify the preliminary design basis events (DBEs) for consideration in the design of the Monitored Geologic Repository (MGR). For external events and natural phenomena (e.g., earthquake), the objective is to identify those initiating events that the MGR will be designed to withstand. Design criteria will ensure that radiological release scenarios resulting from these initiating events are beyond design basis (i.e., have a scenario frequency less than once per million years). For internal (i.e., human-induced and random equipment failures) events, the objective is to identify credible event sequences that result in bounding radiological releases. These sequences will be used to establish the design basis criteria for MGR structures, systems, and components (SSCs) design basis criteria in order to prevent or mitigate radiological releases. The safety strategy presented in this analysis for preventing or mitigating DBEs is based on the preclosure safety strategy outlined in ''Strategy to Mitigate Preclosure Offsite Exposure'' (CRWMS M andO 1998f). DBE analysis is necessary to provide feedback and requirements to the design process, and also to demonstrate compliance with proposed 10 CFR 63 (Dyer 1999b) requirements. DBE analysis is also required to identify and classify the SSCs that are important to safety (ITS)

Risks and safety aspects related to PET/MR examinations

International Nuclear Information System (INIS)

Brix, Gunnar; Nekolla, Elke A.; Nosske, Dietmar; Griebel, Juergen

2009-01-01

The introduction of positron emission tomography (PET)/magnetic resonance (MR) systems into medical practice in the foreseeable future may not only lead to a gain in clinical diagnosis compared to PET/computed tomography (CT) imaging due to the superior soft-tissue contrast of the MR technology but can also substantially reduce exposure of patients to ionizing radiation. On the other hand, there are also risks and health effects associated with the use of diagnostic MR devices that have to be considered carefully. This review article summarizes biophysical and biological aspects, which are of relevance for the assessment of health effects related to the exposure of patients to both ionizing radiation in PET and magnetic and electromagnetic fields in MR. On this basis, some considerations concerning the justification and optimization of PET/MR examinations are presented - as far as this is possible at this very early stage. Current safety standards do not take into account synergistic effects of ionizing radiation and magnetic and electromagnetic fields. In the light of the developing PET/MR technology, there is an urgent need to investigate this aspect in more detail for exposure levels that will occur at PET/MR systems. (orig.)

Nuclear safety regulation on nuclear safety equipment activities in relation to human and organizational factors

International Nuclear Information System (INIS)

Li Tianshu

2013-01-01

Based on years of knowledge in nuclear safety supervision and experience of investigating and dealing with violation events in repair welding of DFHM, this paper analyzes major faults in manufacturing and maintaining activities of nuclear safety equipment in relation to human and organizational factors. It could be deducted that human and organizational factors has definitely become key features in the development of nuclear energy and technology. Some feasible measures to reinforce supervision on nuclear safety equipment activities have also been proposed. (author)

Quantitative safety goals for the regulatory process

International Nuclear Information System (INIS)

Joksimovic, V.; O'Donnell, L.F.

1981-01-01

The paper offers a brief summary of the current regulatory background in the USA, emphasizing nuclear, related to the establishment of quantitative safety goals as a way to respond to the key issue of 'how safe is safe enough'. General Atomic has taken a leading role in advocating the use of probabilistic risk assessment techniques in the regulatory process. This has led to understanding of the importance of quantitative safety goals. The approach developed by GA is discussed in the paper. It is centred around definition of quantitative safety regions. The regions were termed: design basis, safety margin or design capability and safety research. The design basis region is bounded by the frequency of 10 -4 /reactor-year and consequences of no identifiable public injury. 10 -4 /reactor-year is associated with the total projected lifetime of a commercial US nuclear power programme. Events which have a 50% chance of happening are included in the design basis region. In the safety margin region, which extends below the design basis region, protection is provided against some events whose probability of not happening during the expected course of the US nuclear power programme is within the range of 50 to 90%. Setting the lower mean frequency to this region of 10 -5 /reactor-year is equivalent to offering 90% assurance that an accident of given severity will not happen. Rare events with a mean frequency below 10 -5 can be predicted to occur. However, accidents predicted to have a probability of less than 10 -6 are 99% certain not to happen at all, and are thus not anticipated to affect public health and safety. The area between 10 -5 and 10 -6 defines the frequency portion of the safety research region. Safety goals associated with individual risk to a maximum-exposed member of public, general societal risk and property risk are proposed in the paper

Plant functional modelling as a basis for assessing the impact of management on plant safety

International Nuclear Information System (INIS)

Rasmussen, Birgitte; Petersen, Kurt E.

1999-01-01

A major objective of the present work is to provide means for representing a chemical process plant as a socio-technical system, so as to allow hazard identification at a high level in order to identify major targets for safety development. The main phases of the methodology are: (1) preparation of a plant functional model where a set of plant functions describes coherently hardware, software, operations, work organization and other safety related aspects. The basic principle is that any aspect of the plant can be represented by an object based upon an Intent and associated with each Intent are Methods, by which the Intent is realized, and Constraints, which limit the Intent. (2) Plant level hazard identification based on keywords/checklists and the functional model. (3) Development of incident scenarios and selection of hazardous situation with different safety characteristics. (4) Evaluation of the impact of management on plant safety through interviews. (5) Identification of safety critical ways of action in the management system, i.e. identification of possible error- and violation-producing conditions

Using Active Learning to Identify Health Information Technology Related Patient Safety Events.

Science.gov (United States)

Fong, Allan; Howe, Jessica L; Adams, Katharine T; Ratwani, Raj M

2017-01-18

The widespread adoption of health information technology (HIT) has led to new patient safety hazards that are often difficult to identify. Patient safety event reports, which are self-reported descriptions of safety hazards, provide one view of potential HIT-related safety events. However, identifying HIT-related reports can be challenging as they are often categorized under other more predominate clinical categories. This challenge of identifying HIT-related reports is exacerbated by the increasing number and complexity of reports which pose challenges to human annotators that must manually review reports. In this paper, we apply active learning techniques to support classification of patient safety event reports as HIT-related. We evaluated different strategies and demonstrated a 30% increase in average precision of a confirmatory sampling strategy over a baseline no active learning approach after 10 learning iterations.

The Evolution of System Safety at NASA

Science.gov (United States)

Dezfuli, Homayoon; Everett, Chris; Groen, Frank

2014-01-01

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

Safety concerns related to modular/prefabricated building construction.

Science.gov (United States)

Fard, Maryam Mirhadi; Terouhid, Seyyed Amin; Kibert, Charles J; Hakim, Hamed

2017-03-01

The US construction industry annually experiences a relatively high rate of fatalities and injuries; therefore, improving safety practices should be considered a top priority for this industry. Modular/prefabricated building construction is a construction strategy that involves manufacturing of the whole building or some of its components off-site. This research focuses on the safety performance of the modular/prefabricated building construction sector during both manufacturing and on-site processes. This safety evaluation can serve as the starting point for improving the safety performance of this sector. Research was conducted based on Occupational Safety and Health Administration investigated accidents. The study found 125 accidents related to modular/prefabricated building construction. The details of each accident were closely examined to identify the types of injury and underlying causes. Out of 125 accidents, there were 48 fatalities (38.4%), 63 hospitalized injuries (50.4%), and 14 non-hospitalized injuries (11.2%). It was found that, the most common type of injury in modular/prefabricated construction was 'fracture', and the most common cause of accidents was 'fall'. The most frequent cause of cause (underlying and root cause) was 'unstable structure'. In this research, the accidents were also examined in terms of corresponding location, occupation, equipment as well as activities during which the accidents occurred. For improving safety records of the modular/prefabricated construction sector, this study recommends that future research be conducted on stabilizing structures during their lifting, storing, and permanent installation, securing fall protection systems during on-site assembly of components while working from heights, and developing training programmes and standards focused on modular/prefabricated construction.

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...

Resistance ability evaluation of safety-related structures for the simulated aircraft accident

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Kim, Sung Woon; Choi, Jang Kyu [Daewoo E and C Co., Ltd., Suwon (Korea, Republic of)] (and others)

2003-03-15

Aircraft accidents on nuclear safety-related structures can cause severe damage to the safety of NPP(Nuclear Power Plant)s. To assess the safety of nuclear safety-related structures, the local damage and the dynamic response of global structures should be investigated together. This study have compared several local damage assessment formulas suggested for aircraft as an impactor, and have set the assessment system of local damage for impact-proof design of NPP containment buildings. And the local damage of nuclear safety-related structures in operation in Korea for commercial aircraft as impactor have been estimated. Impact load-time functions of the aircraft crash have been decided to assessment the safety of nuclear safety-related structures against the intentional colliding of commercial aircraft. Boeing 747 and Boeing 767 is selected as target aircraft based on the operation frequencies and weights. Comparison of the fire analysis methods showed that the method considering heat convection and radiation is adequate for the temperature analysis of the aircraft fuel fire. Finally, the study covered the analysis of the major structural drawings and design drawings with which three-dimensional finite element model analysis is expected to be performed.

Development of Safety Analysis Technology for Integral Reactor

Energy Technology Data Exchange (ETDEWEB)

Sim, S. K. [Korea Atomic Energy Research Institute, Taejeon (Korea); Seul, K. W.; Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Sin, A. D. [Korea Institute of Nuclear Safety, Taejeon (Korea)

2000-03-01

The Nuclear Desalination Plant(NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated based on the design of foreign and domestic integral reactors. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current and advanced reactor designs, and use requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified. They includes the use of proven technology for new safety systems, the systematic classification and selection of design basis accidents, and the safety assurance of desalination-related systems. These efforts to identify and resolve the safety concerns in the design stage will provide the early confidence of SMART safety to designers, and the technical basis to evaluate the safety to reviewers in the future. 8 refs., 20 figs., 4 tabs. (Author)

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

Reliability of containment and safety-related structures

International Nuclear Information System (INIS)

Nessim, M.A.

1995-09-01

A research program on Reliability of Containment and Safety-related Structures has been developed and is described in this document. This program is designed to support AECB's regulatory activities aimed at ensuring the safety of these structures. These activities include evaluating submissions by operators and requesting special assessments when necessary. The results of the proposed research will also be useful in revising and enhancing the CSA design standards for containment and safety-related structures. The process of developing the research program started with an information collection and review phase. The sources of information included C-FER's previous work in the area, various recent research publications, regulatory documents and relevant design standards, and a detailed discussion with AECB staff. The second step was to outline the process of reliability evaluation, and identify the required models and parameters. Comparison between the required and available information was used to identify gaps in the state-of-the-art, and the research program was designed to fill these gaps. The program is organized in four major topics, namely: development of an approach for reliability analysis; compilation and development of the required analysis tools; application to specific problems related to design, assessment, maintenance and testing of structures; and testing and validation. It is suggested that the program should be supported by an on-going process of communication and consultation between AECB staff and industry experts. This will lend credibility to the results and facilitate their future application. (author). 1 fig

Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

International Nuclear Information System (INIS)

Il'kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I.; Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K.; Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A.; Haire, Jonathan M.; Forsberg, C.W.

2004-01-01

The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism

Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

Energy Technology Data Exchange (ETDEWEB)

Il' kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation); Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K. [All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation); Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A. [All-Russian Research Inst. of Applied Chemistry, Moscow (Russian Federation); Haire, Jonathan M.; Forsberg, C.W. [Oak Ridge National Lab., Oak Ridge (United States)

2004-07-01

The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism.

Prioritization of generic safety issues

International Nuclear Information System (INIS)

Emrit, R.; Minners, W.; VanderMolen, H.

1983-12-01

This report presents the priority rankings for generic safety issues related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The report focuses on the prioritization of generic safety issues. Issues primarily concerned with the licensing process or environmental protection and not directly related to safety have been excluded from prioritization. The prioritized issues include: TMI Action Plan items under development; previously proposed issues covered by Task Action Plans, except issues designated at Unresolved Safety Issues (USIs) which had already been assigned high priority; and newly-proposed issues. Future supplements to this report will include the prioritization of additional issues. The safety priority rankings are HIGH, MEDIUM, LOW, and DROP and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolutions of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative

Safety and Security Interface Technology Initiative

Energy Technology Data Exchange (ETDEWEB)

Dr. Michael A. Lehto; Kevin J. Carroll; Dr. Robert Lowrie

2007-05-01

Safety and Security Interface Technology Initiative Mr. Kevin J. Carroll Dr. Robert Lowrie, Dr. Micheal Lehto BWXT Y12 NSC Oak Ridge, TN 37831 865-576-2289/865-241-2772 carrollkj@y12.doe.gov Work Objective. Earlier this year, the Energy Facility Contractors Group (EFCOG) was asked to assist in developing options related to acceleration deployment of new security-related technologies to assist meeting design base threat (DBT) needs while also addressing the requirements of 10 CFR 830. NNSA NA-70, one of the working group participants, designated this effort the Safety and Security Interface Technology Initiative (SSIT). Relationship to Workshop Theme. “Supporting Excellence in Operations Through Safety Analysis,” (workshop theme) includes security and safety personnel working together to ensure effective and efficient operations. One of the specific workshop elements listed in the call for papers is “Safeguards/Security Integration with Safety.” This paper speaks directly to this theme. Description of Work. The EFCOG Safety Analysis Working Group (SAWG) and the EFCOG Security Working Group formed a core team to develop an integrated process involving both safety basis and security needs allowing achievement of the DBT objectives while ensuring safety is appropriately considered. This effort garnered significant interest, starting with a two day breakout session of 30 experts at the 2006 Safety Basis Workshop. A core team was formed, and a series of meetings were held to develop that process, including safety and security professionals, both contractor and federal personnel. A pilot exercise held at Idaho National Laboratory (INL) in mid-July 2006 was conducted as a feasibility of concept review. Work Results. The SSIT efforts resulted in a topical report transmitted from EFCOG to DOE/NNSA in August 2006. Elements of the report included: Drivers and Endstate, Control Selections Alternative Analysis Process, Terminology Crosswalk, Safety Basis

Canister storage building design basis accident analysis documentation

International Nuclear Information System (INIS)

KOPELIC, S.D.

1999-01-01

This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

Recursion relations for the overlap of a Morse continuum state with a Lanczos basis state

International Nuclear Information System (INIS)

Lutrus, C.K.; Suck Salk, S.H.

1988-01-01

In the resonant reactive scattering theory of Mundel, Berman, and Domcke [Phys. Rev. A 32, 181 (1985)], the overlap of a Morse continuum state and a Lanczos basis state appears in the expression of transition amplitude. In their study, recursion relations for Green's functions in the Lanczos basis were used for computational efficiency. In this paper we derive new recursion relations specifically for the evaluation of overlap between the Morse continuum wave and Lanczos basis state that appears in the transition amplitude of resonant scattering. They are found to be simple to use with great accuracy

Climate and climate-related issues for the safety assessment SR-Can

International Nuclear Information System (INIS)

Naeslund, Jens-Ove

2006-11-01

The purpose of this report is to document current scientific knowledge of the climate-related conditions and processes relevant to the long-term safety of a KBS-3 repository to a level required for an adequate treatment in the safety assessment SR-Can. The report also includes a concise background description of the climate system. The report includes three main chapters: A description of the climate system (Chapter 2); Identification and discussion of climate-related issues (Chapter 3); and, A description of the evolution of climate-related conditions for the safety assessment (Chapter 4). Chapter 2 includes an overview of present knowledge of the Earth climate system and the climate conditions that can be expected to occur in Sweden on a 100,000 year time perspective. Based on this, climate-related issues relevant for the long-term safety of a KBS-3 repository are identified. These are documented in Chapter 3 'Climate-related issues' to a level required for an adequate treatment in the safety assessment. Finally, in Chapter 4, 'Evolution of climate-related conditions for the safety assessment' an evolution for a 120,000 year period is presented, including discussions of identified climate-related issues of importance for repository safety. The documentation is from a scientific point of view not exhaustive, since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of a safety assessment. As further described in the SR-Can Main Report and in the Features Events and Processes report, the content of the present report has been audited by comparison with FEP databases compiled in other assessment projects. This report follows as far as possible the template for documentation of processes regarded as internal to the repository system. However, the term processes is not used in this report, instead the term issue has been used. Each issue includes a set of processes together resulting in the behaviour of a

Climate and climate-related issues for the safety assessment SR-Can

Energy Technology Data Exchange (ETDEWEB)

Naeslund, Jens-Ove (comp.)

2006-11-15

The purpose of this report is to document current scientific knowledge of the climate-related conditions and processes relevant to the long-term safety of a KBS-3 repository to a level required for an adequate treatment in the safety assessment SR-Can. The report also includes a concise background description of the climate system. The report includes three main chapters: A description of the climate system (Chapter 2); Identification and discussion of climate-related issues (Chapter 3); and, A description of the evolution of climate-related conditions for the safety assessment (Chapter 4). Chapter 2 includes an overview of present knowledge of the Earth climate system and the climate conditions that can be expected to occur in Sweden on a 100,000 year time perspective. Based on this, climate-related issues relevant for the long-term safety of a KBS-3 repository are identified. These are documented in Chapter 3 'Climate-related issues' to a level required for an adequate treatment in the safety assessment. Finally, in Chapter 4, 'Evolution of climate-related conditions for the safety assessment' an evolution for a 120,000 year period is presented, including discussions of identified climate-related issues of importance for repository safety. The documentation is from a scientific point of view not exhaustive, since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of a safety assessment. As further described in the SR-Can Main Report and in the Features Events and Processes report, the content of the present report has been audited by comparison with FEP databases compiled in other assessment projects. This report follows as far as possible the template for documentation of processes regarded as internal to the repository system. However, the term processes is not used in this report, instead the term issue has been used. Each issue includes a set of processes together resulting in the

Model review and evaluation for application in DOE safety basis documentation of chemical accidents - modeling guidance for atmospheric dispersion and consequence assessment

Energy Technology Data Exchange (ETDEWEB)

Lazaro, M. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Woodarad, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanna, S. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hesse, D. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Huang, J. -C. [Argonne National Lab. (ANL), Argonne, IL (United States); Lewis, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Mazzola, C. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

1997-09-01

The U.S. Department of Energy (DOE), through its Defense Programs (DP), Office of Engineering and Operations Suppon, established the Accident Phenomenology and Consequence (AP AC) Methodology Evaluation Program to identify and evaluate methodologies and computer codes to support accident phenomenological and consequence calculations for both radiological and nonradiological materials at DOE facilities and to identify development needs. The program is also intended to define and recommend "best or good engineering/safety analysis practices" to be followed in preparing ''design or beyond design basis" assessments to be included in DOE nuclear and nonnuclear facility safety documents. The AP AC effort is intended to provide scientifically sound and more consistent analytical approaches, by identifying model selection procedures and application methodologies, in order to enhance safety analysis activities throughout the DOE complex.

Safety evaluation report of the Waste Isolation Pilot Plant safety analysis report: Contact-handled transuranic waste disposal operations

International Nuclear Information System (INIS)

1997-02-01

DOE 5480.23, Nuclear Safety Analysis Reports, requires that the US Department of Energy conduct an independent, defensible, review in order to approve a Safety Analysis Report (SAR). That review and the SAR approval basis is documented in this formal Safety Evaluation Report (SER). This SER documents the DOE's review of the Waste Isolation Pilot Plant SAR and provides the Carlsbad Area Office Manager, the WIPP SAR approval authority, with the basis for approving the safety document. It concludes that the safety basis documented in the WIPP SAR is comprehensive, correct, and commensurate with hazards associated with planned waste disposal operations

Aging and service wear of air-operated valves used in safety-related systems at nuclear power plants

International Nuclear Information System (INIS)

Cox, D.F.; McElhaney, K.L.; Staunton, R.H.

1995-05-01

Air-operated valves (AOVs) are used in a variety of safety-related applications at nuclear power plants. They are often used where rapid stroke times are required or precise control of the valve obturator is required. They can be designed to operate automatically upon loss of power, which is often desirable when selecting components for response to design basis conditions. The purpose of this report is to examine the reported failures of AOVs and determine whether there are identifiable trends in the failures related to predictable causes. This report examines the specific components that comprise a typical AOV, how those components fail, when they fail, and how such failures are discovered. It also examines whether current testing frequencies and methods are effective in predicting such failures

Safety-related incidents at the Finnish nuclear power plants

International Nuclear Information System (INIS)

Lehtinen, P.

1986-03-01

This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Reports, Operation of Finnish Nuclear Power Plants, which are supplemented by this report intended for experts. (author)

Safety-related incidents at the Finnish nuclear power plants

International Nuclear Information System (INIS)

Lehtinen, P.

1985-01-01

This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Reports, Operation of Finnish Nuclear Power Plants, which are supplemented by this report intended for experts. (author)

Study on some safety-related aspects of tyre use

NARCIS (Netherlands)

Jansen, S.T.H.; Schmeitz, A.J.C.; Maas, S.; Rodarius, C.; Akkermans, L.

2014-01-01

The tyre is a key component that affects road safety. The European commission has posted a tender aimed to study what measures on a European level can be taken in relation to the use of tyres to improve road safety. The results of this study, supported by a cost benefit analyses and carried out by

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

International Nuclear Information System (INIS)

2004-01-01

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

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

International Nuclear Information System (INIS)

2005-01-01

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

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

International Nuclear Information System (INIS)

2000-01-01

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

Can we use IEC 61850 for safety related functions?

Directory of Open Access Journals (Sweden)

Luca Rocca

2016-08-01

Full Text Available Safety is an essential issue for processes that present high risk for human beings and environment. An acceptable level of risk is obtained both with actions on the process itself (risk reduction and with the use of special safety systems that switch the process into safe mode when a fault or an abnormal operation mode happens. These safety systems are today based on digital devices that communicate through digital networks. The IEC 61508 series specifies the safety requirements of all the devices that are involved in a safety function, including the communication network. Also electrical generation and distribution systems are processes that may have a significant level of risk, so the criteria stated by the IEC 61508 applies. Starting from this consideration, the paper analyzes the safety requirement for the communication network and compare them with the services of the communication protocol IEC 61850 that represents the most used protocol for automation of electrical plants. The goal of this job is to demonstrate that, from the technical point of view, IEC 61850 can be used for implementing safety-related functions, even if a formal safety certification is still missing.

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)

FLAMMABLE GAS TECHNICAL BASIS DOCUMENT

Energy Technology Data Exchange (ETDEWEB)

KRIPPS, L.J.

2005-03-03

This document describes the qualitative evaluation of frequency and consequences for DST and SST representative flammable gas accidents and associated hazardous conditions without controls. The evaluation indicated that safety-significant structures, systems and components (SSCs) and/or technical safety requirements (TSRs) were required to prevent or mitigate flammable gas accidents. Discussion on the resulting control decisions is included. This technical basis document was developed to support WP-13033, Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process for the flammable gas representative accidents and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the event frequency and consequence.

Feedback of safety - related operational experience: Lessons learned

Energy Technology Data Exchange (ETDEWEB)

Elias, D [Commonwealth Edison Co. (United States)

1997-09-01

The presentation considers the following aspects of feedback of safety-related operational experience: lessons learned program, objectives, personnel characteristics; three types of documents for transmitting lessons learned issues.

Feedback of safety - related operational experience: Lessons learned

International Nuclear Information System (INIS)

Elias, D.

1997-01-01

The presentation considers the following aspects of feedback of safety-related operational experience: lessons learned program, objectives, personnel characteristics; three types of documents for transmitting lessons learned issues

The basis and safety of food irradiation. Advantages of radiation treatment for food sanitation and storage

Energy Technology Data Exchange (ETDEWEB)

Ito, Hitoshi [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

2001-09-01

The food irradiation has the history of more than 60 years in its development. However, its commercial application has not been promoted well in Japan even though the safety of irradiated foods was confirmed. Recently, relevant authorities in 52 countries have given clearance to many commodities, and irradiated foods are commercially distributed in USA and EU countries. The international situation makes some unavoidable circumstances which can not close the commercialization of food irradiation in Japan. The present report contains the basis and application of food irradiation, and history of development in the World and Japan. Moreover, the safety of irradiated foods are demonstrated from many evidences of researches in animal feeding tests, in analysis of radiolytic products, in nutritional evaluations and in microbiological studies of irradiated foods. Especially, it makes obvious from the results of many researches that unique radiolytic products can not be produced by irradiation of foods. Because main radiation effects are induced by oxidation degradation of food components as similar to natural oxidation by heating or UV light. Radiation engineering for commercial process and identification methods of irradiated foods are also presented. (author)

Canister storage building design basis accident analysis documentation

Energy Technology Data Exchange (ETDEWEB)

KOPELIC, S.D.

1999-02-25

This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

Safety-related occurrences at the Finnish nuclear power plants

International Nuclear Information System (INIS)

Reponen, H.; Viitasaari, O.

1985-04-01

This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Report for this period, Operation of Finnish Nuclear Power Plants (STUK-B-YTO 7), which is supplemented by this report intended for experts. (author)

Staff report on the environmental qualification of safety-related electrical equipment

International Nuclear Information System (INIS)

1977-12-01

The current NRC safety review process for nuclear power plants includes criteria related to the qualification of certain electrical equipment. These criteria require that electrical equipment important to safety must be qualified to function in the environment that might result from various accident conditions. Although such criteria have been applied since the early days of commercial nuclear power, the details of these criteria have been changed over the years. The evolution of environmental qualification of safety-related electrical equipment is described in Appendix A

Sizewell B nuclear power station: the basis for the decision by the Health and Safety Executive to grant consent to load fuel into the reactor

International Nuclear Information System (INIS)

1994-01-01

The licensing and consent process and the basis for granting a consent for Nuclear Electric to load fuel into the Sizewell B reactor in the United Kingdom are explained. Consent was granted by the UK Nuclear Installations Inspectorate on behalf of the Health and Safety Executive on satisfactory completion of construction and those commissioning stages needed to proceed safely, and the production of a satisfactory safety case. A summary of the assessment of the safety case is appended. It covers the reactor core, coolant system structural integrity, engineered safety features, main and essential electrical system, control and instrumentation, radioactive waste management, radiological protection, fuel storage and handling, civil works and structures, fault analysis, human factors, hazard analysis, quality assurance, and decommissioning. (UK)

Commissioning of research reactors. Safety guide

International Nuclear Information System (INIS)

2006-01-01

The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project

Nuclear Safety Bureau: safety objectives and principles for the proposed ANSTO reactor

International Nuclear Information System (INIS)

Westall, D.

1993-01-01

Siting criteria and safety assessment principles were previously promulgated by the Australian Atomic Energy Commission (AAEC), and have been applied by ANSTO and the Nuclear Safety Bureau (NSB). The NSB is revising these criteria and principles to take account of evolving nuclear safety standards and practices. The NSB Safety and Siting Assessment Principles (SSAP) are presented and it is estimated that it will provide a comprehensive basis for the safety assessment of research reactors in Australia, and be applicable to all stages of a reactor project: siting: design and construction; operation; modification; and decommissioning. The SSAP are similar to the principles promulgated by the AAEC, in that probabilistic safety criteria are set for assessment of design, however these criteria are complimentary to a deterministic design basis approach. This is a similar approach to that recently published by the UK Nuclear Installations Inspectorate 4 . Siting principles are now also included, where they were previously separate, and require a consideration of the consequences of severe accidents which are an extension of accidents catered for by the design of the plant. Criteria for radiation doses due to normal operations and design basis accidents are included in the principles for safety assessment. 9 refs

Safety related experience in FFTF startup and operation

International Nuclear Information System (INIS)

Peterson, R.E.; Halverson, T.G.; Daughtry, J.W.

1982-06-01

The Fast Flux Test Facility (FFTF) is a 400 MW(t) sodium cooled fast reactor operating at the Hanford Engineering Development Laboratory, Richland, Washington, to conduct fuels and materials testing in support of the US LMFBR program. Startup and initial power ascension testing of the facility involved a comprehensive series of readiness reviews and acceptance tests, many of which relate to the inherent safety of the plant. Included are physics measurements, natural circulation, integrated containment leakage, shielding effectiveness, fuel failure detection, and plant protection system tests. Described are the measurements taken to confirm the design safety margins upon which the operating authorization of the plant was based. These measurements demonstrate that large margins of safety are available in the FFTF design

A Complete Security of Criminological Safety of Minors

Directory of Open Access Journals (Sweden)

Andrey I. Saveliev

2016-11-01

Full Text Available The article considers questions relating to the diversity of theoretical comprehension levels of criminological safety of minors. The Author analyzes the normative legal basis of activities of subjects of prevention and protection of children's rights

Commercial grade item (CGI) dedication of generators for nuclear safety related applications

International Nuclear Information System (INIS)

Das, R.K.; Hajos, L.G.

1993-01-01

The number of nuclear safety related equipment suppliers and the availability of spare and replacement parts designed specifically for nuclear safety related application are shrinking rapidly. These have made it necessary for utilities to apply commercial grade spare and replacement parts in nuclear safety related applications after implementing proper acceptance and dedication process to verify that such items conform with the requirements of their use in nuclear safety related application. The general guidelines for the commercial grade item (CGI) acceptance and dedication are provided in US Nuclear Regulatory Commission (NRC) Generic Letters and Electric Power Research Institute (EPRI) Report NP-5652, Guideline for the Utilization of Commercial Grade Items in Nuclear Safety Related Applications. This paper presents an application of these generic guidelines for procurement, acceptance, and dedication of a commercial grade generator for use as a standby generator at Salem Generating Station Units 1 and 2. The paper identifies the critical characteristics of the generator which once verified, will provide reasonable assurance that the generator will perform its intended safety function. The paper also delineates the method of verification of the critical characteristics through tests and provide acceptance criteria for the test results. The methodology presented in this paper may be used as specific guidelines for reliable and cost effective procurement and dedication of commercial grade generators for use as standby generators at nuclear power plants

Fast reactor safety program. Progress report, January-March 1980

International Nuclear Information System (INIS)

1980-05-01

The goal of the DOE LMFBR Safety Program is to provide a technology base fully responsive to safety considerations in the design, evaluation, licensing, and economic optimization of LMFBRs for electrical power generation. A strategy is presented that divides safety technology development into seven program elements, which have been used as the basis for the Work Breakdown Structure (WBS) for the Program. These elements include four lines of assurance (LOAs) involving core-related safety considerations, an element supporting non-core-related plant safety considerations, a safety R and D integration element, and an element for the development of test facilities and equipment to be used in Program experiments: LOA-1 (prevent accidents); LOA-2 (limit core damage); LOA-3 (maintain containment integrity); LOA-4 (attenuate radiological consequences); plant considerations; R and D integration; and facility development

Safety-related occurrences at the Finnish nuclear power plants

International Nuclear Information System (INIS)

Viitasaari, O.; Rantavaara, A.

1984-03-01

This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Report for this period, Operation of Finnish Nuclear Power Plants (STL-B-RTO-83/7), which is supplemented by this report intended principally for experts. (author)

Design-Load Basis for LANL Structures, Systems, and Components

Energy Technology Data Exchange (ETDEWEB)

I. Cuesta

2004-09-01

This document supports the recommendations in the Los Alamos National Laboratory (LANL) Engineering Standard Manual (ESM), Chapter 5--Structural providing the basis for the loads, analysis procedures, and codes to be used in the ESM. It also provides the justification for eliminating the loads to be considered in design, and evidence that the design basis loads are appropriate and consistent with the graded approach required by the Department of Energy (DOE) Code of Federal Regulation Nuclear Safety Management, 10, Part 830. This document focuses on (1) the primary and secondary natural phenomena hazards listed in DOE-G-420.1-2, Appendix C, (2) additional loads not related to natural phenomena hazards, and (3) the design loads on structures during construction.

A prioritization of generic safety issues

International Nuclear Information System (INIS)

Emrit, R.; Riggs, R.; Milstead, W.; Pittman, J.

1991-07-01

This report presents the priority rankings for generic safety issues and related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The report focuses on the prioritization of generic safety issues. Issues primarily concerned with the licensing process or environmental protection and not directly related to safety have been excluded from prioritization. The prioritized issues include: TMI Action Plan items under development; previously proposed issues covered by Task Action Plans, except issues designated as Un-resolved Safety Issues (USIs) which had already been assigned high priority; and newly-proposed issues. Future supplements to this report will include the prioritization of additional issues. The safety priority rankings are High, Medium, Low, and Drop and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolutions of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative. 1310 refs

Risk-based evaluation tool for safety-related maintenance involving scaffolding

International Nuclear Information System (INIS)

Stevens, C.; Azizi, M.; Massman, M.

1988-01-01

The US Nuclear Regulatory Commission (NRC) has expressed a general concern that transient materials in and around safety systems at nuclear power plants represent a seismic safety hazard to the plant, in particular, the uncontrolled use of scaffolding during maintenance activities. Currently, most plants perform a seismic safety analysis for all uses of scaffolding near safety-related equipment to determine appropriate tie-down locations, scaffolding reinforcements, etc. This is both time-consuming and, for the most part, unnecessary. A workable engineering solution based on risk analysis techniques has been developed and is being used at the Palo Verde nuclear generating station (PVNGS)

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

CROWE, R.D.; PIEPHO, M.G.

2000-01-01

This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

CROWE, R.D.

1999-01-01

This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

Mixing of incompatible materials in waste tanks technical basis document

International Nuclear Information System (INIS)

SANDGREN, K.R.

2003-01-01

This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA) and describes the risk binning process, the technical basis for assigning risk bins, and the controls selected for the mixing of incompatible materials representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSCs) and/or technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the FR-equency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report

Preventive Effects of Safety Helmets on Traumatic Brain Injury after Work-Related Falls

Directory of Open Access Journals (Sweden)

Sang Chul Kim

2016-10-01

Full Text Available Introduction: Work-related traumatic brain injury (TBI caused by falls is a catastrophic event that leads to disabilities and high socio-medical costs. This study aimed to measure the magnitude of the preventive effect of safety helmets on clinical outcomes and to compare the effect across different heights of fall. Methods: We collected a nationwide, prospective database of work-related injury patients who visited the 10 emergency departments between July 2010 and October 2012. All of the adult patients who experienced work-related fall injuries were eligible, excluding cases with unknown safety helmet use and height of fall. Primary and secondary endpoints were intracranial injury and in-hospital mortality. We calculated adjusted odds ratios (AORs of safety helmet use and height of fall for study outcomes, and adjusted for any potential confounders. Results: A total of 1298 patients who suffered from work-related fall injuries were enrolled. The industrial or construction area was the most common place of fall injury occurrence, and 45.0% were wearing safety helmets at the time of fall injuries. The safety helmet group was less likely to have intracranial injury comparing with the no safety helmet group (the adjusted odds ratios (ORs (95% confidence interval (CI: 0.42 (0.24–0.73, however, there was no statistical difference of in-hospital mortality between two groups (the adjusted ORs (95% CI: 0.83 (0.34–2.03. In the interaction analysis, preventive effects of safety helmet on intracranial injury were significant within 4 m height of fall. Conclusions: A safety helmet is associated with prevention of intracranial injury resulting from work-related fall and the effect is preserved within 4 m height of fall. Therefore, wearing a safety helmet can be an intervention for protecting fall-related intracranial injury in the workplace.

Some Subjects and Relations According to the Act about Safety at Work

OpenAIRE

Marino Đ. Učur

2015-01-01

Complex relations in the field of safety at work could not be present without the subjects which have a specific status and specific rights, obligations and responsibilities regulated by the Occupational Health and Safety Act. This paper deals with: employer’s designated employee for the implementation of occupational health and safety activities, employees’ elected representative for health and safety protection at work, occupational medicine specialist, occupational health and safety sp...

Work stress and patient safety: observer-rated work stressors as predictors of characteristics of safety-related events reported by young nurses.

Science.gov (United States)

Elfering, A; Semmer, N K; Grebner, S

This study investigates the link between workplace stress and the 'non-singularity' of patient safety-related incidents in the hospital setting. Over a period of 2 working weeks 23 young nurses from 19 hospitals in Switzerland documented 314 daily stressful events using a self-observation method (pocket diaries); 62 events were related to patient safety. Familiarity of safety-related events and probability of recurrence, as indicators of non-singularity, were the dependent variables in multilevel regression analyses. Predictor variables were both situational (self-reported situational control, safety compliance) and chronic variables (job stressors such as time pressure, or concentration demands and job control). Chronic work characteristics were rated by trained observers. The most frequent safety-related stressful events included incomplete or incorrect documentation (40.3%), medication errors (near misses 21%), delays in delivery of patient care (9.7%), and violent patients (9.7%). Familiarity of events and probability of recurrence were significantly predicted by chronic job stressors and low job control in multilevel regression analyses. Job stressors and low job control were shown to be risk factors for patient safety. The results suggest that job redesign to enhance job control and decrease job stressors may be an important intervention to increase patient safety.

Commercial-grade motors in safety-related applications: Final report

International Nuclear Information System (INIS)

Holzman, P.M.

1988-04-01

The objective of this project was to discuss the process necessary to utilize commercial grade equipment in safety related applications and to provide utilities with guidance for accepting commercial grade motors for safety-related applications. The generic commercial-grade concepts presented in this report can be successfully applied to motors. Commercial grade item utilization has the greatest applicability to motors in ''mild'' environments, because these motors are essentially similar to commercial grade motors in materials, construction methods, and capabilities. The acceptance process is less applicable to motors that are subject to ''harsh'' environments during postulated accidents, because of the unique design features and testing required to qualify these motors

Development of FPGA-based safety-related I and C systems

Energy Technology Data Exchange (ETDEWEB)

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

2006-07-01

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

Safety Needs Mediate Stressful Events Induced Mental Disorders

Science.gov (United States)

Gu, Simeng; Lei, Yu; Lu, Shanshan

2016-01-01

“Safety first,” we say these words almost every day, but we all take this for granted for what Maslow proposed in his famous theory of Hierarchy of Needs: safety needs come second to physiological needs. Here we propose that safety needs come before physiological needs. Safety needs are personal security, financial security, and health and well-being, which are more fundamental than physiological needs. Safety worrying is the major reason for mental disorders, such as anxiety, phobia, depression, and PTSD. The neural basis for safety is amygdala, LC/NE system, and corticotrophin-releasing hormone system, which can be regarded as a “safety circuitry,” whose major behavior function is “fight or flight” and “fear and anger” emotions. This is similar to the Appraisal theory for emotions: fear is due to the primary appraisal, which is related to safety of individual, while anger is due to secondary appraisal, which is related to coping with the unsafe situations. If coping is good, the individual will be happy; if coping failed, the individual will be sad or depressed. PMID:27738527

Safety Needs Mediate Stressful Events Induced Mental Disorders.

Science.gov (United States)

Zheng, Zheng; Gu, Simeng; Lei, Yu; Lu, Shanshan; Wang, Wei; Li, Yang; Wang, Fushun

2016-01-01

"Safety first," we say these words almost every day, but we all take this for granted for what Maslow proposed in his famous theory of Hierarchy of Needs : safety needs come second to physiological needs. Here we propose that safety needs come before physiological needs. Safety needs are personal security, financial security, and health and well-being, which are more fundamental than physiological needs. Safety worrying is the major reason for mental disorders, such as anxiety, phobia, depression, and PTSD. The neural basis for safety is amygdala, LC/NE system, and corticotrophin-releasing hormone system, which can be regarded as a "safety circuitry," whose major behavior function is "fight or flight" and "fear and anger" emotions. This is similar to the Appraisal theory for emotions: fear is due to the primary appraisal, which is related to safety of individual, while anger is due to secondary appraisal, which is related to coping with the unsafe situations. If coping is good, the individual will be happy; if coping failed, the individual will be sad or depressed.

Safety Needs Mediate Stressful Events Induced Mental Disorders

Directory of Open Access Journals (Sweden)

Zheng Zheng

2016-01-01

Full Text Available “Safety first,” we say these words almost every day, but we all take this for granted for what Maslow proposed in his famous theory of Hierarchy of Needs: safety needs come second to physiological needs. Here we propose that safety needs come before physiological needs. Safety needs are personal security, financial security, and health and well-being, which are more fundamental than physiological needs. Safety worrying is the major reason for mental disorders, such as anxiety, phobia, depression, and PTSD. The neural basis for safety is amygdala, LC/NE system, and corticotrophin-releasing hormone system, which can be regarded as a “safety circuitry,” whose major behavior function is “fight or flight” and “fear and anger” emotions. This is similar to the Appraisal theory for emotions: fear is due to the primary appraisal, which is related to safety of individual, while anger is due to secondary appraisal, which is related to coping with the unsafe situations. If coping is good, the individual will be happy; if coping failed, the individual will be sad or depressed.

Safety balance: Analysis of safety systems

International Nuclear Information System (INIS)

Delage, M.; Giroux, C.

1990-12-01

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

Determination of Design Basis Earthquake ground motion

Energy Technology Data Exchange (ETDEWEB)

Kato, Muneaki [Japan Atomic Power Co., Tokyo (Japan)

1997-03-01

This paper describes principle of determining of Design Basis Earthquake following the Examination Guide, some examples on actual sites including earthquake sources to be considered, earthquake response spectrum and simulated seismic waves. In sppendix of this paper, furthermore, seismic safety review for N.P.P designed before publication of the Examination Guide was summarized with Check Basis Earthquake. (J.P.N.)

Determination of Design Basis Earthquake ground motion

International Nuclear Information System (INIS)

Kato, Muneaki

1997-01-01

This paper describes principle of determining of Design Basis Earthquake following the Examination Guide, some examples on actual sites including earthquake sources to be considered, earthquake response spectrum and simulated seismic waves. In sppendix of this paper, furthermore, seismic safety review for N.P.P designed before publication of the Examination Guide was summarized with Check Basis Earthquake. (J.P.N.)

FLAMMABLE GAS TECHNICAL BASIS DOCUMENT

Energy Technology Data Exchange (ETDEWEB)

KRIPPS, L.J.

2005-02-18

This document describes the qualitative evaluation of frequency and consequences for double shell tank (DST) and single shell tank (SST) representative flammable gas accidents and associated hazardous conditions without controls. The evaluation indicated that safety-significant SSCs and/or TSRS were required to prevent or mitigate flammable gas accidents. Discussion on the resulting control decisions is included. This technical basis document was developed to support of the Tank Farms Documented Safety Analysis (DSA) and describes the risk binning process for the flammable gas representative accidents and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the event frequency and consequence.

RELEASE OF DRIED RADIOACTIVE WASTE MATERIALS TECHNICAL BASIS DOCUMENT

International Nuclear Information System (INIS)

KOZLOWSKI, S.D.

2007-01-01

This technical basis document was developed to support RPP-23429, Preliminary Documented Safety Analysis for the Demonstration Bulk Vitrification System (PDSA) and RPP-23479, Preliminary Documented Safety Analysis for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Facility. The main document describes the risk binning process and the technical basis for assigning risk bins to the representative accidents involving the release of dried radioactive waste materials from the Demonstration Bulk Vitrification System (DBVS) and to the associated represented hazardous conditions. Appendices D through F provide the technical basis for assigning risk bins to the representative dried waste release accident and associated represented hazardous conditions for the Contact-Handled Transuranic Mixed (CH-TRUM) Waste Packaging Unit (WPU). The risk binning process uses an evaluation of the frequency and consequence of a given representative accident or represented hazardous condition to determine the need for safety structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls. A representative accident or a represented hazardous condition is assigned to a risk bin based on the potential radiological and toxicological consequences to the public and the collocated worker. Note that the risk binning process is not applied to facility workers because credible hazardous conditions with the potential for significant facility worker consequences are considered for safety-significant SSCs and/or TSR-level controls regardless of their estimated frequency. The controls for protection of the facility workers are described in RPP-23429 and RPP-23479. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses, as described below

Fundamental safety principles. Safety fundamentals

International Nuclear Information System (INIS)

2007-01-01

This publication states the fundamental safety objective and ten associated safety principles, and briefly describes their intent and purpose. The fundamental safety objective - to protect people and the environment from harmful effects of ionizing radiation - applies to all circumstances that give rise to radiation risks. The safety principles are applicable, as relevant, throughout the entire lifetime of all facilities and activities - existing and new - utilized for peaceful purposes, and to protective actions to reduce existing radiation risks. They provide the basis for requirements and measures for the protection of people and the environment against radiation risks and for the safety of facilities and activities that give rise to radiation risks, including, in particular, nuclear installations and uses of radiation and radioactive sources, the transport of radioactive material and the management of radioactive waste

Fundamental safety principles. Safety fundamentals

International Nuclear Information System (INIS)

2006-01-01

This publication states the fundamental safety objective and ten associated safety principles, and briefly describes their intent and purpose. The fundamental safety objective - to protect people and the environment from harmful effects of ionizing radiation - applies to all circumstances that give rise to radiation risks. The safety principles are applicable, as relevant, throughout the entire lifetime of all facilities and activities - existing and new - utilized for peaceful purposes, and to protective actions to reduce existing radiation risks. They provide the basis for requirements and measures for the protection of people and the environment against radiation risks and for the safety of facilities and activities that give rise to radiation risks, including, in particular, nuclear installations and uses of radiation and radioactive sources, the transport of radioactive material and the management of radioactive waste

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.

International views on nuclear safety

International Nuclear Information System (INIS)

Birkhofer, A.

2002-01-01

Safety has always been an important objective in nuclear technology. Starting with a set of sound physical principles and prudent design approaches, safety concepts have gradually been refined and cover now a wide range of provisions related to design, quality and operation. Research, the evaluation of operating experiences and probabilistic risk assessments constitute an essential basis and international co-operation plays a significant role in that context. Concerning future developments a major objective for new reactor concepts, such as the EPR, is to practically exclude a severe core damage accident with large scale consequences outside the plant. (author)

Performance Monitoring for Nuclear Safety Related Instrumentation at PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2015-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on performance monitoring for nuclear safety related instrumentation in TRIGA PUSPATI Reactor (RTP) of based on various parameter of reactor safety instrument channel such as log power, linear power, Fuel temperature, coolant temperature will take into consideration. Methodology of performance on estimation and monitoring is to evaluate and analysis of reactor parameters which is important of reactor safety and control. And also to estimate power measurement, differential of log and linear power and fuel temperature during reactor start-up, operation and shutdown .This study also focus on neutron power fluctuation from fission chamber during reactor start-up and operation. This work will present result of performance monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that performance of nuclear safety related instrumentation will improved the reactor control and safety parameter during reactor start-up, operation and shutdown. (author)

Range and limits of application of Sec.12, Atomic Energy Act, as a legal basis of the nuclear plant safety ordinance

International Nuclear Information System (INIS)

Schmidt-Preuss, Matthias

2009-01-01

Ensuring plant safety is a key purpose of nuclear law. Sec.7 II No.3, Atomic Energy Act, is considered the basic norm of nuclear legislation. The main requirement this embodies is ensuring 'the provisions against damage arising from construction and operation of a plant as required in accordance with the state of the art'. These normative requirements constitute the strictest yardstick existing in legislation about technology. Putting it into effect has always been the purpose of the set of nuclear rules and regulations constituting the next lower level of legislation, which so far have developed by evolution and are now to be updated comprehensively in the format of so-called modules as provided for in the concept of the Federal Ministry for the Environment, Nature Conservation, and Nuclear Safety (BMU). So far, there has not been a nuclear plant safety ordinance. The Atomic Energy Act has always provided a basis for adopting such an ordinance, especially so in Sec.12 I 1 No.1, Atomic Energy Act. No federal government has so far wanted to make use of it. This makes it all the more remarkable that the BMU took up the subject of a nuclear plant safety ordinance as early as in 2006, starting a dialog with the federal states. This dialog meanwhile has come to a halt. The subject seems to be dormant right now, but certainly has not been shelved. Ensuring plant safety is a key purpose of nuclear law. Sec.7 II No.3, Atomic Energy Act, is considered the basic norm of nuclear legislation. The main requirement this embodies is ensuring 'the provisions against damage arising from construction and operation of a plant as required in accordance with the state of the art'. These normative requirements constitute the strictest yardstick existing in legislation about technology. Putting it into effect has always been the purpose of the set of nuclear rules and regulations constituting the next lower level of legislation, which so far have developed by evolution and are now to be

Safety-related LWR research. Annual report 1989

International Nuclear Information System (INIS)

1990-11-01

The main topics in this annual report 1989 are phenomena of heavy fuel damage and single aspects of a core meltdown accident. The examined single aspects refer to aerosol behavior and filter engineering and to methods for assessment and minimization of the radiological consequences of reactor accidents. Different contributions to selected, safety-related problems of an advanced pressurized-water reactor complete the topic spectrum. The annual report 1989 describes the progress of the research work wich was carried out in the area of safety research by institutes and departments of the KfK, and on behalf of the KfK by external institutions. The individual contributions represent the status of work at the end of the year under review, 1989. (orig./HP) [de

White paper on nuclear safety in 1993. 1993 ed.

International Nuclear Information System (INIS)

1994-01-01

When the development and utilization of nuclear power are advanced, the securing of safety is the prerequisite. The Nuclear Safety Commission has promoted the various important measures which become the basis for securing the safety, such as the execution of strict safety examination, the preparation and perfection of the guidelines which are used for safety examination, the holding of public hearings, the investigation of failures and troubles, and the reflection of the obtained learnings to the countermeasures for securing the safety. Since nuclear power facilities contain radioactive substances, the prevention of their abnormal release to surrounding environment is the fundamental of securing the safety. The policy of coping with severe accidents was determined after Three Mile Island accident and Chernobyl accident, and the concrete investigation is carried out. In this year, the abandonment of radioactive waste in ocean by Russia was a large problem. In this book, the Nuclear Safety Commission in the last one year and the securing of safety related to the utilization of plutonium for power reactors are reported. Various related materials are attached. (K.I.)

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)

1E Qualification of Electrical Equipment - Requirement for Safety Nuclear Power Plants

International Nuclear Information System (INIS)

Geambasu, C.; Segarceanu, D.; Albu, J.

2002-01-01

The paper presents the qualification methods of the safety related equipment according to the safety class 1E. There are presented the qualification principles, procedure and documents, emphasis being laid on the qualification approach by type tests. This approach assumes the equipment test under both normal and accident conditions (design basis events) simulating the operational conditions and covers the largest part of electrical equipment from a nuclear power plant.The safety related equipment is to be qualified is subjected to a sequential test that will be detailed in the paper. (author)

Problems of nuclear energetics safety in the Soviet Union

International Nuclear Information System (INIS)

Kovalevitsh, O.M.

1991-01-01

Authors describe present state of Soviet nuclear energy. They don't cover problems relative to its development and that reasons made so bleak picture of this economic branch. They pay particular attention to low level of nuclear safety in nuclear power plants. The improvement of this situation they see in enacting of atomic low, as quickly as possible, which will make a basis of safety development in nuclear industry

Safety surrogate histograms (SSH): A novel real-time safety assessment of dilemma zone related conflicts at signalized intersections.

Science.gov (United States)

Ghanipoor Machiani, Sahar; Abbas, Montasir

2016-11-01

Drivers' indecisiveness in dilemma zones (DZ) could result in crash-prone situations at signalized intersections. DZ is to the area ahead of an intersection in which drivers encounter a dilemma regarding whether to stop or proceed through the intersection when the signal turns yellow. An improper decision to stop by the leading driver, combined with the following driver deciding to go, can result in a rear-end collision, unless the following driver recognizes a collision is imminent and adjusts his or her behavior at or shortly after the onset of yellow. Considering the significance of DZ-related crashes, a comprehensive safety measure is needed to characterize the level of safety at signalized intersections. In this study, a novel safety surrogate measure was developed utilizing real-time radar field data. This new measure, called safety surrogate histogram (SSH), captures the degree and frequency of DZ-related conflicts at each intersection approach. SSH includes detailed information regarding the possibility of crashes, because it is calculated based on the vehicles conflicts. An example illustrating the application of the new methodology at two study sites in Virginia is presented and discussed, and a comparison is provided between SSH and other DZ-related safety surrogate measures mentioned in the literature. The results of the study reveal the efficacy of the SSH as complementary to existing surrogate measures. Copyright © 2015 Elsevier Ltd. All rights reserved.

Fast reactor safety: proceedings of the international topical meeting. Volume 2

International Nuclear Information System (INIS)

1985-07-01

The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 2 include: safety design concepts; operational transient experiments; analysis of seismic and external events; HCDA-related codes, analysis, and experiments; sodium fires; instrumentation and control/PPS design; whole-core accident analysis codes; and impact of safety design considerations on future LMFBR developments

Aviation Trends Related to Atmospheric Environment Safety Technologies Project Technical Challenges

Science.gov (United States)

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

2014-01-01

Current and future aviation safety trends related to the National Aeronautics and Space Administration's Atmospheric Environment Safety Technologies Project's three technical challenges (engine icing characterization and simulation capability; airframe icing simulation and engineering tool capability; and atmospheric hazard sensing and mitigation technology capability) were assessed by examining the National Transportation Safety Board (NTSB) accident database (1989 to 2008), incidents from the Federal Aviation Administration (FAA) accident/incident database (1989 to 2006), and literature from various industry and government sources. The accident and incident data were examined for events involving fixed-wing airplanes operating under Federal Aviation Regulation (FAR) Parts 121, 135, and 91 for atmospheric conditions related to airframe icing, ice-crystal engine icing, turbulence, clear air turbulence, wake vortex, lightning, and low visibility (fog, low ceiling, clouds, precipitation, and low lighting). Five future aviation safety risk areas associated with the three AEST technical challenges were identified after an exhaustive survey of a variety of sources and include: approach and landing accident reduction, icing/ice detection, loss of control in flight, super density operations, and runway safety.

CONCEPTUALIZATION OF IDEAS OF PSYCHOLOGICAL SAFETY IN SPORTS: PROBLEMS OF EXPERIMENTAL RESEARCH

Directory of Open Access Journals (Sweden)

Yulia Vladimirovna Vardanyan

2013-09-01

Full Text Available This article is devoted to the research of the concept “psychological safety in sports”. On the basis of analysis of ideas about psychological safety in sports and their representation in printed or verbal form the necessity of overcoming the fragmentation and lack of system is substantiated. The authors state that one and the same sports situation can constructively or destructively affect the psychological safety of direct or indirect participants of sports events. In this context, it is important to create the psycholinguistic basis of experimental research of psychological safety in sports. Great attention is paid to systematization of the content of the concept “psychological safety in sports”. The created models of words and expressions that convey ideas about this phenomenon are of particular value. In the structure of the concept the dominant meanings, expressed in the nucleus, and additional meanings, related to the periphery of the concept are distinguished.Purpose: to explore the ideas of psychological safety in sports and their representation in printed or verbal form; to determine ways of overcoming the conceptual psycholinguistic problems in the process of experimental research of psychological safety in sports; to create the model of words and expressions which are used to verbalize the concept “psychological safety in sports”.Methodology: theoretical analysis of psychological and linguistic literature, creation of the psycholinguistic basis of experimental research, modeling of the conceptual ideas of psychological safety in sports.Results: psycholinguistic basis of experimental research of psychological safety in sports, the model of content and structure of the corresponding concept.Practical implications: Pedagogical Psychology, Sports Psychology, Philology, Psycholinguistics.DOI: http://dx.doi.org/10.12731/2218-7405-2013-8-11

Transportation Safety Excellence in Operations Through Improved Transportation Safety Document

International Nuclear Information System (INIS)

Dr. Michael A. Lehto; MAL

2007-01-01

A recent accomplishment of the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) Nuclear Safety analysis group was to obtain DOE-ID approval for the inter-facility transfer of greater-than-Hazard-Category-3 quantity radioactive/fissionable waste in Department of Transportation (DOT) Type A drums at MFC. This accomplishment supported excellence in operations through safety analysis by better integrating nuclear safety requirements with waste requirements in the Transportation Safety Document (TSD); reducing container and transport costs; and making facility operations more efficient. The MFC TSD governs and controls the inter-facility transfer of greater-than-Hazard-Category-3 radioactive and/or fissionable materials in non-DOT approved containers. Previously, the TSD did not include the capability to transfer payloads of greater-than-Hazard-Category-3 radioactive and/or fissionable materials using DOT Type A drums. Previous practice was to package the waste materials to less-than-Hazard-Category-3 quantities when loading DOT Type A drums for transfer out of facilities to reduce facility waste accumulations. This practice allowed operations to proceed, but resulted in drums being loaded to less than the Waste Isolation Pilot Plant (WIPP) waste acceptance criteria (WAC) waste limits, which was not cost effective or operations friendly. An improved and revised safety analysis was used to gain DOE-ID approval for adding this container configuration to the MFC TSD safety basis. In the process of obtaining approval of the revised safety basis, safety analysis practices were used effectively to directly support excellence in operations. Several factors contributed to the success of MFC's effort to obtain approval for the use of DOT Type A drums, including two practices that could help in future safety basis changes at other facilities. (1) The process of incorporating the DOT Type A drums into the TSD at MFC helped to better integrate nuclear 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

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

Energy Technology Data Exchange (ETDEWEB)

CROWE, R.D.

1999-09-09

This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

Integrated Safety Culture Model and Application

Institute of Scientific and Technical Information of China (English)

汪磊; 孙瑞山; 刘汉辉

2009-01-01

A new safety culture model is constructed and is applied to analyze the correlations between safety culture and SMS. On the basis of previous typical definitions, models and theories of safety culture, an in-depth analysis on safety culture's structure, composing elements and their correlations was conducted. A new definition of safety culture was proposed from the perspective of sub-cuhure. 7 types of safety sub-culture, which are safety priority culture, standardizing culture, flexible culture, learning culture, teamwork culture, reporting culture and justice culture were defined later. Then integrated safety culture model (ISCM) was put forward based on the definition. The model divided safety culture into intrinsic latency level and extrinsic indication level and explained the potential relationship between safety sub-culture and all safety culture dimensions. Finally in the analyzing of safety culture and SMS, it concluded that positive safety culture is the basis of im-plementing SMS effectively and an advanced SMS will improve safety culture from all around.

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

Fast reactor safety: proceedings of the international topical meeting. Volume 1

Energy Technology Data Exchange (ETDEWEB)

1985-07-01

The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 1 include: impact of safety and licensing considerations on fast reactor design; safety aspects of innovative designs; intra-subassembly behavior; operational safety; design accommodation of seismic and other external events; natural circulation; safety design concepts; safety implications derived from operational plant data; decay heat removal; and assessment of HCDA consequences.

Fast reactor safety: proceedings of the international topical meeting. Volume 1

International Nuclear Information System (INIS)

1985-07-01

The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 1 include: impact of safety and licensing considerations on fast reactor design; safety aspects of innovative designs; intra-subassembly behavior; operational safety; design accommodation of seismic and other external events; natural circulation; safety design concepts; safety implications derived from operational plant data; decay heat removal; and assessment of HCDA consequences

Cost-benefit evaluation of containment related engineered safety features of Indian pressurized heavy water reactors

International Nuclear Information System (INIS)

Bajaj, S.S.; Bhawal, R.N.; Rustagi, R.S.

1984-01-01

The typical containment system for a commercial nuclear reactor uses several engineered safety features to achieve its objective of limiting the release of radioactive fission products to the environment in the event of postulated accident conditions. The design of containment systems and associated features for Indian Pressurized Heavy Water Reactors (PHWRs) has undergone progressive improvement in successive projects. In particular, the current design adopted for the Narora Atomic Power Project (NAPP) has seen several notable improvements. The paper reports on a cost-benefit study in respect of three containment related engineered safety features and subsystems of NAPP, viz. (i) secondary containment envelope, (ii) primary containment filtration and pump-back system, and (iii) secondary containment filtration, recirculation and purge system. The effect of each of these systems in reducing the environmental releases of radioactivity following a design basis accident is presented. The corresponding reduction in population exposure and the associated monetary value of this reduction in exposure are also given. The costs of the features and subsystem under consideration are then compared with the monetary value of the exposures saved, as well as other non-quantified benefits, to arrive at conclusions regarding the usefulness of each subsystem. This study clearly establishes for the secondary containment envelope the benefit in terms of reduction in public exposure giving a quantitative justification for the costs involved. In the case of the other two subsystems, which involve relatively low costs, while all benefits have not been quantified, their desirability is justified on qualitative considerations. It is concluded that the engineered safety features adopted in the current containment system design of Indian PHWRs contribute to reducing radiation exposures during accident conditions in accordance with the ALARA ('as low as reasonably achievable') principle

Code on the safety of nuclear research reactors: Design

International Nuclear Information System (INIS)

1992-01-01

The main objective of this publication is to provide a safety basis for the design of a research reactor and for the assessment of the design. Another objective is to cover certain aspects related to regulatory supervision, siting and quality assurance, as far as these are related to activities for the design of a research reactor. These objectives are expressed in terms of requirements and recommendations for the design of research reactors. Emphasis is placed on the safety requirements that shall be met rather than on ways in which they can be met. The requirements and recommendations may form the foundation necessary for a Member State to develop specific regulations and safety criteria for its research reactor programme.

78 FR 25488 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

Science.gov (United States)

2013-05-01

... Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for... regulatory guide (DG), DG-1235, ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants... entitled ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants'' is temporarily...

Application of quality assurance program to safety related aging equipment or components

International Nuclear Information System (INIS)

Papaiya, N.C.

1990-01-01

This paper addresses how quality assurance programs and their criteria are applied to safety related and aging equipment or components used in commercial nuclear plant applications. The QA Programs referred to are 10CFR50 Appendix B and EPRI NP-5652. The QA programs as applicable are applied to equipment/component aging qualification, preventive maintenance, surveillance testing and procurement engineering. The intent of this paper is not the technical issues, methods and research of aging. The paper addresses QA program's application to age-related equipment or components in safety related applications. Quality Assurance Program 10CFR50 Appendix B applies to all safety related aging components or equipment related to the qualification program and associated preventive maintenance and surveillance testing programs. Quality Assurance involvement with procurement engineering for age-related commercial grade items supports EPRI NP-5652 and assures that the dedicated OGI is equal to the item purchased as a basic component to 10CFR50 Appendix B requirements

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.

Session 1 theme: Various forms of design basis knowledge and effects of its loss on Safety. Views from EDF

International Nuclear Information System (INIS)

Servière, Georges

2013-01-01

Design basis knowledge - What happens or may happen and corresponding required knowledge: • Unexpected events or failures of equipment; • Spare part issues (no longer availaible,…); • Change in applicable regulations / requirements; • Change of operating conditions; • Change of plant performances; • Evolution of external environment and conditions; • Events and accidents on other plants, worldwide; • New knowledge availaible; • Periodic safety reviews and upgrades; • Extension of plant operation life; • Decommissioning and dismantling; • Some of those you may choose not to do, but most of them have to be faced and need appropriate knowledge

PHWR safety: design, siting and construction

International Nuclear Information System (INIS)

Sharma, V.K.

2002-01-01

In all activities associated with NPPs viz. siting, design, construction, commissioning and operation, safety is given overriding importance. The safety design principles of PHWRs are based on defence-in-depth approach, physical and functional separation between process and safety systems and also among various safety systems, redundancy to meet single failure criteria and postulation of a number of design basis events for which the plant must be designed. Apart from engineered safety systems, PHWRs have inherent characteristics which contribute to safety. In siting of a NPP, it is required to ensure that the given site does not pose undue radiological hazard to public and the environment both during normal operation as well as during and following an accident condition. For this purpose, all site related external events, both natural and man induced, are assessed for their effect on the plant and are considered as part of the design basis. Possible radiological impact of the NPP on environment and surrounding population is assessed and ensured to be within acceptable limits. During construction phase, it is essential that the NPP be built in accordance with design intent and with required quality of workmanship to ensure that the NPP will remain safe during all states of operation. This is achieved through careful execution and QA activities encompassing all aspects of component fabrication at manufacturer works, civil construction, site erection, assembly, and commissioning. Future trends in nuclear safety will continue to be based on existing principles which have proved to be sound. These will be further strengthened by features such as increasing use of passive means of performing safety functions and a more explicit treatment of severe accidents. (author)

Confusion in practice: on nuclear safety responsibility subject of our nation

International Nuclear Information System (INIS)

Wang Jia

2014-01-01

Nuclear safety responsibility subject seems a unquestionable issue, but when I took part in the CNNC searching team of 'nuclear law legislation', I found that there are confusions on understanding of this concept and in application. The paper focuses on the content of nuclear safety responsibility, using legal and practical method to dig out the differences with the related and frequently confusing concepts, on which basis to analyze the situation of nuclear safety responsibility subject of our nation. In conclusion, I give suggestions on who shall be the nuclear safety responsibility subject. (author)

Geo scientific basis for making the safety case for a SF/HL W/IL W repository in Opalinus clay in ne Switzerland (project Entsorgungsnachweis) 1: overview and main conclusions

International Nuclear Information System (INIS)

Gautschi, A.; Lambert, A.; Zuidema, P.

2004-01-01

This paper provides an overview of the geo-scientific basis and the main conclusions concerning the safety case for Project Entsorgungsnachweis (Nagra, 2002a, see first paper). The key geo-scientific input for the safety case is summarised in the following three papers. The data and arguments are discussed in great detail in Nagra (2002b) and in numerous reference reports cited therein. (author)

Evaluating North Carolina Food Pantry Food Safety-Related Operating Procedures.

Science.gov (United States)

Chaifetz, Ashley; Chapman, Benjamin

2015-11-01

Almost one in seven American households were food insecure in 2012, experiencing difficulty in providing enough food for all family members due to a lack of resources. Food pantries assist a food-insecure population through emergency food provision, but there is a paucity of information on the food safety-related operating procedures used in the pantries. Food pantries operate in a variable regulatory landscape; in some jurisdictions, they are treated equivalent to restaurants, while in others, they operate outside of inspection regimes. By using a mixed methods approach to catalog the standard operating procedures related to food in 105 food pantries from 12 North Carolina counties, we evaluated their potential impact on food safety. Data collected through interviews with pantry managers were supplemented with observed food safety practices scored against a modified version of the North Carolina Food Establishment Inspection Report. Pantries partnered with organized food bank networks were compared with those that operated independently. In this exploratory research, additional comparisons were examined for pantries in metropolitan areas versus nonmetropolitan areas and pantries with managers who had received food safety training versus managers who had not. The results provide a snapshot of how North Carolina food pantries operate and document risk mitigation strategies for foodborne illness for the vulnerable populations they serve. Data analysis reveals gaps in food safety knowledge and practice, indicating that pantries would benefit from more effective food safety training, especially focusing on formalizing risk management strategies. In addition, new tools, procedures, or policy interventions might improve information actualization by food pantry personnel.

EDA activities related to safety

International Nuclear Information System (INIS)

Gordon, C.; Raeder, J.

2001-01-01

This article reviews the accomplishments in ITER safety analysis during the course of the Engineering Design Activities (EDA). The key aspects of ITER safety analysis are: effluents and emissions from normal operation, including planned maintenance activities; occupational safety for workers at the facility; radioactive materials and wastes generated during operation and from decommissioning ; potential incidents and accidents and the resulting transients. As a result of the work during the EDA it is concluded that ITER is safe

Official News relating to CERN Safety Rules

CERN Multimedia

HSE Unit

2015-01-01

The CERN Safety Rules listed below have been published on the official CERN Safety Rules website (see here).   Safety Regulation SR-WS Works and services: this SR-WS (version 1) will cancel and replace the corresponding provisions of Safety Instruction IS50 “Safety Coordination on CERN Worksites”. General Safety Instruction GSI-WS-1 Safety coordination for works and services: this GSI-WS-1 (version 1) will cancel and replace the corresponding provisions of Safety Instruction IS39 “Notice of Start of Works (AOC)” and of Safety Instruction IS50 “Safety Coordination on CERN Worksites” ​Specific Safety Instruction SSI-WS-1-1 Safety coordinator for category 1 operations: this SSI-WS-1-4 (version 1) will cancel and replace the corresponding provisions of Safety Instruction IS50 “Safety Coordination on CERN Worksites”.​ ​ In order to limit the impact on the end-of-year technical st...

High level issues in reliability quantification of safety-critical software

International Nuclear Information System (INIS)

Kim, Man Cheol

2012-01-01

For the purpose of developing a consensus method for the reliability assessment of safety-critical digital instrumentation and control systems in nuclear power plants, several high level issues in reliability assessment of the safety-critical software based on Bayesian belief network modeling and statistical testing are discussed. Related to the Bayesian belief network modeling, the relation between the assessment approach and the sources of evidence, the relation between qualitative evidence and quantitative evidence, how to consider qualitative evidence, and the cause-consequence relation are discussed. Related to the statistical testing, the need of the consideration of context-specific software failure probabilities and the inability to perform a huge number of tests in the real world are discussed. The discussions in this paper are expected to provide a common basis for future discussions on the reliability assessment of safety-critical software. (author)

Impact of pre-conditioning on the qualification of safety-related equipment

International Nuclear Information System (INIS)

Isgro, J.R.

1982-01-01

This paper shares some recent experiences on the effects of preconditioning on the qualification of safety-related equipment not located in a harsh environment. Environmental and seismic qualification testing programs were conducted following the guidelines of IEEE 323-1974, IEEE 344-1975 and appropriate IEEE daughter standards, where available. The examples that follow will illustrate the degree of pre-conditioning of safety-related equipment qualified to the requirements of IEEE-323-1974, and its effect on the outcome of the qualification program

Fast reactor safety: proceedings of the international topical meeting. Volume 2. [R

Energy Technology Data Exchange (ETDEWEB)

1985-07-01

The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 2 include: safety design concepts; operational transient experiments; analysis of seismic and external events; HCDA-related codes, analysis, and experiments; sodium fires; instrumentation and control/PPS design; whole-core accident analysis codes; and impact of safety design considerations on future LMFBR developments.

Safety issues relating to the design of fusion power facilities

International Nuclear Information System (INIS)

Stasko, R.R.; Wong, K.Y.; Russell, S.B.

1986-06-01

In order to make fusion power a viable future source of energy, it will be necessary to ensure that the cost of power for fusion electric generation is competitive with advanced fission concepts. In addition, fusion power will have to live up to its original promise of being a more radiologically benign technology than fission, and be able to demonstrate excellent operational safety performance. These two requirements are interrelated, since the selection of an appropriate safety philosophy early in the design phase could greatly reduce or eliminate the capital costs of elaborate safety related and protective sytems. This paper will briefly overview a few of the key safety issues presently recognized as critical to the ultimate achievement of licensable, environmentally safe and socially acceptable fusion power facilities. 12 refs

Development of FPGA-based safety-related instrumentation and control systems

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

Technical Details on Beyond Design Basis Event Pilot Evaluations

Energy Technology Data Exchange (ETDEWEB)

None, None

2013-01-01

The primary focus of the BDBE pilot project was the review of BDBE analysis and mitigation features at four DOE nuclear facilities representing a range of DOE sites, nuclear facility types/activities, and responsible program offices. The pilots looked at (1) how beyond design basis accidents were evaluated and documented in the facility Documented Safety Analysis, (2) potential BDBE vulnerabilities and margins to failure of facility safety features as obtained from general area and specific system walkdowns and design documents reviews, and (3) preparations made in facility and site emergency management programs to respond to severe accidents. It also evaluated whether draft BDBE guidance on safety analysis and emergency management could be used to improve the analysis of and preparations for mitigating severe and beyond design basis accidents. The details of these activities are organized in this report as described below.

Simplifying documentation while approaching site closure: integrated health and safety plans as documented safety analysis

International Nuclear Information System (INIS)

Brown, Tulanda

2003-01-01

At the Fernald Closure Project (FCP) near Cincinnati, Ohio, environmental restoration activities are supported by Documented Safety Analyses (DSAs) that combine the required project-specific Health and Safety Plans, Safety Basis Requirements (SBRs), and Process Requirements (PRs) into single Integrated Health and Safety Plans (I-HASPs). By isolating any remediation activities that deal with Enriched Restricted Materials, the SBRs and PRs assure that the hazard categories of former nuclear facilities undergoing remediation remain less than Nuclear. These integrated DSAs employ Integrated Safety Management methodology in support of simplified restoration and remediation activities that, so far, have resulted in the decontamination and demolition (D and D) of over 150 structures, including six major nuclear production plants. This paper presents the FCP method for maintaining safety basis documentation, using the D and D I-HASP as an example

Nuclear regulatory guides for LWR (PWR) fuel in Japan and some related safety research

International Nuclear Information System (INIS)

Ichikawa, M.

1994-01-01

The general aspects of licensing procedure for NPPs in Japan and regulatory guides are described. The expert committee reports closely related to PWR fuel are reviewed. Some major results of reactor safety research experiments at NSPR (Nuclear Safety Research Reactor of JAERI) used for establishment of related guide, are discussed. It is pointed out that the reactor safety research in Japan supports the regularity activities by establishing and revising guides and preparing the necessary regulatory data as well as improving nuclear safety. 10 figs., 4 refs

Nuclear regulatory guides for LWR (PWR) fuel in Japan and some related safety research

Energy Technology Data Exchange (ETDEWEB)

Ichikawa, M [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

1994-12-31

The general aspects of licensing procedure for NPPs in Japan and regulatory guides are described. The expert committee reports closely related to PWR fuel are reviewed. Some major results of reactor safety research experiments at NSPR (Nuclear Safety Research Reactor of JAERI) used for establishment of related guide, are discussed. It is pointed out that the reactor safety research in Japan supports the regularity activities by establishing and revising guides and preparing the necessary regulatory data as well as improving nuclear safety. 10 figs., 4 refs.

Safety Case Development as an Information Modelling Problem

Science.gov (United States)

Lewis, Robert

This paper considers the benefits from applying information modelling as the basis for creating an electronically-based safety case. It highlights the current difficulties of developing and managing large document-based safety cases for complex systems such as those found in Air Traffic Control systems. After a review of current tools and related literature on this subject, the paper proceeds to examine the many relationships between entities that can exist within a large safety case. The paper considers the benefits to both safety case writers and readers from the future development of an ideal safety case tool that is able to exploit these information models. The paper also introduces the idea that the safety case has formal relationships between entities that directly support the safety case argument using a methodology such as GSN, and informal relationships that provide links to direct and backing evidence and to supporting information.

Report on safety related occurrences and reactor trips July 1, 1976-December 31, 1976

International Nuclear Information System (INIS)

Andermo, L.

1977-04-01

This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1, 1976 to December 31, 1976 inclusive. The facilities involved are Oskarshamn 1 and 2, Ringhals 1 and 2 and Barsebaeck 1. During this period of the 6 months 37 safety related occurrences and 34 reactor trips have been reported to the Nuclear Power Inspectorate. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The fact that even small deviations from prescribed operation results in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The number of reactor trips are almost as low as during the last period, which is a drastic reduction compared to earlier time periods. The greatest outages are caused by occurrences without safety significance.(author)

Manual on quality assurance for computer software related to the safety of nuclear power plants

International Nuclear Information System (INIS)

1988-01-01

The objective of the Manual is to provide guidance in the assurance of quality of specification, design, maintenance and use of computer software related to items and activities important to safety (hereinafter referred to as safety related) in nuclear power plants. This guidance is consistent with, and supplements, the requirements and recommendations of Quality Assurance for Safety in Nuclear Power Plants: A Code of Practice, 50-C-QA, and related Safety Guides on quality assurance for nuclear power plants. Annex A identifies the IAEA documents referenced in the Manual. The Manual is intended to be of use to all those who, in any way, are involved with software for safety related applications for nuclear power plants, including auditors who may be called upon to audit management systems and product software. Figs

Smooth handling: the lack of safety-related consumer information in car advertisements.

Science.gov (United States)

Wilson, Nick; Maher, Anthony; Thomson, George; Keall, Michael

2007-10-01

To examine the content and trends of safety-related consumer information in magazine vehicle advertisements, as a case study within the worldwide marketing of vehicles. Content analysis of popular current affairs magazines in New Zealand for the 5-year period 2001-2005 was undertaken (n = 514 advertisements), supplemented with vehicle data from official websites. Safety information in advertisements for light passenger vehicles was relatively uncommon with only 27% mentioning one or more of nine key safety features examined (average: 1.7 out of nine features in this 27%). Also included were potentially hazardous features of: speed imagery (in 29% of advertisements), power references (14%), and acceleration data (4%). The speed and power aspects became relatively more common over the 5-year period (p advertisements and vehicle marketing - as already occurs with many other consumer products.

The use of probabilistic safety assessment based maintenance indicators to increase the availability of safety related systems in nuclear power plants

International Nuclear Information System (INIS)

Kirchsteiger, C.

1991-04-01

This work describes the theoretical development of a Probabilistic Safety Assessment (PSA) based Performance Indicator (PI) model for a comprehensive Maintenance Efficiency Analysis (MEA) and its practical application to past operational history data of a certain Nuclear Power Plant. Plant specific equipment history and maintenance work order data have been collected and analysed using various advanced statistical procedures (nonparametric methods, multivariate analysis) in order to be able to estimate safety system related equipment and maintenance process trends. The main results of such a MEA case study are the trends in the (in)effectiveness of the performance of a selected safety system and its dominant maintenance related causes of its bad (good) equipment performance. Finally, the therefrom gained results are used to propose a new set of safety system based and maintenance related Performance Indicators, including suggestions for a corresponding plant specific maintenance data collection system. (author)

Interim Safety Basis for Fuel Supply Shutdown Facility

International Nuclear Information System (INIS)

BENECKE, M.W.

2000-01-01

This ISB, in conjunction with the IOSR, provides the required basis for interim operation or restrictions on interim operations and administrative controls for the facility until a SAR is prepared in accordance with the new requirements or the facility is shut down. It is concluded that the risks associated with tha current and anticipated mode of the facility, uranium disposition, clean up, and transition activities required for permanent closure, are within risk guidelines

AEC sets five year nuclear safety research program

International Nuclear Information System (INIS)

Anon.

1976-01-01

The research by the government for the establishment of means of judging the adequacy of safety measures incorporated in nuclear facilities, including setting safety standards and collecting documents of general criteria, and the research by the industry on safety measures and the promotion of safety-related technique are stated in the five year program for 1976-80 reported by subcommittees, Atomic Energy Commission (AEC). Four considerations on the research items incorporated in the program are 1) technical programs relating to the safety of nuclear facilities and the necessary criteria, 2) priority of the relevant items decided according to their impact on circumstances, urgency, the defence-indepth concept and so on, 3) consideration of all relevant data and documents collected, and research subjects necessary to quantify safety measurement, and 4) consideration of technological actualization, the capability of each research body, the budget and the time schedule. In addition, seven major themes decided on the basis of these points are 1) reactivity-initiated accident, 2) LOCA, 3) fuel behavior, 4) structural safety, 5) radioactive release, 6) statistical method of safety evaluation, and 7) seismic characteristics. The committee has deliberated the appropriate division of researches between the government and the industry. A set of tables showing the nuclear safety research plan for 1976-80 are attached. (Iwakiri, K.)

Waste management safety

International Nuclear Information System (INIS)

Boehm, H.

1983-01-01

All studies carried out by competent authors of the safety of a waste management concept on the basis of reprocessing of the spent fuel elements and storage in the deep underground of the radioactive waste show that only a minor technical risk is involved in this step. This also holds true when evaluating the accidents which have occurred in waste management facilities. To explain the risk, first the completely different safety aspects of nuclear power plants, reprocessing plants and repositories are outlined together with the safety related characteristics of these plants. Also this comparison indicates that the risk of waste management facilities is considerably lower than the, already very small, risk of nuclear power plants. For the final storage of waste from reprocessing and for the direct storage of fuel elements, the results of safety analyses show that the radiological exposure following an accident with radioactivity releases, even under conservative assumptions, is considerably below the natural radiation exposure. The very small danger to the environment arising from waste management by reprocessing clearly indicates that aspects of technical safety alone will hardly be a major criterion for the decision in favor of one or the other waste management approach. (orig.) [de

Cold Vacuum Drying Facility Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

PIEPHO, M.G.

1999-01-01

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR

Estimating the CCSD basis-set limit energy from small basis sets: basis-set extrapolations vs additivity schemes

Energy Technology Data Exchange (ETDEWEB)

Spackman, Peter R.; Karton, Amir, E-mail: amir.karton@uwa.edu.au [School of Chemistry and Biochemistry, The University of Western Australia, Perth, WA 6009 (Australia)

2015-05-15

Coupled cluster calculations with all single and double excitations (CCSD) converge exceedingly slowly with the size of the one-particle basis set. We assess the performance of a number of approaches for obtaining CCSD correlation energies close to the complete basis-set limit in conjunction with relatively small DZ and TZ basis sets. These include global and system-dependent extrapolations based on the A + B/L{sup α} two-point extrapolation formula, and the well-known additivity approach that uses an MP2-based basis-set-correction term. We show that the basis set convergence rate can change dramatically between different systems(e.g.it is slower for molecules with polar bonds and/or second-row elements). The system-dependent basis-set extrapolation scheme, in which unique basis-set extrapolation exponents for each system are obtained from lower-cost MP2 calculations, significantly accelerates the basis-set convergence relative to the global extrapolations. Nevertheless, we find that the simple MP2-based basis-set additivity scheme outperforms the extrapolation approaches. For example, the following root-mean-squared deviations are obtained for the 140 basis-set limit CCSD atomization energies in the W4-11 database: 9.1 (global extrapolation), 3.7 (system-dependent extrapolation), and 2.4 (additivity scheme) kJ mol{sup –1}. The CCSD energy in these approximations is obtained from basis sets of up to TZ quality and the latter two approaches require additional MP2 calculations with basis sets of up to QZ quality. We also assess the performance of the basis-set extrapolations and additivity schemes for a set of 20 basis-set limit CCSD atomization energies of larger molecules including amino acids, DNA/RNA bases, aromatic compounds, and platonic hydrocarbon cages. We obtain the following RMSDs for the above methods: 10.2 (global extrapolation), 5.7 (system-dependent extrapolation), and 2.9 (additivity scheme) kJ mol{sup –1}.

Estimating the CCSD basis-set limit energy from small basis sets: basis-set extrapolations vs additivity schemes

International Nuclear Information System (INIS)

Spackman, Peter R.; Karton, Amir

2015-01-01

Coupled cluster calculations with all single and double excitations (CCSD) converge exceedingly slowly with the size of the one-particle basis set. We assess the performance of a number of approaches for obtaining CCSD correlation energies close to the complete basis-set limit in conjunction with relatively small DZ and TZ basis sets. These include global and system-dependent extrapolations based on the A + B/L α two-point extrapolation formula, and the well-known additivity approach that uses an MP2-based basis-set-correction term. We show that the basis set convergence rate can change dramatically between different systems(e.g.it is slower for molecules with polar bonds and/or second-row elements). The system-dependent basis-set extrapolation scheme, in which unique basis-set extrapolation exponents for each system are obtained from lower-cost MP2 calculations, significantly accelerates the basis-set convergence relative to the global extrapolations. Nevertheless, we find that the simple MP2-based basis-set additivity scheme outperforms the extrapolation approaches. For example, the following root-mean-squared deviations are obtained for the 140 basis-set limit CCSD atomization energies in the W4-11 database: 9.1 (global extrapolation), 3.7 (system-dependent extrapolation), and 2.4 (additivity scheme) kJ mol –1 . The CCSD energy in these approximations is obtained from basis sets of up to TZ quality and the latter two approaches require additional MP2 calculations with basis sets of up to QZ quality. We also assess the performance of the basis-set extrapolations and additivity schemes for a set of 20 basis-set limit CCSD atomization energies of larger molecules including amino acids, DNA/RNA bases, aromatic compounds, and platonic hydrocarbon cages. We obtain the following RMSDs for the above methods: 10.2 (global extrapolation), 5.7 (system-dependent extrapolation), and 2.9 (additivity scheme) kJ mol –1

External man-induced events in relation to nuclear power plant design. A Safety Guide. A publication within the NUSS programme

International Nuclear Information System (INIS)

1996-01-01

In order to take account of lessons learned since the first publication of the NUSS programme was issued, it was decided in 1986 to revise and reissue the Codes and Safety Guides. During the original development of these publications, as well as during the revision process, care was taken to ensure that all Member States, in particular those with active nuclear power programmes, could provide their input. Several independent reviews took place including a final one by the Nuclear Safety Standards Advisory Group (NUSSAG). The revised Codes were approved by the Board of Governors in June 1988. In the revision process new developments in technology and methods of analysis have been incorporated on the basis of international consensus. It is hoped that the revised Codes will be used, and that they will be accepted and respected by Member States as a basis for regulation of the safety of power reactors within the national legal and regulatory framework. 28 refs, 9 figs, 1 tab

Cold Vacuum Drying facility design basis accident analysis documentation

International Nuclear Information System (INIS)

CROWE, R.D.

2000-01-01

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls

Cold Vacuum Drying facility design basis accident analysis documentation

Energy Technology Data Exchange (ETDEWEB)

CROWE, R.D.

2000-08-08

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.

A prioritization of generic safety issues. Supplement 21, Revision insertion instructions

Energy Technology Data Exchange (ETDEWEB)

None, None

1996-12-31

The report presents the safety priority ranking for generic safety issues related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The safety priority rankings are HIGH, MEDIUM, LOW, and DROP, and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolution of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative.

A prioritization of generic safety issues. Supplement 21, Revision insertion instructions

International Nuclear Information System (INIS)

1996-01-01

The report presents the safety priority ranking for generic safety issues related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The safety priority rankings are HIGH, MEDIUM, LOW, and DROP, and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolution of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative.

Identification of road user related risk factors, Deliverable 5.1 of the H2020 project SafetyCube (Safety CaUsation, Benefits and Efficiency).

NARCIS (Netherlands)

Filtness, A. & Papadimitriou, E. (Eds.) Leskovšek, B. Focant, N. Martensen, H. Sgarra, V. Usami, D.S. Soteropoulos, A. Stadlbauer, S. Theofilatos, A. Yannis, G. Ziakopoulos, A. Diamandouros, K. Durso, C. Goldenbeld, C. Loenis, B. Schermers, G. Petegem, J.-H. van Elvik, R. Hesjevoll, I.S. Quigley, C. & Papazikou, E.

2017-01-01

The present Deliverable (D5.1) describes the identification and evaluation of infrastructure related risk factors. It outlines the results of Task 5.1 of WP5 of SafetyCube, which aimed to identify and evaluate infrastructure related risk factors and related road safety problems by (i) presenting a

Status of National Nuclear Infrastructure Development (NG-T-3.2). Basis for Evaluation - Legal, safety, security, safeguards issues

International Nuclear Information System (INIS)

Yllera, Javier

2010-01-01

A framework for achieving high levels of nuclear safety and security worldwide Builds upon: Legal Instruments; Use of IAEA SSs and security guidance; Harmonization of national regulations; Exchange of knowledge, experiences & regulatory practices and Multinational cooperation and safety reviews. The IAEA is the depository of many key international conventions and legal agreements. All countries with operating nuclear power plants are now parties to the Convention. The main objective of Convention on Nuclear Safety is to achieve and maintain a high level of nuclear safety worldwide through the enhancement of national measures and international cooperation including, where appropriate, safety related technical co-operation. All practical efforts must be made to prevent and mitigate nuclear or radiation accidents. The primary means of preventing and mitigating the consequences of accidents is “defence in depth”. Safety assessments are to be carried out and documented by the organization responsible for operating the facility, are to be independently verified and are to be submitted to the regulatory body as part of the licensing or authorization process. Licensing process must be well-defined, clear, transparent and traceable. The public should be given an opportunity to provide their views during certain steps of the licensing process

White paper on nuclear safety in 1981

International Nuclear Information System (INIS)

1981-01-01

The measures to research, develop and utilize atomic energy in Japan have been strengthened since the Atomic Energy Act was instituted in 1955, always on the major premise of securing the safety. The Nuclear Safety Commission established in October, 1978, has executed various measures to protect the health and safety of the nation as the center of the atomic energy safety regulation administration in Japan. Now, the Nuclear Safety Commission has published this annual report on atomic energy safety, summarizing various activities for securing the safety of atomic energy since its establishment to the end of March, 1981. This report is the inaugural issue, and the course till the Nuclear Safety Commission has made its start is also described. The report is composed of general remarks, response to the TMI accident, the safety regulation and security of nuclear facilities, the treatment and disposal of radioactive wastes, the investigation of environmental radioactivity, the countermeasures for preventing disasters around nuclear power stations and others, the research on the safety of atomic energy, international cooperation, and the improvement of the basis for securing the safety. Various related materials are attached. (Kako, I.)

Official News relating to CERN Safety Rules

CERN Multimedia

HSE Unit

2015-01-01

The CERN Safety Rules listed below have been published on the HSE website (see here) and entered into force on the 9 June 2015:   Safety Regulation SR-M “Mechanical equipment”: http://cern.ch/safety-rules/SR-M_ENv2.htm; this SR-M (version 2) cancels and replaces SR-M (version 1) and the corresponding provisions of General Safety Instruction GSI-M3 “Special Equipment” (version 1).   General Safety Instruction GSI-M-1 “Lifting equipment and accessories”: http://cern.ch/safety-rules/GSI-M-1_ENv2.htm; this GSI-M-1 (version 2) cancels and replaces GSI-M1 (version 1). Specific Safety Instruction SSI-M-1-1 “Slings and lifting chains”: http://cern.ch/safety-rules/SSI-M-1-1_EN.htm; Specific Safety Instruction SSI-M-1-2 “Cranes, bridge cranes, gantry cranes and power-driven hoists”: http://cern.ch/safety-rules/SSI-M-1-2_EN.htm; Specific Safety Instruction SSI-M-1-3 “Non-f...

Subgrouping Poor Readers on the Basis of Individual Differences in Reading-Related Abilities

OpenAIRE

Catts, Hugh W.; Hogan, Tiffany; Fey, Marc E.

2003-01-01

The present study investigated the use of the Reading Component Model to subgroup poor readers. A large sample of poor readers was identified in second grade and subgrouped on the basis of relative strengths and weaknesses in word recognition and listening comprehension. Although homogeneous subgroups were not identified, poor readers could be classified into four subgroups that differed significantly in reading-related abilities. Further analyses showed that poor readers' strengths and weakn...

Research on integrated managing system based on CIMS for nuclear power plant safety

International Nuclear Information System (INIS)

Zhou Gang

2006-01-01

In order to improve safety, economy and reliability of operation for nuclear power plant (NPP), a novel integrated managing method was proposed based on the ideas of computer and contemporary integrated manufacturing system (CIMS). The application of CIMS to nuclear power plant safety management was researched. In order to design an integrated managing system to meet the needs of NPP safety management, all work related to nuclear safety is divided into different category according to its characters. On basis of this work, general integrated managing system was designed at first. Then subsystems were designed and every subsystem implements a category of nuclear safety management work. All subsystems are independent relatively on the one hand and are interrelated on other hand by global information system. (authors)

Containment-emergency-sump performance. Technical findings related to Unresolved Safety Issue A-43

International Nuclear Information System (INIS)

1983-04-01

This report summarizes key technical findings related to the Unresolved Safety Issue A-43, Containment Emergency Sump Performance, and provides recommendations for resolution of attendant safety issues. The key safety questions relate to: (a) effects of insulation debris on sump performance; (b) sump hydraulic performance as determined by design features, submergence, and plant induced effects, and (c) recirculation pump performance wherein air and/or particulate ingestion can occur. The technical findings presented in this report provide information relevant to the design and performance evaluation of the containment emergency sump

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)

26 CFR 1.1014-4 - Uniformity of basis; adjustment to basis.

Science.gov (United States)

2010-04-01

...) INCOME TAX (CONTINUED) INCOME TAXES Basis Rules of General Application § 1.1014-4 Uniformity of basis... to property acquired by bequest, devise, or inheritance relate back to the death of the decedent... prescribing a general uniform basis rule for property acquired from a decedent is, on the one hand, to tax the...

Report on safety related occurrences and reactor trips July 1, 1979 - December 31, 1979

International Nuclear Information System (INIS)

Olsson, S.; Andermo, L.

1980-01-01

This is a report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1979 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 76 safety related occurrences and 27 reactor trips have been reported to the Nuclear Power Inspectorate. It is to the greatest extent conventional components such as valves and pumps which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant system and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips are normal. The average value for these 6 months is 4.5 trips/unit. Approximetely one half of the reactor trips happened at zero or very low power operation. The fact that even small deviations from prescribed operation result in an automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences without safety significance. (author)

Safety-related decision making at a nuclear power plant

International Nuclear Information System (INIS)

Vaurio, J.K.

1998-01-01

The decision making environment of an operating nuclear power plant is presented. The organizations involved, their roles and interactions as well as the main influencing factors and decision criteria are described. The focus is on safety-related decisions, and the framework is based on the situation at Loviisa power station. The role of probabilistic safety assessment (PSA) is illustrated with decisions concerning plant modifications, optimization, acceptance of temporary configurations and extended repair times. Suggestions are made for rational and flexible risk-based control of allowed times to operate the plant with some components out of service. (orig.)

The concept of underground nuclear heat and power plants (UNHPP) of upgraded safety, developed on the basis of ship-building technologies

International Nuclear Information System (INIS)

Pashin, V.M.; Petrov, Eh.L.; Shalik, G.P.; Khazov, B.S.; Malyshev, S.P.

1996-01-01

A concept of underground nuclear heat and power plants (UNHPP) of upgraded safety on the basis of ship-building technologies is considered, in which the priority is set to population security and environmental protection. Ways of realization of ziro radiation risk for the population residing in a close vicinity of UNHPP are substantiated. basic principles of the concept are formulated which envisage the use of ship propulsion reactor facilities that have been multiply tested in operation. The sources of economic competitiveness of UNHPPs, as compared with the existing NPPs, are shown

Accident beyond the design basis management with the coolant loss at the NPP with WWER

International Nuclear Information System (INIS)

Skalozubov, V.I.; Klyuchnikov, A.A.; Kolykhanov, V.N.

2010-01-01

The analysis of status and experience of development on modelling and accident beyond the design basis management, including the severe accidents, at the nuclear power plants is carried out. The methodical providing of manuals on the accident beyond the design basis management with the coolant loss on the basis of simulated critical system configurations providing the necessary safety function performance on reactor unit is proposed. The project of symptom-oriented manuals on accident beyond the design basis management with the coolant loss on the serial power unit with WWER-1000 on the basis of developed methodical providing and well known results of deepened safety analysis is presented.

Technical basis for instrumentation and control design improvements in WWER-440/230 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1996-01-01

Instrumentation and control (I and C) has been recognized as an area which requires substantial improvements in WWER NNPs, particularly for model 230 plants. Under contract with the IAEA the Spanish company Empresarios Agrupados (EA) developed a basic document proposing a technical basis for improvements related to the following most significant aspects of I and C: criteria for safety classification; remote shutdown panel; I and C support to operation and control room design; instrumentation set point margins; accident monitoring instrumentation. This publication is derived from the original report of EA which was circulated by the IAEA for review by staff members and experts from various Member States. It was finally agreed upon at a Consultants' meeting convened by the IAEA in Vienna in May 1994 with the participation of experts from France, Germany and Spain. The guidance expressed in this report is based on the IAEA/NUSS standards, safety guides and practices, and on regulations in use in various Member States. It is proposed as a way of carrying out the necessary studies to improve safety by upgrading the vital part of instrumentation an control in WWER-440 model 230 nuclear power plants. 28 refs, 3 figs

Replacement cross-site transfer system project W-058 safety class upgrade summary report

International Nuclear Information System (INIS)

Schlosser, R.L.

1998-01-01

This report evaluates the design of the replacement cross-site transfer system structures, systems, and components for safety related applications as defined in the Tank Waste Remediation Systems Basis for Interim Operations

Management of the aging of critical safety-related concrete structures in light-water reactor plants

International Nuclear Information System (INIS)

Naus, D.J.; Oland, C.B.; Arndt, E.G.

1990-01-01

The Structural Aging Program has the overall objective of providing the USNRC with an improved basis for evaluating nuclear power plant safety-related structures for continued service. The program consists of a management task and three technical tasks: materials property data base, structural component assessment/repair technology, and quantitative methodology for continued-service determinations. Objectives, accomplishments, and planned activities under each of these tasks are presented. Major program accomplishments include development of a materials property data base for structural materials as well as an aging assessment methodology for concrete structures in nuclear power plants. Furthermore, a review and assessment of inservice inspection techniques for concrete materials and structures has been complete, and work on development of a methodology which can be used for performing current as well as reliability-based future condition assessment of concrete structures is well under way. 43 refs., 3 tabs

Modeling patient safety incidents knowledge with the Categorial Structure method.

Science.gov (United States)

Souvignet, Julien; Bousquet, Cédric; Lewalle, Pierre; Trombert-Paviot, Béatrice; Rodrigues, Jean Marie

2011-01-01

Following the WHO initiative named World Alliance for Patient Safety (PS) launched in 2004 a conceptual framework developed by PS national reporting experts has summarized the knowledge available. As a second step, the Department of Public Health of the University of Saint Etienne team elaborated a Categorial Structure (a semi formal structure not related to an upper level ontology) identifying the elements of the semantic structure underpinning the broad concepts contained in the framework for patient safety. This knowledge engineering method has been developed to enable modeling patient safety information as a prerequisite for subsequent full ontology development. The present article describes the semantic dissection of the concepts, the elicitation of the ontology requirements and the domain constraints of the conceptual framework. This ontology includes 134 concepts and 25 distinct relations and will serve as basis for an Information Model for Patient Safety.

IEEE Std 382-1980: IEEE standard for qualification of safety-related valve actuators

International Nuclear Information System (INIS)

Anon.

1992-01-01

This standard describes the qualification of all types of power-driven valve actuators, including damper actuators, for safety-related functions in nuclear power generating stations. This standard may also be used to separately qualify actuator components. This standard establishes the minimum requirements for, and guidance regarding, the methods and procedures for qualification of all safety-related functions of power-driven valve actuators

Transient safety performance of the PRISM innovative liquid metal reactor

International Nuclear Information System (INIS)

Magee, P.M.; Dubberley, A.E.; Rhow, S.K.; Wu, T.

1988-01-01

The PRISM sodium-cooled reactor concept utilizes passive safety characteristics and modularity to increase performance margins, improve licensability, reduce owner's risk and reduce costs. The relatively small size of each reactor module (471 MWt) facilitates the use of passive self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. Key to the transient performance is the inherent negative reactivity feedback characteristics of the core design resulting from the use of metal (U-Pu-Zr) swing, and very low control rod runout worth. Selected beyond design basis events relying only on these core design features are analyzed and the design margins summarized to demonstrate the advancement in reactor safety achieved with the PRISM design concept

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 of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition)

International Nuclear Information System (INIS)

2012-01-01

This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

International Nuclear Information System (INIS)

2012-01-01

This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

Analysis of safety supervision effect on coal mine using game theory

Energy Technology Data Exchange (ETDEWEB)

Zhou Min; Xiao Zhong-hai [China University of Mining & Technology, Xuzhou (China). School of Management

2006-07-01

A two-phase game model was constructed based on economic game theory to improve the inefficient supervision conditions of coalmining enterprises' safety production. The inner mechanism of safety supervision system was carefully studied. The strategy set of the dynamic game-theory relations between colliery proprietor and safety supervision department and the strict competitive game-theory relations between safety supervision department and workers at the production line was analyzed. The conditions of achieving the dominant-strategy equilibrium were revealed. Upon this basis, the possible solution set of safety benefit distribution for the cooperative game-theory of triple party was obtained and the effective and cooperative fair solution was formed. Finally, conclusions are drawn as follows: the combination of necessary penalty to supervisors and operators with the fair distribution of safety benefits can realize the zero-defect of operational violation, so that the violation behavior can be completely avoided. 10 refs., 7 figs.

Basic researches on thermo-hydraulic non-equilibrium phenomena related to nuclear reactor safety

International Nuclear Information System (INIS)

Sakurai, Akira; Kataoka, Isao; Aritomi, Masanori.

1989-01-01

A review was made of recent developments of fundamental researches on thermo-hydraulic non-equilibrium phenomena related to light water reactor safety, in relation to problems to be solved for the improvement of safety analysis codes. As for the problems related to flow con ditions, fundamental researches on basic conservation equations and constitutive equations for transient two-phase flow were reviewed. Regarding to the problems related to thermal non-equilibrium phenomena, fundamental researches on film boiling in pool and forced convection, transient boiling heat transfer and flow behavior caused by pressure transients were reviewed. (author)

Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

Energy Technology Data Exchange (ETDEWEB)

Chun, Ji-Han, E-mail: chunjh@kaeri.re.kr; Lim, Sung-Won; Chung, Bub-Dong; Lee, Won-Jae

2015-08-15

Highlights: • Thermal conductivity model of the FCM fuel was developed and adopted in the MARS. • Scoping analysis for candidate FCM FAs was performed to select feasible FA. • Preliminary safety criteria for FCM fuel and SiC/Zr cladding were set up. • Enhanced safety margin and accident tolerance for FCM-SiC/Zr core were demonstrated. - Abstract: The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates. A 16 × 16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

SAFETY PERFORMANCE OF SUBCONTRACTORS IN THE PALESTINIAN CONSTRUCTION INDUSTRY

Directory of Open Access Journals (Sweden)

Adnan Enshassi

2008-06-01

Full Text Available Subcontractors perform most of the construction works and their effect on industry are apparent in different activities of construction. Therefore, subcontractors need more attention from government and contractors union. The aim of this paper is to identify, evaluate, and rank factors that influence safety performance of subcontractors in the Gaza Strip (Palestine according to their relative importance. The study concluded that reported accident rates will decrease among subcontractors and their workers if new workers are trained well in the work site and they are informed about dangerous places, and if a workable safety plan is well preplanned. The results also showed that reported accident rates increased among subcontractors when using old, unsafe equipment and due to the complexity or difficulty in the construction sites features. Owners and general contractors need to stipulate strict clauses for safety in the contract for improving safety record of subcontractors. Construction workers must receive proper job related safety and health training with a safety logbook. It is recommended that the subcontractors and workers should attend continuing safety programs on regular basis as part of their perquisite to work in construction sites.

78 FR 36790 - Commercial Fishing Safety Advisory Committee; Vacancies

Science.gov (United States)

2013-06-19

... position); (c) a representative of education or training professionals related to fishing vessel safety or...) manufacturers of equipment for vessels to which Chapter 45 of Title 46, U.S.C. applies; (3) education or... Homeland Security (DHS) does not discriminate in employment on the basis of race, color, religion, sex...

Transformational leadership and safety performance among nurses: the mediating role of knowledge-related job characteristics.

Science.gov (United States)

Lievens, Ilse; Vlerick, Peter

2014-03-01

To report the impact of transformational leadership on two dimensions of nurses' safety performance (i.e. safety compliance and safety participation) and to study the mediating role of knowledge-related job characteristics in this relationship. Safety performance refers to the behaviours that employees exhibit to adhere to safety guidelines and to promote health and safety at their workplace. Nurses' safety performance is a major challenge for healthcare settings, urging the need to identify the key determinants and psychological mechanisms that influence it. A cross-sectional survey study. The study was carried out in September 2010 in a large Belgian hospital. We used self-administered questionnaires; 152 nurses participated. The hypotheses were tested using hierarchical regression analyses. In line with our first hypothesis, the results show that transformational leadership exerted a significant positive impact on both dimensions of nurses' safety performance. This positive relation was mediated by knowledge-related job characteristics, supporting our second hypothesis. Head nurses' transformational leadership can enhance nurses' compliance with and participation in safety. Furthermore, transformational head nurses are able to influence the perception that their nurses have about the kind and amount of knowledge in their job, which can also lead to increases in both dimensions of nurses' safety performance. This study therefore demonstrates the key impact that transformational head nurses have, both directly and indirectly, on the safety performance of their nurses. © 2013 John Wiley & Sons Ltd.

Light-water reactors. Safety problems and related studies in France

International Nuclear Information System (INIS)

Lelievre, J.

1975-01-01

The program of theoretical and experimental studies developed by the CEA on the safety of PWR reactors is presented: studies relative to the consequences of a LOCA following a rupture of the primary system, studies relative to fuel element behavior, studies on steels, reliability studies and studies of non-destructive testing methods [fr

A prioritization of generic safety issues. Supplement 19, Revision insertion instructions

International Nuclear Information System (INIS)

1995-11-01

The report presents the safety priority ranking for generic safety issues related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The safety priority rankings are HIGH, MEDIUM, LOW, and DROP, and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolution of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative. This document provides revisions and amendments to the report

A prioritization of generic safety issues. Supplement 19, Revision insertion instructions

Energy Technology Data Exchange (ETDEWEB)

None

1995-11-01

The report presents the safety priority ranking for generic safety issues related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of NRC resources for the resolution of those safety issues that have a significant potential for reducing risk. The safety priority rankings are HIGH, MEDIUM, LOW, and DROP, and have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolution of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors. To the extent practical, estimates are quantitative. This document provides revisions and amendments to the report.

Regulation for delivery of subsidies for public relations and safety

International Nuclear Information System (INIS)

1984-01-01

The regulations provide for subsidies for the public relations activities and safety operations carried out by a local government for the local inhabitants in the vicinity of a nuclear power generation, etc. facility. This type of activity includes the dissemination of information on nuclear power, studies on securing the safety of the inhabitants and communication concerning the facility safety. The contents are as follows: limits of the subsidies, terms of subsidy allocations, the application for subsidies, determination of subsidy allocations, withdrawal of applications, the conditions to the allocations, a report on the work proceedings, a report on the results, confirmation on the sum of subsidies, withdrawal of the decision for subsidies, limitations for disposal of the properties, etc. (Mori, K.)

Regulation for delivery of subsidies for public relations and safety

International Nuclear Information System (INIS)

1985-01-01

The regulations provide for subsidies for the public relations activities and safety operations carried out by a local government for the local inhabitants in the vicinity of a nuclear power generation, etc. facility. This type of activity includes the dissemination of information on nuclear power, studies on securing the safety of the inhabitants and communication concerning the facility safety. The contents are as follows : limits of the subsidies, terms of subsidy allocations, the application for subsidies, determination of subsidy allocations, withdrawal of applications, the conditions to the allocations, a report on the work proceedings, a report on the results, confirmation on the sum of subsidies, withdrawal of the decision for subsidies, limitations for disposal of the properties, etc. (Kubozono, M.)

Interim safety basis for fuel supply shutdown facility

International Nuclear Information System (INIS)

Brehm, J.R.; Deobald, T.L.; Benecke, M.W.; Remaize, J.A.

1995-01-01

This ISB in conjunction with the new TSRs, will provide the required basis for interim operation or restrictions on interim operations and administrative controls for the Facility until a SAR is prepared in accordance with the new requirements. It is concluded that the risk associated with the current operational mode of the Facility, uranium closure, clean up, and transition activities required for permanent closure, are within Risk Acceptance Guidelines. The Facility is classified as a Moderate Hazard Facility because of the potential for an unmitigated fire associated with the uranium storage buildings

Fuel safety criteria technical review - Results of OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria

International Nuclear Information System (INIS)

Hollasky, N.; Valtonen, K.; Hache, G.; Gross, H.; Bakker, K.; Recio, M.; Bart, G.; Zimmermann, M.; Van Doesburg, W.; Killeen, J.; Meyer, R.O.; Speis, T.

2000-01-01

With the advent of advanced fuel and core designs, the adoption of more aggressive operational modes and the implementation of more accurate (best estimate or statistical) design and analysis methods, there is a concern if safety margins have remained adequate. Most - if not all - of the currently existing safety criteria were established during the 60's and early 70's, and verified against experiments with fuel that was available at that time, mostly with unirradiated specimens. Verification was of course performed as designs progressed in later years, however mostly with the aim to be able to prove that these designs adequately complied with existing criteria, and not to establish new limits. The OECD/CSNI/PWG2 Task Force on Fuel Safety Criteria (TFFSC) was therefore given the mandate to technically review the existing fuel safety criteria, focusing on the 'new design' elements (new fuel and core design, cladding materials, manufacturing processes, high burnup, MOX, etc.) introduced by the industry. It should also identify if additional efforts may be required (experimental, analytical) to ensure that the basis for fuel safety criteria is adequate to address the relevant safety issues. In this report, fuel-related criteria are discussed without attempting to categorize them according to event type or risk significance. For each of these 20 criteria, we present a brief description of the criterion as it is used in several applications along with the rationale for having such a criterion. New design elements, such as different cladding materials, higher burnup, and the use of MOX fuels, can affect fuel-related margins and, in some cases, the criteria themselves. Some of the more important effects are mentioned in order to indicate whether the criteria need to be re-evaluated. The discussion may not cover all possible effects, but should be sufficient to identify those criteria that need to be addressed. A summary of these discussions is given in Section 7. As part

Compiler issues associated with safety-related software

International Nuclear Information System (INIS)

Feinauer, L.R.

1991-01-01

A critical issue in the quality assurance of safety-related software is the ability of the software to produce identical results, independent of the host machine, operating system, or compiler version under which the software is installed. A study is performed using the VIPRE-0l, FREY-01, and RETRAN-02 safety-related codes. Results from an IBM 3083 computer are compared with results from a CYBER 860 computer. All three of the computer programs examined are written in FORTRAN; the VIPRE code uses the FORTRAN 66 compiler, whereas the FREY and RETRAN codes use the FORTRAN 77 compiler. Various compiler options are studied to determine their effect on the output between machines. Since the Control Data Corporation and IBM machines inherently represent numerical data differently, methods of producing equivalent accuracy of data representation were an important focus of the study. This paper identifies particular problems in the automatic double-precision option (AUTODBL) of the IBM FORTRAN 1.4.x series of compilers. The IBM FORTRAN version 2 compilers provide much more stable, reliable compilation for engineering software. Careful selection of compilers and compiler options can help guarantee identical results between different machines. To ensure reproducibility of results, the same compiler and compiler options should be used to install the program as were used in the development and testing of the program

A new concept of safety parameter display system

International Nuclear Information System (INIS)

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

1986-07-01

A general description of Angra-1 Parameter Display System (SSPA), a real time and on-line computerized monitoring system for the parameters related to the power plant safety is presented. This system has the main purpose of diminish the load on the Angra-1 power plant operators at an emergency event by supplying them with the additional tools serving as the basis for a prompt identification of the accident. The SSPA is a kind of safety parameter display system whose concept was introduced after Three Mile Island accident in USA. The SSPA comprises two nuclear applications independently considered. They are included into the Parameters Monitoring Integrated System (SIMP) and the safety critical function system (SFCS). (Author) [pt

Safety of Research Reactors. Specific Safety Requirements (French Edition)

International Nuclear Information System (INIS)

2017-01-01

This Safety Requirements publication establishes requirements for all main areas of safety for research reactors, with particular emphasis on requirements for design and operation. It explains the safety objectives and concepts that form the basis for safety and safety assessment for all stages in the lifetime of a research reactor. Technical and administrative requirements for the safety of new research reactors are established in accordance with these objectives and concepts, and they are to be applied to the extent practicable for existing research reactors. The safety requirements established in this publication for the management of safety and regulatory supervision apply to site evaluation, design, manufacturing, construction, commissioning, operation (including utilization and modification), and planning for decommissioning of research reactors (including critical assemblies and subcritical assemblies). The publication is intended for use by regulatory bodies and other organizations with responsibilities in these areas and in safety analysis, verification and review, and the provision of technical support.

The use of probabilistic safety assessment (PSA) based maintenance indicators to increase the availability of safety related systems in nuclear power plants

International Nuclear Information System (INIS)

Kirchsteiger, C.

1991-04-01

This work describes the theoretical development of a Probabilistic Safety Assessment (PSA) based Performance Indicator (PI) model for a comprehensive Maintenance Efficiency Analysis (MEA) and its practical application to past operational history data of a certain nuclear power plant. Plant specific equipment history and maintenance work on data have been collected and analysed using various advanced statistical procedures (nonparametric methods, multivariate analysis in order to be able to estimate safety system related equipment and maintenance process trends. The main results of such a MEA case study are the trends in the (in)effectiveness of the performance of a selected safety system and its dominant components as well as the detection of the dominant maintenance related causes of its bad (good) equipment performance. Finally, the therefrom gained results are used to propose a new set of safety system-based and maintenance-related performance indicators, including suggestions for a corresponding plant specific maintenance data collection system. (author)

Design and installation of advanced computer safety related instrumentation

International Nuclear Information System (INIS)

Koch, S.; Andolina, K.; Ruether, J.

1993-01-01

The rapidly developing area of computer systems creates new opportunities for commercial utilities operating nuclear reactors to improve plant operation and efficiency. Two of the main obstacles to utilizing the new technology in safety-related applications is the current policy of the licensing agencies and the fear of decision making managers to introduce new technologies. Once these obstacles are overcome, advanced diagnostic systems, CRT-based displays, and advanced communication channels can improve plant operation considerably. The article discusses outstanding issues in the area of designing, qualifying, and licensing of computer-based instrumentation and control systems. The authors describe the experience gained in designing three safety-related systems, that include a Programmable Logic Controller (PLC) based Safeguard Load Sequencer for NSP Prairie Island, a digital Containment Isolation monitoring system for TVA Browns Ferry, and a study that was conducted for EPRI/NSP regarding a PLC-based Reactor Protection system. This article presents the benefits to be gained in replacing existing, outdated equipment with new advanced instrumentation

Design basis document open-item resolution and reportability

International Nuclear Information System (INIS)

Gambhir, S.K.; Livingston, B.R.; Purcell, J.J.; Erickson, E.A.

1989-01-01

In the process of reconstituting the design bases for older nuclear power plants, information or references may not be available to fully define the design requirements or to document and verify the adequacy of the design. Also, information that is in conflict with other data is identified. The missing and conflicting information must be reconstituted in order to adequately document the design bases of the plant. For these operating facilities, the identification, tracking, and resolution of missing or conflicting information is very important when the reporting requirements stipulated by 10CFR21, 10CFR50.72, and 10CFR50.73 are considered. Additionally, controlled documentation (calculations, drawings, etc.) used to develop the design basis documents may contain conflicting data. In some cases, conflicts between the as-built design and licensing or design basis requirements established in specific commitments to the U.S. Nuclear Regulatory Commission may be identified. Furthermore, concerns regarding the adequacy of safety-related systems or components to perform their required function may be identified that would warrant prompt action by the licensee. The approach discussed in this paper was used by Omaha Public Power District for the ongoing design basis reconstitution effort at the Fort Calhoun nuclear plant

Technical basis for the aboveground structure failure and associated represented hazardous conditions

International Nuclear Information System (INIS)

GOETZ, T.G.

2003-01-01

This technical basis document describes the risk binning process and the technical basis for assigning risk bins for the aboveground structure failure representative accident and associated represented hazardous conditions. This document was developed to support the documented safety analysis

Synergistic behaviour of nuclear radiation, temperature-humidity extremes and LOCA situation on safety and safety-related equipment in Indian nuclear power plants

International Nuclear Information System (INIS)

Kulkarni, R.D.; Bora, J.S.; Prakash, Ravi; Agarwal, Vivek; Sundersingh, V.P.

2002-01-01

Full text: The general philosophy for the instrumentation in nuclear power plants is based on the use of equipment/instruments which are capable of continuous satisfactory operation over a long period of time with minimum attention. Long term reliability under varying service conditions is of prime importance. The reliability of nuclear power plant depends on the reliability of safety and safety-related electronic instruments/ equipment used for performing the crucial tasks. The electrical and electronic systems/ circuits/ components of the equipment used in reactor safety systems (e.g. reactor protection system, emergency core cooling system, etc.) and reactor safety-related systems (e.g. reactor containment isolation and cooling system, reactor shutdown system, etc.) are responsible for safe and reliable operation of a nuclear power plant. The performance of reactor safety and safety-related equipment/instruments viz. pressure and differential pressure transmitter, amplifier for ion chamber, etc. has been evaluated under synergistic atmosphere including LOCA to find out the critical link in the circuits and subsequent modifications are suggested. The mathematical representation of the generated database has been done to estimate the life span of the instruments and accordingly the guidelines has been prepared for the operational staff to avoid the forced outage of the plant. All the details are included and mathematical models are presented to predict the future performances

Safety culture in design. Final report

International Nuclear Information System (INIS)

Macchi, L.; Pietikaeinen, E.; Liinasuo, M.; Savioja, P.; Reiman, T.; Wahlstroem, M.; Kahlbom, U.; Rollenhagen, C.

2013-04-01

In this report we approach design from a safety culture approach As this research area is new and understudied, we take a wide scope on the issue. Different theoretical perspectives that can be taken when improving safety of the design process are considered in this report. We suggest that in the design context the concept of safety culture should be expanded from an organizational level to the level of the network of organizations involved in the design activity. The implication of approaching the design process from a safety culture perspective are discussed and the results of the empirical part of the research are presented. In the interview study in Finland and Sweden we identified challenges and opportunities in the design process from safety culture perspective. Also, a small part of the interview study concentrated on state of the art human factors engineering (HFE) practices in Finland and the results relating to that are presented. This report provide a basis for future development of systematic good design practices and for providing guidelines that can lead to safe and robust technical solutions. (Author)

Safety culture in design. Final report

Energy Technology Data Exchange (ETDEWEB)

Macchi, L.; Pietikaeinen, E.; Liinasuo, M.; Savioja, P.; Reiman, T.; Wahlstroem, M. [VTT Technical Research Centre of Finland, Espoo (Finland); Kahlbom, U. [Risk Pilot AB, Stockholm (Sweden); Rollenhagen, C. [Vattenfall, Stockholm, (Sweden)

2013-04-15

In this report we approach design from a safety culture approach As this research area is new and understudied, we take a wide scope on the issue. Different theoretical perspectives that can be taken when improving safety of the design process are considered in this report. We suggest that in the design context the concept of safety culture should be expanded from an organizational level to the level of the network of organizations involved in the design activity. The implication of approaching the design process from a safety culture perspective are discussed and the results of the empirical part of the research are presented. In the interview study in Finland and Sweden we identified challenges and opportunities in the design process from safety culture perspective. Also, a small part of the interview study concentrated on state of the art human factors engineering (HFE) practices in Finland and the results relating to that are presented. This report provide a basis for future development of systematic good design practices and for providing guidelines that can lead to safe and robust technical solutions. (Author)

Regulatory instrument review: Management of aging of LWR [light water reactor] major safety-related components

International Nuclear Information System (INIS)

Werry, E.V.

1990-10-01

This report comprises Volume 1 of a review of US nuclear plant regulatory instruments to determine the amount and kind of information they contain on managing the aging of safety-related components in US nuclear power plants. The review was conducted for the US Nuclear Regulatory Commission (NRC) by the Pacific Northwest Laboratory (PNL) under the NRC Nuclear Plant Aging Research (NPAR) Program. Eight selected regulatory instruments, e.g., NRC Regulatory Guides and the Code of Federal Regulations, were reviewed for safety-related information on five selected components: reactor pressure vessels, steam generators, primary piping, pressurizers, and emergency diesel generators. Volume 2 will be concluded in FY 1991 and will also cover selected major safety-related components, e.g., pumps, valves and cables. The focus of the review was on 26 NPAR-defined safety-related aging issues, including examination, inspection, and maintenance and repair; excessive/harsh testing; and irradiation embrittlement. The major conclusion of the review is that safety-related regulatory instruments do provide implicit guidance for aging management, but include little explicit guidance. The major recommendation is that the instruments be revised or augmented to explicitly address the management of aging

Cold Vacuum Drying (CVD) Facility Design Basis Accident Analysis Documentation

Energy Technology Data Exchange (ETDEWEB)

PIEPHO, M.G.

1999-10-20

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR.

Relational approach in managing construction project safety: a social capital perspective.

Science.gov (United States)

Koh, Tas Yong; Rowlinson, Steve

2012-09-01

Existing initiatives in the management of construction project safety are largely based on normative compliance and error prevention, a risk management approach. Although advantageous, these approaches are not wholly successful in further lowering accident rates. A major limitation lies with the approaches' lack of emphasis on the social and team processes inherent in construction project settings. We advance the enquiry by invoking the concept of social capital and project organisational processes, and their impacts on project safety performance. Because social capital is a primordial concept and affects project participants' interactions, its impact on project safety performance is hypothesised to be indirect, i.e. the impact of social capital on safety performance is mediated by organisational processes in adaptation and cooperation. A questionnaire survey was conducted within Hong Kong construction industry to test the hypotheses. 376 usable responses were received and used for analyses. The results reveal that, while the structural dimension is not significant, the mediational thesis is generally supported with the cognitive and relational dimensions affecting project participants' adaptation and cooperation, and the latter two processes affect safety performance. However, the cognitive dimension also directly affects safety performance. The implications of these results for project safety management are discussed. Copyright © 2011 Elsevier Ltd. All rights reserved.

15 CFR 970.205 - Vessel safety.

Science.gov (United States)

2010-01-01

... 15 Commerce and Foreign Trade 3 2010-01-01 2010-01-01 false Vessel safety. 970.205 Section 970.205... safety. In order to provide a basis for the necessary determinations with respect to the safety of life... Safety of Life at Sea, 1974 (SOLAS 74) possesses current valid SOLAS 74 certificates; (2) That any...

Technical basis document for external events

International Nuclear Information System (INIS)

OBERG, B.D.

2003-01-01

This document supports the Tank Farms Documented Safety Analysis and presents the technical basis for the FR-equencies of externally initiated accidents. The consequences of externally initiated events are discussed in other documents that correspond to the accident that was caused by the external event. The external events include aircraft crash, vehicle accident, range fire, and rail accident

Safety problems relating to plutonium recycling in light water reactors

International Nuclear Information System (INIS)

Devillers, C.; Frison, J.M.; Mercier, J.P.; Revais, J.P.

1991-01-01

This paper describes the specific nature, as regards safety, of the mixed oxide (MOX) fuel cycle, with the exception of safety problems relating to the operation of nuclear power plants. These specific characteristics are due mainly to the presence of plutonium in fresh fuel and to the higher plutonium and transuranic element content in spent fuel assemblies. The fuel cycle steps analysed here are the transport of plutonium oxide, the manufacture of MOX fuel assemblies, the transport of fresh and spent fuel assemblies and the processing of spent fuel assemblies

Safety problems relating to plutonium recycling in light water reactors

International Nuclear Information System (INIS)

Devillers, C.; Frison, J.M.; Mercier, J.P.; Revais, J.P

1991-01-01

This paper describes the specific nature, as regards safety, of the mixed oxide (MOX) fuel cycle, with the exception of safety problems relating to the operation of nuclear power plants. These specific characteristics are due mainly to the presence of plutonium in fresh fuel and to the higher plutonium and transuranic element content in spent fuel assemblies. The fuel cycle steps analysed here are the transport of plutonium oxide, the manufacture of MOX fuel assemblies, the transport of fresh and spent fuel assemblies and the processing of spent fuel assemblies. (author) [fr

Effect of snowboard-related concussion safety education for recognizing possible concussions.

Science.gov (United States)

Koh, J O

2011-12-01

The aim of this study was to examine the understanding of snowboard-related concussion and to measure the recognition of possible concussion occurrence after an intervention of snowboard-related concussion safety education in snowboarding. Incidence cohort design. 2008-2009 season Gangwon-do Ski resorts, South Korea. A total of 208 university students (female-72; male-136; age-18 to 32) who registered for a snowboarding class and received credit participated in this project. Snowboard-related concussion safety education class was administered for 30 minutes before the snowboard class began. The knowledge of snowboard-related concussion before and after the safety education was evaluated. Concussion data were collected via a self-report case form at the last day of snowboarding class. The incidence of possible concussion and factors associated with concussions were analyzed by χ2 test. The mean score of snowboard-related concussion knowledge improved from fifteen points to eighteen points out of 20 total points possible. Overall the incidence of concussion was 10 per 100 snowboarder-exposures. χ2 tests showed concussion rates to be significantly different in female snowboarders (P=0.00) and in helmet users (P=0.02). The incidence of possible concussion is high among snowboarding class participants. Emphasis should be given for instituting pre-participation balance training, especially for females to reduce falling in snowboarding. To verify the effects of pre-participation balance training and falling results in a concussion, more research is needed in the future.

Development of methodology for the analysis of fuel behavior in light water reactor in design basis accidents

International Nuclear Information System (INIS)

Salatov, A. A.; Goncharov, A. A.; Eremenko, A. S.; Kuznetsov, V. I.; Bolnov, V. A.; Gusev, A. S.; Dolgov, A. B.; Ugryumov, A. V.

2013-01-01

The report attempts to analyze the current experience of the safety fuel for light-water reactors (LWRs) under design-basis accident conditions in terms of its compliance with international requirements for licensing nuclear power plants. The components of fuel behavior analysis methodology in design basis accidents in LWRs were considered, such as classification of design basis accidents, phenomenology of fuel behavior in design basis accidents, system of fuel safety criteria and their experimental support, applicability of used computer codes and input data for computational analysis of the fuel behavior in accidents, way of accounting for the uncertainty of calculation models and the input data. A brief history of the development of probabilistic safety analysis methodology for nuclear power plants abroad is considered. The examples of a conservative approach to safety analysis of VVER fuel and probabilistic approach to safety analysis of fuel TVS-K are performed. Actual problems in development of the methodology of analyzing the behavior of VVER fuel at the design basis accident conditions consist, according to the authors opinion, in following: 1) Development of a common methodology for analyzing the behavior of VVER fuel in the design basis accidents, implementing a realistic approach to the analysis of uncertainty - in the future it is necessary for the licensing of operating VVER fuel abroad; 2) Experimental and analytical support to the methodology: experimental studies to identify and study the characteristics of the key uncertainties of computational models of fuel and the cladding, development of computational models of key events in codes, validation code on the basis of integral experiments

Using field feedback to estimate failure rates of safety-related systems

International Nuclear Information System (INIS)

Brissaud, Florent

2017-01-01

The IEC 61508 and IEC 61511 functional safety standards encourage the use of field feedback to estimate the failure rates of safety-related systems, which is preferred than generic data. In some cases (if “Route 2_H” is adopted for the 'hardware safety integrity constraints”), this is even a requirement. This paper presents how to estimate the failure rates from field feedback with confidence intervals, depending if the failures are detected on-line (called 'detected failures', e.g. by automatic diagnostic tests) or only revealed by proof tests (called 'undetected failures'). Examples show that for the same duration and number of failures observed, the estimated failure rates are basically higher for “undetected failures” because, in this case, the duration observed includes intervals of time where it is unknown that the elements have failed. This points out the need of using a proper approach for failure rates estimation, especially for failures that are not detected on-line. Then, this paper proposes an approach to use the estimated failure rates, with their uncertainties, for PFDavg and PFH assessment with upper confidence bounds, in accordance with IEC 61508 and IEC 61511 requirements. Examples finally show that the highest SIL that can be claimed for a safety function can be limited by the 90% upper confidence bound of PFDavg or PFH. The requirements of the IEC 61508 and IEC 61511 relating to the data collection and analysis should therefore be properly considered for the study of all safety-related systems. - Highlights: • This paper deals with requirements of the IEC 61508 and IEC 61511 for using field feedback to estimate failure rates of safety-related systems. • This paper presents how to estimate the failure rates from field feedback with confidence intervals for failures that are detected on-line. • This paper presents how to estimate the failure rates from field feedback with confidence intervals for failures that are only revealed by

Bowtie Risk Management methodology and Modern Nuclear Safety Reports

International Nuclear Information System (INIS)

Ilizastigui Pérez, F.

2016-01-01

The Safety Report (SR) plays a crucial role within the nuclear licensing regime as the principal means for demonstrating the adequacy of safety analysis for a nuclear facility to ensure that it can be constructed, operated, maintained, shut down, and decommissioned safely and in compliance with applicable laws and regulations. It serves as the basis for granting authorizations for the commencement of the main stages of the facility’s life cycle as well as decision-making processes related to safety. Historically, the majority of nuclear safety reports have operated under rather prescriptive regimes, with emphasis placed on demonstrations of the robustness of the facility’s design (design safety) against prescriptive technical requirements set by the regulatory body, and less attention paid to demonstrating the adequacy and effectiveness of Operator’s management system for managing risks to daily operation.

Technical basis for the ITER final design report, cost review and safety analysis (FDR)

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-12-01

The ITER final design report, cost review and safety analysis (FDR) is the 4th major milestone, representing the progress made in the ITER Engineering Design Activities. With the approval of the Detailed Design Report (DDR), the design work was concentrated on the requirements of operation, with only relatively minor changes to design concepts of major components. The FDR is the culmination of almost 6 years collaborative design and supporting technical work by the ITER Joint Central Team and Home Teams under the terms of the ITER EDA Agreement. Refs, figs, tabs

Technical basis for the ITER final design report, cost review and safety analysis (FDR)

International Nuclear Information System (INIS)

1998-01-01

The ITER final design report, cost review and safety analysis (FDR) is the 4th major milestone, representing the progress made in the ITER Engineering Design Activities. With the approval of the Detailed Design Report (DDR), the design work was concentrated on the requirements of operation, with only relatively minor changes to design concepts of major components. The FDR is the culmination of almost 6 years collaborative design and supporting technical work by the ITER Joint Central Team and Home Teams under the terms of the ITER EDA Agreement

Hospital nurses' working conditions in relation to motivation and patient safety.

Science.gov (United States)

Toode, Kristi; Routasalo, Pirkko; Helminen, Mika; Suominen, Tarja

2015-03-01

There is a lack of empirical knowledge about nurses' perceptions of their workplace characteristics and conditions, such as level of autonomy and decision authority, work climate, teamwork, skill exploitation and learning opportunities, and their work motivation in relation to practice outputs such as patient safety. Such knowledge is needed particularly in countries, such as Estonia, where hospital systems for preventing errors and improving patient safety are in the early stages of development. This article reports the findings from a cross-sectional survey of hospital nurses in Estonia that was aimed at determining their perceptions of workplace characteristics, working conditions, work motivation and patient safety, and at exploring the relationship between these. Results suggest that perceptions of personal control over their work can affect nurses' motivation, and that perceptions of work satisfaction might be relevant to patient safety improvement work.

Safety guidebook relative to the disposal of radioactive wastes in deep geologic formation

International Nuclear Information System (INIS)

2008-01-01

The French nuclear safety authority (ASN) initiated in 2003 a revision process of the objectives to be considered during the research and work steps of the implementation of a radioactive waste storage facility in deep geologic formations. The purpose of this document is to define the safety objectives that have to be retained at each step of this implementation, from the site characterization to the closure of the facility. This update takes into account the works carried out by the ANDRA (French national agency of radioactive wastes) in the framework of the law from December 30, 1991, and the advices of the permanent experts group about these works. It takes also into consideration the international research works in this domain and the choices defined in the program law no 2006-739 from June 28, 2006 relative to the sustainable management of radioactive materials and wastes. The main modifications concern: the notion of reversibility, the definition of the safety functions of disposal components, the safety goals and the design principles assigned to waste packages, the control of nuclear materials and the monitoring objectives of the facility. The documents treats of the following points: 1 - the objectives of public health and environment protection; 2 - the safety principles and the safety-related design bases of the facility; and 3 - the method used for demonstrating the disposal safety. (J.S.)

Contractor-, steward-, and coworker-safety practice: associations with musculoskeletal pain and injury-related absence among construction apprentices.

Science.gov (United States)

Kim, Seung-Sup; Dutra, Lauren M; Okechukwu, Cassandra A

2014-07-01

This paper sought to assess organizational safety practices at three different levels of hierarchical workplace structure and to examine their association with injury outcomes among construction apprentices. Using a cross-sectional sample of 1,775 construction apprentices, three measures of organizational safety practice were assessed: contractor-, steward-, and coworker-safety practice. Each safety practice measure was assessed using three similar questions (i.e., on-the-job safety commitment, following required or recommended safe work practices, and correcting unsafe work practices); the summed average of the responses ranged from 1 to 4, with a higher score indicating poorer safety practice. Outcome variables included the prevalence of four types of musculoskeletal pain (i.e., neck, shoulder, hand, and back pain) and injury-related absence. In adjusted analyses, contractor-safety practice was associated with both hand pain (OR: 1.27, 95 % CI: 1.04, 1.54) and back pain (OR: 1.40, 95 % CI: 1.17, 1.68); coworker-safety practice was related to back pain (OR: 1.42, 95 % CI: 1.18, 1.71) and injury-related absence (OR: 1.36, 95 % CI: 1.11, 1.67). In an analysis that included all three safety practice measures simultaneously, the association between coworker-safety practice and injury-related absence remained significant (OR: 1.68, 95 % CI: 1.20, 2.37), whereas all other associations became non-significant. This study suggests that organizational safety practice, particularly coworker-safety practice, is associated with injury outcomes among construction apprentices.

Development of safety analysis technology for integral reactor; evaluation on safety concerns of integral reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hee Chul; Kim, Woong Sik; Lee, J. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)

2002-03-01

The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in this study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They include the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. The study presents the general safety requirements applicable to licensing of an integral reactor and suggests additional regulatory requirements, which need to be developed, based on the direction to resolution of the safety concerns. The efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. Suggestion on the development of additional regulatory requirements will contribute for the regulator to taking actions for licensing of an integral reactor. 66 refs., 5 figs., 24 tabs. (Author)

Common safety approach for future pressurized water reactors in France and Germany

International Nuclear Information System (INIS)

Queniart, D.; Gros, G.

1994-01-01

In France and Germany all major activities related to future pressurized water reactors are now proceeding in a coordinated way among the two countries. This holds for utilities and industry in the development of a joint PWR project, the ''European Pressurized Water Reactor (EPR)'' by Electricite de France (EDF), German utilities, Nuclear Power International (NPI), Framatome and Siemens as well as for the technical safety objectives for future evolutionary reactors on the basis of a common safety approach adopted by the safety authorities of both countries for plants to operate form the beginning of the next century. The proposed paper covers this common development of a safety approach and particular technical safety objectives. (authors). 5 refs. 1 fig

IGSC - Integration Group for the Safety Case

International Nuclear Information System (INIS)

2015-01-01

Countries that rely on nuclear energy and materials have an ethical obligation to manage radioactive waste in a safe and environmentally responsible manner. For society to support the sustainable solutions envisaged, disposal concepts must be technologically sound and the safety of these concepts must be convincingly demonstrated. The Nuclear Energy Agency's Integration Group for the Safety Case (IGSC) establishes and documents the technical and scientific basis for developing and reviewing safety cases as a platform for dialogue among technical experts and as a tool for decision making. The IGSC addresses various strategic and policy aspects of radioactive waste management as the technical advisory body to the NEA Radioactive Waste Management Committee (RWMC) for all issues related to repository development. For more than two decades, the IGSC and its predecessor technical groups have promoted the exchange of national experience in evaluating and implementing geological repositories. IGSC activities foster consensus on best practices and encourage the development of innovative, advanced approaches covering the technical aspects at all stages of repository implementation, including: - strategies to characterise and evaluate potential disposal sites; - methods to design and test engineered barrier systems; - priorities for research and development programmes to improve the understanding of important processes and interactions; - tools for safety assessments; - techniques for the effective presentation and communication of the results of safety cases and other factors that provide the basis for increased confidence in the safety of geological disposal facilities. The IGSC has been instrumental in further developing the 'modern safety case', a concept that originally emerged from NEA work in the 1990's. Cooperation with the International Atomic Energy Agency (IAEA) and the European Commission (EC) has led to the worldwide adoption of this safety

Safety strategy

International Nuclear Information System (INIS)

Schultheiss, G.F.

1980-01-01

The basis for safety strategy in nuclear industry and especially nuclear power plants is the prevention of radioactivity release inside or outside of the technical installation. Therefore either technical or administrative measures are combined to a general strategy concept. This introduction will explain in more detail the following topics: - basic principles of safety - lines of assurance (LOA) - defense in depth - deterministic and probabilistic methods. This presentation is seen as an introduction to the more detailed discussion following in this course, nevertheless some selected examples will be used to illustrate the aspects of safety strategy development although they might be repeated later on. (orig.)

General principles of nuclear safety management related to research reactor decommissioning

International Nuclear Information System (INIS)

Banciu, Ortenzia; Vladescu, Gabriela

2003-01-01

The paper contents the general principles applicable to the decommissioning of research reactors to ensure a proper nuclear safety management, during both decommissioning activities and post decommissioning period. The main objective of decommissioning is to ensure the protection of workers, population and environment against all radiological and non-radiological hazards that could result after a reactor shutdown and dismantling. In the same time, it is necessary, by some proper provisions, to limit the effect of decommissioning for the future generation, according to the new Romanian, IAEA and EU Norms and Regulations. Assurance of nuclear safety during decommissioning process involves, in the first step, to establish of some safety principles and requirements to be taken into account during whole process. In the same time, it is necessary to perform a series of analyses to ensure that the whole process is conducted in a planned and safe manner. The general principles proposed for a proper management of safety during research reactor decommissioning are as follows: - Set-up of all operations included in a Decommissioning Plan; - Set-up and qualitative evaluation of safety problems, which could appear during normal decommissioning process, both radiological and nonradiological risks for workers and public; - Set-up of accident list related to decommissioning process the events that could appear both due to some abnormal working conditions and to some on-site and off-site events like fires, explosions, flooding, earthquake, etc.); - Development and qualitative/ quantitative evaluation of scenarios for each incidents; - Development (and evaluation) of safety indicator system. The safety indicators are the most important tools used to assess the level of nuclear safety during decommissioning process, to discover the weak points and to establish safety measures. The paper contains also, a safety case evaluation (description of facility according to the decommissioning

The safety related aspects of pressure components in nuclear power plants

International Nuclear Information System (INIS)

Lindackers, K.H.

1979-01-01

Over the last two years the safety philosophy for nuclear power plants in the Federal Republic of Germany has changed considerably, as everyone working in the field perceives. The original and appropriate philosophy of risk minimalisation through graduated safety barriers has been more and more replaced by the utopian goal of total prevention of any damage. The reasons for this development are discussed briefly especially regarding pressure components. The very numerous pressure components of a nuclear power station are not all of equal importance with respect to safety. Although considerable efforts have been made, it has not been possible, to date, to achieve an agreement between operators, manufacturers, licensing authorities, independent experts, and other specialists about the safety related classification of the manifold pressure bearing parts in nuclear power stations. The background of this extremely regrettable situation is explained. In the last part of the paper the author suggests a simple and clear safety philosophy for pressure components in nuclear power stations. This philosophy is orientated both on Safety Regulations of the Radiation Protection Decree ('Strahlenschutzverordnung') of the 13th October 1976 and on the Safety Criteria for Nuclear Power Stations from 21st October 1977. Only a simple, clear framework can make a contribution to the further improvement of the already exceptional safety of nuclear facilities and to the removal of obstacles in the licensing procedure which, taken as a whole, tie up skilled personnel to a senseless degree, involve considerable financial expenditure, and have no relevance for the safety of nuclear power plants. (orig.) [de

Safety in transport and storage of radioactive materials

International Nuclear Information System (INIS)

Mezrahi, A.; Xavier, A.M.

1987-01-01

The increasing utilization of radioisotopes in Industrial, Medical and Research Facilities as well as the processing of Nuclear Materials involve transport activities in a routine basis. The present work has the following main objectives: I) the identification of the safety aspects related to handling, transport and storage of radioactive materials; II) the orientation of the personnel responsible for the radiological safety of Radioactive Installations viewing the elaboration and implementation of procedures to minimize accidents; III) the report of case-examples of accidents that have occured in Brazil due to non-compliance with Transport Regulations. (author) [pt

Safety design

International Nuclear Information System (INIS)

Kunitomi, Kazuhiko; Shiozawa, Shusaku

2004-01-01

JAERI established the safety design philosophy of the HTTR based on that of current reactors such as LWR in Japan, considering inherent safety features of the HTTR. The strategy of defense in depth was implemented so that the safety engineering functions such as control of reactivity, removal of residual heat and confinement of fission products shall be well performed to ensure safety. However, unlike the LWR, the inherent design features of the high-temperature gas-cooled reactor (HTGR) enables the HTTR meet stringent regulatory criteria without much dependence on active safety systems. On the other hand, the safety in an accident typical to the HTGR such as the depressurization accident initiated by a primary pipe rupture shall be ensured. The safety design philosophy of the HTTR considers these unique features appropriately and is expected to be the basis for future Japanese HTGRs. This paper describes the safety design philosophy and safety evaluation procedure of the HTTR especially focusing on unique considerations to the HTTR. Also, experiences obtained from an HTTR safety review and R and D needs for establishing the safety philosophy for the future HTGRs are reported

TH-C-18C-01: MRI Safety

Energy Technology Data Exchange (ETDEWEB)

Pooley, R [Mayo Clinic, Jacksonville, FL (United States); Bernstein, M; Shu, Y; Gorny, K; Felmlee, J [Mayo Clinic, Rochester, MN (United States); Panda, A [Mayo Clinic, Arizona, Scottsdale, AZ (United States)

2014-06-15

Clinical diagnostic medical physicists may be responsible for implementing and maintaining a comprehensive MR safety program. Accrediting bodies including the ACR, IAC, Radsite and The Joint Commission each include aspects of MR Safety into their imaging accreditation programs; MIPPA regulations further raise the significance of non-compliance. In addition, The Joint Commission recently announced New and Revised Diagnostic Imaging Standards for accredited health care organizations which include aspects of MR Safety. Hospitals and clinics look to the physicist to understand guidelines, regulations and accreditation requirements related to MR safety. The clinical medical physicist plays a significant role in a clinical practice by understanding the physical basis for the risks and acting as a facilitator to successfully implement a safety program that provides well-planned siting, allows for the safe scanning of certain implanted devices, and helps radiologists manage specific patient exams. The MRI scanning of specific devices will be discussed including cardiac pacemakers and neurostimulators such as deep brain stimulators. Furthermore for sites involved in MR guided interventional procedures, the MR physicist plays an essential role to establish safe practices. Creating a framework for a safe MRI practice includes the review of actual safety incidents or close calls to determine methods for prevention in the future. Learning Objectives: Understand the requirements and recommendations related to MR safety from accrediting bodies and federal regulations. Understand the Medical Physicist's roles to ensure MR Safety. Identify best practices for dealing with implanted devices, including pacemakers and deep brain stimulators. Review aspects of MR safety involved in an MR guided interventional environment. Understand the important MR safety aspects in actual safety incidents or near misses.

[A systemic risk analysis of hospital management processes by medical employees--an effective basis for improving patient safety].

Science.gov (United States)

Sobottka, Stephan B; Eberlein-Gonska, Maria; Schackert, Gabriele; Töpfer, Armin

2009-01-01

Due to the knowledge gap that exists between patients and health care staff the quality of medical treatment usually cannot be assessed securely by patients. For an optimization of safety in treatment-related processes of medical care, the medical staff needs to be actively involved in preventive and proactive quality management. Using voluntary, confidential and non-punitive systematic employee surveys, vulnerable topics and areas in patient care revealing preventable risks can be identified at an early stage. Preventive measures to continuously optimize treatment quality can be defined by creating a risk portfolio and a priority list of vulnerable topics. Whereas critical incident reporting systems are suitable for continuous risk assessment by detecting safety-relevant single events, employee surveys permit to conduct a systematic risk analysis of all treatment-related processes of patient care at any given point in time.

Ultrasonic data acquisition installation for basis and in-service testing of nuclear pressure vessels

International Nuclear Information System (INIS)

Gutmann, G.; Engl, G.

1976-01-01

The safety of nuclear installations requires continuous safety inspections during construction and operation. Essential parts of this safety inspection are the basis and in-line inspections. For this purpose installation systems are used which allow an optimal statement to be made regarding the conditions of tested components

Safety assessment for facilities and activities. General safety requirements. Pt. 4

International Nuclear Information System (INIS)

2009-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF 6 ; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

International Nuclear Information System (INIS)

2009-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

International Nuclear Information System (INIS)

2010-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

International Nuclear Information System (INIS)

2009-01-01

The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation.? read more The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are

Food safety performance indicators to benchmark food safety output of food safety management systems.

Science.gov (United States)

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

2010-07-31

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

Reactor system safety assurance

International Nuclear Information System (INIS)

Mattson, R.J.

1984-01-01

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

Unites States position paper on sodium fires. Design basis and testing

International Nuclear Information System (INIS)

Lancet, R.T.; Johnson, R.P.; Matlin, E.; Vaughan, E.U.; Fields, D.E.; Glueckler, E.; McCormack, J.D.; Miller, C.W.; Pedersen, D.R.

1989-01-01

This paper focuses on designs, analyses, and tests performed since the last Sodium Fires Meeting of the IAEA International Working Group on Fast Reactors in May 1982. Since the U.S. Liquid Metal Reactor (LMR) program is focused on the two advanced LMRs, SAFR and PRISM, the paper relates this work to these designs. First, the design philosophy and approach taken by these advanced pool reactors are described. This includes methods of leak detection, the design basis leaks, and passive accommodation of sodium fires. Then the small- and large-scale sodium fire tests performed in support of the Clinch River Breeder Reactor Plant (CRBRP) program, including post-accident cleanup, are presented and related to the advanced LMR designs. Next, the assessment and behavior of the aerosols generated are discussed including generation rate, behavior within structures, release and dispersal, and deposition on safety-grade equipment. Finally, the impact of these aerosols on the performance of safety-grade decay heat removal heat exchange surfaces is discussed including some test results as well as planned tests. (author)

Investigation on regulatory requirements for radiation safety management

International Nuclear Information System (INIS)

Han, Eun Ok; Choi, Yoon Seok; Cho, Dae Hyung

2013-01-01

NRC recognizes that efficient management of radiation safety plan is an important factor to achieve radiation safety service. In case of Korea, the contents to perform the actual radiation safety management are legally contained in radiation safety management reports based on the Nuclear Safety Act. It is to prioritize the importance of safety regulations in each sector in accordance with the current situation of radiation and radioactive isotopes-used industry and to provide a basis for deriving safety requirements and safety regulations system maintenance by the priority of radiation safety management regulations. It would be helpful to achieve regulations to conform to reality based on international standards if consistent safety requirements is developed for domestic users, national standards and international standards on the basis of the results of questions answered by radiation safety managers, who lead on-site radiation safety management, about the priority of important factors in radioactive sources use, sales, production, moving user companies, to check whether derived configuration requirements for radiation safety management are suitable for domestic status

International exchange of safety and licensing information

International Nuclear Information System (INIS)

Lafleur, J.D. Jr.; Hauber, R.D.; Chenier, D.M.

1977-01-01

A network of formal and informal bilateral arrangements for the exchange of nuclear safety information is being established by the U.S. Nuclear Regulatory Commission. For developing countries, such arrangements can provide ready access to the extensive, fully documented safety analyses and safety research results that NRC has accumulated. NRC has been receiving foreign visitors at a rate of about 500 per year, largely for discussions of safety and licensing questions related to light water reactors. Exchanges also are taking place on the safety of advanced reactors. A special interest of the NRC is in providing for reciprocal communicaion, at the earliest possible time, of important problems, decisions and other actions on nuclear safety matters. For example, it is essential that a newly-discovered problem in a nuclear reactor be brought immediately to the attention of other governments which are responsible for the safety of similar reactors. Definite progress has been made in the U.S. Freedom of Information Act. Certain exchanges have taken place on this basis. Experience in the establishment and operation of NRC's bilateral exchange arrangements is summarized. A typical exchange with the regulatory authority of country building its first power reactor is described

Report on safety related occurrences and reactor trips July 1, 1977 - December 31, 1977

International Nuclear Information System (INIS)

Andermo, L.; Sundman, B.

1974-04-01

This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1977 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 48 safety related occurrences and 49 reactor trips have been reported to the Nuclear Power Inspectorate. Included is also one incident June 21 in Barsebaeck 2 which was not included in the last compilation of occurrences. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips have increased nearly 30% since the last period. Those occurred during power operation however, were less. More than 50% of the reactor trips happened in the shutdown condition. The fact that even small deviations from prescribed operation result in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences withou02068NRM 0000169 450

Using naturalistic driving data to explore the association between traffic safety-related events and crash risk at driver level.

Science.gov (United States)

Wu, Kun-Feng; Aguero-Valverde, Jonathan; Jovanis, Paul P

2014-11-01

There has been considerable research conducted over the last 40 years using traffic safety-related events to support road safety analyses. Dating back to traffic conflict studies from the 1960s these observational studies of driver behavior have been criticized due to: poor quality data; lack of available and useful exposure measures linked to the observations; the incomparability of self-reported safety-related events; and, the difficulty in assessing culpability for safety-related events. This study seeks to explore the relationships between driver characteristics and traffic safety-related events, and between traffic safety-related events and crash involvement while mitigating some of those limitations. The Virginia Tech Transportation Institute 100-Car Naturalistic Driving Study dataset, in which the participants' vehicles were instrumented with various cameras and sensors during the study period, was used for this study. The study data set includes 90 drivers observed for 12-13 months driving. This study focuses on single vehicle run-off-road safety-related events only, including 14 crashes and 182 safety-related events (30 near crashes, and 152 crash-relevant incidents). Among the findings are: (1) drivers under age 25 are significantly more likely to be involved in safety-related events and crashes; and (2) significantly positive correlations exist between crashes, near crashes, and crash-relevant incidents. Although there is still much to learn about the factors affecting the positive correlation between safety-related events and crashes, a Bayesian multivariate Poisson log-normal model is shown to be useful to quantify the associations between safety-related events and crash risk while controlling for driver characteristics. Copyright © 2014 Elsevier Ltd. All rights reserved.

Scientific and technical basis of safety increase measures at NPPs with WWER

International Nuclear Information System (INIS)

Skalozubov, V.I.; Klyuchnikov, A.A.; Komarov, Yu.A.; Shavlakov, A.V.

2010-01-01

This monograph presents the original development of the authors on scientific and technical substantiation of foreground modern measures on safety increase at nuclear power plants with water-water reactors: development and implementation of operative diagnostic system for thermo acoustical instability of reactor core, substantituation of performance capacity and reliability of fast-acting reducing units systems and regulation systems of reactor emergency cooling at control of dominant for safety accidents.