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Sample records for safety studies assessment

  1. Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

    International Nuclear Information System (INIS)

    Oh, Kyemin; Kang, Myoung-suk; Heo, Gyunyoung; Kim, Hyoung-chan

    2014-01-01

    Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

  2. Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Kyemin; Kang, Myoung-suk [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Heo, Gyunyoung, E-mail: gheo@khu.ac.kr [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Kim, Hyoung-chan [National Fusion Research Institute, Daejeon-si 305-333 (Korea, Republic of)

    2014-10-15

    Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

  3. Human factors in safety assessment. Safety culture assessment

    International Nuclear Information System (INIS)

    Zhang Li; Deng Zhiliang; Wang Yiqun; Huang Weigang

    1996-01-01

    This paper analyses the present conditions and problems in enterprises safety assessment, and introduces the characteristics and effects of safety culture. The authors think that safety culture must be used as a 'soul' to form the pattern of modern safety management. Furthermore, they propose that the human safety and synthetic safety management assessment in a system should be changed into safety culture assessment. Finally, the assessment indicators are discussed

  4. Safety Assessment for Decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-06-15

    In the past few decades, international guidance has been developed on methods for assessing the safety of predisposal and disposal facilities for radioactive waste. More recently, it has been recognized that there is also a need for specific guidance on safety assessment in the context of decommissioning nuclear facilities. The importance of safety during decommissioning was highlighted at the International Conference on Safe Decommissioning for Nuclear Activities held in Berlin in 2002 and at the First Review Meeting of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management in 2003. At its June 2004 meeting, the Board of Governors of the IAEA approved the International Action Plan on Decommissioning of Nuclear Facilities (GOV/2004/40), which called on the IAEA to: ''establish a forum for the sharing and exchange of national information and experience on the application of safety assessment in the context of decommissioning and provide a means to convey this information to other interested parties, also drawing on the work of other international organizations in this area''. In response, in November 2004, the IAEA launched the international project Evaluation and Demonstration of Safety for Decommissioning of Facilities Using Radioactive Material (DeSa) with the following objectives: -To develop a harmonized approach to safety assessment and to define the elements of safety assessment for decommissioning, including the application of a graded approach; -To investigate the practical applicability of the methodology and performance of safety assessments for the decommissioning of various types of facility through a selected number of test cases; -To investigate approaches for the review of safety assessments for decommissioning activities and the development of a regulatory approach for reviewing safety assessments for decommissioning activities and as a basis for regulatory decision making; -To provide a forum

  5. A study on safety assessment methodology for a vitrification plant

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Y. C.; Lee, G. S.; Choi, Y. C.; Kim, G. H. [Yonsei Univ., Seoul (Korea, Republic of)

    2002-03-15

    In this study, the technical and regulatory status of radioactive waste vitrification technologies in foreign and domestic plants is investigated and analyzed, and then significant factors are suggested which must be contained in the final technical guideline or standard for the safety assessment of vitrification plants. Also, the methods to estimate the stability of vitrified waste forms are suggested with property analysis of them. The contents and scope of the study are summarized as follows : survey of the status on radioactive waste vitrification technologies in foreign and domestic plants, survey of the characterization methodology for radioactive waste form, analysis of stability for vitrified waste forms, survey and analysis of technical standards and regulations concerned with them in foreign and domestic plants, suggestion of significant factors for the safety assessment of vitrification plants, submission of regulated technical standard on radioactive waste vitrification plats.

  6. Correlation between safety climate and contractor safety assessment programs in construction.

    Science.gov (United States)

    Sparer, Emily H; Murphy, Lauren A; Taylor, Kathryn M; Dennerlein, Jack T

    2013-12-01

    Contractor safety assessment programs (CSAPs) measure safety performance by integrating multiple data sources together; however, the relationship between these measures of safety performance and safety climate within the construction industry is unknown. Four hundred and one construction workers employed by 68 companies on 26 sites and 11 safety managers employed by 11 companies completed brief surveys containing a nine-item safety climate scale developed for the construction industry. CSAP scores from ConstructSecure, Inc., an online CSAP database, classified these 68 companies as high or low scorers, with the median score of the sample population as the threshold. Spearman rank correlations evaluated the association between the CSAP score and the safety climate score at the individual level, as well as with various grouping methodologies. In addition, Spearman correlations evaluated the comparison between manager-assessed safety climate and worker-assessed safety climate. There were no statistically significant differences between safety climate scores reported by workers in the high and low CSAP groups. There were, at best, weak correlations between workers' safety climate scores and the company CSAP scores, with marginal statistical significance with two groupings of the data. There were also no significant differences between the manager-assessed safety climate and the worker-assessed safety climate scores. A CSAP safety performance score does not appear to capture safety climate, as measured in this study. The nature of safety climate in construction is complex, which may be reflective of the challenges in measuring safety climate within this industry. Am. J. Ind. Med. 56:1463-1472, 2013. © 2013 Wiley Periodicals, Inc. © 2013 Wiley Periodicals, Inc.

  7. Development of a New Safety Culture Assessment Method for Nuclear Power Plants (NPPs) (A study to suggest a new safety culture assessment method in nuclear power plants)

    International Nuclear Information System (INIS)

    Han, Sang Min; Seong, Poong Hyun

    2014-01-01

    This study is conducted to suggest a new safety culture assessment method in nuclear power plants. Criteria with various existing safety culture analysis methods are united, and reliability analysis methods are applied. The concept of the most representative methods, Fault Tree Analysis (FTA) and Failure Mode and Effect Analysis (FMEA), are adopted to assess safety culture. Through this application, it is expected that the suggested method will bring results with convenience and objectiveness

  8. Development of a New Safety Culture Assessment Method for Nuclear Power Plants (NPPs) (A study to suggest a new safety culture assessment method in nuclear power plants)

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sang Min; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

    2014-08-15

    This study is conducted to suggest a new safety culture assessment method in nuclear power plants. Criteria with various existing safety culture analysis methods are united, and reliability analysis methods are applied. The concept of the most representative methods, Fault Tree Analysis (FTA) and Failure Mode and Effect Analysis (FMEA), are adopted to assess safety culture. Through this application, it is expected that the suggested method will bring results with convenience and objectiveness.

  9. A Methodology To Incorporate The Safety Culture Into Probabilistic Safety Assessments

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sunghyun; Kim, Namyeong; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

    2015-10-15

    In order to incorporate organizational factors into PSA, a methodology needs to be developed. Using the AHP to weigh organizational factors as well as the SLIM to rate those factors, a methodology is introduced in this study. The safety issues related to nuclear safety culture have occurred increasingly. The quantification tool has to be developed in order to include the organizational factor into Probabilistic Safety Assessments. In this study, the state-of-the-art for the organizational evaluation methodologies has been surveyed. This study includes the research for organizational factors, maintenance process, maintenance process analysis models, a quantitative methodology using Analytic Hierarchy Process, Success Likelihood Index Methodology. The purpose of this study is to develop a methodology to incorporate the safety culture into PSA for obtaining more objective risk than before. The organizational factor considered in nuclear safety culture might affect the potential risk of human error and hardware-failure. The safety culture impact index to monitor the plant safety culture can be assessed by applying the developed methodology into a nuclear power plant.

  10. A Methodology To Incorporate The Safety Culture Into Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Park, Sunghyun; Kim, Namyeong; Jae, Moosung

    2015-01-01

    In order to incorporate organizational factors into PSA, a methodology needs to be developed. Using the AHP to weigh organizational factors as well as the SLIM to rate those factors, a methodology is introduced in this study. The safety issues related to nuclear safety culture have occurred increasingly. The quantification tool has to be developed in order to include the organizational factor into Probabilistic Safety Assessments. In this study, the state-of-the-art for the organizational evaluation methodologies has been surveyed. This study includes the research for organizational factors, maintenance process, maintenance process analysis models, a quantitative methodology using Analytic Hierarchy Process, Success Likelihood Index Methodology. The purpose of this study is to develop a methodology to incorporate the safety culture into PSA for obtaining more objective risk than before. The organizational factor considered in nuclear safety culture might affect the potential risk of human error and hardware-failure. The safety culture impact index to monitor the plant safety culture can be assessed by applying the developed methodology into a nuclear power plant

  11. Assessment of Electrical Safety Beliefs and Practices: A Case Study

    Directory of Open Access Journals (Sweden)

    S. Boubaker

    2017-12-01

    Full Text Available In this paper, the electrical safety beliefs and practices in Hail region, Saudi Arabia, have been assessed. Based on legislative recommendations and rules applied in Saudi Arabia, on official statistics regarding the electricity-caused accidents and on the analysis of more than 200 photos captured in Hail (related to electrical safety, a questionnaire composed of 36 questions (10 for the respondents information, 16 for the home safety culture and 10 for the electrical devices purchasing culture has been devised and distributed to residents. 228 responses have been collected and analyzed. Using a scale similar to the one adopted for a university student GPA calculation, the electrical safety level (ESL in Hail region has been found to be 0.76 (in a scale of 4 points which is a very low score and indicates a poor electrical safety culture. Several recommendations involving different competent authorities have been proposed. Future work will concern the assessment of safety in industrial companies in Hail region.

  12. HSE's safety assessment principles for criticality safety

    International Nuclear Information System (INIS)

    Simister, D N; Finnerty, M D; Warburton, S J; Thomas, E A; Macphail, M R

    2008-01-01

    The Health and Safety Executive (HSE) published its revised Safety Assessment Principles for Nuclear Facilities (SAPs) in December 2006. The SAPs are primarily intended for use by HSE's inspectors when judging the adequacy of safety cases for nuclear facilities. The revised SAPs relate to all aspects of safety in nuclear facilities including the technical discipline of criticality safety. The purpose of this paper is to set out for the benefit of a wider audience some of the thinking behind the final published words and to provide an insight into the development of UK regulatory guidance. The paper notes that it is HSE's intention that the Safety Assessment Principles should be viewed as a reflection of good practice in the context of interpreting primary legislation such as the requirements under site licence conditions for arrangements for producing an adequate safety case and for producing a suitable and sufficient risk assessment under the Ionising Radiations Regulations 1999 (SI1999/3232 www.opsi.gov.uk/si/si1999/uksi_19993232_en.pdf). (memorandum)

  13. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    This publication supports the Safety Requirements on the Safety of Nuclear Power Plants: Design. This Safety Guide was prepared on the basis of a systematic review of all the relevant publications including the Safety Fundamentals, Safety of Nuclear Power Plants: Design, current and ongoing revisions of other Safety Guides, INSAG reports and other publications that have addressed the safety of nuclear power plants. This Safety Guide also provides guidance for Contracting Parties to the Convention on Nuclear Safety in meeting their obligations under Article 14 on Assessment and Verification of Safety. The Safety Requirements publication entitled Safety of Nuclear Power Plants: Design states that a comprehensive safety assessment and an independent verification of the safety assessment shall be carried out before the design is submitted to the regulatory body. This publication provides guidance on how this requirement should be met. This Safety Guide provides recommendations to designers for carrying out a safety assessment during the initial design process and design modifications, as well as to the operating organization in carrying out independent verification of the safety assessment of new nuclear power plants with a new or already existing design. The recommendations for performing a safety assessment are suitable also as guidance for the safety review of an existing plant. The objective of reviewing existing plants against current standards and practices is to determine whether there are any deviations which would have an impact on plant safety. The methods and the recommendations of this Safety Guide can also be used by regulatory bodies for the conduct of the regulatory review and assessment. Although most recommendations of this Safety Guide are general and applicable to all types of nuclear reactors, some specific recommendations and examples apply mostly to water cooled reactors. Terms such as 'safety assessment', 'safety analysis' and 'independent

  14. A Methodology for Safety Culture Impact Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Han, Kiyoon; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of)

    2014-05-15

    The purpose of this study is to develop methodology for assessing safety culture impact on nuclear power plants. A new methodology for assessing safety culture impact index has been developed and applied for the reference nuclear power plants. The developed SCII model might contribute to comparing the level of safety culture among nuclear power plants as well as to improving the safety of nuclear power plants. Safety culture is defined to be fundamental attitudes and behaviors of the plant staff which demonstrate that nuclear safety is the most important consideration in all activities conducted in nuclear power operation. Through several accidents of nuclear power plant including the Fukusima Daiichi in 2011 and Chernovyl accidents in 1986, the safety of nuclear power plant is emerging into a matter of interest. From the accident review report, it can be easily found out that safety culture is important and one of dominant contributors to accidents. However, the impact methodology for assessing safety culture has not been established analytically yet. It is difficult to develop the methodology for assessing safety culture impact quantitatively.

  15. A Methodology for Safety Culture Impact Assessment

    International Nuclear Information System (INIS)

    Han, Kiyoon; Jae, Moosung

    2014-01-01

    The purpose of this study is to develop methodology for assessing safety culture impact on nuclear power plants. A new methodology for assessing safety culture impact index has been developed and applied for the reference nuclear power plants. The developed SCII model might contribute to comparing the level of safety culture among nuclear power plants as well as to improving the safety of nuclear power plants. Safety culture is defined to be fundamental attitudes and behaviors of the plant staff which demonstrate that nuclear safety is the most important consideration in all activities conducted in nuclear power operation. Through several accidents of nuclear power plant including the Fukusima Daiichi in 2011 and Chernovyl accidents in 1986, the safety of nuclear power plant is emerging into a matter of interest. From the accident review report, it can be easily found out that safety culture is important and one of dominant contributors to accidents. However, the impact methodology for assessing safety culture has not been established analytically yet. It is difficult to develop the methodology for assessing safety culture impact quantitatively

  16. A new assessment method for demonstrating the sufficiency of the safety assessment and the safety margins of the geological disposal system

    International Nuclear Information System (INIS)

    Ohi, Takao; Kawasaki, Daisuke; Chiba, Tamotsu; Takase, Toshio; Hane, Koji

    2013-01-01

    A new method for demonstrating the sufficiency of the safety assessment and safety margins of the geological disposal system has been developed. The method is based on an existing comprehensive sensitivity analysis method and can systematically identify the successful conditions, under which the dose rate does not exceed specified safety criteria, using analytical solutions for nuclide migration and the results of a statistical analysis. The successful conditions were identified using three major variables. Furthermore, the successful conditions at the level of factors or parameters were obtained using relational equations between the variables and the factors or parameters making up these variables. In this study, the method was applied to the safety assessment of the geological disposal of transuranic waste in Japan. Based on the system response characteristics obtained from analytical solutions and on the successful conditions, the classification of the analytical conditions, the sufficiency of the safety assessment and the safety margins of the disposal system were then demonstrated. A new assessment procedure incorporating this method into the existing safety assessment approach is proposed in this study. Using this procedure, it is possible to conduct a series of safety assessment activities in a logical manner. (author)

  17. Assessment of radiation safety awareness among nuclear medicine nurses: a pilot study

    International Nuclear Information System (INIS)

    Yunus, N A; Abdullah, M H R O; Said, M A; Ch'ng, P E

    2014-01-01

    All nuclear medicine nurses need to have some knowledge and awareness on radiation safety. At present, there is no study to address this issue in Malaysia. The aims of this study were (1) to determine the level of knowledge and awareness on radiation safety among nuclear medicine nurses at Putrajaya Hospital in Malaysia and (2) to assess the effectiveness of a training program provided by the hospital to increase the knowledge and awareness of the nuclear medicine nurses. A total of 27 respondents attending a training program on radiation safety were asked to complete a questionnaire. The questionnaire consists 16 items and were categorized into two main areas, namely general radiation knowledge and radiation safety. Survey data were collected before and after the training and were analyzed using descriptive statistics and paired sample t-test. Respondents were scored out of a total of 16 marks with 8 marks for each area. The findings showed that the range of total scores obtained by the nuclear medicine nurses before and after the training were 6-14 (with a mean score of 11.19) and 13-16 marks (with a mean score of 14.85), respectively. Findings also revealed that the mean score for the area of general radiation knowledge (7.59) was higher than that of the radiation safety (7.26). Currently, the knowledge and awareness on radiation safety among the nuclear medicine nurses are at the moderate level. It is recommended that a national study be conducted to assess and increase the level of knowledge and awareness among all nuclear medicine nurses in Malaysia

  18. Assessment of radiation safety awareness among nuclear medicine nurses: a pilot study

    Science.gov (United States)

    Yunus, N. A.; Abdullah, M. H. R. O.; Said, M. A.; Ch'ng, P. E.

    2014-11-01

    All nuclear medicine nurses need to have some knowledge and awareness on radiation safety. At present, there is no study to address this issue in Malaysia. The aims of this study were (1) to determine the level of knowledge and awareness on radiation safety among nuclear medicine nurses at Putrajaya Hospital in Malaysia and (2) to assess the effectiveness of a training program provided by the hospital to increase the knowledge and awareness of the nuclear medicine nurses. A total of 27 respondents attending a training program on radiation safety were asked to complete a questionnaire. The questionnaire consists 16 items and were categorized into two main areas, namely general radiation knowledge and radiation safety. Survey data were collected before and after the training and were analyzed using descriptive statistics and paired sample t-test. Respondents were scored out of a total of 16 marks with 8 marks for each area. The findings showed that the range of total scores obtained by the nuclear medicine nurses before and after the training were 6-14 (with a mean score of 11.19) and 13-16 marks (with a mean score of 14.85), respectively. Findings also revealed that the mean score for the area of general radiation knowledge (7.59) was higher than that of the radiation safety (7.26). Currently, the knowledge and awareness on radiation safety among the nuclear medicine nurses are at the moderate level. It is recommended that a national study be conducted to assess and increase the level of knowledge and awareness among all nuclear medicine nurses in Malaysia.

  19. Safety assessment in plant layout design using indexing approach: implementing inherent safety perspective. Part 2-Domino Hazard Index and case study.

    Science.gov (United States)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-12-15

    The design of layout plans requires adequate assessment tools for the quantification of safety performance. The general focus of the present work is to introduce an inherent safety perspective at different points of the layout design process. In particular, index approaches for safety assessment and decision-making in the early stages of layout design are developed and discussed in this two-part contribution. Part 1 (accompanying paper) of the current work presents an integrated index approach for safety assessment of early plant layout. In the present paper (Part 2), an index for evaluation of the hazard related to the potential of domino effects is developed. The index considers the actual consequences of possible escalation scenarios and scores or ranks the subsequent accident propagation potential. The effects of inherent and passive protection measures are also assessed. The result is a rapid quantification of domino hazard potential that can provide substantial support for choices in the early stages of layout design. Additionally, a case study concerning selection among various layout options is presented and analyzed. The case study demonstrates the use and applicability of the indices developed in both parts of the current work and highlights the value of introducing inherent safety features early in layout design.

  20. The role of risk assessment and safety analysis in integrated safety assessments

    International Nuclear Information System (INIS)

    Niall, R.; Hunt, M.; Wierman, T.E.

    1990-01-01

    To ensure that the design and operation of both nuclear and non- nuclear hazardous facilities is acceptable, and meets all societal safety expectations, a rigorous deterministic and probabilistic assessment is necessary. An approach is introduced, founded on the concept of an ''Integrated Safety Assessment.'' It merges the commonly performed safety and risk analyses and uses them in concert to provide decision makers with the necessary depth of understanding to achieve ''adequacy.'' 3 refs., 1 fig

  1. A 2-year study of patient safety competency assessment in 29 clinical laboratories.

    Science.gov (United States)

    Reed, Robyn C; Kim, Sara; Farquharson, Kara; Astion, Michael L

    2008-06-01

    Competency assessment is critical for laboratory operations and is mandated by the Clinical Laboratory Improvement Amendments of 1988. However, no previous reports describe methods for assessing competency in patient safety. We developed and implemented a Web-based tool to assess performance of 875 laboratory staff from 29 laboratories in patient safety. Question categories included workplace culture, categorizing error, prioritization of patient safety interventions, strength of specific interventions, and general patient safety concepts. The mean score was 85.0%, with individual scores ranging from 56% to 100% and scores by category from 81.3% to 88.6%. Of the most difficult questions (laboratory technologists. Computer-based competency assessments help laboratories identify topics for continuing education in patient safety.

  2. Uncertainty analysis in safety assessment

    International Nuclear Information System (INIS)

    Lemos, Francisco Luiz de; Sullivan, Terry

    1997-01-01

    Nuclear waste disposal is a very complex subject which requires the study of many different fields of science, like hydro geology, meteorology, geochemistry, etc. In addition, the waste disposal facilities are designed to last for a very long period of time. Both of these conditions make safety assessment projections filled with uncertainty. This paper addresses approaches for treatment of uncertainties in the safety assessment modeling due to the variability of data and some current approaches used to deal with this problem. (author)

  3. A study on the methodology of probabilistic safety assessment for KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwan Seong; Kwon, Young Min; Lee, Yong Bum; Jeong, Hae Yong; Yang, Joon Eon; Ha, Kyu Suk; Hahn, Do Hee [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    Existing Probabilistic Safety Assessment(PSA) is a method for Light Water Reactor or Pressurized Heavy Water Reactor. Because KALIMER is different from these reactor, the new methodology of PSA need to be developed. In this paper, the PSA of Power Reactor Inherently Safety Module(PRISM) is analyzed, and Initiating Event such as Experiential Assessment, Logical Assessment and Failure Mode Effect Analysis(FMEA) is reviewed. Also, Pipe Damage Frequency Method is suggested for KALIMER. And the Reliability Physical method of Passive System, which is a chief safety system of KALIMER, is reviewed and its applicability is investigated. Finally, for the Preliminary PSA of KALIMER, Intermediate Heat Transfer System is analyzed. 23 refs., 10 figs., 13 tabs. (Author)

  4. A probabilistic safety assessment PEER review: Case study on the use of probabilistic safety assessment for safety decisions

    International Nuclear Information System (INIS)

    1989-10-01

    The purpose of this case study is to illustrate, using an actual example, the organizing and carrying out of an independent peer review of a draft full-scope (level 3) probabilistic safety assessment. The specific findings of the peer review are of less importance than the approach taken, the interaction between sponsor and study team, and the technical and administrative issues that can arise during a peer review. This case study will examine the following issues: how the scope of the peer review was established, based on how it was to be used by the review sponsoring body; how the level of effort was determined, and what this determination meant for the technical quality of the review; how the team of peer reviewers was selected; how the review itself was carried out; what findings were made; what was done with these findings by both the review sponsoring body and the PSA analysis team. 9 refs, 2 figs, 1 tab

  5. Probabilistic safety assessment

    International Nuclear Information System (INIS)

    Hoertner, H.; Schuetz, B.

    1982-09-01

    For the purpose of assessing applicability and informativeness on risk-analysis methods in licencing procedures under atomic law, the choice of instruments for probabilistic analysis, the problems in and experience gained in their application, and the discussion of safety goals with respect to such instruments are of paramount significance. Naturally, such a complex field can only be dealt with step by step, making contribution relative to specific problems. The report on hand shows the essentials of a 'stocktaking' of systems relability studies in the licencing procedure under atomic law and of an American report (NUREG-0739) on 'Quantitative Safety Goals'. (orig.) [de

  6. Assessment of Safety Culture

    International Nuclear Information System (INIS)

    Bilic Zabric, T.; Kavsek, D.

    2006-01-01

    A strong safety culture leads to more effective conduct of work and a sense of accountability among managers and employees, who should be given the opportunity to expand skills by training. The resources expended would thus result in tangible improvements in working practices and skills, which encourage further improvement of safety culture. In promoting an improved safety culture, NEK has emphasized both national and organizational culture with an appropriate balance of behavioural sciences and quality management systems approaches. In recent years there has been particular emphasis put on an increasing awareness of the contribution that human behavioural sciences can make to develop good safety practices. The purpose of an assessment of safety culture is to increase the awareness of the present culture, to serve as a basis for improvement and to keep track of the effects of change or improvement over a longer period of time. There is, however, no single approach that is suitable for all purposes and which can measure, simultaneously, all the intangible aspects of safety culture, i.e. the norms, values, beliefs, attitudes or the behaviours reflecting the culture. Various methods have their strengths and weaknesses. To prevent significant performance problems, self-assessment is used. Self-assessment is the process of identifying opportunities for improvement actively or, in some cases, weaknesses that could cause more serious errors or events. Self-assessments are an important input to the corrective action programme. NEK has developed questionnaires for safety culture self-assessment to obtain information that is representative of the whole organization. Questionnaires ensure a greater degree of anonymity, and create a less stressful situation for the respondent. Answers to questions represent the more apparent and conscious values and attitudes of the respondent. NEK proactively co-operates with WANO, INPO, IAEA in the areas of Safety Culture and Human

  7. Safety assessment, safety performance indicators at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Baji, C.; Vamos, G.; Toth, J.

    2001-01-01

    The Paks Nuclear Power Plant has been using different methods of safety assessment (event analysis, self-assessment, probabilistic safety analysis), including performance indicators characterizing both operational and safety performance since the early years of operation of the plant. Regarding the safety performance, the indicators include safety system performance, number of scrams, release of radioactive materials, number of safety significant events, industrial safety indicator, etc. The Paks NPP also reports a set of ten indicators to WANO Performance Indicator Programme which, among others, include safety related indicators as well. However, a more systematic approach to structuring and trending safety indicators is needed so that they can contribute to the enhancement of the operational safety. A more comprehensive set of indicators and a systematic evaluation process was introduced in 1996. The performance indicators framework proposed by the IAEA was adapted to Paks in this year to further improve the process. Safety culture assessment and characterizing safety culture is part of the assessment process. (author)

  8. Safety assessment in plant layout design using indexing approach: Implementing inherent safety perspective

    International Nuclear Information System (INIS)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-01-01

    Layout planning plays a key role in the inherent safety performance of process plants since this design feature controls the possibility of accidental chain-events and the magnitude of possible consequences. A lack of suitable methods to promote the effective implementation of inherent safety in layout design calls for the development of new techniques and methods. In the present paper, a safety assessment approach suitable for layout design in the critical early phase is proposed. The concept of inherent safety is implemented within this safety assessment; the approach is based on an integrated assessment of inherent safety guideword applicability within the constraints typically present in layout design. Application of these guidewords is evaluated along with unit hazards and control devices to quantitatively map the safety performance of different layout options. Moreover, the economic aspects related to safety and inherent safety are evaluated by the method. Specific sub-indices are developed within the integrated safety assessment system to analyze and quantify the hazard related to domino effects. The proposed approach is quick in application, auditable and shares a common framework applicable in other phases of the design lifecycle (e.g. process design). The present work is divided in two parts: Part 1 (current paper) presents the application of inherent safety guidelines in layout design and the index method for safety assessment; Part 2 (accompanying paper) describes the domino hazard sub-index and demonstrates the proposed approach with a case study, thus evidencing the introduction of inherent safety features in layout design

  9. IAEA safety requirements for safety assessment of fuel cycle facilities and activities

    International Nuclear Information System (INIS)

    Jones, G.

    2013-01-01

    The IAEA's Statute authorises the Agency to establish standards of safety for protection of health and minimisation of danger to life and property. In that respect, the IAEA has established a Safety Fundamentals publication which contains ten safety principles for ensuring the protection of workers, the public and the environment from the harmful effects of ionising radiation. A number of these principles require safety assessments to be carried out as a means of evaluating compliance with safety requirements for all nuclear facilities and activities and to determine the measures that need to be taken to ensure safety. The safety assessments are required to be carried out and documented by the organisation responsible for operating the facility or conducting the activity, are to be independently verified and are to be submitted to the regulatory body as part of the licensing or authorisation process. In addition to the principles of the Safety Fundamentals, the IAEA establishes requirements that must be met to ensure the protection of people and the environment and which are governed by the principles in the Safety Fundamentals. The IAEA's Safety Requirements publication 'Safety Assessment for Facilities and Activities', establishes the safety requirements that need to be fulfilled in conducting and maintaining safety assessments for the lifetime of facilities and activities, with specific attention to defence in depth and the requirement for a graded approach to the application of these safety requirements across the wide range of fuel cycle facilities and activities. Requirements for independent verification of the safety assessment that needs to be carried out by the operating organisation, including the requirement for the safety assessment to be periodically reviewed and updated are also covered. For many fuel cycle facilities and activities, environmental impact assessments and non-radiological risk assessments will be required. The

  10. Uncertainty analysis in safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lemos, Francisco Luiz de [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Sullivan, Terry [Brookhaven National Lab., Upton, NY (United States)

    1997-12-31

    Nuclear waste disposal is a very complex subject which requires the study of many different fields of science, like hydro geology, meteorology, geochemistry, etc. In addition, the waste disposal facilities are designed to last for a very long period of time. Both of these conditions make safety assessment projections filled with uncertainty. This paper addresses approaches for treatment of uncertainties in the safety assessment modeling due to the variability of data and some current approaches used to deal with this problem. (author) 13 refs.; e-mail: lemos at bnl.gov; sulliva1 at bnl.gov

  11. Experiences in assessing safety culture

    International Nuclear Information System (INIS)

    Spitalnik, J.

    2002-01-01

    Based on several Safety Culture self-assessment applications in nuclear organisations, the paper stresses relevant aspects to be considered when programming an assessment of this type. Reasons for assessing Safety Culture, basic principles to take into account, necessary resources, the importance of proper statistical analyses, the feed-back of results, and the setting up of action plans to enhance Safety Culture are discussed. (author)

  12. OSART Independent Safety Culture Assessment (ISCA) Guidelines

    International Nuclear Information System (INIS)

    2016-01-01

    Safety culture is understood as an important part of nuclear safety performance. This has been demonstrated by the analysis of significant events such as Chernobyl, Davis Besse, Vandellos II, Asco, Paks, Mihamma and Forsmark, among others. In order to enhance safety culture, one essential activity is to perform assessments. IAEA Safety Standard Series No. GS-R-3, The Management System for Facilitites and Activities, states requirements for continuous improvement of safety culture, of which self, peer and independent safety culture assessments constitute an essential part. In line with this requirement, the Independent Safety Culture Assessment (ISCA) module is offered as an add-on module to the IAEA Operational Safety Review Team (OSART) programme. The OSART programme provides advice and assistance to Member States to enhance the safety of nuclear power plants during commissioning and operation. By including the ISCA module in an OSART mission, the receiving organization benefits from the synergy between the technical and the safety culture aspects of the safety review. The joint operational safety and safety culture assessment provides the organization with the opportunity to better understand the interactions between technical, human, organizational and cultural aspects, helping the organization to take a systemic approach to safety through identifying actions that fully address the root causes of any identified issue. Safety culture assessments provide insight into the fundamental drivers that shape organizational patterns of behaviour, safety consciousness and safety performance. The complex nature of safety culture means that the analysis of the results of such assessments is not as straightforward as for other types of assessment. The benefits of the results of nuclear safety culture assessments are maximized only if appropriate tools and guidance for these assessments is used; hence, this comprehensive guideline has been developed. The methodology explained

  13. Metrics design for safety assessment

    NARCIS (Netherlands)

    Luo, Yaping; van den Brand, M.G.J.

    2016-01-01

    Context:In the safety domain, safety assessment is used to show that safety-critical systems meet the required safety objectives. This process is also referred to as safety assurance and certification. During this procedure, safety standards are used as development guidelines to keep the risk at an

  14. Safety assessment principles for nuclear plants

    International Nuclear Information System (INIS)

    1992-01-01

    The present Safety Assessment Principles result from the revision of those which were drawn up following a recommendation arising from the Sizewell-B enquiry. The principles presented here relate only to nuclear safety; there is a section on risks from normal operation and accident conditions and the standards against which those risks are assessed. A major part of the document deals with the principles that cover the design of nuclear plants. The revised Safety assessment principles are aimed primarily at the safety assessment of new nuclear plants but they will also be used in assessing existing plants. (UK)

  15. 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)

  16. Thinking of the safety assessment of HLW disposal

    International Nuclear Information System (INIS)

    Li Honghui; Zhao Shuaiwei; Liu Jianqin; Liu Wei; Wan Lei; Yang Zhongtian; An Hongxiang; Sun Qinghong

    2014-01-01

    The function and the research methods of safety assessment are discussed. Two methods about safety assessment and the requirement of safety assessment are introduced. The key parameters and influence factors in nuclide transport of safety assessment are specialized. The works will be done on safety assessment is discussed which will give some suggests for the development of safety assessment. (authors)

  17. ALARP considerations in criticality safety assessments

    International Nuclear Information System (INIS)

    Bowden, Russell L.; Barnes, Andrew; Thorne, Peter R.; Venner, Jack

    2003-01-01

    Demonstrating that the risk to the public and workers is As Low As Reasonably Practicable (ALARP) is a fundamental requirement of safety cases for nuclear facilities in the United Kingdom. This is embodied in the Safety Assessment Principles (SAPs) published by the Regulator, the essence of which is incorporated within the safety assessment processes of the various nuclear site licensees. The concept of ALARP within criticality safety assessments has taken some time to establish in the United Kingdom. In principle, the licensee is obliged to search for a deterministic criticality safety solution, such as safe geometry vessels and passive control features, rather than placing reliance on active measurement devices and plant administrative controls. This paper presents a consideration of some ALARP issues in relation to the development of criticality safety cases. The paper utilises some idealised examples covering a range of issues facing the criticality safety assessor, including new plant design, operational plant and decommissioning activities. These examples are used to outline the elements of the criticality safety cases and present a discussion of ALARP in the context of criticality safety assessments. (author)

  18. Safety culture assessment developed by JANTI

    International Nuclear Information System (INIS)

    Hamada, Jun

    2009-01-01

    Japan's JCO accident in September 1999 provided a real-life example of what can happen when insufficient attention is paid to safety culture. This accident brought to light the importance of safety culture and reinforced the movement to foster a safety culture. Despite this, accidents and inappropriate conduct have continued to occur. Therefore, there is a strong demand to instill a safety culture throughout the nuclear power industry. In this context, Japan's nuclear power regulator, the Nuclear and Industrial Safety Agency (NISA), decided to include in its safety inspections assessments of the safety culture found in power utilities' routine safety operations to get signs of deterioration in the organizational climate. In 2007, NISA constructed guidelines for their inspectors to carry out these assessments. At the same time, utilities have embarked on their own independent safety culture initiatives, such as revising their technical specifications and building effective PDCA cycle to promote safety culture. In concert with these developments, JANTI has also instituted safety culture assessments. (author)

  19. Healthcare professionals’ views of feedback on patient safety culture assessment.

    OpenAIRE

    Zwijnenberg, N.C.; Hendriks, M.; Hoogervorst-Schilp, J.; Wagner, C.

    2016-01-01

    Background: By assessing patient safety culture, healthcare providers can identify areas for improvement in patient safety culture. To achieve this, these assessment outcomes have to be relevant and presented clearly. The aim of our study was to explore healthcare professionals’ views on the feedback of a patient safety culture assessment. Methods: Twenty four hospitals participated in a patient safety culture assessment in 2012. Hospital departments received feedback in a report and on a web...

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

    International Nuclear Information System (INIS)

    Jeong, Jongtae; Choi, Jongwon; Kang, Chulhyung

    2012-04-01

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

  1. A study on methodologies for assessing safety critical network's risk impact on Nuclear Power Plant

    International Nuclear Information System (INIS)

    Lim, T. J.; Lee, H. J.; Park, S. K.; Seo, S. J.

    2006-08-01

    The objectives of this project is to investigate and study existing reliability analysis techniques for communication networks in order to develop reliability analysis models for Nuclear Power Plant's safety-critical networks. It is necessary to make a comprehensive survey of current methodologies for communication network reliability. Major outputs of the first year study are design characteristics of safety-critical communication networks, efficient algorithms for quantifying reliability of communication networks, and preliminary models for assessing reliability of safety-critical communication networks

  2. Safety and reliability assessment

    International Nuclear Information System (INIS)

    1979-01-01

    This report contains the papers delivered at the course on safety and reliability assessment held at the CSIR Conference Centre, Scientia, Pretoria. The following topics were discussed: safety standards; licensing; biological effects of radiation; what is a PWR; safety principles in the design of a nuclear reactor; radio-release analysis; quality assurance; the staffing, organisation and training for a nuclear power plant project; event trees, fault trees and probability; Automatic Protective Systems; sources of failure-rate data; interpretation of failure data; synthesis and reliability; quantification of human error in man-machine systems; dispersion of noxious substances through the atmosphere; criticality aspects of enrichment and recovery plants; and risk and hazard analysis. Extensive examples are given as well as case studies

  3. Safety assessment in plant layout design using indexing approach: Implementing inherent safety perspective

    International Nuclear Information System (INIS)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-01-01

    The design of layout plans requires adequate assessment tools for the quantification of safety performance. The general focus of the present work is to introduce an inherent safety perspective at different points of the layout design process. In particular, index approaches for safety assessment and decision-making in the early stages of layout design are developed and discussed in this two-part contribution. Part 1 (accompanying paper) of the current work presents an integrated index approach for safety assessment of early plant layout. In the present paper (Part 2), an index for evaluation of the hazard related to the potential of domino effects is developed. The index considers the actual consequences of possible escalation scenarios and scores or ranks the subsequent accident propagation potential. The effects of inherent and passive protection measures are also assessed. The result is a rapid quantification of domino hazard potential that can provide substantial support for choices in the early stages of layout design. Additionally, a case study concerning selection among various layout options is presented and analyzed. The case study demonstrates the use and applicability of the indices developed in both parts of the current work and highlights the value of introducing inherent safety features early in layout design

  4. Feasibility studies of safety assessment methods for programmable automation systems. Final report of the AVV project

    International Nuclear Information System (INIS)

    Haapanen, P.; Maskuniitty, M.; Pulkkinen, U.; Heikkinen, J.; Korhonen, J.; Tuulari, E.

    1995-10-01

    Feasibility studies of two different groups of methodologies for safety assessment of programmable automation systems has been executed at the Technical Research Centre of Finland (VTT). The studies concerned the dynamic testing methods and the fault tree (FT) and failure mode and effects analysis (FMEA) methods. In order to get real experience in the application of these methods, an experimental testing of two realistic pilot systems were executed and a FT/FMEA analysis of a programmable safety function accomplished. The purpose of the studies was not to assess the object systems, but to get experience in the application of methods and assess their potentials and development needs. (46 refs., 21 figs.)

  5. Health, safety and environmental unit performance assessment model under uncertainty (case study: steel industry).

    Science.gov (United States)

    Shamaii, Azin; Omidvari, Manouchehr; Lotfi, Farhad Hosseinzadeh

    2017-01-01

    Performance assessment is a critical objective of management systems. As a result of the non-deterministic and qualitative nature of performance indicators, assessments are likely to be influenced by evaluators' personal judgments. Furthermore, in developing countries, performance assessments by the Health, Safety and Environment (HSE) department are based solely on the number of accidents. A questionnaire is used to conduct the study in one of the largest steel production companies in Iran. With respect to health, safety, and environment, the results revealed that control of disease, fire hazards, and air pollution are of paramount importance, with coefficients of 0.057, 0.062, and 0.054, respectively. Furthermore, health and environment indicators were found to be the most common causes of poor performance. Finally, it was shown that HSE management systems can affect the majority of performance safety indicators in the short run, whereas health and environment indicators require longer periods of time. The objective of this study is to present an HSE-MS unit performance assessment model in steel industries. Moreover, we seek to answer the following question: what are the factors that affect HSE unit system in the steel industry? Also, for each factor, the extent of impact on the performance of the HSE management system in the organization is determined.

  6. Safety assessment for radioactive waste disposal facility

    International Nuclear Information System (INIS)

    Thanaletchumy Karuppiah; Mohd Abdul Wahab Yusof; Nik Marzuki Nik Ibrahim; Nurul Wahida Ahmad Khairuddin

    2008-08-01

    Safety assessments are used to evaluate the performance of a radioactive waste disposal facility and its impact on human health and the environment. This paper presents the overall information and methodology to carry out the safety assessment for a long term performance of a disposal system. A case study was also conducted to gain hands-on experience in the development and justification of scenarios, the formulation and implementation of models and the analysis of results. AMBER code using compartmental modeling approach was used to represent the migration and fate of contaminants in this training. This safety assessment is purely illustrative and it serves as a starting point for each development stage of a disposal facility. This assessment ultimately becomes more detail and specific as the facility evolves. (Author)

  7. Living probabilistic safety assessment (LPSA)

    International Nuclear Information System (INIS)

    1999-08-01

    Over the past few years many nuclear power plant organizations have performed probabilistic safety assessments (PSAs) to identify and understand key plant vulnerabilities. As a result of the availability of these PSA studies, there is a desire to use them to enhance plant safety and to operate the nuclear stations in the most efficient manner. PSA is an effective tool for this purpose as it assists plant management to target resources where the largest benefit to plant safety can be obtained. However, any PSA which is to be used in this way must have a credible and defensible basis. Thus, it is very important to have a high quality 'living PSA' accepted by the plant and the regulator. With this background in mind, the IAEA has prepared this report on Living Probabilistic Safety Assessment (LPSA) which addresses the updating, documentation, quality assurance, and management and organizational requirements for LPSA. Deficiencies in the areas addressed in this report would seriously reduce the adequacy of the LPSA as a tool to support decision making at NPPs. This report was reviewed by a working group during a Technical Committee Meeting on PSA Applications to Improve NPP Safety held in Madrid, Spain, from 23 to 27 February 1998

  8. 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.

  9. Complementary safety assessments - Report by the French Nuclear Safety Authority

    International Nuclear Information System (INIS)

    2011-12-01

    As an immediate consequence of the Fukushima accident, the French Authority of Nuclear Safety (ASN) launched a campaign of on-site inspections and asked operators (mainly EDF, AREVA and CEA) to make complementary assessments of the safety of the nuclear facilities they manage. The approach defined by ASN for the complementary safety assessments (CSA) is to study the behaviour of nuclear facilities in severe accidents situations caused by an off-site natural hazard according to accident scenarios exceeding the current baseline safety requirements. This approach can be broken into 2 phases: first conformity to current design and secondly an approach to the beyond design-basis scenarios built around the principle of defence in depth. 38 inspections were performed on issues linked to the causes of the Fukushima crisis. It appears that some sites have to reinforce the robustness of the heat sink. The CSA confirmed that the processes put into place at EDF to detect non-conformities were satisfactory. The complementary safety assessments demonstrated that the current seismic margins on the EDF nuclear reactors are satisfactory. With regard to flooding, the complementary safety assessments show that the complete reassessment carried out following the flooding of the Le Blayais nuclear power plant in 1999 offers the installations a high level of protection against the risk of flooding. Concerning the loss of electrical power supplies and the loss of cooling systems, the analysis of EDF's CSA reports showed that certain heat sink and electrical power supply loss scenarios can, if nothing is done, lead to core melt in just a few hours in the most unfavourable circumstances. As for nuclear facilities that are not power or experimental reactors, some difficulties have appeared to implement the CSA approach that was initially devised for reactors. Generally speaking, ASN considers that the safety of nuclear facilities must be made more robust to improbable risks which are not

  10. 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

  11. 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

  12. 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

  13. 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

  14. Safety Culture Monitoring: How to Assess Safety Culture in Real Time?

    International Nuclear Information System (INIS)

    Zronek, B.; Maryska, J.; Treslova, L.

    2016-01-01

    Do you know what is current level of safety culture in your company? Are you able to follow trend changes? Do you know what your recent issues are? Since safety culture is understood as vital part of nuclear industry daily life, it is crucial to know what the current level is. It is common to perform safety culture survey or ad hoc assessment. This contribution shares Temelin NPP, CEZ approach how to assess safety culture level permanently. Using behavioral related outputs of gap solving system, observation program, dedicated surveys, regulatory assessment, etc., allows creating real time safety culture monitoring without the need to perform any other activities. (author)

  15. Assessment of safety regulation using an artificial society

    International Nuclear Information System (INIS)

    Furuta, Kazuo; Nagase, Masaya

    2005-01-01

    This study proposes using an artificial society to assess impacts of safety regulation on the society. The artificial society used in this study is a multi-agent system, which consists of many agents representing companies. The agents cannot survive unless they get profits by producing some products. Safety regulation functions as the business environment, which the agents will evolve to fit to. We modeled this process of survival and adaptation by the genetic algorithm. Using the proposed model, case simulations were performed to compare various regulation styles, and some interesting insights were obtained how regulation style influences behavior of the agents and then productivity and safety level of the industry. In conclusion, an effective method for assessment of safety regulation has been developed, and then several insights were shown in this study

  16. Regulatory review of safety cases and safety assessments - associated challenges

    International Nuclear Information System (INIS)

    Bennett, D.G.; Ben Belfadhel, M.; Metcalf, P.E.

    2006-01-01

    Regulatory reviews of safety cases and safety assessments are essential for credible decision making on the licensing or authorization of radioactive waste disposal facilities. Regulatory review also plays an important role in developing the safety case and in establishing stakeholders' confidence in the safety of the facility. Reviews of safety cases for radioactive waste disposal facilities need to be conducted by suitably qualified and experienced staff, following systematic and well planned review processes. Regulatory reviews should be sufficiently comprehensive in their coverage of issues potentially affecting the safety of the disposal system, and should assess the safety case against clearly established criteria. The conclusions drawn from a regulatory review, and the rationale for them should be reproducible and documented in a transparent and traceable way. Many challenges are faced when conducting regulatory reviews of safety cases. Some of these relate to issues of project and programme management, and resources, while others derive from the inherent difficulties of assessing the potential long term future behaviour of engineered and environmental systems. The paper describes approaches to the conduct of regulatory reviews and discusses some of the challenges faced. (author)

  17. Intermediate probabilistic safety assessment approach for safety critical digital systems

    International Nuclear Information System (INIS)

    Taeyong, Sung; Hyun Gook, Kang

    2001-01-01

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

  18. Risk assessment of safety violations for coal mines

    Energy Technology Data Exchange (ETDEWEB)

    Megan Orsulaka; Vladislav Kecojevicb; Larry Graysona; Antonio Nietoa [Pennsylvania State University, University Park, PA (United States). Dept of Energy and Mineral Engineering

    2010-09-15

    This article presents an application of a risk assessment approach in characterising the risks associated with safety violations in underground bituminous mines in Pennsylvania using the Mine Safety and Health Administration (MSHA) citation database. The MSHA database on citations provides an opportunity to assess risks in mines through scrutiny of violations of mandatory safety standards. In this study, quantitative risk assessment is performed, which allows determination of the frequency of occurrence of safety violations (through associated citations) as well as the consequences of them in terms of penalty assessments. Focus is on establishing risk matrices on citation experiences of mines, which can give early indication of emerging potentially serious problems. The resulting frequency, consequence and risk rankings present valuable tools for prioritising resource allocations, determining control strategies, and could potentially contribute to more proactive prevention of incidents and injuries.

  19. Safety assessments for potential exposures

    International Nuclear Information System (INIS)

    Dunn, D.I.

    2012-04-01

    Safety Assessment of potential exposures have been carried out in major practices, namely: industrial radiography, gamma irradiators and electron accelerators used in industry and research, and radiotherapy. This paper focuses on reviewing safety assessment methodologies and using developed software to analyse radiological accidents, also review, and discuss these past accidents.The primary objective of the assessment is to assess the adequacy of planned or existing measures for protection and safety and to identify any additional measures that should be put in place. As such, both routine use of the source and the probability and magnitude of potential exposures arising from accidents or incidents should be considered. Where the assessment indicates that there is a realistic possibility of an accident affecting workers or members of the public or having consequences for the environment, the registrant or licensee should prepare a suitable emergency plan. A safety assessment for normal operation addresses all the conditions under which the radiation source operates as expected, including all phases of the lifetime of the source. Due account needs to be taken of the different factors and conditions that will apply during non-operational phases, such as installation, commissioning and maintenance. (author)

  20. Development of a Safety Assessment Information System for the Management of Periodic Safety Assessment Activities

    International Nuclear Information System (INIS)

    Song, Tae Young

    2007-01-01

    At present, the 10-year Periodic Safety Review(PSR) has been performing to confirm all the aspects of safety issues for all the operating plants in compliance with domestic nuclear law of article 23, subarticle 3. For each plant, in addition, Probabilistic Safety Assessment(PSA) and Severe Accident Management Guideline(SAMG) are being implemented and revised periodically to reflect the latest safety level according to principle fulfillment of severe accident policy statement. The assessment reports, as one of outcomes from these activities, are submitted into and reviewed by domestic regulatory body. During reviewing (in-office duty) and licensing (regulatory duty) process, a large number of outcomes of which most are the formal technical reports and licensing materials, are inevitably produced. Moreover, repeated review process over the plants can make them accumulated and produce a variety of documents additionally. This circumstance motivates to develop effective tool or system for the management of these reports and related technical documents for the future use in licensing process and for subsequent plant assessments. This paper presents the development status of Safety Assessment Information System(SAIS) which manages safety-related documents of PSR, PSA and SAMG for practical use for experienced engineers in charge of these areas

  1. Development of a Safety Assessment Information System for the Management of Periodic Safety Assessment Activities

    Energy Technology Data Exchange (ETDEWEB)

    Song, Tae Young [Nuclear Engineering and Technology Institute, Daejeon (Korea, Republic of)

    2007-07-01

    At present, the 10-year Periodic Safety Review(PSR) has been performing to confirm all the aspects of safety issues for all the operating plants in compliance with domestic nuclear law of article 23, subarticle 3. For each plant, in addition, Probabilistic Safety Assessment(PSA) and Severe Accident Management Guideline(SAMG) are being implemented and revised periodically to reflect the latest safety level according to principle fulfillment of severe accident policy statement. The assessment reports, as one of outcomes from these activities, are submitted into and reviewed by domestic regulatory body. During reviewing (in-office duty) and licensing (regulatory duty) process, a large number of outcomes of which most are the formal technical reports and licensing materials, are inevitably produced. Moreover, repeated review process over the plants can make them accumulated and produce a variety of documents additionally. This circumstance motivates to develop effective tool or system for the management of these reports and related technical documents for the future use in licensing process and for subsequent plant assessments. This paper presents the development status of Safety Assessment Information System(SAIS) which manages safety-related documents of PSR, PSA and SAMG for practical use for experienced engineers in charge of these areas.

  2. Safety assessment for spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Practice has been prepared as part of the IAEA's programme on the safety assessment of interim spent fuel storage facilities which are not an integral part of an operating nuclear power plant. This report provides general guidance on the safety assessment process, discussing both deterministic and probabilistic assessment methods. It describes the safety assessment process for normal operation and anticipated operational occurrences and also related to accident conditions. 10 refs, 2 tabs

  3. Safety assessment of smoke flavouring primary products by the European Food Safety Authority

    NARCIS (Netherlands)

    Theobald, A.; Arcella, D.; Carere, A.; Croera, C.; Engel, K.H.; Gott, D.; Gurtler, R.; Meier, D.; Pratt, I.; Rietjens, I.M.C.M.; Simon, R.; Walker, R.

    2012-01-01

    This paper summarises the safety assessments of eleven smoke flavouring primary products evaluated by the European Food Safety Authority (EFSA). Data on chemical composition, content of polyaromatic hydrocarbons and results of genotoxicity tests and subchronic toxicity studies are presented and

  4. Healthcare professionals? views on feedback of a patient safety culture assessment

    OpenAIRE

    Zwijnenberg, Nicolien C.; Hendriks, Michelle; Hoogervorst-Schilp, Janneke; Wagner, Cordula

    2016-01-01

    Background By assessing patient safety culture, healthcare providers can identify areas for improvement in patient safety culture. To achieve this, these assessment outcomes have to be relevant and presented clearly. The aim of our study was to explore healthcare professionals? views on the feedback of a patient safety culture assessment. Methods Twenty four hospitals participated in a patient safety culture assessment in 2012. Hospital departments received feedback in a report and on a websi...

  5. Consideration of aging in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Titina, B.; Cepin, M.

    2007-01-01

    Probabilistic safety assessment is a standardised tool for assessment of safety of nuclear power plants. It is a complement to the safety analyses. Standard probabilistic models of safety equipment assume component failure rate as a constant. Ageing of systems, structures and components can theoretically be included in new age-dependent probabilistic safety assessment, which generally causes the failure rate to be a function of age. New age-dependent probabilistic safety assessment models, which offer explicit calculation of the ageing effects, are developed. Several groups of components are considered which require their unique models: e.g. operating components e.g. stand-by components. The developed models on the component level are inserted into the models of the probabilistic safety assessment in order that the ageing effects are evaluated for complete systems. The preliminary results show that the lack of necessary data for consideration of ageing causes highly uncertain models and consequently the results. (author)

  6. Promoting and assessment of safety culture within regulatory body

    International Nuclear Information System (INIS)

    Awasthi, Sumit; Bhattacharya, D.; Koley, J.; Krishnamurthy, P.R.

    2015-01-01

    Regulators have an important role to play in assisting organizations under their jurisdiction to develop positive safety cultures. It is therefore essential for the regulator to have a robust safety culture as an inherent strategy and communication of this strategy to the organizations it supervises. Atomic Energy Regulatory Board (AERB) emphasizes every utility to institute a good safety culture during various stages of a NPP. The regulatory requirement for establishing organisational safety culture within utility at different stages are delineated in the various AERB safety codes which are presented in the paper. Although the review and assessment of the safety culture is a part of AERB’s continual safety supervision through existing review mechanism, AERB do not use any specific indicators for safety culture assessment. However, establishing and nurturing a good safety culture within AERB helps in encouraging the utility to institute the same. At the induction level AERB provides training to its staffs for regulatory orientation which include a specific course on safety culture. Subsequently, the junior staffs are mentored by seniors while involving them in various regulatory processes and putting them as observers during regulatory decision making process. Further, AERB established a formal procedure for assessing and improving safety culture within its staff as a management system process. The paper describes as a case study the above safety culture assessment process established within AERB

  7. Nirex safety assessment research programme: 1987/88

    International Nuclear Information System (INIS)

    George, D.; Hodgkinson, D.P.

    1987-01-01

    The Nirex Safety Assessment Research programme's objective is to provide information for the radiological safety case for disposing low-level and intermediate-level radioactive wastes in underground repositories. The programme covers a wide range of experimental studies and mathematical modelling for the near and far field. It attempts to develop a quantitative understanding of events and processes which have an impact on the safety of radioactive waste disposal. (U.K.)

  8. Regulatory review of safety cases and safety assessments for near surface

    International Nuclear Information System (INIS)

    Nys, V.

    2003-01-01

    The activities of the ASAM Regulatory Review Working Group are presented. Regulatory review of the safety assessment is made. It includes the regulatory review of post-closure safety assessment; safety case development and confidence building. The ISAM methodology is reviewed and SA system description is presented. Recommendations on the review process management are given

  9. Safety Auditing and Assessments

    Science.gov (United States)

    Goodin, James Ronald (Ronnie)

    2005-01-01

    Safety professionals typically do not engage in audits and independent assessments with the vigor as do our quality brethren. Taking advantage of industry and government experience conducting value added Independent Assessments or Audits benefits a safety program. Most other organizations simply call this process "internal audits." Sources of audit training are presented and compared. A relation of logic between audit techniques and mishap investigation is discussed. An example of an audit process is offered. Shortcomings and pitfalls of auditing are covered.

  10. Confidence building in safety assessments

    International Nuclear Information System (INIS)

    Grundfelt, Bertil

    1999-01-01

    Future generations should be adequately protected from damage caused by the present disposal of radioactive waste. This presentation discusses the core of safety and performance assessment: The demonstration and building of confidence that the disposal system meets the safety requirements stipulated by society. The major difficulty is to deal with risks in the very long time perspective of the thousands of years during which the waste is hazardous. Concern about these problems has stimulated the development of the safety assessment discipline. The presentation concentrates on two of the elements of safety assessment: (1) Uncertainty and sensitivity analysis, and (2) validation and review. Uncertainty is associated both with respect to what is the proper conceptual model and with respect to parameter values for a given model. A special kind of uncertainty derives from the variation of a property in space. Geostatistics is one approach to handling spatial variability. The simplest way of doing a sensitivity analysis is to offset the model parameters one by one and observe how the model output changes. The validity of the models and data used to make predictions is central to the credibility of safety assessments for radioactive waste repositories. There are several definitions of model validation. The presentation discusses it as a process and highlights some aspects of validation methodologies

  11. Safety assessment of novel foods and strategies to determine their safety in use

    International Nuclear Information System (INIS)

    Edwards, Gareth

    2005-01-01

    Safety assessment of novel foods requires a different approach to that traditionally used for the assessment of food chemicals. A case-by-case approach is needed which must be adapted to take account of the characteristics of the individual novel food. A thorough appraisal is required of the origin, production, compositional analysis, nutritional characteristics, any previous human exposure and the anticipated use of the food. The information should be compared with a traditional counterpart of the food if this is available. In some cases, a conclusion about the safety of the food may be reached on the basis of this information alone, whereas in other cases, it will help to identify any nutritional or toxicological testing that may be required to further investigate the safety of the food. The importance of nutritional evaluation cannot be over-emphasised. This is essential for the conduct of toxicological studies in order to avoid dietary imbalances, etc., that might lead to interpretation difficulties, but also in the context of its use as food and to assess the potential impact of the novel food on the human diet. The traditional approach used for chemicals, whereby an acceptable daily intake (ADI) is established with a large safety margin relative to the expected exposure, cannot be applied to foods. The assessment of safety in use should be based upon a thorough knowledge of the composition of the food, evidence from nutritional, toxicological and human studies, expected use of the food and its expected consumption. Safety equates to a reasonable certainty that no harm will result from intended uses under the anticipated conditions of consumption

  12. Assessment of Safety Condition in One of the Teaching Hospitals in Kermanshah (2015: A Case Study

    Directory of Open Access Journals (Sweden)

    Masod Ghanbari Kakavand

    2016-09-01

    Full Text Available Background & Aims of the Study: Many working conditions-related stress factors that can produce injuries and illnesses are important in hospital environments. So, the health and safety of nurses and patients from workplace-induced injuries and illnesses is important. In this study, we have assessed the safety condition of one of the teaching hospitals in Kermanshah (2015. Materials and Methods: This descriptive and cross-sectional study was conducted in one of the teaching hospital of Kermanshah University of medical sciences. For this aim a checklist was prepared based on the Occupational Safety and Health Administration's standards and Part 3 of the manual of National Building Regulations. These checklists comprised (The final checklist had 239 questions of 9 dimensions various sections of safety including; fire safety, building safety, electrical safety, emergency exit routes safety, heating and cooling equipment safety, operating room and laundry room and salty home safety. Eventually, using SPSS 16 and descriptive statistics, data were analyzed. Results: According to the results of this study, 66.6% of the units had poor safety and 33.4% of them were moderately safe. As well as, only ICU and CCU unit, heating and cooling equipment and operational room showed moderate compliance with safety requirements and other sections were poorly complied. Conclusion: The results of this study showed that safety conditions of hospital were not at favorable level. These poor safety statues can jeopardize patients and hospital personnel. Thus some interventions such as improvement of working conditions, compliance with safety acts and implementation of health, safety and environmental management system would be necessary.

  13. Risk assessment of safety data link and network communication in digital safety feature control system of nuclear power plant

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  14. Fundamental study on applicability of resilience index for system safety assessment

    International Nuclear Information System (INIS)

    Suzuki, Masaaki; Demachi, Kazuyuki; Murakami, Kenta

    2015-01-01

    We have developed a new index called Resilience index, which evaluate the reliability of system safety of nuclear power plant under severe accident by considering the capability to recover from the situation the system safety function was lost. In this paper, a detailed evaluation procedure for the Resilience index was described. System safety of a PWR plant under severe accident was then assessed according to the Resilience index concept to discuss applicability of the index. We found that the Resilience index successfully visualize the management capability, and therefore, resilience capability of a nuclear power plant. (author)

  15. Cannabis for the Management of Pain: Assessment of Safety Study (COMPASS).

    Science.gov (United States)

    Ware, Mark A; Wang, Tongtong; Shapiro, Stan; Collet, Jean-Paul

    2015-12-01

    Cannabis is widely used as a self-management strategy by patients with a wide range of symptoms and diseases including chronic non-cancer pain. The safety of cannabis use for medical purposes has not been systematically evaluated. We conducted a prospective cohort study to describe safety issues among individuals with chronic non-cancer pain. A standardized herbal cannabis product (12.5% tetrahydrocannabinol) was dispensed to eligible individuals for a 1-year period; controls were individuals with chronic pain from the same clinics who were not cannabis users. The primary outcome consisted of serious adverse events and non-serious adverse events. Secondary safety outcomes included pulmonary and neurocognitive function and standard hematology, biochemistry, renal, liver, and endocrine function. Secondary efficacy parameters included pain and other symptoms, mood, and quality of life. Two hundred and fifteen individuals with chronic pain were recruited to the cannabis group (141 current users and 58 ex-users) and 216 controls (chronic pain but no current cannabis use) from 7 clinics across Canada. The median daily cannabis dose was 2.5 g/d. There was no difference in risk of serious adverse events (adjusted incidence rate ratio = 1.08, 95% confidence interval = .57-2.04) between groups. Medical cannabis users were at increased risk of non-serious adverse events (adjusted incidence rate ratio = 1.73, 95% confidence interval = 1.41-2.13); most were mild to moderate. There were no differences in secondary safety assessments. Quality-controlled herbal cannabis, when used by patients with experience of cannabis use as part of a monitored treatment program over 1 year, appears to have a reasonable safety profile. Longer-term monitoring for functional outcomes is needed. The study was registered with www.controlled-trials.com (ISRCTN19449752). This study evaluated the safety of cannabis use by patients with chronic pain over 1 year. The study found that there was a higher

  16. Safety Management and Safety Culture Self Assessment of Kartini Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip, S., E-mail: syarip@batan.go.id [Centre for Accelerator and Material Process Technology, National Nuclear Energy Agency (BATAN), Yogyakarta (Indonesia)

    2014-10-15

    The self-assessment of safety culture and safety management status of Kartini research reactor is a step to foster safety culture and management by identifying good practices and areas for improvement, and also to improve reactor safety in a whole. The method used in this assessment is based on questionnaires provided by the Forum for Nuclear Cooperation in Asia (FNCA), then reviewed by experts. Based on the assessment and evaluation results, it can be concluded that there were several good practices in maintaining the safety status of Kartini reactor such as: reactor operators and radiation protection workers were aware and knowledgeable of the safety standards and policies that apply to their operation, readily accept constructive criticism from their management and from the inspectors of regulatory body that address safety performance. As a proof, for the last four years the number of inspection/audit findings from Regulatory Body (BAPETEN) tended to decrease while the reactor utilization and its operating hour increased. On the other hands there were also some comments and recommendations for improvement of reactor safety culture, such as that there should be more frequent open dialogues between employees and managers, to grow and attain a mutual support to achieve safety goals. (author)

  17. Preliminary investigation on reliability assessment of passive safety system

    International Nuclear Information System (INIS)

    Huang Changfan; Kuang Bo

    2012-01-01

    The reliability evaluation of passive safety system plays an important part in probabilistic safety assessment (PSA) of nuclear power plant applying passive safety design, which depends quantitatively on reliabilities of passive safety system. According to the object of reliability assessment of passive safety system, relevant parameters are identified. Then passive system behavior during accident scenarios are studied. A practical example of this method is given for the case of reliability assessment of AP1000 passive heat removal system in loss of normal feedwater accident. Key and design parameters of PRHRS are identified and functional failure criteria are established. Parameter combinations acquired by Latin hyper~ cube sampling (LHS) in possible parametric ranges are input and calculations of uncertainty propagation through RELAP5/MOD3 code are carried out. Based on the calculations, sensitivity assessment on PRHRS functional criteria and reliability evaluation of the system are presented, which might provide further PSA with PRHR system reliability. (authors)

  18. Determination of Safety Performance Grade of NPP Using Integrated Safety Performance Assessment (ISPA) Program

    International Nuclear Information System (INIS)

    Chung, Dae Wook

    2011-01-01

    Since the beginning of 2000, the safety regulation of nuclear power plant (NPP) has been challenged to be conducted more reasonable, effective and efficient way using risk and performance information. In the United States, USNRC established Reactor Oversight Process (ROP) in 2000 for improving the effectiveness of safety regulation of operating NPPs. The main idea of ROP is to classify the NPPs into 5 categories based on the results of safety performance assessment and to conduct graded regulatory programs according to categorization, which might be interpreted as 'Graded Regulation'. However, the classification of safety performance categories is highly comprehensive and sensitive process so that safety performance assessment program should be prepared in integrated, objective and quantitative manner. Furthermore, the results of assessment should characterize and categorize the actual level of safety performance of specific NPP, integrating all the substantial elements for assessing the safety performance. In consideration of particular regulatory environment in Korea, the integrated safety performance assessment (ISPA) program is being under development for the use in the determination of safety performance grade (SPG) of a NPP. The ISPA program consists of 6 individual assessment programs (4 quantitative and 2 qualitative) which cover the overall safety performance of NPP. Some of the assessment programs which are already implemented are used directly or modified for incorporating risk aspects. The others which are not existing regulatory programs are newly developed. Eventually, all the assessment results from individual assessment programs are produced and integrated to determine the safety performance grade of a specific NPP

  19. Fire safety assessment of tunnel structures

    DEFF Research Database (Denmark)

    Gkoumas, Konstantinos; Giuliani, Luisa; Petrini, Francesco

    2011-01-01

    .g. structural and non structural, organizational, human behavior). This is even more truth for the fire safety design of such structures. Fire safety in tunnels is challenging because of the particular environment, bearing in mind also that a fire can occur in different phases of the tunnel’s lifecycle. Plans...... for upgrading fire safety provisions and tunnel management are also important for existing tunnels. In this study, following a brief introduction of issues regarding the above mentioned aspects, the structural performance of a steel rib for a tunnel infrastructure subject to fire is assessed by means...

  20. Safety assessment of complex engineered and natural systems: radioactive waste disposal

    International Nuclear Information System (INIS)

    McNeish, J.A.; Vallikat, V.; Atkins, J.; Balady, M.A.

    1997-01-01

    Evaluation of deep, geologic disposal of nuclear waste requires the probabilistic safety assessment of a complex system from the coupling of various processes and sub-systems, parameter and model uncertainties, spatial and temporal variabilities, and the multiplicity of designs and scenarios. Both the engineered and natural system are included in the evaluation. Each system has aspects with considerable uncertainty both in important parameters and in overall conceptual models. The study represented herein provides a probabilistic safety assessment of a potential respository system for multiple engineered barrier system (EBS) design and conceptual model configurations (CRWMS M and O, 1996a) and considers the effects of uncertainty on the overall results. The assessment is based on data and process models available at the time of the study and doesnt necessarily represent the current safety evaluation. In fact, the percolation flux through the repository system is now expected to be higher than the estimate used for this study. The potential effects of higher percolation fluxes are currently under study. The safety of the system was assessed for both 10,000 and 1,000,000 years. Use of alternative conceptual models also produced major improvement in safety. For example, use of a more realistic engineered system release model produced improvement of over an order of magnitude in safety. Alternative measurement locations for the safety assessment produced substantial increases in safety, through the results are based on uncertain dilution factors in the transporting groundwater. (Author)

  1. Assessment of the safety of Ulchin nuclear power plant in the event of tsunami using parametric study

    International Nuclear Information System (INIS)

    Kim, Ji Young; Kang, Keum Seok

    2011-01-01

    Previous evaluations of the safety of the Ulchin Nuclear Power Plant in the event of a tsunami have the shortcoming of uncertainty of the tsunami sources. To address this uncertainty, maximum and minimum wave heights at the intake of Ulchin NPP have been estimated through a parametric study, and then assessment of the safety margin for the intake has been carried out. From the simulation results for the Ulchin NPP site, it can be seen that the coefficient of eddy viscosity considerably affects wave height at the inside of the breakwater. In addition, assessment of the safety margin shows that almost all of the intake water pumps have a safety margin over 2 m, and Ulchin NPP site seems to be safe in the event of a tsunami according to this parametric study, although parts of the CWPs rarely have a margin for the minimum wave height

  2. Quantitative risk assessment of digitalized safety systems

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  3. Assessing the validity of road safety evaluation studies by analysing causal chains.

    Science.gov (United States)

    Elvik, Rune

    2003-09-01

    This paper discusses how the validity of road safety evaluation studies can be assessed by analysing causal chains. A causal chain denotes the path through which a road safety measure influences the number of accidents. Two cases are examined. One involves chemical de-icing of roads (salting). The intended causal chain of this measure is: spread of salt --> removal of snow and ice from the road surface --> improved friction --> shorter stopping distance --> fewer accidents. A Norwegian study that evaluated the effects of salting on accident rate provides information that describes this causal chain. This information indicates that the study overestimated the effect of salting on accident rate, and suggests that this estimate is influenced by confounding variables the study did not control for. The other case involves a traffic club for children. The intended causal chain in this study was: join the club --> improve knowledge --> improve behaviour --> reduce accident rate. In this case, results are rather messy, which suggests that the observed difference in accident rate between members and non-members of the traffic club is not primarily attributable to membership in the club. The two cases show that by analysing causal chains, one may uncover confounding factors that were not adequately controlled in a study. Lack of control for confounding factors remains the most serious threat to the validity of road safety evaluation studies.

  4. Fusion blanket inherent safety assessment

    International Nuclear Information System (INIS)

    Sze, D.K.; Jung, J.; Cheng, E.T.

    1986-01-01

    Fusion has significant potential safety advantages. There is a strong incentive for designing fusion plants to ensure that inherent safety will be achieved. Accordingly, both the Tokamak Power Systems Studies and MINIMARS have identified inherent safety as a design goal. A necessary condition is for the blanket to maintain its configuration and integrity under all credible accident conditions. A main problem is caused by afterheat removal in an accident condition. In this regard, it is highly desirable to achieve the required level of protection of the plant capital investment and limitation of radioactivity release by systems that rely only on inherent properties of matter (e.g., thermal conductivity, specific heat, etc.) and without the use of active safety equipment. This paper assesses the conditions under which inherent safety is feasible. Three types of accident conditions are evaluated for two blankets. The blankets evaluated are a self cooled vanadium/lithium blanket and a self-cooled vanadium/Flibe blanket. The accident conditions evaluated are: (1) loss-of-flow accident; (2) loss-of-coolant accident (LOCA); and (3) partial loss-of-coolant accident

  5. A Computer Program for Assessing Nuclear Safety Culture Impact

    Energy Technology Data Exchange (ETDEWEB)

    Han, Kiyoon; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of)

    2014-10-15

    Through several accidents of NPP including the Fukushima Daiichi in 2011 and Chernobyl accidents in 1986, a lack of safety culture was pointed out as one of the root cause of these accidents. Due to its latent influences on safety performance, safety culture has become an important issue in safety researches. Most of the researches describe how to evaluate the state of the safety culture of the organization. However, they did not include a possibility that the accident occurs due to the lack of safety culture. Because of that, a methodology for evaluating the impact of the safety culture on NPP's safety is required. In this study, the methodology for assessing safety culture impact is suggested and a computer program is developed for its application. SCII model which is the new methodology for assessing safety culture impact quantitatively by using PSA model. The computer program is developed for its application. This program visualizes the SCIs and the SCIIs. It might contribute to comparing the level of the safety culture among NPPs as well as improving the management safety of NPP.

  6. The role of probabilistic safety assessment and probabilistic safety criteria in nuclear power plant safety

    International Nuclear Information System (INIS)

    1992-01-01

    The purpose of this Safety Report is to provide guidelines on the role of probabilistic safety assessment (PSA) and a range of associated reference points, collectively referred to as probabilistic safety criteria (PSC), in nuclear safety. The application of this Safety Report and the supporting Safety Practice publication should help to ensure that PSA methodology is used appropriately to assess and enhance the safety of nuclear power plants. The guidelines are intended for use by nuclear power plant designers, operators and regulators. While these guidelines have been prepared with nuclear power plants in mind, the principles involved have wide application to other nuclear and non-nuclear facilities. In Section 2 of this Safety Report guidelines are established on the role PSA can play as part of an overall safety assurance programme. Section 3 summarizes guidelines for the conduct of PSAs, and in Section 4 a PSC framework is recommended and guidance is provided for the establishment of PSC values

  7. Healthcare professionals’ views of feedback on patient safety culture assessment.

    NARCIS (Netherlands)

    Zwijnenberg, N.C.; Hendriks, M.; Hoogervorst-Schilp, J.; Wagner, C.

    2016-01-01

    Background: By assessing patient safety culture, healthcare providers can identify areas for improvement in patient safety culture. To achieve this, these assessment outcomes have to be relevant and presented clearly. The aim of our study was to explore healthcare professionals’ views on the

  8. Reactor Safety Assessment System

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  9. Reactor safety assessment system

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  10. Independent assessment for new nuclear reactor safety

    Directory of Open Access Journals (Sweden)

    D'Auria Francesco

    2017-01-01

    Full Text Available A rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On the one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs. Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry. The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty approach.

  11. Independent assessment for new nuclear reactor safety

    International Nuclear Information System (INIS)

    D'Auria, F.; Glaeser, H.; Debrecin, N.

    2017-01-01

    A rigorous framework for safety assessment is established in all countries where nuclear technology is used for the production of electricity. On one side, industry, i.e. reactor designers, vendors and utilities perform safety analysis and demonstrate consistency between results of safety analyses and requirements. On the other side, regulatory authorities perform independent assessment of safety and confirm the acceptability of safety of individual reactor units. The process of comparing results from analyses by reactor utilities and regulators is very complex. The process is also highly dependent upon mandatory approaches pursued for the analysis and from very many details which required the knowledge of sensitive proprietary data (e.g. spacer designs). Furthermore, all data available for the design, construction and operation of reactors produced by the nuclear industry are available to regulators. Two areas for improving the process of safety assessment for individual Nuclear Power Plant Units are identified: New details introduced by industry are not always and systematically requested by regulators for the independent assessment; New analytical techniques and capabilities are not necessarily used in the analyses by regulators (and by the industry). The established concept of independent assessment constitutes the way for improving the process of safety assessment. This is possible, or is largely facilitated, by the recent availability of the so-called Best Estimate Plus Uncertainty (BEPU) approach. (authors)

  12. Research on advanced system safety assessment procedures (4)

    International Nuclear Information System (INIS)

    Suzuki, Kazuhiko; Shimada, Yukiyasu

    2001-03-01

    The past research reports in the area of safety engineering proposed the Computer-aided HAZOP system to be applied to Nuclear Reprocessing Facilities. Automated HAZOP system has great advantage compared with human analysts in terms of accuracy of the results, and time required to conduct HAZOP studies. This report surveys the literature on risk assessment and safety design based on the concept of independent protection layers (IPLs). Furthermore, to improve HAZOP System, tool is proposed to construct the basic model and the internal state model. Such HAZOP system is applied to analyze two kinds of processes, where the ability of the proposed system is verified. In addition, risk assessment support system is proposed to integrate safety design environment and assessment result to be used by other plants as well as to enable the underline plant to use other plants' information. This technique can be implemented using web-based safety information systems. (author)

  13. A study on a reliability assessment methodology for the VHTR safety systems

    International Nuclear Information System (INIS)

    Lee, Hyung Sok

    2012-02-01

    The passive safety system of a 300MWt VHTR (Very High Temperature Reactor)which has attracted worldwide attention recently is actively considered for designing the improvement in the safety of the next generation nuclear power plant. The passive system functionality does not rely on an external source of the electrical support system,but on an intelligent use of the natural phenomena, such as convection, conduction, radiation, and gravity. It is not easy to evaluate quantitatively the reliability of the passive safety for the risk analysis considering the existing active system failure since the classical reliability assessment method could not be applicable. Therefore a new reliability methodology needs to be developed and applied for evaluating the reliability of the conceptual designed VHTR in this study. The preliminary evaluation and conceptualization are performed using the concept of the load and capacity theory related to the reliability physics model. The method of response surface method (RSM) is also utilized for evaluating the maximum temperature of nuclear fuel in this study. The significant variables and their correlation are considered for utilizing the GAMMA+ code. The proposed method might contribute to designing the new passive system of the VHTR

  14. Assessing the general safety and tolerability of vildagliptin: value of pooled analyses from a large safety database versus evaluation of individual studies

    Science.gov (United States)

    Schweizer, Anja; Dejager, Sylvie; Foley, James E; Kothny, Wolfgang

    2011-01-01

    Aim: Analyzing safety aspects of a drug from individual studies can lead to difficult-to-interpret results. The aim of this paper is therefore to assess the general safety and tolerability, including incidences of the most common adverse events (AEs), of vildagliptin based on a large pooled database of Phase II and III clinical trials. Methods: Safety data were pooled from 38 studies of ≥12 to ≥104 weeks’ duration. AE profiles of vildagliptin (50 mg bid; N = 6116) were evaluated relative to a pool of comparators (placebo and active comparators; N = 6210). Absolute incidence rates were calculated for all AEs, serious AEs (SAEs), discontinuations due to AEs, and deaths. Results: Overall AEs, SAEs, discontinuations due to AEs, and deaths were all reported with a similar frequency in patients receiving vildagliptin (69.1%, 8.9%, 5.7%, and 0.4%, respectively) and patients receiving comparators (69.0%, 9.0%, 6.4%, and 0.4%, respectively), whereas drug-related AEs were seen with a lower frequency in vildagliptin-treated patients (15.7% vs 21.7% with comparators). The incidences of the most commonly reported specific AEs were also similar between vildagliptin and comparators, except for increased incidences of hypoglycemia, tremor, and hyperhidrosis in the comparator group related to the use of sulfonylureas. Conclusions: The present pooled analysis shows that vildagliptin was overall well tolerated in clinical trials of up to >2 years in duration. The data further emphasize the value of a pooled analysis from a large safety database versus assessing safety and tolerability from individual studies. PMID:21415917

  15. AGR core safety assessment methodologies

    International Nuclear Information System (INIS)

    McLachlan, N.; Reed, J.; Metcalfe, M.P.

    1996-01-01

    To demonstrate the safety of its gas-cooled graphite-moderated AGR reactors, nuclear safety assessments of the cores are based upon a methodology which demonstrates no component failures, geometrical stability of the structure and material properties bounded by a database. All AGRs continue to meet these three criteria. However, predictions of future core behaviour indicate that the safety case methodology will eventually need to be modified to deal with new phenomena. A new approach to the safety assessment of the cores is currently under development, which can take account of these factors while at the same time providing the same level of protection for the cores. This approach will be based on the functionality of the core: unhindered movement of control rods, continued adequate cooling of the fuel and the core, continued ability to charge and discharge fuel. (author). 5 figs

  16. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Complementary evaluations of safety

    International Nuclear Information System (INIS)

    Neall, Fiona; Pastina, Barbara; Snellman, Margit; Smith, Paul; Gribi, P.; Johnson, Lawrence

    2008-12-01

    The KBS-3H design is a variant of the more general KBS-3 method for the geological disposal of spent nuclear fuel in Finland and Sweden. In the KBS-3H design, multiple assemblies containing spent fuel are emplaced horizontally in parallel, approximately 300 m long, slightly inclined deposition drifts. The copper canisters, each with a surrounding layer of bentonite clay, are placed in perforated steel shells prior to deposition in the drifts; the assembly is called the 'supercontainer'. The other KBS-3 variant is the KBS-3V design, in which the copper canisters are emplaced vertically in individual deposition holes surrounded by bentonite clay but without steel supercontainer shells. SKB and Posiva have conducted a Research, Development and Demonstration programme over the period 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to KBS-3V. As part of this programme, the long-term safety of a KBS-3H repository has been assessed in the KBS-3H safety studies. In order to focus the safety studies, the Olkiluoto site in the municipality of Eurajoki, which is the proposed site for a spent fuel repository in Finland, was used as a hypothetical site for a KBS-3H repository. The present report is part of a portfolio of reports discussing the long-term safety of the KBS-3H repository. The overall outcome of the KBS-3H safety studies is documented in the summary report, 'Safety assessment for a KBS-3H repository for spent nuclear fuel at Olkiluoto'. The purpose and scope of the KBS-3H complementary evaluations of safety report is provided in Posiva's Safety Case Plan, which is based on Regulatory Guide YVL 8.4 and on international guidelines on complementary lines of argument to long-term safety that are considered an important element of a post-closure safety case for geological repositories. Complementary evaluations of safety require the use of evaluations, evidence and qualitative supporting arguments that lie outside the

  17. Initial development of a practical safety audit tool to assess fleet safety management practices.

    Science.gov (United States)

    Mitchell, Rebecca; Friswell, Rena; Mooren, Lori

    2012-07-01

    Work-related vehicle crashes are a common cause of occupational injury. Yet, there are few studies that investigate management practices used for light vehicle fleets (i.e. vehicles less than 4.5 tonnes). One of the impediments to obtaining and sharing information on effective fleet safety management is the lack of an evidence-based, standardised measurement tool. This article describes the initial development of an audit tool to assess fleet safety management practices in light vehicle fleets. The audit tool was developed by triangulating information from a review of the literature on fleet safety management practices and from semi-structured interviews with 15 fleet managers and 21 fleet drivers. A preliminary useability assessment was conducted with 5 organisations. The audit tool assesses the management of fleet safety against five core categories: (1) management, systems and processes; (2) monitoring and assessment; (3) employee recruitment, training and education; (4) vehicle technology, selection and maintenance; and (5) vehicle journeys. Each of these core categories has between 1 and 3 sub-categories. Organisations are rated at one of 4 levels on each sub-category. The fleet safety management audit tool is designed to identify the extent to which fleet safety is managed in an organisation against best practice. It is intended that the audit tool be used to conduct audits within an organisation to provide an indicator of progress in managing fleet safety and to consistently benchmark performance against other organisations. Application of the tool by fleet safety researchers is now needed to inform its further development and refinement and to permit psychometric evaluation. Copyright © 2012 Elsevier Ltd. All rights reserved.

  18. Assessing the general safety and tolerability of vildagliptin: value of pooled analyses from a large safety database versus evaluation of individual studies

    Directory of Open Access Journals (Sweden)

    Schweizer A

    2011-02-01

    Full Text Available Anja Schweizer1, Sylvie Dejager2, James E Foley3, Wolfgang Kothny31Novartis Pharma AG, Basel, Switzerland; 2Novartis Pharma SAS, Rueil-Malmaison, France; 3Novartis Pharmaceuticals Corporation, East Hanover, NJ, USAAim: Analyzing safety aspects of a drug from individual studies can lead to difficult-to-interpret results. The aim of this paper is therefore to assess the general safety and tolerability, including incidences of the most common adverse events (AEs, of vildagliptin based on a large pooled database of Phase II and III clinical trials.Methods: Safety data were pooled from 38 studies of ≥12 to ≥104 weeks' duration. AE profiles of vildagliptin (50 mg bid; N = 6116 were evaluated relative to a pool of comparators (placebo and active comparators; N = 6210. Absolute incidence rates were calculated for all AEs, serious AEs (SAEs, discontinuations due to AEs, and deaths.Results: Overall AEs, SAEs, discontinuations due to AEs, and deaths were all reported with a similar frequency in patients receiving vildagliptin (69.1%, 8.9%, 5.7%, and 0.4%, respectively and patients receiving comparators (69.0%, 9.0%, 6.4%, and 0.4%, respectively, whereas drug-related AEs were seen with a lower frequency in vildagliptin-treated patients (15.7% vs 21.7% with comparators. The incidences of the most commonly reported specific AEs were also similar between vildagliptin and comparators, except for increased incidences of hypoglycemia, tremor, and hyperhidrosis in the comparator group related to the use of sulfonylureas.Conclusions: The present pooled analysis shows that vildagliptin was overall well tolerated in clinical trials of up to >2 years in duration. The data further emphasize the value of a pooled analysis from a large safety database versus assessing safety and tolerability from individual studies.Keywords: type 2 diabetes, dipeptidyl peptidase-4, edema, safety, vildagliptin

  19. Procedures for self-assessment of operational safety

    International Nuclear Information System (INIS)

    1997-08-01

    Self-assessment processes have been continuously developed by nuclear organizations, including nuclear power plants. Currently, the nuclear industry and governmental organizations are showing an increasing interest in the implementation of this process as an effective way for improving safety performance. Self-assessment involves the use of different types of tools and mechanisms to assist the organizations in assessing their own safety performance against given standards. This helps to enhance the understanding of the need for improvements, the feeling of ownership in achieving them and and the safety culture as a whole. The concepts developed in this report present the basic approach to self-assessment taking into consideration experience gained during Operational Safety Review Team (OSART) missions, from organizations and utilities which have successfully implemented parts of a self-assessment programme and from meetings organized to discuss the subject

  20. Assessment of Safety Standards for Automotive Electronic Control Systems

    Science.gov (United States)

    2016-06-01

    This report summarizes the results of a study that assessed and compared six industry and government safety standards relevant to the safety and reliability of automotive electronic control systems. These standards include ISO 26262 (Road Vehicles - ...

  1. The network to review natural analogue studies and their applications to repository safety assessment and public communication (NAnet)

    Energy Technology Data Exchange (ETDEWEB)

    Miller, W.M.; Hooker, P.J. [ENVIROS Consulting ltd, 61, the Shore Leith, UK-0 EH6 6RA Edinburgh (United Kingdom)

    2004-07-01

    Analogue information can increase our conceptual understanding of long-term repository behaviour in support of post-closure performance assessment (PA), provide quantitative data for PA models and provide ways of communicating safety information to non-specialist audiences. These functions of analogue studies have, however, received too little attention in PA reports and safety cases. Many analogue studies have been undertaken in the last two decades costing tens of millions of euros, and these have covered a wide range of phenomena such as uranium ore deposition, natural fission reactors, natural nuclide migration, contaminant containment by clays and sediments, preservation of ancient fossil trees and buried artefacts etc. The different uses of analogues would be easier to manage if a single database of quality approved analogue information were to be created. NAnet, a Thematic Network within the 5. EURATOM FP is aiming to promote more considered applications of analogues in performance and safety assessments and in audience dialogue. NAnet intends critically to review a number of analogue studies in terms of their relevance and limitations to different repository concepts and environments and with regard to their applications in performance assessments, safety cases and communication. On the basis of these reviews, a simple digital database is being developed for the PA community which will allow PA modelers to make quicker and wider use of natural analogue information in performance and safety assessments. It is expected that some of these tools will help radioactive waste institutions to make better use of natural analogue information for communication with different audiences, including the public. (authors)

  2. NUMO's approach for long-term safety assessment - 59404

    International Nuclear Information System (INIS)

    Ebashi, Takeshi; Kaku, Kenichi; Ishiguro, Katsuhiko

    2012-01-01

    One of NUMO's policies for ensuring safety is staged and flexible project implementation and decision-making based on iterative confirmation of safety. The safety assessment takes the central role in multiple lines of reasoning and argumentation by providing a quantitative evaluation of long-term safety; a key aspect is uncertainty management. This paper presents NUMO's basic strategies for long-term safety assessment based on the above policy. NUMO's approach considering Japanese boundary conditions is demonstrated as a starting-point for evaluating the long-term safety of an actual site. In Japan, the Act on Final Disposal of Specified Radioactive Waste states that the siting process shall consist of three stages. The Nuclear Waste Management Organization of Japan (NUMO) is responsible for geological disposal of vitrified high-level waste and some types of TRU waste. NUMO has chosen to implement a volunteer approach to siting. NUMO decided to prepare the so-called 2010 technical report, which sets out three safety policies, one of which is staged project implementation and decision-making based on iterative confirmation of safety. Based on this policy, NUMO will gradually integrate relevant interdisciplinary knowledge to build a safety case when a formal volunteer application is received that would allow site investigations to be initiated. The safety assessment takes the central role in multiple lines of reasoning and argumentation by providing a quantitative evaluation of long-term safety; one of a key aspect is uncertainty management. This paper presents the basic strategies for NUMO's long-term safety assessment based on the above policy. In concrete terms, the common procedures involved in safety assessment are applied in a stepwise manner, based on integration of knowledge obtained from site investigations/evaluations and engineered measures. The results of the safety assessment are then reflected in the planning of site investigations and engineered

  3. Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Complementary evaluations of safety

    Energy Technology Data Exchange (ETDEWEB)

    Neall, Fiona; Pastina, Barbara; Snellman, Margit; Smith, Paul; Gribi, P.; Johnson, Lawrence

    2008-12-15

    The KBS-3H design is a variant of the more general KBS-3 method for the geological disposal of spent nuclear fuel in Finland and Sweden. In the KBS-3H design, multiple assemblies containing spent fuel are emplaced horizontally in parallel, approximately 300 m long, slightly inclined deposition drifts. The copper canisters, each with a surrounding layer of bentonite clay, are placed in perforated steel shells prior to deposition in the drifts; the assembly is called the 'supercontainer'. The other KBS-3 variant is the KBS-3V design, in which the copper canisters are emplaced vertically in individual deposition holes surrounded by bentonite clay but without steel supercontainer shells. SKB and Posiva have conducted a Research, Development and Demonstration programme over the period 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to KBS-3V. As part of this programme, the long-term safety of a KBS-3H repository has been assessed in the KBS-3H safety studies. In order to focus the safety studies, the Olkiluoto site in the municipality of Eurajoki, which is the proposed site for a spent fuel repository in Finland, was used as a hypothetical site for a KBS-3H repository. The present report is part of a portfolio of reports discussing the long-term safety of the KBS-3H repository. The overall outcome of the KBS-3H safety studies is documented in the summary report, 'Safety assessment for a KBS-3H repository for spent nuclear fuel at Olkiluoto'. The purpose and scope of the KBS-3H complementary evaluations of safety report is provided in Posiva's Safety Case Plan, which is based on Regulatory Guide YVL 8.4 and on international guidelines on complementary lines of argument to long-term safety that are considered an important element of a post-closure safety case for geological repositories. Complementary evaluations of safety require the use of evaluations, evidence and qualitative supporting arguments

  4. Human reliability analysis methods for probabilistic safety assessment

    International Nuclear Information System (INIS)

    Pyy, P.

    2000-11-01

    Human reliability analysis (HRA) of a probabilistic safety assessment (PSA) includes identifying human actions from safety point of view, modelling the most important of them in PSA models, and assessing their probabilities. As manifested by many incidents and studies, human actions may have both positive and negative effect on safety and economy. Human reliability analysis is one of the areas of probabilistic safety assessment (PSA) that has direct applications outside the nuclear industry. The thesis focuses upon developments in human reliability analysis methods and data. The aim is to support PSA by extending the applicability of HRA. The thesis consists of six publications and a summary. The summary includes general considerations and a discussion about human actions in the nuclear power plant (NPP) environment. A condensed discussion about the results of the attached publications is then given, including new development in methods and data. At the end of the summary part, the contribution of the publications to good practice in HRA is presented. In the publications, studies based on the collection of data on maintenance-related failures, simulator runs and expert judgement are presented in order to extend the human reliability analysis database. Furthermore, methodological frameworks are presented to perform a comprehensive HRA, including shutdown conditions, to study reliability of decision making, and to study the effects of wrong human actions. In the last publication, an interdisciplinary approach to analysing human decision making is presented. The publications also include practical applications of the presented methodological frameworks. (orig.)

  5. Institutionalization of safety re-assessment system for operating nuclear power plants

    International Nuclear Information System (INIS)

    Kim, H. J.; Cho, J. C.; Min, B. K.; Park, J. S.; Jung, H. D.; Oh, K. M.; Kim, W. K.; Lim, J. H.

    1999-01-01

    In this study, in-depth reviews of the foreign countries' experiences and practices in applications of the periodic safety review (PSR), backfitting and license renewal systems as well as the current status of nuclear power safety assurance programs and activities in Korea have been performed to investigate the necessity and feasibility of the application of the systems for the domestic operating nuclear power plants and to establish effective strategy and methodology for the institutionalization of a periodic safety re-assessment system appropriate to both the domestic and international nuclear power environments by incorporating the PSR with the backfitting and license renewal systems. For these purposes, the regulatory policy, fundamental principles and detailed requirements for the institutionalization of the safety re-assessment system and the effective measures for active implementation of the backfitting program have been developed and then a comparative study of benefits and shortcomings has been conducted for the three different models of the periodic safety re-assessment system incorporated with either the license renewal or life extension process, which have been considered as practicable ones in the domestic situation. The model chosen in this study as the most appropriate safety re-assessment system is the one that the re-assessments are performed at the interval of ten years throughout the service life of nuclear power plant and the ten-year license renewal or life extension after the expiration of design life can be permitted based on the regulatory review of the re-assessment results and follow-up measures. Finally, this paper has discussed on the details of the requirements, approach and procedures established for the institutionalization of the periodic safety re-assessment system chosen as the most appropriate one for domestic applications

  6. Additional safety assessment of ITER - Addition safety investigation of the INB ITER

    International Nuclear Information System (INIS)

    2012-01-01

    This assessment aims at re-assessing safety margins in the light of events which occurred in Fukushima Daiichi, i.e. extreme natural events challenging the safety of installations. After a presentation of some characteristics of the ITER installation (location, activities, buildings, premise detritiation systems, electric supply, handling means, radioactive materials, chemical products, nuclear risks, specific risks), the report addresses the installation robustness by identifying cliff-edge effect risks which can be related to a loss of confinement of radioactive materials, explosions, a significant increase of exposure level, a possible effect on water sheets, and so on. The next part addresses the various aspects related to a seismic risk: installation sizing (assessment methodology, seismic risk characterization in Cadarache), sizing protection measures, installation compliance, and margin assessment. External flooding is the next addressed risk: installation sizing with respect to this specific risk, protection measures, installation compliance, margin assessment, and studied additional measures. Other extreme natural phenomena are considered (meteorological conditions, earthquake and flood) which may have effects on other installations (dam, canal). Then, the report addresses technical risks like the loss of electric supplies and cooling systems, the way a crisis is managed in terms of technical and human means and organization in different typical accidental cases. Subcontracting practices are also discussed. A synthesis proposes an overview of this additional safety assessment and discusses the impact which could have additional measures which could be implemented

  7. LWR safety studies. Analyses and further assessments relating to the German Risk Assessment Study on Nuclear Power Plants. Vol. 3

    International Nuclear Information System (INIS)

    1983-01-01

    Critical review of the analyses of the German Risk Assessment Study on Nuclear Power Plants (DRS) concerning the reliability of the containment under accident conditions and the conditions of fission product release (transport and distribution in the environment). Main point of interest in this context is an explosion in the steam section and its impact on the containment. Critical comments are given on the models used in the DRS for determining the accident consequences. The analyses made deal with the mathematical models and database for propagation calculations, the methods of dose computation and assessment of health hazards, and the modelling of protective and safety measures. Social impacts of reactor accidents are also considered. (RF) [de

  8. Development and applications of a safety assessment system for promoting safety culture in nuclear power plants

    International Nuclear Information System (INIS)

    Takano, Ken-ichi; Hasegawa, Naoko; Hirose, Ayako; Hayase, Ken-ichi

    2004-01-01

    For past five years, CRIEPI has been continuing efforts to develop and make applications of a 'safety assessment system' which enable to measure the safety level of organization. This report describe about frame of the system, assessment results and its reliability, and relation between labor accident rate in the site and total safety index (TSI), which can be obtained by the principal factors analysis. The safety assessment in this report is based on questionnaire survey of employee. The format and concrete questionnaires were developed using existing literatures including organizational assessment tools. The tailored questionnaire format involved 124 questionnaire items. The assessment results could be considered as a well indicator of the safety level of organization, safety management, and safety awareness of employee. (author)

  9. Safety culture' is integrating 'human' into risk assessment

    International Nuclear Information System (INIS)

    Sugimoto, Taiji

    2014-01-01

    Significance of Fukushima nuclear power accident requested reconsideration of safety standards, of which we had usually no doubt. Risk assessment standard (JIS B 9702), Which was used for repetition of database preparation and cumulative assessment, defined allowable risk and residual risk. However, work site and immediate assessment was indispensable beside such assessment so as to ensure safety. Risk of casualties was absolutely not acceptable in principle and judgments to approve allowable risk needed accountability, which was reminded by safety culture proposed by IAEA and also identified by investigation of organizational cause of Columbia accident. Actor of safety culture would be organization and individual, and mainly individual. Realization of safety culture was conducted by personnel having moral consciousness and firm sense of mission in the course of jobs and working daily with sweat pouring. Safety engineering/technology should have framework integrating human as such totality. (T. Tanaka)

  10. A novel safety assessment strategy applied to non-selective extracts.

    Science.gov (United States)

    Koster, Sander; Leeman, Winfried; Verheij, Elwin; Dutman, Ellen; van Stee, Leo; Nielsen, Lene Munch; Ronsmans, Stefan; Noteborn, Hub; Krul, Lisette

    2015-06-01

    A main challenge in food safety research is to demonstrate that processing of foodstuffs does not lead to the formation of substances for which the safety upon consumption might be questioned. This is especially so since food is a complex matrix in which the analytical detection of substances, and consequent risk assessment thereof, is difficult to determine. Here, a pragmatic novel safety assessment strategy is applied to the production of non-selective extracts (NSEs), used for different purposes in food such as for colouring purposes, which are complex food mixtures prepared from reference juices. The Complex Mixture Safety Assessment Strategy (CoMSAS) is an exposure driven approach enabling to efficiently assess the safety of the NSE by focussing on newly formed substances or substances that may increase in exposure during the processing of the NSE. CoMSAS enables to distinguish toxicologically relevant from toxicologically less relevant substances, when related to their respective levels of exposure. This will reduce the amount of work needed for identification, characterisation and safety assessment of unknown substances detected at low concentration, without the need for toxicity testing using animal studies. In this paper, the CoMSAS approach has been applied for elderberry and pumpkin NSEs used for food colouring purposes. Copyright © 2015 Elsevier Ltd. All rights reserved.

  11. Probabilistic Safety Assessment of Waste from PyroGreen Processes

    International Nuclear Information System (INIS)

    Ju, Hee Jae; Ham, In hye; Hwang, Il Soon

    2016-01-01

    The main object of PyroGreen processes is decontaminating SNFs into intermediate level waste meeting U.S. WIPP contact-handled (CH) waste characteristics to achieve long-term radiological safety of waste disposal. In this paper, radiological impact of PyroGreen waste disposal is probabilistically assessed using domestic input parameters for safety assessment of disposal. PyroGreen processes is decontamination technology using pyro-chemical process developed by Seoul National University in collaboration with KAERI, Chungnam University, Korea Hydro-Nuclear Power and Yonsei University. Advanced Korean Reference Disposal System (A-KRS) design for vitrified waste is applied to develop safety assessment model using GoldSim software. The simulation result shows that PyroGreen vitrified waste is expected to satisfy the regulatory dose limit criteria, 0.1 mSv/yr. With small probability, however, radiological impact to public can be higher than the expected value after 2E5-year. Although the result implies 100 times safety margin even in that case, further study will be needed to assess the sensitivity of other input parameters which can affect the radiological impact for long-term.

  12. Probabilistic Safety Assessment of Waste from PyroGreen Processes

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Hee Jae; Ham, In hye; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    2016-05-15

    The main object of PyroGreen processes is decontaminating SNFs into intermediate level waste meeting U.S. WIPP contact-handled (CH) waste characteristics to achieve long-term radiological safety of waste disposal. In this paper, radiological impact of PyroGreen waste disposal is probabilistically assessed using domestic input parameters for safety assessment of disposal. PyroGreen processes is decontamination technology using pyro-chemical process developed by Seoul National University in collaboration with KAERI, Chungnam University, Korea Hydro-Nuclear Power and Yonsei University. Advanced Korean Reference Disposal System (A-KRS) design for vitrified waste is applied to develop safety assessment model using GoldSim software. The simulation result shows that PyroGreen vitrified waste is expected to satisfy the regulatory dose limit criteria, 0.1 mSv/yr. With small probability, however, radiological impact to public can be higher than the expected value after 2E5-year. Although the result implies 100 times safety margin even in that case, further study will be needed to assess the sensitivity of other input parameters which can affect the radiological impact for long-term.

  13. Quantitative safety assessment of air traffic control systems through system control capacity

    Science.gov (United States)

    Guo, Jingjing

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

  14. The practice of pre-marketing safety assessment in drug development.

    Science.gov (United States)

    Chuang-Stein, Christy; Xia, H Amy

    2013-01-01

    The last 15 years have seen a substantial increase in efforts devoted to safety assessment by statisticians in the pharmaceutical industry. While some of these efforts were driven by regulations and public demand for safer products, much of the motivation came from the realization that there is a strong need for a systematic approach to safety planning, evaluation, and reporting at the program level throughout the drug development life cycle. An efficient process can help us identify safety signals early and afford us the opportunity to develop effective risk minimization plan early in the development cycle. This awareness has led many pharmaceutical sponsors to set up internal systems and structures to effectively conduct safety assessment at all levels (patient, study, and program). In addition to process, tools have emerged that are designed to enhance data review and pattern recognition. In this paper, we describe advancements in the practice of safety assessment during the premarketing phase of drug development. In particular, we share examples of safety assessment practice at our respective companies, some of which are based on recommendations from industry-initiated working groups on best practice in recent years.

  15. Model quality and safety studies

    DEFF Research Database (Denmark)

    Petersen, K.E.

    1997-01-01

    The paper describes the EC initiative on model quality assessment and emphasizes some of the problems encountered in the selection of data from field tests used in the evaluation process. Further, it discusses the impact of model uncertainties in safety studies of industrial plants. The model...... that most of these have never been through a procedure of evaluation, but nonetheless are used to assist in making decisions that may directly affect the safety of the public and the environment. As a major funder of European research on major industrial hazards, DGXII is conscious of the importance......-tain model is appropriate for use in solving a given problem. Further, the findings from the REDIPHEM project related to dense gas dispersion will be highlighted. Finally, the paper will discuss the need for model quality assessment in safety studies....

  16. Safety assessment of foods derived from genetically modified crops

    NARCIS (Netherlands)

    Kleter, G.A.; Kuiper, H.A.

    2003-01-01

    The pre-market safety assessment of foods derived from genetically modified crops is carried out according to the consensus approach of "substantial equivalence", in other words: the comparative safety assessment. Currently, the safety assessment of genetically modified foods is harmonized at the

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

    International Nuclear Information System (INIS)

    2003-01-01

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

  18. Assessment of safety culture: Changing regulatory approach in Hungary

    International Nuclear Information System (INIS)

    Ronaky, Jozsef; Toth, Andras

    2002-01-01

    Hungarian Atomic Energy Authority (HAEA) is changing its inspection practice and assessment methods of safety performance and safety culture in operating nuclear facilities. The new approach emphasises integrated team inspection of safety cornerstones and systematic assessment of safety performance of operators. (author)

  19. Tolerability of risk, safety assessment principles and their implications for probabilistic safety analysis

    International Nuclear Information System (INIS)

    Ewing, D.J.F.; Campbell, J.F.

    1994-01-01

    This paper gives a regulatory view of probabilistic safety assessment as seen by the Nuclear Installations Inspectorate (NII) and in the light of the general regulatory risk aims set out in the Health and Safety Executive's (HSE) The tolerability of risk from nuclear power stations (TOR) and in Safety assessment principles for nuclear plants (SAPs), prepared by NII on behalf of the HSE. Both of these publications were revised and republished in 1992. This paper describes the SAPs, together with the historical background, the motivation for review, the effects of the Sizewell and Hinkley Point C public inquiries, changes since the original versions, comparison with international standards and use in assessment. For new plant, probabilistic safety analysis (PSA) is seen as an essential tool in balancing the safety of the design and in demonstrating compliance with TOR and the SAPs. (Author)

  20. Topical session proceedings of the 5. IGSC meeting on: observations regarding the safety case in recent safety assessment studies

    International Nuclear Information System (INIS)

    Hooper, Alan J.; Voinis, Sylvie; Van Luik, Abraham E.

    2004-01-01

    Within the NEA, the IGSC (Integration Group for the Safety Case) has, as an essential role, to develop common views on such key aspects of the safety case. Therefore, since the inauguration of the IGSC in 2000, four meetings were organised with topical sessions to explore various of these key aspects. This is a report on the fifth such topical session, held as part of the 5. plenary meeting of the IGSC. The session was attended by 36 participants, representing waste management organisations and regulatory authorities from 16 NEA member countries, the IAEA and the European Commission. The purpose of this topical session was to provide support to the finalising of the IGSC safety case brochure by getting a description of the safety case content of the IAEA Draft Safety Requirements document and by getting an overview of progress that could be observed from national organisations on developing their cases for system safety and/or developing the required methodologies. The objective was that the IGSC safety case brochure should be supportive of the IAEA/NEA document, and be reflective of the experience of the IGSC member programmes and organisations. The topical session was mainly aimed at exchanging information on: - The safety case related content of the proposed IAEA/NEA document (currently titled: 'IAEA Safety Standards Series, Geological Disposal of Radioactive Waste, Draft Safety Requirements (DS-154)'). - National programmes where safety assessments have recently been completed, e.g. ONDRAF/NIRAS, Nagra and Andra. - Feedback from international peer reviews, e.g. the Andra Dossier 2001 Argile, the Belgian SAFIR 2 report, the SR 97 report and the US-DOE Yucca Mountain TSPA. - The evolution of some national assessment methods and approaches e.g. SKB and Nagra. - The content of the draft IGSC safety case brochure entitled: 'The Nature and Purpose of the Post-closure Safety Case in Geological Disposal'. This document presents the various

  1. Preclosure radiological safety assessment for the ground support system in the exploratory studies facility

    International Nuclear Information System (INIS)

    Smith, A.J.; Tsai, F.C.

    1995-01-01

    An initial probabilistic safety assessment was performed for the exploratory studies facility underground opening to determine whether the ground support system should be classified as an item important to safety. The initiating event was taken to be a rock fall in an operational facility impacting a loaded waste transporter. Rock fall probability rates were estimated from data reported by commercial mining operations. This information was retrieved from the data base compiled by the Mining Safety and Health Administration from the mandatory reporting of incidents. The statistical distribution of the rock fall magnitude was estimated from the horizontal and vertical spacing fractures measured at the Yucca Mountain repository horizon. Simple models were developed to estimate container deformation and radionuclide releases arising from the projected distribution of impacts. Accepted techniques were used to calculate atmospheric dispersion and obtain the committed dose to individuals

  2. The radiation safety self-assessment program of Ontario Hydro

    International Nuclear Information System (INIS)

    Armitage, G.; Chase, W.J.

    1987-01-01

    Ontario Hydro has developed a self-assessment program to ensure that high quality in its radiation safety program is maintained. The self-assessment program has three major components: routine ongoing assessment, accident/incident investigation, and detailed assessments of particular radiation safety subsystems or of the total radiation safety program. The operation of each of these components is described

  3. On the fundamentals of nuclear reactor safety assessment. Inherent threats and their implications

    Energy Technology Data Exchange (ETDEWEB)

    Hyvaerinen, J. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland). Nuclear Safety Dept.

    1996-12-01

    The thesis addresses some fundamental questions related to implementation and assessment of nuclear safety. The safety principles and assessment methods are described, followed by descriptions of selected novel technical challenges to nuclear safety. The novel challenges encompass a wide variety of technical issues, thus providing insights on the limitations of conventional safety assessment methods. Study of the limitations suggests means to improve nuclear reactor design criteria and safety assessment practices. The novel safety challenges discussed are (1) inherent boron dilution in PWRs, (2) metallic insulation performance with respect to total loss of emergency cooling systems in a loss-of-coolant accident, and (3) horizontal steam generator heat transfer performance at natural circulation conditions. (50 refs.).

  4. On the fundamentals of nuclear reactor safety assessment. Inherent threats and their implications

    International Nuclear Information System (INIS)

    Hyvaerinen, J.

    1996-12-01

    The thesis addresses some fundamental questions related to implementation and assessment of nuclear safety. The safety principles and assessment methods are described, followed by descriptions of selected novel technical challenges to nuclear safety. The novel challenges encompass a wide variety of technical issues, thus providing insights on the limitations of conventional safety assessment methods. Study of the limitations suggests means to improve nuclear reactor design criteria and safety assessment practices. The novel safety challenges discussed are (1) inherent boron dilution in PWRs, (2) metallic insulation performance with respect to total loss of emergency cooling systems in a loss-of-coolant accident, and (3) horizontal steam generator heat transfer performance at natural circulation conditions. (50 refs.)

  5. Probabilistic safety assessment for seismic events

    International Nuclear Information System (INIS)

    1993-10-01

    This Technical Document on Probabilistic Safety Assessment for Seismic Events is mainly associated with the Safety Practice on Treatment of External Hazards in PSA and discusses in detail one specific external hazard, i.e. earthquakes

  6. Safety analysis and risk assessment handbook

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  7. The Nirex safety assessment research programme: annual report for 1986/87

    International Nuclear Information System (INIS)

    Cooper, M.J.; Hodgkinson, D.P.

    1987-05-01

    This report describes research relating to the underground disposal of low-level and intermediate-level radioactive wastes, to provide information for post-emplacement radiological safety assessment. Topics reported are solubility and sorption, organic degradation, microbial activity, leaching, the corrosion of containers, and radionuclide migration studies. Properties of clays, slates, colloids and uranium disequilibrium are studied. Mathematical modelling to support the safety assessment of radioactive waste disposal is also studied. (U.K.)

  8. Study on systematic integration technology of design and safety assessment for HLW geological disposal. 2. Research document

    International Nuclear Information System (INIS)

    Ishihara, Yoshinao; Fukui, Hiroshi; Sagawa, Hiroshi; Matsunaga, Kenichi; Ito, Takaya; Kohanawa, Osamu; Kuwayama, Yuki

    2003-02-01

    The present study was carried out relating to basic design of the Geological Disposal Technology Integration System' that will be systematized as knowledge base for design analysis and safety assessment of HLW geological disposal system by integrating organically and hierarchically various technical information in three study field. The key conclusions are summarized as follows: (1) As referring to the current performance assessment report, the technical information for R and D program of HLW geological disposal system was systematized hierarchically based on summarized information in a suitable form between the work flow (work item) and processes/characteristic flow (process item). (2) As the result of the systematized technical information, database structure and system functions necessary for development and construction to the computer system were clarified in order to secure the relation between technical information and data set for assessment of HLW geological disposal system. (3) The control procedure for execution of various analysis code used by design and safety assessment in HLW geological disposal study was arranged possibility in construction of 'Geological Disposal Technology Integration System' after investigating the distributed computing technology. (author)

  9. Safety Assessment of Talc as Used in Cosmetics.

    Science.gov (United States)

    Fiume, Monice M; Boyer, Ivan; Bergfeld, Wilma F; Belsito, Donald V; Hill, Ronald A; Klaassen, Curtis D; Liebler, Daniel C; Marks, James G; Shank, Ronald C; Slaga, Thomas J; Snyder, Paul W; Andersen, F Alan

    2015-01-01

    The Cosmetic Ingredient Review Expert Panel (Panel) assessed the safety of talc for use in cosmetics. The safety of talc has been the subject of much debate through the years, partly because the relationship between talc and asbestos is commonly misunderstood. Industry specifications state that cosmetic-grade talc must contain no detectable fibrous, asbestos minerals. Therefore, the large amount of available animal and clinical data the Panel relied on in assessing the safety of talc only included those studies on talc that did not contain asbestos. The Panel concluded that talc is safe for use in cosmetics in the present practices of use and concentration (some cosmetic products are entirely composed of talc). Talc should not be applied to the skin when the epidermal barrier is missing or significantly disrupted. © The Author(s) 2015.

  10. Probabilistic safety assessment in nuclear power plant management

    International Nuclear Information System (INIS)

    Holloway, N.J.

    1989-06-01

    Probabilistic Safety Assessment (PSA) techniques have been widely used over the past few years to assist in understanding how engineered systems respond to abnormal conditions, particularly during a severe accident. The use of PSAs in the design and operation of such systems thus contributes to the safety of nuclear power plants. Probabilistic safety assessments can be maintained to provide a continuous up-to-date assessment (Living PSA), supporting the management of plant operations and modifications

  11. LWR safety studies. Analyses and further assessments relating to the German Risk Assessment Study on Nuclear Power Plants. Vol. 1

    International Nuclear Information System (INIS)

    1983-01-01

    This documentation of the activities of the Oeko-Institut is intended to show errors made and limits encountered in the experimental approaches and in results obtained by the work performed under phase A of the German Risk Assessment Study on Nuclear Power Plants (DRS). Concern is expressed and explained relating to the risk definition used in the Study, and the results of other studies relied on; specific problems of methodology are discussed with regard to the value of fault-tree/accident analyses for describing the course of safety-related events, and to the evaluations presented in the DRS. The Markov model is explained as an approach offering alternative solutions. The identification and quantification of common-mode failures is discussed. Origin, quality and methods of assessing the reliability characteristics used in the DRS as well as the statistical models for describing failure scenarios of reactor components and systems are critically reviewed. (RF) [de

  12. Types of safety assessments of near surface repository for radioactive waste

    International Nuclear Information System (INIS)

    Mateeva, M.

    2004-01-01

    The purpose of this article is to presents the classification of different types safety assessments of near surface repository for low and intermediate level radioactive waste substantiated with results of safety assessments generated in Bulgaria. The different approach of safety assessments applied for old existing repository as well as for site selection for construction new repository is outlined. The regulatory requirements in Bulgaria define three main types of assessments: Safety assessment; Technical substation of repository safety; Assessment of repository influence on environment that is in form of report prepared from the Ministry of environment and waters on the base of results obtained in two first types of assessments. Additionally first type is subdivided in three categories - preliminary safety assessment, safety assessment and post closure safety assessment, which are generated using deterministic approach. The technical substation of repository safety is generated using probabilistic approach. Safety assessment results that are presented here are based on evaluation of existing old repository type 'Radon' in Novi Han and real site selection procedure for new near surface repository for low and intermediate level radioactive waste from nuclear power station in Kozloduy. The important role of safety assessment for improvement the repository safety as well as for repository licensing, correct site selection and right choice of engineer barriers and repository design is discussed using generated results. (author)

  13. Data used for safety assessment of reprocessing facilities

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Suzuki, Atsuyuki; Kanagawa, Akira

    1990-08-01

    For safety assessment of a reprocessing facility, it is important to know performance of radioactive materials in their accidental release and transfer. Accordingly, it is necessary to collect and prepare data for use in analyses for their performance. In JAERI, experiments such as for data acquisition, for source-term evaluation and for radioactive material transfer, are now planned to be performed. Prior to these experiments, it is decided to investigate data in use for accidental safety assessment of reprocessing plants and their based experimental data, thus to make it possible to recommend reasonable values for safety analysis parameters by evaluating the investigated results, to select the experimental items, to edit a safety assessment handbook and so on. In this line of objectives, JAERI rewarded a two-year contract of investigation to Nuclear Safety Research Association, to make a working group under a special committee on data investigation for reprocessing facility safety assessment. This report is a collection of results reviewed and checked by the working group. The contents consist of two parts, one for investigation and review of data used for safety assessment of domestic or oversea reprocessing facilities, and the other for investigation, review and evaluation of ANSI recommended American standard data reported by E. Walker together with their based experimental data resorting to the original referred reports. (author)

  14. Safety functions and safety function indicators - key elements in SKB'S methodology for assessing long-term safety of a KBS-3 repository

    International Nuclear Information System (INIS)

    Hedin, A.

    2008-01-01

    The application of so called safety function indicators in SKB safety assessment of a KBS-3 repository for spent nuclear fuel is presented. Isolation and retardation are the two main safety functions of the KBS-3 concept. In order to quantitatively evaluate safety on a sub-system level, these functions need to be differentiated, associated with quantitative measures and, where possible, with quantitative criteria relating to the fulfillment of the safety functions. A safety function is defined as a role through which a repository component contributes to safety. A safety function indicator is a measurable or calculable property of a repository component that allows quantitative evaluation of a safety function. A safety function indicator criterion is a quantitative limit such that if the criterion is fulfilled, the corresponding safety function is upheld. The safety functions and their associated indicators and criteria developed for the KBS-3 repository are primarily related to the isolating potential and to physical states of the canister and the clay buffer surrounding the canister. They are thus not directly related to release rates of radionuclides. The paper also describes how the concepts introduced i) aid in focussing the assessment on critical, safety related issues, ii) provide a framework for the accounting of safety throughout the different time frames of the assessment and iii) provide key information in the selection of scenarios for the safety assessment. (author)

  15. Bridging probabilistic safety assessment studies with information Management System

    International Nuclear Information System (INIS)

    Luanco, E. M.

    2010-01-01

    Probabilistic Safety Assessment (PSA) is a critical business often known in conjunction with either new build or life extension of nuclear power plant. However, it is not so often referred to the operation phase of the plant, although it could bring a lot of long term benefits to the operator. The purpose of this paper is to discuss the potential contribution of PSA with day to day operation in bridging the deficiencies and specific failures characteristics of critical Structure System and Component (SSC) with the results of PSA studies. From and Information System prospective, the use of Information Management system (IMS) -also known as EAM solution -widely used by the majority of nuclear operators- is the potential vehicle to bridge the 2 worlds of PSA and daily operation. Most EAM solution get reliability management functionalities which are not really integrated with PSA tools and data and thus cannot provide the anticipated benefits of addressing typical aging phenomena beyond the only predictive models used by the PSA studies. The paper will also discuss potential integration scenario between PSA tools and EAM solutions. (authors)

  16. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1990-01-01

    Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

  17. Safety assessment as basis for the decision making process

    International Nuclear Information System (INIS)

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

    2005-01-01

    This paper deals with the safety assessment for a new near surface repository, particularly for the early stage of repository development using ISAM (Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities) safety assessment methodology. In this stage of the repository life cycle the main purpose of the safety assessment is to demonstrate that the plant is capable to be constructed and operated safely. The paper is based on development of the ASAM (Application of the Safety Assessment Methodologies for Near-Surface Disposal Facilities) Decision Support Subgroup of the Common Aspects Working Group. The implications of decision making for the application of the ISAM methodology on post-closure safety assessment are analysed. Some important elements of the decision-making process with impact on key components of the ISAM process are described. Following the development of Decision Support Subgroup of the ASAM Common Aspects Working Group the proposed change of ISAM methodology is analysed. This approach puts all activities in a decision context where the first iteration of the safety assessment is based on the existing state of knowledge and the initial engineering design. Confidence in the process is accomplished through the direct inclusion of all decision makers and stakeholders in the formulation of decisions, the definition of the state of knowledge, and decision making activities. The decision process is developed in context of undertaking assessments with little site-specific information, this situation is specifically for new planned repository. Limited site-specific information can result in a high degree of uncertainty, therefore it is important first of all to identify the sources of uncertainty arising from the limited nature of the site-specific information and then to apply appropriate approaches to manage the uncertainties and to determine whether the uncertainties are important to the overall safety of the disposal facility

  18. Assessment of safety culture at INPP

    International Nuclear Information System (INIS)

    Lesin, S.

    2002-01-01

    Safety Culture covers all main directions of plant activities and the plant departments involved through integration into the INPP Quality Assurance System. Safety Culture is represented by three components. The first is the clear INPP Safety and Quality Assurance Policy. Based on the Policy INPP is safely operated and managers' actions firstly aim at safety assurance. The second component is based on personal responsibility for safety and attitude of each employee of the plant. The third component is based on commitment to safety and competence of managers and employees of the plant. This component links the first two to ensure efficient management of safety at the plant. The above mentioned components including the elements which may significantly affect Safety Culture are also presented in the attachment. The concept of such model implies understanding of effect of different factors on the level of Safety Culture in the organization. In order to continuously correct safety problems, self-assessment of the Safety Culture level is performed at regular intervals. (author)

  19. Flamanville 3 EPR, Safety Assessment and On-site Inspections

    International Nuclear Information System (INIS)

    Piedagnel, Corinne; Tarallo, Francois; Monnot, Bernard

    2011-01-01

    As a Technical Support Organisation of the French Safety Authority (ASN), the IRSN carries out the safety assessment of EPR project design and participates in the ASN inspections performed at the construction site and in factories. The design assessment consists in defining the safety functions which should be ensured by civil structures, evaluating the EPR Technical Code for Civil works (ETC-C) in which EdF has defined design criteria and construction rules, and carrying out a detailed assessment of a selection of safety-related structures. Those detailed assessments do not consist of a technical control but of an analysis whose objectives are to ensure that design and demonstrations are robust, in accordance with safety and regulatory rules. Most assessments led IRSN to ask EdF to provide additional justification sometimes involving significant modifications. In the light of those complementary justifications and modifications, IRSN concluded that assessments carried out on design studies were globally satisfactory. The participation of IRSN to the on-site inspections led by ASN is a part of the global control of the compliance of the reactor with its safety objectives. For that purpose IRSN has defined a methodology and an inspection program intended to ASN: based on safety functions associated with civil works (confinement and resistance to aggressions), the corresponding behaviour requirements are identified and linked to a list of main civil works elements. During the inspections, deviations to the project's technical specifications or to the rules of the art were pointed out by IRSN. Those deviations cover various items, such as concrete fabrication, concrete pouring methodology, lack of reinforcement in some structures, unadapted welding procedures of the containment leak-tight steel liner and unsatisfactory treatment of concreting joints. The analysis of those problems has revealed flaws in the organisation of the contractors teams together with an

  20. Preliminary safety assessment of the WIPP facility

    International Nuclear Information System (INIS)

    Balestri, R.J.; Torres, B.W.; Pahwa, S.B.; Brannen, J.P.

    1979-01-01

    This paper summarizes the efforts to perform a safety assessment of the Waste Isolation Pilot Plant (WIPP) facility being proposed for southeastern New Mexico. This preliminary safety assessment is limited to a consequence assessment in terms of the dose to a maximally exposed individual as a result of introducing the radionuclides into the biosphere. The extremely low doses to the organs as a result of the liquid breach scenarios are contrasted with the background radiation

  1. Safety assessment in plant layout design using indexing approach: implementing inherent safety perspective. Part 1 - guideword applicability and method description.

    Science.gov (United States)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-12-15

    Layout planning plays a key role in the inherent safety performance of process plants since this design feature controls the possibility of accidental chain-events and the magnitude of possible consequences. A lack of suitable methods to promote the effective implementation of inherent safety in layout design calls for the development of new techniques and methods. In the present paper, a safety assessment approach suitable for layout design in the critical early phase is proposed. The concept of inherent safety is implemented within this safety assessment; the approach is based on an integrated assessment of inherent safety guideword applicability within the constraints typically present in layout design. Application of these guidewords is evaluated along with unit hazards and control devices to quantitatively map the safety performance of different layout options. Moreover, the economic aspects related to safety and inherent safety are evaluated by the method. Specific sub-indices are developed within the integrated safety assessment system to analyze and quantify the hazard related to domino effects. The proposed approach is quick in application, auditable and shares a common framework applicable in other phases of the design lifecycle (e.g. process design). The present work is divided in two parts: Part 1 (current paper) presents the application of inherent safety guidelines in layout design and the index method for safety assessment; Part 2 (accompanying paper) describes the domino hazard sub-index and demonstrates the proposed approach with a case study, thus evidencing the introduction of inherent safety features in layout design.

  2. LNG Safety Assessment Evaluation Methods

    Energy Technology Data Exchange (ETDEWEB)

    Muna, Alice Baca [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); LaFleur, Angela Christine [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-05-01

    Sandia National Laboratories evaluated published safety assessment methods across a variety of industries including Liquefied Natural Gas (LNG), hydrogen, land and marine transportation, as well as the US Department of Defense (DOD). All the methods were evaluated for their potential applicability for use in the LNG railroad application. After reviewing the documents included in this report, as well as others not included because of repetition, the Department of Energy (DOE) Hydrogen Safety Plan Checklist is most suitable to be adapted to the LNG railroad application. This report was developed to survey industries related to rail transportation for methodologies and tools that can be used by the FRA to review and evaluate safety assessments submitted by the railroad industry as a part of their implementation plans for liquefied or compressed natural gas storage ( on-board or tender) and engine fueling delivery systems. The main sections of this report provide an overview of various methods found during this survey. In most cases, the reference document is quoted directly. The final section provides discussion and a recommendation for the most appropriate methodology that will allow efficient and consistent evaluations to be made. The DOE Hydrogen Safety Plan Checklist was then revised to adapt it as a methodology for the Federal Railroad Administration’s use in evaluating safety plans submitted by the railroad industry.

  3. Probabilistic safety assessment for research reactors

    International Nuclear Information System (INIS)

    1986-12-01

    Increasing interest in using Probabilistic Safety Assessment (PSA) methods for research reactor safety is being observed in many countries throughout the world. This is mainly because of the great ability of this approach in achieving safe and reliable operation of research reactors. There is also a need to assist developing countries to apply Probabilistic Safety Assessment to existing nuclear facilities which are simpler and therefore less complicated to analyse than a large Nuclear Power Plant. It may be important, therefore, to develop PSA for research reactors. This might also help to better understand the safety characteristics of the reactor and to base any backfitting on a cost-benefit analysis which would ensure that only necessary changes are made. This document touches on all the key aspects of PSA but placed greater emphasis on so-called systems analysis aspects rather than the in-plant or ex-plant consequences

  4. Road safety impact assessment RIA : a proposal for tools and procedures for a RIA. A study commissioned by the European Commission DG VII.

    NARCIS (Netherlands)

    Wegman, F.C.M. Roszbach, R. Mulder, J.A.G. Schoon, C.C. & Poppe, F.

    1994-01-01

    This study develops a Road Safety Impact Assessment (RIA) tool. RIAs could be made on a more strategic level, and on an individual project or scheme level. On a strategic level, the study suggests to assess safety consequences of changes of traffic over a road network due to infrastructural projects

  5. Safety assessment of radioactive wastes storage 'Mironova Gora'

    International Nuclear Information System (INIS)

    Serbryakov, B.; Karamushka, V.; Ostroborodov, V.

    2000-01-01

    A project of transforming the radioactive wastes storage 'Mironova Gora' is under development. A safety assessment of this storage facility was performed to gain assurance on the design decision. The assessment, which was based on the safety assessment methods developed for radioactive wastes repositories, is presented in this paper. (author)

  6. The probability safety assessment impact on the BR2 refurbishment

    International Nuclear Information System (INIS)

    Pouleur, Yvan

    1995-01-01

    The probabilistic safety assessment (PSA) study has proven its worth by establishing a sensitive safety screening of the reactor. It has focused engineering forces to technically improve safety systems and to measure the influence of functional modifications. In the future, the project will be developed in a living way, to reinforce the present structure along with continuous safety monitoring of the reactor and to develop engineers and operators safety skills. This paper presents the PSA impact on the BR2 (Belgian Reactor Two) refurbishment. (author)

  7. NASA Aviation Safety Program Systems Analysis/Program Assessment Metrics Review

    Science.gov (United States)

    Louis, Garrick E.; Anderson, Katherine; Ahmad, Tisan; Bouabid, Ali; Siriwardana, Maya; Guilbaud, Patrick

    2003-01-01

    The goal of this project is to evaluate the metrics and processes used by NASA's Aviation Safety Program in assessing technologies that contribute to NASA's aviation safety goals. There were three objectives for reaching this goal. First, NASA's main objectives for aviation safety were documented and their consistency was checked against the main objectives of the Aviation Safety Program. Next, the metrics used for technology investment by the Program Assessment function of AvSP were evaluated. Finally, other metrics that could be used by the Program Assessment Team (PAT) were identified and evaluated. This investigation revealed that the objectives are in fact consistent across organizational levels at NASA and with the FAA. Some of the major issues discussed in this study which should be further investigated, are the removal of the Cost and Return-on-Investment metrics, the lack of the metrics to measure the balance of investment and technology, the interdependencies between some of the metric risk driver categories, and the conflict between 'fatal accident rate' and 'accident rate' in the language of the Aviation Safety goal as stated in different sources.

  8. Audit of data and code use in the SR-Can safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Hicks, T.W.; Baldwin, T.D. [Galson Sciences Ltd, 5 Grosvenor House, Melton R oad, Oakham, Rutland LE15 6AX (United Kingdom)

    2008-03-15

    Building on the findings of previous studies on data and code quality assurance (QA) in safety assessments, this report provides a review of data and code QA in the SR-Can safety assessment. The data quality audit aimed to check that the selection and use of data in the SR-Can safety assessment was appropriate, focusing on the data that underpin representations of and assumptions about canister, insert, buffer, and backfill behaviour. The SR-Can Data Report provided the initial focus for examining the traceability and reliability of data used in the safety assessment; the Data Report is one of the series of SR-Can safety assessment reports and, in this review, it was anticipated that it would provide the primary source of data on the canister, insert, buffer, and backfill. However, other safety assessment reports (the SR-Can Main Report, the Initial State Report, the Fuel and Canister Process Report, and the Buffer and Backfill Process Report) were found to provide key information on data used in the safety assessment. The quality audit of codes aimed to check that code use in the SR-Can safety assessment has been justified through a transparent and traceable process of code development and selection. The Model Summary Report provided the focus for reviewing the QA status of the codes used in the safety assessment. As well as highlighting a number of concerns regarding QA aspects of specific data sets, parameter values, and codes used in the SR-Can safety assessment (which are presented in the report), the review has led to several general observations on data and code QA that should be considered by SKB in the development and implementation of a QA system for the SR-Site safety assessment: - The SR-Site safety assessment and associated QA records should include information that demonstrates that a full QA system has been implemented in order to build confidence in the validity of the assessment. - The data and parameter values used directly in the safety

  9. Self-assessment of operational safety for nuclear power plants

    International Nuclear Information System (INIS)

    1999-12-01

    Self-assessment processes have been continuously developed by nuclear organizations, including nuclear power plants. Currently, the nuclear industry and governmental organizations are showing an increasing interest in the implementation of this process as an effective way for improving safety performance. Self-assessment involves the use of different types of tools and mechanisms to assist the organizations in assessing their own safety performance against given standards. This helps to enhance the understanding of the need for improvements, the feeling of ownership in achieving them and the safety culture as a whole. Although the primary beneficiaries of the self-assessment process are the plant and operating organization, the results of the self-assessments are also used, for example, to increase the confidence of the regulator in the safe operation of an installation, and could be used to assist in meeting obligations under the Convention on Nuclear Safety. Such considerations influence the form of assessment, as well as the type and detail of the results. The concepts developed in this report present the basic approach to self-assessment, taking into consideration experience gained during Operational Safety Review Team (OSART) missions, from organizations and utilities which have successfully implemented parts of a self-assessment programme and from meetings organized to discuss the subject. This report will be used in IAEA sponsored workshops and seminars on operational safety that include the topic of self-assessment

  10. Analysis of truncation limit in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Cepin, Marko

    2005-01-01

    A truncation limit defines the boundaries of what is considered in the probabilistic safety assessment and what is neglected. The truncation limit that is the focus here is the truncation limit on the size of the minimal cut set contribution at which to cut off. A new method was developed, which defines truncation limit in probabilistic safety assessment. The method specifies truncation limits with more stringency than presenting existing documents dealing with truncation criteria in probabilistic safety assessment do. The results of this paper indicate that the truncation limits for more complex probabilistic safety assessments, which consist of larger number of basic events, should be more severe than presently recommended in existing documents if more accuracy is desired. The truncation limits defined by the new method reduce the relative errors of importance measures and produce more accurate results for probabilistic safety assessment applications. The reduced relative errors of importance measures can prevent situations, where the acceptability of change of equipment under investigation according to RG 1.174 would be shifted from region, where changes can be accepted, to region, where changes cannot be accepted, if the results would be calculated with smaller truncation limit

  11. Efficacy and safety assessment of microbiological feed additive for chicken broilers in tolerance studies

    Directory of Open Access Journals (Sweden)

    Kupryś-Caruk Marta

    2018-03-01

    Full Text Available Introduction: One aim of the study was to evaluate the impact when added to feed of the two potentially probiotic strains of lactic acid bacteria (LAB Lactobacillus plantarum K KKP 593/p and Lactobacillus rhamnosus KKP 825 on production performance, health, and the composition of gut microbiota. The complementary aim was to assess the safety of these strains in broiler rearing.

  12. Economic aspects of risk assessment in chemical safety

    Energy Technology Data Exchange (ETDEWEB)

    Drummond, M F; Shannon, H S

    1986-05-01

    This paper considers how the economic aspects of risk assessment in chemical safety can be strengthened. Its main focus is on how economic appraisal techniques, such as cost-benefit and cost-effectiveness analysis, can be adapted to the requirements of the risk-assessment process. Following a discussion of the main methodological issues raised by the use of economic appraisal, illustrated by examples from the health and safety field, a number of practical issues are discussed. These include the consideration of the distribution of costs, effects and benefits, taking account of uncertainty, risk probabilities and public perception, making the appraisal techniques useful to the early stages of the risk-assessment process and structuring the appraisal to permit continuous feedback to the participants in the risk-assessment process. It is concluded that while the way of thinking embodied in economic appraisal is highly relevant to the consideration of choices in chemical safety, the application of these principles in formal analysis of risk reduction procedures presents a more mixed picture. The main suggestions for improvement in the analyses performed are the undertaking of sensitivity analyses of study results to changes in the key assumptions, the presentation of the distribution of costs and benefits by viewpoint, the comparison of health and safety measures in terms of their incremental cost per life-year (or quality-adjusted life-year) gained and the more frequent retrospective review and revision of the economic analyses that are undertaken.

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

    International Nuclear Information System (INIS)

    1994-01-01

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

  14. Assessment of safety culture from the INB organization: A case study for nuclear fuel cycle industry

    International Nuclear Information System (INIS)

    Goncalves, J.S.; Barreto, A.C.

    2002-01-01

    The present article describes strategies, methodologies and first results on the Safety Culture Self-assessment Project under way at INB since August 2001. As a Brazilian Government company in charge of the nuclear fuel cycle activities,. the main purposes of the Project is to evaluate the present status of its safety culture and to propose actions to ensure continuous safety improvement at management level of its industrial processes. The proposed safety culture assessment describes INB's various production sites taking into account the different aspects of their activities, such as regional, social and technical issues. The survey was performed in March/2002 very good attendance (about 80%) the employees. The first global survey results are presented in item 4. (author)

  15. Safety Assessment of Polyether Lanolins as Used in Cosmetics.

    Science.gov (United States)

    Becker, Lillian C; Bergfeld, Wilma F; Belsito, Donald V; Hill, Ronald A; Klaassen, Curtis D; Liebler, Daniel C; Marks, James G; Shank, Ronald C; Slaga, Thomas J; Snyder, Paul W; Andersen, F Alan; Heldreth, Bart

    The Cosmetic Ingredient Review (CIR) Expert Panel (Panel) assessed the safety of 39 polyether lanolin ingredients as used in cosmetics. These ingredients function mostly as hair conditioning agents, skin conditioning agent-emollients, and surfactant-emulsifying agents. The Panel reviewed available animal and clinical data, from previous CIR safety assessments of related ingredients and components. The similar structure, properties, functions, and uses of these ingredients enabled grouping them and using the available toxicological data to assess the safety of the entire group. The Panel concluded that these polyether lanolin ingredients are safe in the practices of use and concentration as given in this safety assessment.

  16. The German nuclear power plant safety study

    International Nuclear Information System (INIS)

    1979-01-01

    With this study a new approach has been chosen, taking nuclear power plants as an example to assess and to describe the risks arising from the use of modern technology, including those hazards emanating from the rather hypothetical possibility of occurrence of very serious accidents. Following the definition of basic concepts and methods to be applied in risk assessment studied, as well as a brief account of the design and operating mode of nuclear power plants with PWRs', accidents and failures to be considered in a safety study are described. Using the course-of-event and fault tree analysis, the probability of fission product release as a consequence of failures in safety systems or of core meltdown is evaluated. Subsequently, the theoretical model for assessment of reactor accident consequences is presented, discussing such aspects as the dispersion of radioactivity in the atmosphere, the radiation dose model, safety and countermeasures, the model for the evaluation of health hazards as well as methods and calculations for estimating the reliability of risk assessments together with the remaining uncertainties. In an appendix to this study, the analyses presented in the study are discussed in the light of the TMI-2 event. This safety study showing the possibilities of detecting, keeping in check and minimizing harmful effects, can be regarded as a contribution to a better understanding of our modern, highly industrialised society, and eventually to an improvement of the quality of life. (GL) 891 GL/GL 892 MB [de

  17. Probabilistic safety assessment as a standpoint for decision making

    International Nuclear Information System (INIS)

    Cepin, M.

    2001-01-01

    This paper focuses on the role of probabilistic safety assessment in decision-making. The prerequisites for use of the results of probabilistic safety assessment and the criteria for the decision-making based on probabilistic safety assessment are discussed. The decision-making process is described. It provides a risk evaluation of impact of the issue under investigation. Selected examples are discussed, which highlight the described process. (authors)

  18. The Safety Assessment Framework Tool (SAFRAN) - Description, Overview and Applicability

    International Nuclear Information System (INIS)

    Alujevic, Luka

    2014-01-01

    The SAFRAN tool (Safety Assessment Framework) is a user-friendly software application that incorporates the methodologies developed in the SADRWMS (Safety Assessment Driven Radioactive Waste Management Solutions) project. The International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of all types of radioactive waste, including disused sources, small volumes, legacy and decommissioning waste, operational waste, and large volume naturally occurring radioactive material residues. SAFRAN provides aid in: Describing the predisposal RW management activities in a systematic way, Conducting the SA (safety assessment) with clear documentation of the methodology, assumptions, input data and models, Establishing a traceable and transparent record of the safety basis for decisions on the proposed RW management solutions, Demonstrating clear consideration of and compliance with national and international safety standards and recommendations. The SAFRAN tool allows the user to visibly, systematically and logically address predisposal radioactive waste management and decommissioning challenges in a structured way. It also records the decisions taken in such a way that it constitutes a justifiable safety assessment of the proposed management solutions. The objective of this paper is to describe the SAFRAN architecture and features, properly define the terms safety case and safety assessment, and to predict the future development of the SAFRAN tool and assess its applicability to the construction of a future LILW (Low and Intermediate Level Waste) storage facility and repository in Croatia, taking into account all the capabilities and modelling features of the SAFRAN tool. (author)

  19. Safety factors for neutron fluences in NPP safety assessment

    International Nuclear Information System (INIS)

    Demekhin, V.L.; Bukanov, V.N.; Il'kovich, V.V.; Pugach, A.M.

    2016-01-01

    In accordance with global practice and a number of existing regulations, the use of conservative approach is required for the calculations related to nuclear safety assessment of NPP. It implies the need to consider the determination of neutron fluence errors that is rather complicated. It is proposed to carry out the consideration by the way of multiplying the neutron fluences obtained with transport calculations by safety factors. The safety factor values are calculated by the developed technique based on the theory of errors, features of the neutron transport calculation code and the results obtained with the code. It is shown that the safety factor value is equal 1.18 with the confidence level of not less than 0.95 for the majority of VVER-1000 reactor places where neutron fluences are determined by MCPV code, and its maximum value is 1.25

  20. Need for an "integrated safety assessment" of GMOs, linking food safety and environmental considerations.

    Science.gov (United States)

    Haslberger, Alexander G

    2006-05-03

    Evidence for substantial environmental influences on health and food safety comes from work with environmental health indicators which show that agroenvironmental practices have direct and indirect effects on human health, concluding that "the quality of the environment influences the quality and safety of foods" [Fennema, O. Environ. Health Perspect. 1990, 86, 229-232). In the field of genetically modified organisms (GMOs), Codex principles have been established for the assessment of GM food safety and the Cartagena Protocol on Biosafety outlines international principles for an environmental assessment of living modified organisms. Both concepts also contain starting points for an assessment of health/food safety effects of GMOs in cases when the environment is involved in the chain of events that could lead to hazards. The environment can act as a route of unintentional entry of GMOs into the food supply, such as in the case of gene flow via pollen or seeds from GM crops, but the environment can also be involved in changes of GMO-induced agricultural practices with relevance for health/food safety. Examples for this include potential regional changes of pesticide uses and reduction in pesticide poisonings resulting from the use of Bt crops or influences on immune responses via cross-reactivity. Clearly, modern methods of biotechnology in breeding are involved in the reasons behind the rapid reduction of local varieties in agrodiversity, which constitute an identified hazard for food safety and food security. The health/food safety assessment of GM foods in cases when the environment is involved needs to be informed by data from environmental assessment. Such data might be especially important for hazard identification and exposure assessment. International organizations working in these areas will very likely be needed to initiate and enable cooperation between those institutions responsible for the different assessments, as well as for exchange and analysis of

  1. Ensuring the quality of occupational safety risk assessment.

    Science.gov (United States)

    Pinto, Abel; Ribeiro, Rita A; Nunes, Isabel L

    2013-03-01

    In work environments, the main aim of occupational safety risk assessment (OSRA) is to improve the safety level of an installation or site by either preventing accidents and injuries or minimizing their consequences. To this end, it is of paramount importance to identify all sources of hazards and assess their potential to cause problems in the respective context. If the OSRA process is inadequate and/or not applied effectively, it results in an ineffective safety prevention program and inefficient use of resources. An appropriate OSRA is an essential component of the occupational safety risk management process in industries. In this article, we performed a survey to elicit the relative importance for identified OSRA tasks to enable an in-depth evaluation of the quality of risk assessments related to occupational safety aspects on industrial sites. The survey involved defining a questionnaire with the most important elements (tasks) for OSRA quality assessment, which was then presented to safety experts in the mining, electrical power production, transportation, and petrochemical industries. With this work, we expect to contribute to the main question of OSRA in industries: "What constitutes a good occupational safety risk assessment?" The results obtained from the questionnaire showed that experts agree with the proposed OSRA process decomposition in steps and tasks (taxonomy) and also with the importance of assigning weights to obtain knowledge about OSRA task relevance. The knowledge gained will enable us, in the near future, to build a framework to evaluate OSRA quality for industrial sites. © 2012 Society for Risk Analysis.

  2. Ventilatory function assessment in safety pharmacology: Optimization of rodent studies using normocapnic or hypercapnic conditions

    International Nuclear Information System (INIS)

    Goineau, Sonia; Rompion, Sonia; Guillaume, Philippe; Picard, Sandra

    2010-01-01

    Although the whole body plethysmography for unrestrained animals is the most widely used method to assess the respiratory risk of new drugs in safety pharmacology, non-appropriate experimental conditions may mask deleterious side effects of some substances. If stimulant or bronchodilatory effects can be easily evidenced in rodents under standard experimental conditions, i.e. normal air breathing and diurnal phase, drug-induced respiratory depression remains more difficult to detect. This study was aimed at comparing the responsiveness of Wistar rats, Duncan Hartley guinea-pigs or BALB/c mice to the respiratory properties of theophylline (50 or 100 mg/kg p.o.) or morphine (30 mg/kg i.p.) under varying conditions (100% air versus 5% CO 2 -enriched air, light versus dark day phase), in order to select the most appropriate experimental conditions to each species for safety airway investigations. Our results showed that under normocapnia the ventilatory depressant effects of morphine can be easily evidenced in mice, slightly observed in guinea-pigs and not detected in rats in any day phase. Slight hypercapnic conditions enhanced the responsiveness of rats to morphine but not that of guinea-pigs and importantly they did not blunt the airway responsiveness of rats to the stimulation and bronchodilation evoked by theophylline, the most widely used reference agent in safety pharmacology studies. In conclusion, hypercapnic conditions associated with the non-invasive whole body plethysmography should be considered for optimizing the assessment of both the ventilatory depressant potential of morphine-like substances or the respiratory stimulant effects of new drugs in the rat, the most extensively used species in rodent safety and toxicological investigations.

  3. Exploiting data from safety investigations and processes to assess performance of safety management aspects

    NARCIS (Netherlands)

    Karanikas, Nektarios

    2016-01-01

    This paper presents an alternative way to use records from safety investigations as a means to support the evaluation of safety management (SM) aspects. Datasets from safety investigation reports and progress records of an aviation organization were analyzed with the scope of assessing safety

  4. NPP Krsko periodic safety review. Safety assessment and analyses

    International Nuclear Information System (INIS)

    Basic, I.; Spiler, J.; Thaulez, F.

    2002-01-01

    Definition of a PSR (Periodic Safety Review) project is a comprehensive safety review of a plant after ten years of operation. The objective is a verification by means of a comprehensive review using current methods that the plant remains safe when judged against current safety objectives and practices and that adequate arrangements are in place to maintain plant safety. The overall goals of the NEK PSR Program are defined in compliance with the basic role of a PSR and the current practice typical for most of the countries in EU. This practice is described in the related guides and good practice documents issued by international organizations. The overall goals of the NEK PSR are formulated as follows: to demonstrate that the plant is as safe as originally intended; to evaluate the actual plant status with respect to aging and wear-out identifying any structures, systems or components that could limit the life of the plant in the foreseeable future, and to identify appropriate corrective actions, where needed; to compare current level of safety in the light of modern standards and knowledge, and to identify where improvements would be beneficial for minimizing deviations at justifiable costs. The Krsko PSR will address the following safety factors: Operational Experience, Safety Assessment, EQ and Aging Management, Safety Culture, Emergency Planning, Environmental Impact and Radioactive Waste.(author)

  5. Operational safety assessment of underground test facilities for mined geologic waste disposal

    International Nuclear Information System (INIS)

    Elder, H.K.

    1993-01-01

    This paper describes the operational safety assessment for the underground facilities for the exploratory studies facility (ESF) at the Yucca Mountain Project. The systematic identification and evaluation of hazards related to the ESF is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach based on the analysis of potential accidents was used since radiological safety analysis was not required. The risk assessment summarized credible accident scenarios and the design provides mitigation of the risks to a level that the facility can be constructed and operated with an adequate level of safety. The risk assessment also provides reasonable assurance that all identifiable major accident scenarios have been reviewed and design mitigation features provided to ensure an adequate level of safety

  6. Safety assessment of a lithium target

    International Nuclear Information System (INIS)

    Burgazzi, Luciano; Roberta, Ferri; Barbara, Giannone

    2006-01-01

    This paper addresses the safety assessment of the lithium target of the International Fusion Materials Irradiation Facility (IFMIF) through evaluating the most important risk factors related to system operation and verifying the fulfillment of the safety criteria. The hazard assessment is based on using a well-structured Failure Mode and Effect Analysis (FMEA) procedure by detailing on a component-by-component basis all the possible failure modes and identifying their effects on the plant. Additionally, a systems analysis, applying the fault tree technique, is performed in order to evaluate, from a probabilistic standpoint, all the relevant and possible failures of each component required for safe system operation and assessing the unavailability of the lithium target system. The last task includes the thermal-hydraulic transient analysis of the target lithium loop, including operational and accident transients. A lithium target loop model is developed, using the RELAP5/Mod3.2 thermal-hydraulic code, which has been modified to include specific features of IFMIF itself. The main conclusions are that target safety is fulfilled, the hazards associated with lithium operation are confined within the IFMIF security boundaries, the environmental impact is negligible, and the plant responds to the simulated transients by being able to reach steady conditions in a safety situation

  7. Environment, safety and health progress assessment manual

    International Nuclear Information System (INIS)

    1992-12-01

    On June 27, 1989, the Secretary of Energy announced a 1O-Point Initiative to strengthen environment,safety, and health (ES ampersand H) programs, and waste management activities at involved conducting DOE production, research, and testing facilities. One of the points independent Tiger Team Assessments of DOE operating facilities. The Office of Special Projects (OSP), EH-5, in the Office of the Assistant Secretary for Environment, Safety and Health, EH-1, was assigned the responsibility to conduct the Tiger Team Assessments. Through June 1992, a total of 35 Tiger Team Assessments were completed. The Secretary directed that Corrective Action Plans be developed and implemented to address the concerns identified by the Tiger Teams. In March 1991, the Secretary approved a plan for assessments that are ''more focused, concentrating on ES ampersand H management, ES ampersand H corrective actions, self-assessment programs, and root-cause related issues.'' In July 1991, the Secretary approved the initiation of ES ampersand H Progress Assessments, as a followup to the Tiger Team Assessments, and in the continuing effort to institutionalize the self-assessment process and line management accountability in the ES ampersand H areas. This volume contains appendices to the Environment, Safety and Health Progress Assessment Manual

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

    International Nuclear Information System (INIS)

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

    1998-10-01

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

  9. The role of natural analogues in safety assessment and acceptability

    International Nuclear Information System (INIS)

    Papp, Toenis

    1987-01-01

    The safety assessment must evaluate the level of safety for a repository, the confidence that can be placed on the assessment and how well the repository can meet the acceptance criteria of the society. Many of the processes and phenomena that govern the long term performance of a deep geologic repository for radioactive waste also take place in nature. To investigate these natural analogues and try to validate the models on which the safety assessment are based is a main task in the effort to build of confidence in the safety assessments. The assessment of the safety of a repository can, however, not only be based on good models. The possible role of natural analogues or natural evidence in other parts of the safety assessment is discussed. Specially with regard to - the need to demonstrate that all relevant processes have been taken into account, and that the important ones have been validated to an acceptable level for relevant parameters spans, -the definition and analysis of external scenarios for the safety assessment and for the claim that all reasonable scenarios have been addressed, - the public confidence in the long-term relevance of the acceptance criteria. (author)

  10. Safety assessment of HLW geological disposal system

    International Nuclear Information System (INIS)

    Naito, Morimasa

    2006-01-01

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

  11. Environmental Restoration Disposal Facility (Project W-296) Safety Assessment

    International Nuclear Information System (INIS)

    Armstrong, D.L.

    1994-08-01

    This Safety Assessment is based on information derived from the Conceptual Design Report for the Environmental Restoration Disposal Facility (DOE/RL 1994) and ancillary documentation developed during the conceptual design phase of Project W-296. The Safety Assessment has been prepared to support the Solid Waste Burial Ground Interim Safety Basis document. The purpose of the Safety Assessment is to provide an evaluation of the design to determine if the process, as proposed, will comply with US Department of Energy (DOE) Limits for radioactive and hazardous material exposures and be acceptable from an overall health and safety standpoint. The evaluation considered affects on the worker, onsite personnel, the public, and the environment

  12. Environmental Restoration Disposal Facility (Project W-296) Safety Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, D.L.

    1994-08-01

    This Safety Assessment is based on information derived from the Conceptual Design Report for the Environmental Restoration Disposal Facility (DOE/RL 1994) and ancillary documentation developed during the conceptual design phase of Project W-296. The Safety Assessment has been prepared to support the Solid Waste Burial Ground Interim Safety Basis document. The purpose of the Safety Assessment is to provide an evaluation of the design to determine if the process, as proposed, will comply with US Department of Energy (DOE) Limits for radioactive and hazardous material exposures and be acceptable from an overall health and safety standpoint. The evaluation considered affects on the worker, onsite personnel, the public, and the environment.

  13. RSAS: a Reactor Safety Assessment System

    International Nuclear Information System (INIS)

    Sebo, D.E.; Dixon, B.W.; Bray, M.A.

    1985-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is being developed for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system generates situation assessments for the NRC Reactor Safety Team based on a limited number of plant parameters, known operator actions, and plant status data. The RSAS rule base currently covers one reactor type. The extension of the rule base to other reactor types is also discussed

  14. Assessment of patient safety culture in private and public hospitals in Peru.

    Science.gov (United States)

    Arrieta, Alejandro; Suárez, Gabriela; Hakim, Galed

    2018-04-01

    To assess the patient safety culture in Peruvian hospitals from the perspective of healthcare professionals, and to test for differences between the private and public healthcare sectors. Patient safety is defined as the avoidance and prevention of patient injuries or adverse events resulting from the processes of healthcare delivery. A non-random cross-sectional study conducted online. An online survey was administered from July to August 2016, in Peru. This study reports results from Lima and Callao, which are the capital and the port region of Peru. A total of 1679 healthcare professionals completed the survey. Participants were physicians, medical residents and nurses working in healthcare facilities from the private sector and public sector. Assessment of the degree of patient safety and 12 dimensions of patient safety culture in hospital units as perceived by healthcare professionals. Only 18% of healthcare professionals assess the degree of patient safety in their unit of work as excellent or very good. Significant differences are observed between the patient safety grades in the private sector (37%) compared to the public sub-sectors (13-15%). Moreover, in all patient safety culture dimensions, healthcare professionals from the private sector give more favorable responses for patient safety, than those from the public sub-systems. The most significant difference in support comes from patient safety administrators through communication and information about errors. Overall, the degree of patient safety in Peru is low, with significant gaps that exist between the private and the public sectors.

  15. Safety assessment and detection methods of genetically modified organisms.

    Science.gov (United States)

    Xu, Rong; Zheng, Zhe; Jiao, Guanglian

    2014-01-01

    Genetically modified organisms (GMOs), are gaining importance in agriculture as well as the production of food and feed. Along with the development of GMOs, health and food safety concerns have been raised. These concerns for these new GMOs make it necessary to set up strict system on food safety assessment of GMOs. The food safety assessment of GMOs, current development status of safety and precise transgenic technologies and GMOs detection have been discussed in this review. The recent patents about GMOs and their detection methods are also reviewed. This review can provide elementary introduction on how to assess and detect GMOs.

  16. Safety assessment for the underground disposal of radioactive wastes

    International Nuclear Information System (INIS)

    1981-01-01

    This document is addressed to authorities and specialists responsible for or involved in planning, performing and reviewing safety assessments of underground radioactive waste repositories. It introduces and discusses in a general manner approaches and areas to be considered in making such safety assessments; its emphasis is on repositories for long-lived radioactive wastes in deep geological formations. It is hoped that this document will contribute to providing a base for a common understanding among the authorities and specialists concerned with the numerous studies involving a variety of scientific disciplines. While providing guidance, the document is also intended to stimulate further international discussion on this subject. It is the intention of the IAEA to develop more specific reports providing examples for the application of safety analyses for underground waste disposal

  17. Safety assessment for the underground disposal of radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-01

    This document is addressed to authorities and specialists responsible for or involved in planning, performing and reviewing safety assessments of underground radioactive waste repositories. It introduces and discusses in a general manner approaches and areas to be considered in making such safety assessments; its emphasis is on repositories for long-lived radioactive wastes in deep geological formations. It is hoped that this document will contribute to providing a base for a common understanding among the authorities and specialists concerned with the numerous studies involving a variety of scientific disciplines. While providing guidance, the document is also intended to stimulate further international discussion on this subject. It is the intention of the IAEA to develop more specific reports providing examples for the application of safety analyses for underground waste disposal.

  18. A cross-sectional study to assess the patient safety culture in the Palestinian hospitals: a baseline assessment for quality improvement.

    Science.gov (United States)

    Elsous, Aymen; Akbari Sari, Ali; Rashidian, Arash; Aljeesh, Yousef; Radwan, Mahmoud; AbuZaydeh, Hatem

    2016-12-01

    To measure and establish a baseline assessment of the patient safety culture in the Palestinian hospitals. A cross-sectional descriptive study using the Arabic version of the Safety Attitude Questionnaire (Short Form 2006). A total of 339 nurses and physicians returned the questionnaire out of 370 achieving a response rate of 91.6%. Four public general hospitals in the Gaza Strip, Palestine. Nurses and physicians were randomly selected using a proportionate random sampling. Data analysis performed using Statistical Package for the Social Sciences software version 20, and p value less than 0.05 was statistically significant. Current status of patient safety culture among healthcare providers and percentage of positive attitudes. Male to female ratio was 2.16:1, and mean age was 36.5 ± 9.4 years. The mean score of Arabic Safety Attitude Questionnaire across the six dimensions on 100-point scale ranged between 68.5 for Job Satisfaction and 48.5 for Working Condition. The percentage of respondents holding a positive attitude was 34.5% for Teamwork Climate, 28.4% for Safety Climate, 40.7% for Stress Recognition, 48.8% for Job Satisfaction, 11.3% for Working Conditions and 42.8% for Perception of Management. Healthcare workers holding positive attitudes had better collaboration with co-workers than those without positive attitudes. Findings are useful to formulate a policy on patient safety culture and targeted a specific safety culture dimension to improve the safety of patients and improve the clinical outcomes within healthcare organisations.

  19. Guidelines for Self-assessment of Research Reactor Safety

    International Nuclear Information System (INIS)

    2018-01-01

    Self-assessment is an organization’s internal process to review its current status, processes and performance against predefined criteria and thereby to provide key elements for the organization’s continual development and improvement. Self-assessment helps the organization to think through what it is expected to do, how it is performing in relation to these expectations, and what it needs to do to improve performance, fulfil the expectations and achieve better compliance with the predefined criteria. This publication provides guidelines for a research reactor operating organization to perform a self-assessment of the safety management and the safety of the facility and to identify gaps between the current situation and the IAEA safety requirements for research reactors. These guidelines also provide a methodology for Member States, regulatory bodies and operating organizations to perform a self-assessment of their application of the provisions of the Code of Conduct on the Safety of Research Reactors. This publication also addresses planning, implementation and follow-up of actions to enhance safety and strengthen application of the Code. The guidelines are applicable to all types of research reactor and critical and subcritical assemblies, at all stages in their lifetimes, and to States, regulatory bodies and operating organizations throughout all phases of research reactor programmes. Research reactor operating organizations can use these guidelines at any time to support self-assessments conducted in accordance with the organization’s integrated management system. These guidelines also serve as a tool for an organization to prepare to receive an IAEA Integrated Safety Assessment of Research Reactors (INSARR) mission. An important result of this is the opportunity for an operating organization to identify focus areas and make safety improvements in advance of an INSARR mission, thereby increasing the effectiveness of the mission and efficiency of the

  20. Safety Culture Assessment at Regulatory Body - PNRA Experience of Implementing IAEA Methodology for Safety Culture Self Assessment

    International Nuclear Information System (INIS)

    Bhatti, S.A.N.; Arshad, N.

    2016-01-01

    The prevalence of a good safety culture is equally important for all kind of organizations involved in nuclear business including operating organizations, designers, regulator, etc., and this should be reflected through all the processes and activities of these organizations. The need for inculcating safety culture into regulatory processes and practices is gradually increasing since the major accident at Fukushima. Accordingly, several international fora in last few years repeatedly highlighted the importance of prevalence of safety culture in regulatory bodies as well. The utilisation of concept of safety culture always remained applicable in regulatory activities of PNRA in the form of core values. After the Fukushima accident, PNRA considered it important to check the extent of utilisation of safety culture concept in organizational activities and decided to conduct its “Safety Culture Self-Assessment (SCSA)” for presenting itself as a role model in-order to endorse the fact that safety culture at regulatory authority plays an important role to influence safety culture at licenced facilities.

  1. Preparing suitable climate scenario data to assess impacts on local food safety

    NARCIS (Netherlands)

    Liu, C.; Hofstra, N.; Leemans, R.

    2015-01-01

    Quantification of climate change impacts on food safety requires food safety assessment with different past and future climate scenario data to compare current and future conditions. This study presents a tool to prepare climate and climate change data for local food safety scenario analysis and

  2. ASCOT guidelines revised 1996 edition. Guidelines for organizational self-assessment of safety culture and for reviews by the assessment of safety culture in organizations team

    International Nuclear Information System (INIS)

    1996-01-01

    In order to properly assess safety culture, it is necessary to consider the contribution of all organizations which have an impact on it. Therefore, while assessing the safety culture in an operating organization it is necessary to address at least its interfaces with the local regulatory agency, utility corporate headquarters and supporting organizations. These guidelines are primarily intended for use by any organization wishing to conduct a self-assessment of safety culture. They should also serve as a basis for conducting an international peer review of the organization's self-assessment carried out by an ASCOT (Assessment of Safety Culture in Organizations Team) mission

  3. Safety assessment of Superba™ krill powder: Subchronic toxicity study in rats

    Directory of Open Access Journals (Sweden)

    Kjetil Berge

    2015-01-01

    Full Text Available The safety of krill powder was assessed in a subchronic 13-week toxicity study where rats were fed krill powder or control diets. The krill powder inclusion in the test diet was 9.67% (w/w. There were no differences noted in body weight or food consumption in either gender. Differences in clinical chemistry values were noted in the krill powder-treated animals, but these findings were of no toxicological significance. A significant decrease in absolute heart weight, but not relative heart weight, was observed in both sexes given krill powder, although no corresponding histological changes were observed. Hepatocyte vacuolation was noted histologically in males fed krill powder. This finding was not associated with other indications of hepatic dysfunction. The no observed adverse effect level (NOAEL for the conditions of this study was considered to be 9.67% krill powder.

  4. Application of fuzzy set theory for safety culture and safety management assessment of Kartini research reactor

    International Nuclear Information System (INIS)

    Syarip; Hauptmanns, U.

    2000-01-01

    The safety culture status of nuclear power plant is usually assessed through interview and/or discussions with personnel and management in plant, and an assessment of the pertinent documentation. The approach for safety culture assessment described in IAEA Safety Series, make uses of a questionnaire composed of questions which require 'Yes' or 'No' as an answer. Hence, it is basically a check-list approach which is quite common for safety assessments in industry. Such a procedure ignores the fact that the expert answering the question usually has knowledge which goes far beyond a mere binary answer. Additionally, many situations cannot readily be described in such restricted terms. Therefore, it was developed a checklist consisting of questions which are formulated such that they require more than a simple 'yes' or 'no' as an answer. This allows one to exploit the expert knowledge of the analyst appropriately by asking him to qualify the degree of compliance of each of the topics examined. The method presented has proved useful in assessing the safety culture and quality of safety management of the research reactor. The safety culture status and the quality of safety management of Kartini research reactor is rated as 'average'. The method is also flexible and allows one to add questions to existing areas or to introduce new areas covering related topics

  5. Criticality safety evaluations - a open-quotes stalking horseclose quotes for integrated safety assessment

    International Nuclear Information System (INIS)

    Williams, R.A.

    1995-01-01

    The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility's criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE

  6. IRSN-ANCCLI partnership. Work session on Complementary safety assessments - November 2011

    International Nuclear Information System (INIS)

    Lachaume, Jean-Luc; Lheureux, Yves; Sene, Monique; Sene, Raymond; Jorel, Martial; Lavarenne, Caroline; Rousseau, Jean-Marie; Rebour, Vincent; Baumont, David; Dupuy, Patricia

    2011-11-01

    After an overview by the ASN of complementary safety assessments and an assessment of 'post-Fukushima' inspections of basic nuclear installations, the contributions (Power Point presentations) of this seminar proposed: the opinion of the Gravelines CLI (local information commission) on the Gravelines complementary safety assessment report, an analysis and discussion by the GSIEN on reports of complementary assessment of safety of nuclear installations with respect to the Fukushima accident, an analysis by the IRSN of complementary safety assessments performed by operators, the IRSN approach to analyze complementary safety assessments, reports on installation conditions, external flooding and seismic hazard, 'meltdown prevention' aspects in the management of accidental situations in EDF reactors

  7. Preliminary study on improving safety culture in Malaysian nuclear industries

    International Nuclear Information System (INIS)

    Ibrahim, Sabariah Kader; Lee, Y. E.

    2012-01-01

    This paper presents preliminary study on safety culture and its implementation in Malaysian nuclear industries by realizing the importance of safety culture; identification of important safety culture attributes; safety culture assessment and the practices to incorporate the identified safety culture attributes in organization. The first section of this paper explains the terms and definitions related to safety culture. Second, for the realization of importance of safety culture in organization, the international operational experiences emphasizing the importance of safety culture are described. Third, important safety culture attributes which are frequently cited in literature are provided. Fourth, methods to assess safety culture in operating organization are described. Finally, the practices to enhance the safety culture in an organization are discussed

  8. Preliminary study on improving safety culture in Malaysian nuclear industries

    Energy Technology Data Exchange (ETDEWEB)

    Ibrahim, Sabariah Kader [KAIST, Daejeon (Korea, Republic of); Lee, Y. E. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-10-15

    This paper presents preliminary study on safety culture and its implementation in Malaysian nuclear industries by realizing the importance of safety culture; identification of important safety culture attributes; safety culture assessment and the practices to incorporate the identified safety culture attributes in organization. The first section of this paper explains the terms and definitions related to safety culture. Second, for the realization of importance of safety culture in organization, the international operational experiences emphasizing the importance of safety culture are described. Third, important safety culture attributes which are frequently cited in literature are provided. Fourth, methods to assess safety culture in operating organization are described. Finally, the practices to enhance the safety culture in an organization are discussed.

  9. Safety distance assessment of industrial toxic releases based on frequency and consequence: A case study in Shanghai, China

    International Nuclear Information System (INIS)

    Yu, Q.; Zhang, Y.; Wang, X.; Ma, W.C.; Chen, L.M.

    2009-01-01

    A case study on the safety distance assessment of a chemical industry park in Shanghai, China, is presented in this paper. Toxic releases were taken into consideration. A safety criterion based on frequency and consequence of major hazard accidents was set up for consequence analysis. The exposure limits for the accidents with the frequency of more than 10 -4 , 10 -5 -10 -4 and 10 -6 -10 -5 per year were mortalities of 1% (or SLOT), 50% (SLOD) and 75% (twice of SLOD) respectively. Accidents with the frequency of less than 10 -6 per year were considered incredible and ignored in the consequence analysis. Taking the safety distance of all the hazard installations in a chemical plant into consideration, the results based on the new criterion were almost smaller than those based on LC50 or SLOD. The combination of the consequence and risk based results indicated that the hazard installations in two of the chemical plants may be dangerous to the protection targets and measurements had to be taken to reduce the risk. The case study showed that taking account of the frequency of occurrence in the consequence analysis would give more feasible safety distances for major hazard accidents and the results were more comparable to those calculated by risk assessment.

  10. Animal-Free Chemical Safety Assessment

    Directory of Open Access Journals (Sweden)

    George D Loizou

    2016-07-01

    Full Text Available The exponential growth of the Internet of Things and the global popularity and remarkable decline in cost of the mobile phone is driving the digital transformation of medical practice. The rapidly maturing digital, nonmedical world of mobile (wireless devices, cloud computing and social networking is coalescing with the emerging digital medical world of omics data, biosensors and advanced imaging which offers the increasingly realistic prospect of personalized medicine. Described as a potential seismic shift from the current healthcare model to a wellness paradigm that is predictive, preventative, personalized and participatory, this change is based on the development of increasingly sophisticated biosensors which can track and measure key biochemical variables in people. Additional key drivers in this shift are metabolomic and proteomic signatures, which are increasingly being reported as pre-symptomatic, diagnostic and prognostic of toxicity and disease. These advancements also have profound implications for toxicological evaluation and safety assessment of pharmaceuticals and environmental chemicals. An approach based primarily on human in vivo and high-throughput in vitro human cell-line data is a distinct possibility. This would transform current chemical safety assessment practise which operates in a human data poor to a human data rich environment. This could also lead to a seismic shift from the current animal-based to an animal-free chemical safety assessment paradigm.

  11. 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

  12. 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

  13. Additional safety assessments. Report by the Nuclear Safety Authority - December 2011

    International Nuclear Information System (INIS)

    2011-12-01

    The first part of this voluminous report proposes an assessment of targeted audits performed in French nuclear installations (water pressurized reactors on the one hand, laboratories, factories and waste and dismantling installations on the other hand) on issues related to the Fukushima accident. The examined issues were the protection against flooding and against earthquake, and the loss of electricity supplies and of cooling sources. The second part addresses the additional safety assessments of the reactors and the European resistance tests: presentation of the French electronuclear stock, earthquake, flooding and natural hazards (installation sizing, safety margin assessment), loss of electricity supplies and cooling systems, management of severe accidents, subcontracting conditions. The third part addresses the same issues for nuclear installations other than nuclear power reactors

  14. Probabilistic safety assessment of the Fugen NPS

    International Nuclear Information System (INIS)

    Sotsu, Masutake; Iguchi, Yukihiro; Mizuno, Kouichi; Sato, Shinichirou; Shimizu, Miwako

    1999-01-01

    We performed a probabilistic safety assessment (PSA) on the Fugen NPS. The main topic of assessment was internal factors. We assessment core damage frequency (level 1 PSA) and containment damage frequency (level 2 PSA) during rated operation, and core damage frequency during shutdown (PSA during shutdowns). Our assessment showed that the core damage frequency of Fugen is well below the IAEA criteria for existing plants, that the conditional containment damage during shutdown is almost the target value of 0.1, and that the core damage frequency during shutdown is almost the same as that assessed during operation. These results confirm that the Fugen plant maintains a sufficient safety margin during shutdowns for regular inspections and for refueling. We developed and verified the effectiveness of an accident management plan incorporating the results of the assessment. (author)

  15. Probabilistic safety assessment based expert systems in support of dynamic risk assessment

    International Nuclear Information System (INIS)

    Varde, P.V.; Sharma, U.L.; Marik, S.K.; Raina, V.K.; Tikku, A.C.

    2006-01-01

    Probabilistic Safety Assessment (PSA) studies are being performed, world over as part of integrated risk assessment for Nuclear Power Plants and in many cases PSA insight is utilized in support of decision making. Though the modern plants are built with inherent safety provisions, particularly to reduce the supervisory requirements during initial period into the accident, it is always desired to develop an efficient user friendly real-time operator advisory system for handling of plant transients/emergencies which would be of immense benefit for the enhancement of operational safety of the plant. This paper discusses an integrated approach for the development of operator support system. In this approach, PSA methodology and the insight obtained from PSA has been utilized for development of knowledge based or rule based experts system. While Artificial Neural Network (ANN) approach has been employed for transient identification, rule-base expert system shell environment was used for the development of diagnostic module in this system. Attempt has been made to demonstrate that this approach offers an efficient framework for addressing requirements related to handling of real-time/dynamic scenario. (author)

  16. Safety standards for near surface disposal and the safety case and supporting safety assessment for demonstrating compliance with the standards

    International Nuclear Information System (INIS)

    Metcalf, P.

    2003-01-01

    The report presents the safety standards for near surface disposal (ICRP guidance and IAEA standards) and the safety case and supporting safety assessment for demonstrating compliance with the standards. Special attention is paid to the recommendations for disposal of long-lived solid radioactive waste. The requirements are based on the principle for the same level of protection of future individuals as for the current generation. Two types of exposure are considered: human intrusion and natural processes and protection measures are discussed. Safety requirements for near surface disposal are discussed including requirements for protection of human health and environment, requirements or safety assessments, waste acceptance and requirements etc

  17. Failure rate data for fusion safety and risk assessment

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    1993-01-01

    The Fusion Safety Program (FSP) at the Idaho National Engineering Laboratory (INEL) conducts safety research in materials, chemical reactions, safety analysis, risk assessment, and in component research and development to support existing magnetic fusion experiments and also to promote safety in the design of future experiments. One of the areas of safety research is applying probabilistic risk assessment (PRA) methods to fusion experiments. To apply PRA, we need a fusion-relevant radiological dose code and a component failure rate data base. This paper describes the FSP effort to develop a failure rate data base for fusion-specific components

  18. Validation study of safety assessment model for radioactive waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Munakata, Masahiro; Takeda, Seiji; Kimura, Hideo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    The JAERI-AECL collaboration research program has been conducted to validate a groundwater flow and radionuclide transport models for safety assessment. JAERI have developed a geostatistical model for radionuclide transport through a heterogeneous geological media and verify using experimental results of field tracer tests. The simulated tracer plumes explain favorably the experimental tracer plumes. A regional groundwater flow and transport model using site-scale parameter obtained from tracer tests have been verified by comparing simulation results with observation ones of natural environmental tracer. (author)

  19. Mathematical Safety Assessment Approaches for Thermal Power Plants

    Directory of Open Access Journals (Sweden)

    Zong-Xiao Yang

    2014-01-01

    Full Text Available How to use system analysis methods to identify the hazards in the industrialized process, working environment, and production management for complex industrial processes, such as thermal power plants, is one of the challenges in the systems engineering. A mathematical system safety assessment model is proposed for thermal power plants in this paper by integrating fuzzy analytical hierarchy process, set pair analysis, and system functionality analysis. In the basis of those, the key factors influencing the thermal power plant safety are analyzed. The influence factors are determined based on fuzzy analytical hierarchy process. The connection degree among the factors is obtained by set pair analysis. The system safety preponderant function is constructed through system functionality analysis for inherence properties and nonlinear influence. The decision analysis system is developed by using active server page technology, web resource integration, and cross-platform capabilities for applications to the industrialized process. The availability of proposed safety assessment approach is verified by using an actual thermal power plant, which has improved the enforceability and predictability in enterprise safety assessment.

  20. Buffer and backfill process report for the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Sellin, Patrik (comp.)

    2006-09-15

    This document compiles information on processes in the buffer and deposition tunnel backfill relevant for long-term safety of a KBS-repository. It supports the safety assessment SR-Can, which is a preparatory step for a safety assessment that will support the licence application for a final repository in Sweden. The purpose of the process reports is to document the scientific knowledge of the processes to a level required for an adequate treatment of the processes in the safety assessment. The documentation is not exhaustive from a scientific point of view, since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of an assessment. However, it must be sufficiently detailed to motivate, by arguments founded on scientific understanding, the treatment of each process in the safety assessment. The purpose is further to determine how to handle each process in the safety assessment at an appropriate degree of detail, and to demonstrate how uncertainties are taken care of, given the suggested handling.

  1. Buffer and backfill process report for the safety assessment SR-Can

    International Nuclear Information System (INIS)

    Sellin, Patrik

    2006-09-01

    This document compiles information on processes in the buffer and deposition tunnel backfill relevant for long-term safety of a KBS-repository. It supports the safety assessment SR-Can, which is a preparatory step for a safety assessment that will support the licence application for a final repository in Sweden. The purpose of the process reports is to document the scientific knowledge of the processes to a level required for an adequate treatment of the processes in the safety assessment. The documentation is not exhaustive from a scientific point of view, since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of an assessment. However, it must be sufficiently detailed to motivate, by arguments founded on scientific understanding, the treatment of each process in the safety assessment. The purpose is further to determine how to handle each process in the safety assessment at an appropriate degree of detail, and to demonstrate how uncertainties are taken care of, given the suggested handling

  2. The Nirex Safety Assessment Research Programme; annual report for 1988/89

    International Nuclear Information System (INIS)

    Cooper, M.J.

    1989-07-01

    This report summarises progress of the Nirex Safety Assessment Research Programme during 1988/89, in support of assessments of the post-emplacement radiological safety of a repository for the disposal of low-level and intermediate-level radioactive waste. During this period the assessments were concentrating on a comparative study of concepts and areas for deep emplacement of waste, in order to assist in the selection of preferred sites, and the research programme was therefore focussed on providing the data necessary for broad comparisons between different options to be made. (author)

  3. Safety assessment of the liquid-fed ceramic melter process

    International Nuclear Information System (INIS)

    Buelt, J.L.; Partain, W.L.

    1980-08-01

    As part of its development program for the solidification of high-level nuclear waste, Pacific Northwest Laboratory assessed the safety issues for a complete liquid-fed ceramic melter (LFCM) process. The LFCM process, an adaption of commercial glass-making technology, is being developed to convert high-level liquid waste from the nuclear fuel cycle into glass. This safety assessment uncovered no unresolved or significant safety problems with the LFCM process. Although in this assessment the LFCM process was not directly compared with other solidification processes, the safety hazards of the LFCM process are comparable to those of other processes. The high processing temperatures of the glass in the LFCM pose no additional significant safety concerns, and the dispersible inventory of dried waste (calcine) is small. This safety assessment was based on the nuclear power waste flowsheet, since power waste is more radioactive than defense waste at the time of solidification, and all accident conditions for the power waste would have greater radiological consequences than those for defense waste. An exhaustive list of possible off-standard conditions and equipment failures was compiled. These accidents were then classified according to severity of consequence and type of accident. Radionuclide releases to the stack were calculated for each group of accidents using conservative assumptions regarding the retention and decontamination features of the process and facility. Two recommendations that should be considered by process designers are given in the safety assessment

  4. A Study on Safety and Risk Assessment of Dangerous Cargo Operations in Oil/Chemical Tankers

    OpenAIRE

    Cenk ŞAKAR; Yusuf ZORBA

    2017-01-01

    The safety and risk assessment of dangerous cargo operations in oil and chemical tankers is a necessary process to prevent possible accidents during these operations. Fire and explosion are the major accidents encountered in tanker operations. In this study, a model was constructed through the Fuzzy Bayes Network Method for the probabilistic relationships between the causes of fire and explosion accidents that could occur during the tank cleaning process. The study is composed of two stages. ...

  5. LANL Safety Conscious Work Environment (SCWE) Self-Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Hargis, Barbara C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-01-29

    On December 21, 2012 Secretary of Energy Chu transmitted to the Defense Nuclear Facilities Safety Board (DNFSB) revised commitments on the implementation plan for Safety Culture at the Waste Treatment and Immobilization Plant. Action 2-5 was revised to require contractors and federal organizations to complete Safety Conscious Work Environment (SCWE) selfassessments and provide reports to the appropriate U.S. Department of Energy (DOE) - Headquarters Program Office by September 2013. Los Alamos National Laboratory (LANL) planned and conducted a Safety Conscious Work Environment (SCWE) Self-Assessment over the time period July through August, 2013 in accordance with the SCWE Self-Assessment Guidance provided by DOE. Significant field work was conducted over the 2-week period August 5-16, 2013. The purpose of the self-assessment was to evaluate whether programs and processes associated with a SCWE are in place and whether they are effective in supporting and promoting a SCWE.

  6. Safety assessment of high consequence robotics system

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  7. A safety assessment of the SEAFP fuel cycle systems

    International Nuclear Information System (INIS)

    Natalizio, A.; Kalyanam, K.; Ciattaglia, S.; Pace, L. di

    1995-01-01

    CFFTP and ENEA participated in a joint safety assessment of the fuel cycle design developed for the SEAFP fusion power reactor study (SEAFP: Safety and Environmental Assessment of Fusion Power). The assessment considered both conventional (deflagation/detonation) and radioactive hazards associated with the handling of significant quantities of hydrogen isotopes (H, D and T). Accordingly, the assessment focused on systems or equipment where either the flow rate, or inventory, of hydrogen isotopes was large. A systematic and thorough assessment of initiating events that can lead to an accidental release of tritium into the environment was the first step of the analysis process. This review demonstrated that, in all cases, there are at least two lines of defence available for mitigating the consequences of such accidents -i.e., secondary confinement (glove box, second pipe, caisson, etc.) and the building confinement, backed-up by an air detritiation capability. Therefore, large releases of tritium to the environment will occur only at very low frequencies. (orig.)

  8. Progress in the development of methodology for fusion safety systems studies

    International Nuclear Information System (INIS)

    Ho, S.K.; Cambi, G.; Ciattaglia, S.; Fujii-e, Y.; Seki, Y.

    1994-01-01

    The development of fusion safety systems-study methodology, including the aspects of schematic classification of overall fusion safety system, qualitative assessment of fusion system for identification of critical accident scenarios, quantitative analysis of accident consequences and risk for safety design evaluation, and system-level analysis of accident consequences and risk for design optimization, by a consortium of international efforts is presented. The potential application of this methodology into reactor design studies will facilitate the systematic assessment of safety performance of reactor designs and enhance the impacts of safety considerations on the selection of design configurations

  9. Safety Assessment Methodologies and Their Application in Development of Near Surface Waste Disposal Facilities--ASAM Project

    International Nuclear Information System (INIS)

    Batandjieva, B.; Metcalf, P.

    2003-01-01

    Safety of near surface disposal facilities is a primary focus and objective of stakeholders involved in radioactive waste management of low and intermediate level waste and safety assessment is an important tool contributing to the evaluation and demonstration of the overall safety of these facilities. It plays significant role in different stages of development of these facilities (site characterization, design, operation, closure) and especially for those facilities for which safety assessment has not been performed or safety has not been demonstrated yet and the future has not been decided. Safety assessments also create the basis for the safety arguments presented to nuclear regulators, public and other interested parties in respect of the safety of existing facilities, the measures to upgrade existing facilities and development of new facilities. The International Atomic Energy Agency (IAEA) has initiated a number of research coordinated projects in the field of development and improvement of approaches to safety assessment and methodologies for safety assessment of near surface disposal facilities, such as NSARS (Near Surface Radioactive Waste Disposal Safety Assessment Reliability Study) and ISAM (Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities) projects. These projects were very successful and showed that there is a need to promote the consistent application of the safety assessment methodologies and to explore approaches to regulatory review of safety assessments and safety cases in order to make safety related decisions. These objectives have been the basis of the IAEA follow up coordinated research project--ASAM (Application of Safety Assessment Methodologies for Near Surface Disposal Facilities), which will commence in November 2002 and continue for a period of three years

  10. Adaptation of the ToxRTool to Assess the Reliability of Toxicology Studies Conducted with Genetically Modified Crops and Implications for Future Safety Testing.

    Science.gov (United States)

    Koch, Michael S; DeSesso, John M; Williams, Amy Lavin; Michalek, Suzanne; Hammond, Bruce

    2016-01-01

    To determine the reliability of food safety studies carried out in rodents with genetically modified (GM) crops, a Food Safety Study Reliability Tool (FSSRTool) was adapted from the European Centre for the Validation of Alternative Methods' (ECVAM) ToxRTool. Reliability was defined as the inherent quality of the study with regard to use of standardized testing methodology, full documentation of experimental procedures and results, and the plausibility of the findings. Codex guidelines for GM crop safety evaluations indicate toxicology studies are not needed when comparability of the GM crop to its conventional counterpart has been demonstrated. This guidance notwithstanding, animal feeding studies have routinely been conducted with GM crops, but their conclusions on safety are not always consistent. To accurately evaluate potential risks from GM crops, risk assessors need clearly interpretable results from reliable studies. The development of the FSSRTool, which provides the user with a means of assessing the reliability of a toxicology study to inform risk assessment, is discussed. Its application to the body of literature on GM crop food safety studies demonstrates that reliable studies report no toxicologically relevant differences between rodents fed GM crops or their non-GM comparators.

  11. Study of parameters for safety assessment of sub-surface disposal. Tunnel-excavating speed and thickness of additional soil in residential land development by filling

    International Nuclear Information System (INIS)

    Ishitoya, Kimihide; Sugaya, Toshikatsu; Funabashi, Hideyuki

    2012-02-01

    Japan Atomic Energy Agency (JAEA) is making preparations for the sub-surface disposal of own low level radioactive wastes. In order to carry out the disposal, it is necessary to confirm the safety of the disposal. Nuclear Safety Commission of Japan (NSC) issued 'Policy of the Safety Assessment of Sub-surface Disposal after the Period for Active Control' (April 1, 2010). Then, we investigated the parameters for dose assessment in tunnel excavation scenario and large-scale land use scenario which were described in the 'Policy of the Safety Assessment', in order to perform the assessment based on actual conditions. To be concrete, we investigated the tunnel excavating speeds in Japan for the former scenario, and investigated technical standards of the filling for the latter scenario. We studied the realistic parameters for the dose assessment with the results of those investigations. (author)

  12. Research on fuzzy comprehensive assessment method of nuclear power plant safety culture

    International Nuclear Information System (INIS)

    Xiang Yuanyuan; Chen Xukun; Xu Rongbin

    2012-01-01

    Considering the traits of safety culture in nuclear plant, 38 safety culture assessment indexes are established from 4 aspects such as safety values, safety institution, safety behavior and safety sub- stances. Based on it, a comprehensive assessment method for nuclear power plant safety culture is constructed by using AHP (Analytic Hierarchy Process) approach and fuzzy mathematics. The comprehensive assessment method has the quality of high precision and high operability, which can support the decision making of safety culture development. (authors)

  13. Safety/security interface assessments at commercial nuclear power plants

    International Nuclear Information System (INIS)

    Byers, K.R.; Brown, P.J.; Norderhaug, L.R.

    1985-01-01

    The findings of the Haynes Task Force Committee (NUREG-0992) are used as the basis for defining safety/security assessment team activities at commercial nuclear power plants in NRC Region V. A safety/security interface assessment outline and the approach used for making the assessments are presented along with the composition of team members. As a result of observing simulated plant emergency conditions during scheduled emergency preparedness exercises, examining security and operational response procedures, and interviewing plant personnel, the team has identified instances where safety/security conflicts can occur

  14. Safety/security interface assessments at commercial nuclear power plants

    International Nuclear Information System (INIS)

    Byers, K.R.; Brown, P.J.; Norderhaug, L.R.

    1985-07-01

    The findings of the Haynes Task Force Committee (NUREG-0992) are used as the basis for defining safety/security assessment team activities at commercial nuclear power plants in NRC Region V. A safety/security interface assessment outline and the approach used for making the assessments are presented along with the composition of team members. As a result of observing simulated plant emergency conditions during scheduled emergency preparedness exercises, examining security and operational response procedures, and interviewing plant personnel, the team has identified instances where safety/security conflicts can occur. 2 refs

  15. Safety assessment for Generation IV nuclear systems

    International Nuclear Information System (INIS)

    Leahy, T.J.

    2012-01-01

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

  16. Method of operator safety assessment for underground mobile mining equipment

    Science.gov (United States)

    Działak, Paulina; Karliński, Jacek; Rusiński, Eugeniusz

    2018-01-01

    The paper presents a method of assessing the safety of operators of mobile mining equipment (MME), which is adapted to current and future geological and mining conditions. The authors focused on underground mines, with special consideration of copper mines (KGHM). As extraction reaches into deeper layers of the deposit it can activate natural hazards, which, thus far, have been considered unusual and whose range and intensity are different depending on the field of operation. One of the main hazards that affect work safety and can become the main barrier in the exploitation of deposits at greater depths is climate threat. The authors have analysed the phenomena which may impact the safety of MME operators, with consideration of accidents that have not yet been studied and are not covered by the current safety standards for this group of miners. An attempt was made to develop a method for assessing the safety of MME operators, which takes into account the mentioned natural hazards and which is adapted to current and future environmental conditions in underground mines.

  17. Method of operator safety assessment for underground mobile mining equipment

    Directory of Open Access Journals (Sweden)

    Działak Paulina

    2018-01-01

    Full Text Available The paper presents a method of assessing the safety of operators of mobile mining equipment (MME, which is adapted to current and future geological and mining conditions. The authors focused on underground mines, with special consideration of copper mines (KGHM. As extraction reaches into deeper layers of the deposit it can activate natural hazards, which, thus far, have been considered unusual and whose range and intensity are different depending on the field of operation. One of the main hazards that affect work safety and can become the main barrier in the exploitation of deposits at greater depths is climate threat. The authors have analysed the phenomena which may impact the safety of MME operators, with consideration of accidents that have not yet been studied and are not covered by the current safety standards for this group of miners. An attempt was made to develop a method for assessing the safety of MME operators, which takes into account the mentioned natural hazards and which is adapted to current and future environmental conditions in underground mines.

  18. Safety assessment requirements for onsite transfers of radioactive material

    International Nuclear Information System (INIS)

    Opperman, E.K.; Jackson, E.J.; Eggers, A.G.

    1992-05-01

    This document contains the requirements for developing a safety assessment document for an onsite package containing radioactive material. It also provides format and content guidance to establish uniformity in the safety assessment documentation and to ensure completeness of the information provided

  19. Geosphere process report for the safety assessment SR-Site

    International Nuclear Information System (INIS)

    Skagius, Kristina

    2010-11-01

    This report documents geosphere processes identified as relevant to the long-term safety of a KBS-3 repository, and forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

  20. Geosphere process report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    Skagius, Kristina (ed.) (Kemakta Konsult AB, Stockholm (Sweden))

    2010-11-15

    This report documents geosphere processes identified as relevant to the long-term safety of a KBS-3 repository, and forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

  1. A study on the assessment of safety culture impacts on risk of nuclear power plants using common uncertainty source model

    International Nuclear Information System (INIS)

    Lee, Yong Suk; Bang, Young Suk; Chung, Chang Hyun; Jeong, Ji Hwan

    2004-01-01

    Since International Safety Advisory Group (INSAG) introduced term 'safety culture', it has been widely recognized that safety culture has an important role in safety of nuclear power plants. Research on the safety culture can be divided in the following two parts. 1) Assessment of safety culture (by interview, questionnaire, etc.) 2) Assessment of link between safety culture and safety of nuclear power plants. There is a substantial body of literature that addresses the first part, but there is much less work that addresses the second part. To address the second part, most work focused on the development of model incorporating safety culture into Probabilistic Safety Assessment (PSA). One of the most advanced methodology in the area of incorporating safety culture quantitatively into PSA is System Dynamics (SD) model developed by Kwak et al. It can show interactions among various factors which affect employees' productivity and job quality. Also various situations in nuclear power plant can be simulated and time-dependent risk can be recalculated with this model. But this model does not consider minimal cut set (MCS) dependency and uncertainty of risk. Another well-known methodology is Work Process Analysis Model (WPAM) developed by Davoudian. It considers MCS dependency by modifying conditional probability values using SLI methodology. But we found that the modified conditional probability values in WPAM are somewhat artificial and have no sound basis. WPAM tend to overestimate conditional probability of hardware failure, because it uses SLI methodology which is normally used in Human Reliability Analysis (HRA). WPAM also does not consider uncertainty of risk. In this study, we proposed methodology to incorporate safety culture into PSA quantitatively that can deal with MCS dependency and uncertainty of risk by applying the Common Uncertainty Source (CUS) model developed by Zhang. CUS is uncertainty source that is common to basic events, and this can be physical

  2. Assessing safety culture using RADAR matrix

    International Nuclear Information System (INIS)

    Mariscal-Saldana, M. a.; Garcia-Herrero, S.; Toca-Otero, A.

    2009-01-01

    Santa Maria de Garona nuclear power plant, in collaboration with Burgos University, has proceeded to conduct a pilot project aimed at seeing the possibilities for the RADAR (Results, Approach, Development, Assessment and review) logic of EFQM model, as a tool for self evaluation of Safety Culture in a nuclear power plant. In the work it has sought evidences of Safety culture implanted in the plant, and identify strengths and areas for improvement regarding this Culture. the score obtained by analyzing these strengths and areas for improvements has served to prioritize actions implemented. The nuclear power plant has been submitted voluntarily to the mission SCART (Safety Culture Assessment Review Team), an international review being done for the first time in the world at a plant in operation and the team of experts led by International Agency of Atomic Energy (IAEA) has identified this project as a good practice, an innovative process implemented in the plant, that must be transmitted to other plants. (Author) 10 refs

  3. Probability safety assessment of the Kozloduy-5 and Kozloduy-6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Boyadzhiev, A; Manchev, B [Risk Engineering Ltd., Sofia (Bulgaria)

    1996-12-31

    A probability safety assessment (PSA) of Level 1 (assessment of plant failures leading to the determination of core damage frequency) has been carried out for the NPP Kozloduy Units 5 and 6 (reactors WWER-1000). The scope of the study includes all significant accident initiators including seismic (earthquake) and fire initiators. Event trees for all initiators and fault trees for front line systems, support systems and major safety systems have been built. A distribution of the different initiators has been established as follows: internal initiators - 85%, seismic initiators - 5%, fire initiators- 10%. The loss of offsite power was identified as main contributor from the internal initiators with frequency 1,1.10{sup -4}/y. It is concluded that the safety functions of WWER-1000 are adequately covered by the safety systems. 4 refs., 2 tabs.

  4. Complementary safety assessment assessment of nuclear facilities - La Hague plant - AREVA

    International Nuclear Information System (INIS)

    2011-01-01

    This complementary safety assessment analyses the robustness of La Hague plant to extreme situations such as those that led to the Fukushima accident. Robustness is the ability for the plant to withstand events beyond which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accident sequence. Moreover, safety is not only a matter of design or engineered systems but also a matter of organizing: task organization (including subcontracting) as well as the setting of emergency plans or the inventory of nuclear materials are taken into consideration in this assessment. This report is divided into 10 main chapters: 1) the feedback experience of the Fukushima accident; 2) description of the site; 3) featuring the activities and installations; 4) accidental sequences 5) protection from the earthquake; 6) protection from the flood; 7) protection from other extreme natural disasters; 8) the loss of electrical power and of the heat sink; 9) the management of severe accidents; and 10) subcontracting policy. This study shows a globally good robustness of the plant for the considered risks and, in the case of a severe accident, specified remedial actions can be brought into play by the staff to secure the installations. (A.C.)

  5. Developing IAM for Life Cycle Safety Assessment

    NARCIS (Netherlands)

    Toxopeus, Marten E.; Lutters, Diederick; Nee, Andrew Y.C.; Song, Bin; Ong, Soh-Khim

    2013-01-01

    This publication discusses aspects of the development of an impact assessment method (IAM) for safety. Compared to the many existing IAM’s for environmentally oriented LCA, this method should translate the impact of a product life cycle on the subject of safety. Moreover, the method should be

  6. Probabilistic assessment methods as a tool for developing nations to make safety decisions

    International Nuclear Information System (INIS)

    Gumley, P.; Inamdar, S.V.

    1985-01-01

    This paper advocates the use of probabilistic safety assessment methods in making safety decisions. It discusses the question of adequate safety - what it means to a country buying a nuclear power plant, and how probabilistic safety assessment studies of the reference plant can be used for ensuring this adequate safety. It is proposed that adequate safety means ensuring that the plant would behave, in accident conditions, in a manner similar to the way it is expected to behave were it in the country of origin. For this one needs to know how the plant responds under somewhat altered conditions. These altered conditions can arise from such factors as varying reliability of electrical grids, different manufacturing technology, local systems design and operator capability. In the design of nuclear power plants, the traditional approach to safety has led to the belief that availability and effectiveness of safety systems alone are all that is required to ensure plant safety. This belief can result in design oversights leading to potential problems arising from the power production systems and the service systems. Participation by the buying country in the design of such systems, and understanding the safety implications thereof, can be facilitated by probabilistic safety assessment methods. This philosophy is illustrated in this paper by examples. (author)

  7. Framework for Continuous Assessment and Improvement of Occupational Health and Safety Issues in Construction Companies

    OpenAIRE

    Mahmoudi, Shahram; Ghasemi, Fakhradin; Mohammadfam, Iraj; Soleimani, Esmaeil

    2014-01-01

    Background: Construction industry is among the most hazardous industries, and needs a comprehensive and simple-to-administer tool to continuously assess and promote its health and safety performance. Methods: Through the study of various standard systems (mainly Health, Safety, and Environment Management System; Occupational Health and Safety Assessment Series 180001; and British Standard, occupational health and safety management systems-Guide 8800), seven main elements were determined fo...

  8. Safety management system needs assessment.

    Science.gov (United States)

    2016-04-01

    The safety of the traveling public is critical as each year there are approximately 200 highway fatalities in Nebraska and numerous crash injuries. The objective of this research was to conduct a needs assessment to identify the requirements of a sta...

  9. Using driving simulators to assess driving safety.

    Science.gov (United States)

    Boyle, Linda Ng; Lee, John D

    2010-05-01

    Changes in drivers, vehicles, and roadways pose substantial challenges to the transportation safety community. Crash records and naturalistic driving data are useful for examining the influence of past or existing technology on drivers, and the associations between risk factors and crashes. However, they are limited because causation cannot be established and technology not yet installed in production vehicles cannot be assessed. Driving simulators have become an increasingly widespread tool to understand evolving and novel technologies. The ability to manipulate independent variables in a randomized, controlled setting also provides the added benefit of identifying causal links. This paper introduces a special issue on simulator-based safety studies. The special issue comprises 25 papers that demonstrate the use of driving simulators to address pressing transportation safety problems and includes topics as diverse as neurological dysfunction, work zone design, and driver distraction. Copyright (c) 2010 Elsevier Ltd. All rights reserved.

  10. Criticality safety evaluations - a {open_quotes}stalking horse{close_quotes} for integrated safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R.A. [Westinghouse Electric Corp., Columbia, SC (United States)

    1995-12-31

    The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility`s criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE.

  11. Safety and security risk assessments--now demystified!

    Science.gov (United States)

    White, Donald E

    2011-01-01

    Safety/security risk assessments no longer need to spook nor baffle healthcare safety/security managers. This grid template provides at-at-glance quick lookup of the possible threats, the affected people and things, a priority ranking of these risks, and a workable solution for each risk. Using the standard document, spreadsheet, or graphics software already available on your computer, you can easily use a scientific method to produce professional looking risk assessments that get quickly understood by both senior managers and first responders alike!

  12. Visualization of Safety Assessment Result Using GIS in SITES

    International Nuclear Information System (INIS)

    Yun, Bong-Yo; Park, Joo Wan; Park, Se-Moon; Kim, Chang-Lak

    2006-01-01

    Site Information and Total Environmental database management System (SITES) is an integrated program for overall data analysis, environmental monitoring, and safety analysis that are produced from the site investigation and environmental assessment of the relevant nuclear facility. SITES is composed of three main modules such as Site Environment Characterization database for Unified and Reliable Evaluation system (SECURE), Safety Assessment INTegration system (SAINT) and Site Useful Data Analysis and ALarm system (SUDAL). The visualization function of safety assessment and environmental monitoring results is designed. This paper is to introduce the visualization design method using Geographic Information System (GIS) for SITES

  13. Radiological safety and risk assessment

    International Nuclear Information System (INIS)

    Hunter, P.H.; Barg, D.C.; Baird, R.D.; Card, D.H.; de Souza, F.; Elder, J.; Felthauser, K.; Jensen, C.; Winkler, V.

    1982-02-01

    A brief radiological safety and risk assessment of a nuclear power generation center with an adjacent on-site waste disposal facility at a specific site in the State of Utah is presented. The assessment was conducted to assist in determining the feasibility and practicality of developing a nuclear energy center (NEC) in Utah consisting of nine 1250 MWe nuclear pressurized water reactor (PWR) electrical generating units arranged in 3 clusters of 3 units each known as triads. The site selected for this conceptual study is in the Horse Bench area about 15 miles directly south of the town of Green River, Utah. The radiological issues included direct radiation exposures to on-site workers and the off-site population, release of radioactive material, and effects of these releases for both normal operations and accidental occurrences. The basic finding of this study is that the concept of an NEC in the Green River area, specifically at the Horse Bench site, is radiologically feasible

  14. Swedish REGULATORY APPROACH TO SAFETY Assessment AND SEVERE ACCIDENT MANAGEMENT

    International Nuclear Information System (INIS)

    Frid, W.; Sandervaag, O.

    1997-01-01

    The Swedish regulatory approach to safety assessment and severe accident management is briefly described. The safety assessment program, which focuses on prevention of incidents and accidents, has three main components: periodic safety reviews, probabilistic safety analysis, and analysis of postulated disturbances and accident progression sequences. Management and man-technology-organisation issues, as well as inspections, play a key role in safety assessment. Basis for severe accident management were established by the Government decisions in 1981 and 1986. By the end of 1988, the severe accident mitigation systems and emergency operating procedures were implemented at all Swedish reactors. The severe accident research has continued after 1988 for further verification of the protection provided by the systems and reduction of remaining uncertainties in risk dominant phenomena

  15. Complementary assessment of the safety of French nuclear power plants

    International Nuclear Information System (INIS)

    Camarcat, N.; Pouget-Abadie, X.

    2011-01-01

    As an immediate consequence of the Fukushima accident the French nuclear safety Authority (ASN) asked EDF to perform a complementary safety assessment for each nuclear power plant dealing with 3 points: 1) the consequences of exceptional natural disasters, 2) the consequences of total loss of electrical power, and 3) the management of emergency situations. The safety margin has to be assessed considering 3 main points: first a review of the conformity to the initial safety requirements, secondly the resistance to events overdoing what the facility was designed to stand for, and the feasibility of any modification susceptible to improve the safety of the facility. This article details the specifications of such assessment, the methodology followed by EDF, the task organization and the time schedule. (A.C.)

  16. Safety assessment of pipes with multiple local wall thinning defects under pressure and bending moment

    International Nuclear Information System (INIS)

    Peng Jian; Zhou Changyu; Xue Jilin; Dai Qiao; He Xiaohua

    2011-01-01

    The safety assessment of pipes with local wall thinning defects is highly important in engineering. Most attention has been paid on the safety assessment of pipe with single local wall thinning defect, while the studies about multiple local wall thinning defects are not nearly enough. However, the interaction of multiple local wall thinning defects in some conditions is great, and may have a great impact on the safety assessment. In the present standard API 579/ASME FFS, the safety assessment of pipes with multiple local wall thinning defects is given, while as well as the influence of load condition, the influences of arrangement and relative depth of defects are ignored, which may influence the safety assessment considerably. In this paper, the influence of the interaction between multiple local wall thinning defects on the remaining strength of pipes at different arrangements and depths of defects under different load conditions (pressure, tension-bending moment and compression-bending moment) are studied. A quantified index is defined to describe the interaction between defects quantitatively. For different arrangements and relative depths of defects, based on a limit value 0.05 of the quantified index of the interaction between defects, a relatively systematic safety assessment of pipes with multiple local wall thinning defects under different load conditions has been proposed.

  17. Assessment of Health, Safety and Environmental Risks of Zahedan City Gasoline Stations

    Directory of Open Access Journals (Sweden)

    S. Y. Far

    2018-04-01

    Full Text Available The purpose of this study was to assess the risk and determine the health, safety and environmental status of fuel stations in Zahedan. In this study, failure mode and effects analysis (FMEA method was used for risk assessment in accordance with the HSE guidelines, national and international standards and laws. In this cross-sectional study, 2 governmental stations and 6 active private stations were evaluated after the necessary coordination with the relevant units. As a result of risk assessment, 27 health risks, 55 safety risks and 22 environmental risks were identified. From among all the identified risks, 67 risks had a Risk Priority Number (RPN of less than 91, 31 risks had an RPN ranging between 91 and 201, and 6 risks had an RPN of over 201. The findings of the study indicated that compliance with the HSE requirements was 51.85%, in the area of health, 47.57% in the area of safety and 27.45% in the environmental area. Overall compliance with the HSE requirements was 42.54%. In order to distribute fuel considering health, reducing risk and increasing compliance with the requirements for safety improvement, health and environmental conditions of fuel supplies are essential.

  18. The role of engineering judgement, safety culture, and organizational factors in risk assessment

    International Nuclear Information System (INIS)

    Muzumdar, Ajit; Professor, Visiting

    1996-01-01

    This paper reviews the role of engineering judgement, safety culture, and organizational factors in risk assessment by examining the reasons for human-based error. The need for more emphasis on producing engineers with good engineering judgement is described. The progress in quantifying the role of safety culture and organizational factors in risk assessment studies is summarized

  19. Progress on the European Safety and Environmental Assessment of Fusion Power (SEAFP)

    International Nuclear Information System (INIS)

    Cook, I.

    1994-01-01

    The Safety and Environmental Assessment of Fusion Power (SEAFP) project was set up by the European Community Fusion Programme in response to recommendations made by a high level Fusion Programme Evaluation Board. The Evaluation Board stated that fusion potentially possesses ''inherent environmental and safety advantages over all current alternatives for base load electricity generation'', but that a ''convincing demonstration'' of these potential advantages is necessary. SEAFP is undertaken by three main participants: the NET Team, The Euratom/UKAEA Association, and European industry. Other EC fusion laboratories also participate. The work embraces the outline design of fusion power stations, the safety and environmental assessment of those designs, and interactions between design and assessment to improve the design. The project began in April 1992 and will report in December 1994. In the first year of the project, five candidate blanket concepts were developed in parallel. Other aspects of design were developed as far as possible independently of the blanket designs. Assessments were made of the technical merits of the candidate designs, and scoping calculations were used to provide preliminary assessments of their accident and waste management characteristics. Accident identification studies were used to select the bounding sequences to be analysed later in detail. Targets for safety and environmental performance were developed. This phase of the study culminated, in August 1993, in the selection of two plant models, one based on a water/martensitic steel/lithium-lead blanket, the other based on a helium/vanadium alloy/lithium oxide blanket, to be developed and assessed in more detail. Other design variants will be assessed through sensitivity studies. ((orig.))

  20. Integrated Deterministic-Probabilistic Safety Assessment Methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Kudinov, P.; Vorobyev, Y.; Sanchez-Perea, M.; Queral, C.; Jimenez Varas, G.; Rebollo, M. J.; Mena, L.; Gomez-Magin, J.

    2014-02-01

    IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) is a family of methods which use tightly coupled probabilistic and deterministic approaches to address respective sources of uncertainties, enabling Risk informed decision making in a consistent manner. The starting point of the IDPSA framework is that safety justification must be based on the coupling of deterministic (consequences) and probabilistic (frequency) considerations to address the mutual interactions between stochastic disturbances (e.g. failures of the equipment, human actions, stochastic physical phenomena) and deterministic response of the plant (i.e. transients). This paper gives a general overview of some IDPSA methods as well as some possible applications to PWR safety analyses. (Author)

  1. Human reliability in probabilistic safety assessments

    International Nuclear Information System (INIS)

    Nunez Mendez, J.

    1989-01-01

    Nowadays a growing interest in medioambiental aspects is detected in our country. It implies an assessment of the risk involved in the industrial processess and installations in order to determine if those are into the acceptable limits. In these safety assessments, among which PSA (Probabilistic Safety Assessments), can be pointed out the role played by the human being in the system is one of the more relevant subjects. (This relevance has been demostrated in the accidents happenned). However in Spain there aren't manuals specifically dedicated to asses the human contribution to risk in the frame of PSAs. This report aims to improve this situation providing: a) a theoretical background to help the reader in the understanding of the nature of the human error, b) a guide to carry out a Human Reliability Analysis and c) a selected overwiev of the techniques and methodologies currently applied in this area. (Author)

  2. Probabilistic assessment of NPP safety under aircraft impact

    International Nuclear Information System (INIS)

    Birbraer, A.N.; Roleder, A.J.; Arhipov, S.B.

    1999-01-01

    Methodology of probabilistic assessment of NPP safety under aircraft impact is described below. The assessment is made taking into account not only the fact of aircraft fall onto the NPP building, but another casual parameters too, namely an aircraft class, velocity and mass, as well as point and angle of its impact with the building structure. This analysis can permit to justify the decrease of the required structure strength and dynamic loads on the NPP equipment. It can also be especially useful when assessing the safety of existing NPP. (author)

  3. Nirex Safety Assessment Research Programme bibliography, 1990

    International Nuclear Information System (INIS)

    Cooper, M.J.

    1990-10-01

    This bibliography lists reports and papers written as part of the Nirex Safety Assessment Research Programme, which is concerned with disposal of low-level and intermediate-level waste (LLW and ILW) and associated radiological assessments. (author)

  4. Understanding and assessing safety culture

    International Nuclear Information System (INIS)

    Dalling, Ian

    1997-01-01

    The 'Dalling' integrated model of organisational performance is introduced and described. A principal element of this model is culture, which is dynamically contrasted with the five other interacting critical elements, which comprise: the management system, the knowledge base, corporate leadership, stakeholders and consciousness. All six of these principal driving elements significantly influence health, safety, environmental, security, or any other aspect of organisational performance. It is asserted that the elements of organisational performance must be clearly defined and understood if meaningful measurements are to be carried out and sustained progress made in improving the knowledge of organisational performance. AEA Technology's safety culture research programme is then described together with the application of a safety culture assessment tool to organisations in the nuclear, electricity, transport, and oil and gas industries, both within and outside of the United Kingdom. (author)

  5. Complementary safety assessment assessment of nuclear facilities - Tricastin facility - AREVA

    International Nuclear Information System (INIS)

    2011-01-01

    This complementary safety assessment analyses the robustness of the Areva part of the Tricastin nuclear site to extreme situations such as those that led to the Fukushima accident. This study includes the following facilities: Areva NC Pierrelatte, EURODIF production, Comurhex Pierrelatte, Georges Besse II plant and Socatri. Robustness is the ability for the plant to withstand events beyond which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accidental sequence. Moreover, safety is not only a matter of design or engineered systems but also a matter of organizing: task organization (including subcontracting) as well as the setting of emergency plans or the inventory of nuclear materials are taken into consideration in this assessment. This report is divided into 10 main chapters: 1) the feedback experience of the Fukushima accident; 2) description of the site and its surroundings; 3) featuring of the site's activities and installations; 4) accidental sequences; 5) protection from earthquakes; 6) protection from floods; 7) protection from other extreme natural disasters; 8) the loss of electrical power and of the heat sink; 9) the management of severe accidents; and 10) subcontracting policy. This analysis has identified 5 main measures to be taken to limit the risks linked to natural disasters: -) continuing the program for replacing the current conversion plant and the enrichment plant; -) renewing the storage of hydrofluoric acid at the de-fluorination workshop; -) assessing the seismic behaviour of some parts of the de-fluorination workshop and of the fluorine fabrication workshop; -) improving the availability of warning and information means in case of emergency; and -) improving the means to mitigate accidental gaseous releases. (A.C.)

  6. Safety and immunotoxicity assessment of immunomodulatory monoclonal antibodies

    Science.gov (United States)

    Morton, Laura Dill; Spindeldreher, Sebastian; Kiessling, Andrea; Allenspach, Roy; Hey, Adam; Muller, Patrick Y; Frings, Werner; Sims, Jennifer

    2010-01-01

    Most therapeutic monoclonal antibodies (mAbs) licensed for human use or in clinical development are indicated for treatment of patients with cancer and inflammatory/autoimmune disease and as such, are designed to directly interact with the immune system. A major hurdle for the development and early clinical investigation of many of these immunomodulatory mAbs is their inherent risk for adverse immune-mediated drug reactions in humans such as infusion reactions, cytokine storms, immunosuppression and autoimmunity. A thorough understanding of the immunopharmacology of a mAb in humans and animals is required to both anticipate the clinical risk of adverse immunotoxicological events and to select a safe starting dose for first-in-human (FIH) clinical studies. This review summarizes the most common adverse immunotoxicological events occurring in humans with immunomodulatory mAbs and outlines non-clinical strategies to define their immunopharmacology and assess their immunotoxic potential, as well as reduce the risk of immunotoxicity through rational mAb design. Tests to assess the relative risk of mAb candidates for cytokine release syndrome, innate immune system (dendritic cell) activation and immunogenicity in humans are also described. The importance of selecting a relevant and sensitive toxicity species for human safety assessment in which the immunopharmacology of the mAb is similar to that expected in humans is highlighted, as is the importance of understanding the limitations of the species selected for human safety assessment and supplementation of in vivo safety assessment with appropriate in vitro human assays. A tiered approach to assess effects on immune status, immune function and risk of infection and cancer, governed by the mechanism of action and structural features of the mAb, is described. Finally, the use of immunopharmacology and immunotoxicity data in determining a minimum anticipated biologic effect Level (MABEL) and in the selection of safe human

  7. The use of probabilistic safety assessments for improving nuclear safety in Europe

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1992-01-01

    The political changes in Europe broadened the scope of international nuclear safety matters considerably. The Western world started to receive reliable and increasingly detailed information on Eastern European nuclear technology and took note of a broad range of technical and administrative problems relevant for nuclear safety in these countries. Reunification made Germany a focus of information exchange on these matters. Here, cooperation with the former German Democratic Republic and with other Eastern European countries as well as safety analyses of Soviet-built nuclear power plants started rather early. Meanwhile, these activities are progressing toward all-European cooperation in the nuclear safety sector. This cooperation includes the use of probabilistic safety assessments (PSAs) addressing applications in both Western and Eastern Europe as well as the further development of this methodology in a converging Europe

  8. Savannah River Site K-Reactor Probabilistic Safety Assessment

    International Nuclear Information System (INIS)

    Brandyberry, M.D.; Bailey, R.T.; Baker, W.H.; Kearnaghan, D.P.; O'Kula, K.R.; Wittman, R.S.; Woody, N.D.; Amos, C.N.; Weingardt, J.J.

    1992-12-01

    This report gives the results of a Savannah River Site (SRS) K-Reactor Probabilistic Safety Assessment (PSA). Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide useful information to the U. S. Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other DOE programs in Heavy Water Reactor safety

  9. Nuclear Safety Culture Assessment for a Newcomer Country: Case Study of Jordan

    International Nuclear Information System (INIS)

    Khasawneh, Khalid; Park, Yun Woon

    2016-01-01

    For countries initiating or considering to start their nuclear power programs; developing a successful safety culture is of a great challenge, owing to lack of experience and the sensitive nature of the nuclear industry in general. The Jordanian case was chosen since Jordan is in the early stages of its nuclear program and the establishment of an effective safety culture is crucial to guarantee the safe operation of its future nuclear facilities. It also should be noted that Fukushima accident has adversely affected the progress of the Jordanian nuclear program driven by the negative public opinion. The government shifts the policies toward enhancing the nuclear safety by enforcing the communication between the engaged parties and openness and transparency with public. In the wake of Fukushima accident the Jordanian government reassured the appropriate siting criteria and siting review, the leadership and the organizations commitment to nuclear safety by adopting advanced reactor technology, the consideration of modern operator accident mitigation strategies and the increased and close cooperation with IAEA and adherence to evolving international safety standards. The progress in the Jordanian nuclear power project in order to satisfy the IAEA requirements was quantified and ranked. A good progress was shown in some aspects, for example in the multicultural and multi-national elements and the establishment of an independent and effective regulatory body. However, some elements, concerning the understanding of the safety culture, management system of the regulatory body and the cultural assessment was not satisfied and an urgent need to focus on and enhance those aspects are required by the Jordanian government. Some elements, for example the leadership, communication and competence, have partial fulfillment of the IAEA requirements. However enhancing those aspects is required in the short and the mid-term in order to guarantee a well-established nuclear power

  10. Nuclear Safety Culture Assessment for a Newcomer Country: Case Study of Jordan

    Energy Technology Data Exchange (ETDEWEB)

    Khasawneh, Khalid; Park, Yun Woon [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    For countries initiating or considering to start their nuclear power programs; developing a successful safety culture is of a great challenge, owing to lack of experience and the sensitive nature of the nuclear industry in general. The Jordanian case was chosen since Jordan is in the early stages of its nuclear program and the establishment of an effective safety culture is crucial to guarantee the safe operation of its future nuclear facilities. It also should be noted that Fukushima accident has adversely affected the progress of the Jordanian nuclear program driven by the negative public opinion. The government shifts the policies toward enhancing the nuclear safety by enforcing the communication between the engaged parties and openness and transparency with public. In the wake of Fukushima accident the Jordanian government reassured the appropriate siting criteria and siting review, the leadership and the organizations commitment to nuclear safety by adopting advanced reactor technology, the consideration of modern operator accident mitigation strategies and the increased and close cooperation with IAEA and adherence to evolving international safety standards. The progress in the Jordanian nuclear power project in order to satisfy the IAEA requirements was quantified and ranked. A good progress was shown in some aspects, for example in the multicultural and multi-national elements and the establishment of an independent and effective regulatory body. However, some elements, concerning the understanding of the safety culture, management system of the regulatory body and the cultural assessment was not satisfied and an urgent need to focus on and enhance those aspects are required by the Jordanian government. Some elements, for example the leadership, communication and competence, have partial fulfillment of the IAEA requirements. However enhancing those aspects is required in the short and the mid-term in order to guarantee a well-established nuclear power

  11. Suggestions on the Development of Safety Culture Assessment Method

    International Nuclear Information System (INIS)

    Choi, Young Sung; Choi, Kwang Sik; Kim, Woong Sik

    2006-01-01

    Several efforts have been made to assess safety culture of organization that operates nuclear power plants in Korea. The MOST and KINS played a major role to develop assessment methods and KHNP applied them to its NPPs. This paper explains the two methods developed by KINS briefly and presents the insights obtained from the two different applications. It concludes with some suggestions for safety culture assessment based on the insights

  12. Safety assessment of primary system components at the USNRC

    Energy Technology Data Exchange (ETDEWEB)

    Serpan, C Z; Chen, C Y; Taboada, A

    1988-12-31

    This document deals with the safety assessment in nuclear reactor components at the USNRC. The USNRC regulations and requirements concerning nuclear reactor design and operations are presented, together with guides and standards which describe how the actions should be implemented. The safety assessment relies on fracture analysis and Non Destructive Examination (NDE). (TEC).

  13. Uncertainty estimation in nuclear power plant probabilistic safety assessment

    International Nuclear Information System (INIS)

    Guarro, S.B.; Cummings, G.E.

    1989-01-01

    Probabilistic Risk Assessment (PRA) was introduced in the nuclear industry and the nuclear regulatory process in 1975 with the publication of the Reactor Safety Study by the U.S. Nuclear Regulatory Commission. Almost fifteen years later, the state-of-the-art in this field has been expanded and sharpened in many areas, and about thirty-five plant-specific PRAs (Probabilistic Risk Assessments) have been performed by the nuclear utility companies or by the U.S. Nuclear Regulatory commission. Among the areas where the most evident progress has been made in PRA and PSA (Probabilistic Safety Assessment, as these studies are more commonly referred to in the international community outside the U.S.) is the development of a consistent framework for the identification of sources of uncertainty and the estimation of their magnitude as it impacts various risk measures. Techniques to propagate uncertainty in reliability data through the risk models and display its effect on the top level risk estimates were developed in the early PRAs. The Seismic Safety Margin Research Program (SSMRP) study was the first major risk study to develop an approach to deal explicitly with uncertainty in risk estimates introduced not only by uncertainty in component reliability data, but by the incomplete state of knowledge of the assessor(s) with regard to basic phenomena that may trigger and drive a severe accident. More recently NUREG-1150, another major study of reactor risk sponsored by the NRC, has expanded risk uncertainty estimation and analysis into the realm of model uncertainty related to the relatively poorly known post-core-melt phenomena which determine the behavior of the molten core and of the rector containment structures

  14. Concept of risk: risk assessment and nuclear safety

    International Nuclear Information System (INIS)

    Thompson, P.B.

    1980-01-01

    The dissertation is a critical examination of risk assessment and its role in public policy. Nuclear power safety safety issues are selected as the primary source of illustrations and examples. The dissertation examines how risk assessment studies develop a concept of risk which becomes decisive for policy choices. Risk-assessment techniques are interpreted as instruments which secure an evaluation of risk which, in turn, figures prominently in technical reports on nuclear power. The philosophical critique is mounted on two levels. First, an epistemological critique surveys distinctions between the technical concept of risk and more familiar senses of risk. The critique shows that utilization of risk assessment re-structures the concept of risk. The technical concept is contrasted to the function of risk within a decision-maker's conceptual agenda and hierarchy of values. Second, an ethical critique exposes the value commitments of risk assessment recommendations. Although some of these values might be defended for policy decisions, the technical character of risk assessment obfuscates normative issues. Risk assessment is shown to be a form of factual enquiry which, nonetheless, represents a commitment to a specific selection of ethical and social values. Risk assessment should not be interpreted as a primary guide to decision unless the specific values incorporated into its concept of risk are stated explicitly and justified philosophically. Such a statement would allow value questions which have been sublimated by the factual tone of the analytic techniques to be debated on clear, social and ethical grounds

  15. Operational safety review programmes for nuclear power plants. Guidelines for assessment

    International Nuclear Information System (INIS)

    2002-01-01

    The IAEA has been offering the Operational Safety Review Team (OSART) programme to provide advice and assistance to Member States in enhancing the operational safety of nuclear power plants (NPPs). Simultaneously, the IAEA has encouraged self-assessment and review by Member States of their own nuclear power plants to continuously improve nuclear safety. Currently, some utilities have been implementing safety review programmes to independently review their own plants. Corporate or national operational safety review programmes may be compliance or performance based. Successful utilities have found that both techniques are necessary to provide assurance that (i) as a minimum the NPP meets specific corporate and legal requirements and (ii) management at the NPP is encouraged to pursue continuous improvement principles. These programmes can bring nuclear safety benefits to the plants and utilities. The IAEA has conducted two pilot missions to assess the effectiveness of the operational review programme. Based on these missions and on the experience gained during OSART missions, this document has been developed to provide guidance on and broaden national/corporate safety review programmes in Member States, and to assist in maximizing their benefits. These guidelines are intended primarily for the IAEA team to conduct assessment of a national/corporate safety review programme. However, this report may also be used by a country or utility to establish its own national/corporate safety review programme. The guidelines may likewise be used for self-assessment or for establishing a baseline when benchmarking other safety review programmes. This report consists of four parts. Section 2 addresses the planning and preparation of an IAEA assessment mission and Sections 3 and 4 deal with specific guidelines for conducting the assessment mission itself

  16. Making the message meaningful: a qualitative assessment of media promoting all-terrain vehicle safety.

    Science.gov (United States)

    Brann, Maria; Mullins, Samantha Hope; Miller, Beverly K; Eoff, Shane; Graham, James; Aitken, Mary E

    2012-08-01

    Millions of all-terrain vehicles (ATV) are used around the world for recreation by both adults and youth. This increase in use has led to a substantial increase in the number of injuries and fatalities each year. Effective strategies for reducing this incidence are clearly needed; however, minimal research exists regarding effective educational interventions. This study was designed to assess rural ATV riders' preferences for and assessment of safety messages. 13 focus group discussions with youth and adult ATV riders were conducted. 88 formative research participants provided feedback on existing ATV safety materials, which was used to develop more useful ATV safety messages. 60 evaluative focus group participants critiqued the materials developed for this project. Existing ATV safety materials have limited effectiveness, in part because they may not address the content or design needs of the target population. ATV riders want educational and action-oriented safety messages that inform youth and adult riders about their responsibilities to learn, educate and implement safety behaviours (eg, appropriate-sized ATV, safety gear, solo riding, speed limits, riding locations). In addition, messages should be clear, realistic, visually appealing and easily accessible. Newly designed ATV safety materials using the acronym TRIPSS (training, ride off-road, impairment, plan ahead, safety gear, single rider) meet ATV riders' safety messaging needs. To reach a target population, it is crucial to include them in the development and assessment of safety messages. Germane to this particular study, ATV riders provided essential information for creating useful ATV safety materials.

  17. Prospects for probabilistic safety assessment

    International Nuclear Information System (INIS)

    Hirschberg, S.

    1992-01-01

    This article provides some reflections on future developments of Probabilistic Safety Assessment (PSA) in view of the present state of the art and evaluates current trends in the use of PSA for safety management. The main emphasis is on Level 1 PSA, although Level 2 aspects are also highlighted to some extent. As a starting point, the role of PSA is outlined from a historical perspective, demonstrating the rapid expansion of the uses of PSA. In this context the wide spectrum of PSA applications and the associated benefits to the users are in focus. It should be kept in mind, however, that PSA, in spite of its merits, is not a self-standing safety tool. It complements deterministic analysis and thus improves understanding and facilitating prioritization of safety issues. Significant progress in handling PSA limitations - such as reliability data, common-cause failures, human interactions, external events, accident progression, containment performance, and source-term issues - is described. This forms a background for expected future developments of PSA. Among the most important issues on the agenda for the future are PSA scope extensions, methodological improvements and computer code advancements, and full exploitation of the potential benefits of applications to operational safety management. Many PSA uses, if properly exercised, lead to safety improvements as well as major burden reductions. The article provides, in addition, International Atomic Energy Agency (IAEA) perspective on the topics covered, as reflected in the current PSA programs of the agency. 74 refs., 6 figs., 1 tab

  18. Training courses on integrated safety assessment modelling for waste repositories

    International Nuclear Information System (INIS)

    Mallants, D.

    2007-01-01

    Near-surface or deep repositories of radioactive waste are being developed and evaluated all over the world. Also, existing repositories for low- and intermediate-level waste often need to be re-evaluated to extend their license or to obtain permission for final closure. The evaluation encompasses both a technical feasibility as well as a safety analysis. The long term safety is usually demonstrated by means of performance or safety assessment. For this purpose computer models are used that calculate the migration of radionuclides from the conditioned radioactive waste, through engineered barriers to the environment (groundwater, surface water, and biosphere). Integrated safety assessment modelling addresses all relevant radionuclide pathways from source to receptor (man), using in combination various computer codes in which the most relevant physical, chemical, mechanical, or even microbiological processes are mathematically described. SCK-CEN organizes training courses in Integrated safety assessment modelling that are intended for individuals who have either a controlling or supervising role within the national radwaste agencies or regulating authorities, or for technical experts that carry out the actual post-closure safety assessment for an existing or new repository. Courses are organised by the Department of Waste and Disposal

  19. An approach for assessing ALWR passive safety system reliability

    International Nuclear Information System (INIS)

    Hake, T.M.

    1991-01-01

    Many of the advanced light water reactor (ALWR) concepts proposed for the next generation of nuclear power plants rely on passive rather than active systems to perform safety functions. Despite the reduced redundancy of the passive systems as compared to active systems in current plants, the assertion is that the overall safety of the plant is enhanced due to the much higher expected reliability of the passive systems. In order to investigate this assertion, a study is being conducted at Sandia National Laboratories to evaluate the reliability of ALWR passive safety features in the context of probabilistic risk assessment (PRA). The purpose of this paper is to provide a brief overview of the approach to this study. The quantification of passive system reliability is not as straightforward as for active systems, due to the lack of operating experience, and to the greater uncertainty in the governing physical phenomena. Thus, the adequacy of current methods for evaluating system reliability must be assessed, and alternatives proposed if necessary. For this study, the Westinghouse Advanced Passive 600 MWe reactor (AP600) was chosen as the advanced reactor for analysis, because of the availability of AP600 design information. This study compares the reliability of AP600 emergency cooling system with that of corresponding systems in a current generation reactor

  20. Human Reliability in Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Nunez Mendez, J.

    1989-01-01

    Nowadays a growing interest in environmental aspects is detected in our country. It implies an assessment of the risk involved in the industrial processes and installations in order to determine if those are into the acceptable limits. In these safety assessments, among which PSA (Probabilistic Safety Assessments), can be pointed out the role played by the human being in the system is one of the more relevant subjects (This relevance has been demonstrated in the accidents happened) . However, in Spain there aren't manuals specifically dedicated to asses the human contribution to risk in the frame of PSAs. This report aims to improve this situation providing: a) a theoretical background to help the reader in the understanding of the nature of the human error, b) a quid to carry out a Human Reliability Analysis and c) a selected overview of the techniques and methodologies currently applied in this area. (Author) 20 refs

  1. Interim process report for the safety assessment SR-Can

    International Nuclear Information System (INIS)

    Sellin, Patrick

    2004-08-01

    This report is a documentation of buffer processes identified as relevant to the long-term safety of a KBS-3 repository. The report is part of the interim reporting of the safety assessment SR-Can, see further the Interim main report. The final SR-Can reporting will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of this report is to document the scientific knowledge of the processes to a level required for an adequate treatment in the safety assessment. The documentation is thus from a scientific point of not exhaustive since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of an assessment. The purpose is further to determine the handling of each process in the safety assessment and to demonstrate how uncertainties are taken care of, given the suggested handling. The process documentation in the SR 97 version of the Process report is a starting point for this SR-Can interim version. As further described in the Interim main report, the list of relevant processes has been reviewed and slightly extended by comparison to other databases. Furthermore, the backfill has been included as a system part of its own, rather than being described together with the buffer as in SR 97. Apart from giving an interim account of the documentation and handling of buffer processes in SR-Can, this report is meant to serve as a template for the forthcoming documentation of processes occurring in other parts of the repository system. A complete list of processes can be found in the Interim FEP report for the safety assessment SR-Can. All material presented in this document is preliminary in nature and will possibly be updated as the SR-Can project progresses

  2. Interim process report for the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Sellin, Patrick (ed.)

    2004-08-01

    This report is a documentation of buffer processes identified as relevant to the long-term safety of a KBS-3 repository. The report is part of the interim reporting of the safety assessment SR-Can, see further the Interim main report. The final SR-Can reporting will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of this report is to document the scientific knowledge of the processes to a level required for an adequate treatment in the safety assessment. The documentation is thus from a scientific point of not exhaustive since such a treatment is neither necessary for the purposes of the safety assessment nor possible within the scope of an assessment. The purpose is further to determine the handling of each process in the safety assessment and to demonstrate how uncertainties are taken care of, given the suggested handling. The process documentation in the SR 97 version of the Process report is a starting point for this SR-Can interim version. As further described in the Interim main report, the list of relevant processes has been reviewed and slightly extended by comparison to other databases. Furthermore, the backfill has been included as a system part of its own, rather than being described together with the buffer as in SR 97. Apart from giving an interim account of the documentation and handling of buffer processes in SR-Can, this report is meant to serve as a template for the forthcoming documentation of processes occurring in other parts of the repository system. A complete list of processes can be found in the Interim FEP report for the safety assessment SR-Can. All material presented in this document is preliminary in nature and will possibly be updated as the SR-Can project progresses.

  3. Probabilistic safety assessment goals in Canada

    International Nuclear Information System (INIS)

    Snell, V.G.

    1986-01-01

    CANDU safety philosphy, both in design and in licensing, has always had a strong bias towards quantitative probabilistically-based goals derived from comparative safety. Formal probabilistic safety assessment began in Canada as a design tool. The influence of this carried over later on into the definition of the deterministic safety guidelines used in CANDU licensing. Design goals were further developed which extended the consequence/frequency spectrum of 'acceptable' events, from the two points defined by the deterministic single/dual failure analysis, to a line passing through lower and higher frequencies. Since these were design tools, a complete risk summation was not necessary, allowing a cutoff at low event frequencies while preserving the identification of the most significant safety-related events. These goals gave a logical framework for making decisions on implementing design changes proposed as a result of the Probabilistic Safety Analysis. Performing this analysis became a regulatory requirement, and the design goals remained the framework under which this was submitted. Recently, there have been initiatives to incorporate more detailed probabilistic safety goals into the regulatory process in Canada. These range from far-reaching safety optimization across society, to initiatives aimed at the nuclear industry only. The effectiveness of the latter is minor at very low and very high event frequencies; at medium frequencies, a justification against expenditures per life saved in other industries should be part of the goal setting

  4. Electronuclear's safety culture assessment and enhancement program

    International Nuclear Information System (INIS)

    Selvatici, E.; Diaz-Francisco, J.M.; Diniz de Souza, V.

    2002-01-01

    The present paper describes the Eletronuclear's safety culture assessment and enhancement program. The program was launched by the company's top management one year after the creation of Eletronuclear in 1997, from the merging of two companies with different organizational cultures, the design and engineering company Nuclen and the nuclear directorate of the Utility Furnas, Operator of the Angra1 NPP. The program consisted of an assessment performed internally in 1999 with the support and advice of the IAEA. This assessment, performed with the help of a survey, pooled about 80% of the company's employees. The overall result of the assessment was that a satisfactory level of safety culture existed; however, a number of points with a considerable margin for improvement were also identified. These points were mostly related with behavioural matters such as motivation, stress in the workplace, view of mistakes, handling of conflicts, and last but not least a view by a considerable number of employees that a conflict between safety and production might exist. An Action Plan was established by the company managers to tackle these weak points. This Plan was issued as company guideline by the company's Directorate. The subsequent step was to detail and implement the different actions of the Plan, which is the phase that we are at present. In the detailing of the Action Plan, special care was taken to sum up efforts, avoiding duplication of work or competition with already existing programs. In this process it was identified that the company had a considerable number of initiatives directly related to organizational and safety culture improvement, already operational. These initiatives have been integrated in the detailed Action Plan. A new assessment, for checking the effectiveness of the undertaken actions, is planned for 2003. (author)

  5. Safety testing of monoclonal antibodies in non-human primates: Case studies highlighting their impact on human risk assessment.

    Science.gov (United States)

    Brennan, Frank R; Cavagnaro, Joy; McKeever, Kathleen; Ryan, Patricia C; Schutten, Melissa M; Vahle, John; Weinbauer, Gerhard F; Marrer-Berger, Estelle; Black, Lauren E

    2018-01-01

    Monoclonal antibodies (mAbs) are improving the quality of life for patients suffering from serious diseases due to their high specificity for their target and low potential for off-target toxicity. The toxicity of mAbs is primarily driven by their pharmacological activity, and therefore safety testing of these drugs prior to clinical testing is performed in species in which the mAb binds and engages the target to a similar extent to that anticipated in humans. For highly human-specific mAbs, this testing often requires the use of non-human primates (NHPs) as relevant species. It has been argued that the value of these NHP studies is limited because most of the adverse events can be predicted from the knowledge of the target, data from transgenic rodents or target-deficient humans, and other sources. However, many of the mAbs currently in development target novel pathways and may comprise novel scaffolds with multi-functional domains; hence, the pharmacological effects and potential safety risks are less predictable. Here, we present a total of 18 case studies, including some of these novel mAbs, with the aim of interrogating the value of NHP safety studies in human risk assessment. These studies have identified mAb candidate molecules and pharmacological pathways with severe safety risks, leading to candidate or target program termination, as well as highlighting that some pathways with theoretical safety concerns are amenable to safe modulation by mAbs. NHP studies have also informed the rational design of safer drug candidates suitable for human testing and informed human clinical trial design (route, dose and regimen, patient inclusion and exclusion criteria and safety monitoring), further protecting the safety of clinical trial participants.

  6. A study on quantitative V and V of safety-critical software

    International Nuclear Information System (INIS)

    Eom, H. S.; Kang, H. G.; Chang, S. C.; Ha, J. J.; Son, H. S.

    2004-03-01

    Recently practical needs have required quantitative features for the software reliability for Probabilistic Safety Assessment which is one of the important methods being used in assessing the overall safety of nuclear power plant. But the conventional assessment methods of software reliability could not provide enough information for PSA of NPP, therefore current assessments of a digital system which includes safety-critical software usually exclude the software part or use arbitrary values. This paper describes a Bayesian Belief Networks based method that models the rule-based qualitative software assessment method for a practical use and can produce quantitative results for PSA. The framework was constructed by utilizing BBN that can combine the qualitative and quantitative evidence relevant to the reliability of safety-critical software and can infer a conclusion in a formal and a quantitative way. The case study was performed by applying the method for assessing the quality of software requirement specification of safety-critical software that will be embedded in reactor protection system

  7. Selection of important initiating events for Level 1 probabilistic safety assessment study at Puspati TRIGA Reactor

    International Nuclear Information System (INIS)

    Maskin, M.; Charlie, F.; Hassan, A.; Prak Tom, P.; Ramli, Z.; Mohamed, F.

    2016-01-01

    Highlights: • Identifying possible important initiating events (IEs) for Level 1 probabilistic safety assessment performed on research nuclear reactor. • Methods in screening and grouping IEs are addressed. • Focusing only on internal IEs due to random failures of components. - Abstract: This paper attempts to present the results in identifying possible important initiating events (IEs) as comprehensive as possible to be applied in the development of Level-1 probabilistic safety assessment (PSA) study. This involves the approaches in listing and the methods in screening and grouping IEs, by focusing only on the internal IEs due to random failures of components and human errors with full power operational conditions and reactor core as the radioactivity source. Five approaches were applied in listing the IEs and each step of the methodology was described and commented. The criteria in screening and grouping the IEs were also presented. The results provided the information on how the Malaysian PSA team applied the approaches in selecting the most probable IEs as complete as possible in order to ensure the set of IEs was identified systematically and as representative as possible, hence providing confidence to the completeness of the PSA study. This study is perhaps one of the first to address classic comprehensive steps in identifying important IEs to be used in a Level-1 PSA study.

  8. Towards understanding work-as-done in air traffic management safety assessment and design

    International Nuclear Information System (INIS)

    Woltjer, Rogier; Pinska-Chauvin, Ella; Laursen, Tom; Josefsson, Billy

    2015-01-01

    This paper describes the approach taken and the results to develop guidance, to include Resilience Engineering principles in methodology for safety assessment of functional changes, in Air Traffic Management (ATM). It summarizes the process of deriving resilience principles for ATM, originating from Resilience Engineering concepts and transposed into ATM operations. These principles are the foundation for guidance material incorporating Resilience Engineering (RE) concepts into safety assessment methodology. The guidance material provides a method using workshops generating qualitative descriptions of RE principles applied to ATM services of everyday work, as done currently and as envisioned after introduction of a new technology or way of working. The guidance material has been proposed as part of the safety assessment methodology of SESAR (Single European Sky ATM Research), and as stand-alone guidance for ATM design processes. The methodology was validated via a test case on the i4D/CTA (Controlled Time of Arrival) concept. Operational examples from the application of the developed guidance to the i4D/CTA concept are provided. Initial evaluation of the guidance suggests that the methodology (1) provides a narrative, vocabulary and documentation means of project discussions on resilience; (2) brings the discussions of safety and resilience closer to operational practice; (3) facilitates a broader systemic and integrative perspective on operational, management, business, safety, environmental, and human performance aspects; and (4) can extend the vocabulary of safety assessment to include the description of emergent properties, to better support functional changes in ATM. - Highlights: • Guidance material for safety assessment based on systemic thinking is proposed. • It operationalizes Resilience Engineering principles in Air Traffic Management, including a case study. • It enables description of expected changes in work-as-done when introducing a new

  9. Development of a quality assurance safety assessment database for near surface radioactive waste disposal

    International Nuclear Information System (INIS)

    Park, J. W.; Kim, C. L.; Park, J. B.; Lee, E. Y.; Lee, Y. M.; Kang, C. H.; Zhou, W.; Kozak, M. W.

    2003-01-01

    A quality assurance safety assessment database, called QUARK (QUality Assurance program for Radioactive waste management in Korea), has been developed to manage both analysis information and parameter database for safety assessment of Low- and Intermediate-Level radioactive Waste (LILW) disposal facility in Korea. QUARK is such a tool that serves QA purposes for managing safety assessment information properly and securely. In QUARK, the information is organized and linked to maximize the integrity of information and traceability. QUARK provides guidance to conduct safety assessment analysis, from scenario generation to result analysis, and provides a window to inspect and trace previous safety assessment analysis and parameter values. QUARK also provides default database for safety assessment staff who construct input data files using SAGE(Safety Assessment Groundwater Evaluation), a safety assessment computer code

  10. Safety assessments for deep geological disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Lyon, R.B.

    1984-01-01

    The objective of safety assessment for deep geological disposal of radioactive wastes is to evaluate how well the engineered barriers and geological setting inhibit radionuclide migration and prevent radiation dose to man. Safety assessment is influenced through interaction with the regulatory agencies, research groups, the public and the various levels of government. Under the auspices of the IAEA, a generic disposal system description has been developed to facilitate international exchange and comparison of data and results, and to enable development and comparison of performance for all components of the disposal system. It is generally accepted that a systems modelling approach is required and that safety assessment can be considered on two levels. At the systems level, all components of the system are taken into account to evaluate the risk to man. At the systems level, critical review and quality assurance on software provide the major validation techniques. Risk is a combination of dose estimate and probability of that dose. For analysis of the total system to be practical, the components are usually represented by simplified models. Recently, assessments have been taking uncertainties in the input data into account. At the detailed level, large-scale, complex computer programs model components of the system in sufficient detail that validation by comparison with field and laboratory measurements is possible. For example, three-dimensional fluid-flow, heat-transport and solute-transport computer programs have been used. Approaches to safety assessment are described, with illustrations from safety assessments performed in a number of countries. (author)

  11. New safety performance indicators for safety assessment of radioactive waste disposal facilities. Cuban experience

    International Nuclear Information System (INIS)

    Peralta Vital, J.L.; Castillo, R.G.; Olivera, J.

    2002-01-01

    The paper shows the Cuban experience on implementing geological disposal of radioactive waste and the necessity for identifying new safety performance indicators for the safety assessment (SA) of radioactive waste disposal facilities. The selected indicator was the concentration of natural radioactive elements (U, Ra, Th, K) in the Cuban geologic environment. We have carried out a group of investigations, which have allowed characterising the concentration for the whole Country, creating a wide database where this indicator is associated with the lithology. The main lithologies in Cuba are: the sedimentary rocks (70 percent of national occurrence), which are present in the three regions (limestone and lutite), and finally the igneous and metamorphic rocks. The results show the concentrations ranges of the natural radionuclides associated fundamentally to the variation in the lithology and geographical area of the Country. In Cuba, the higher concentration (ppm) of Uranium and Radium are referenced to the Central region associated to Skarn, while for Thorium (ppm) and Potassium (%), in the East region the concentration peaks in Tuffs have been found. The concentrations ranges obtained are preliminary, they characterise the behaviour of this parameter for the Cuban geology, but they do not represent limits for safety assessment purposes yet. Also other factors should be taken into account as the assessment context, time scales and others assumptions before establishing the final concentration limits for the natural radionuclides as a radiological and nuclear safety performance indicator complementary to dose and risk for safety assessment for radiological and nuclear facilities. (author)

  12. EFFICIENT QUANTITATIVE RISK ASSESSMENT OF JUMP PROCESSES: IMPLICATIONS FOR FOOD SAFETY

    OpenAIRE

    Nganje, William E.

    1999-01-01

    This paper develops a dynamic framework for efficient quantitative risk assessment from the simplest general risk, combining three parameters (contamination, exposure, and dose response) in a Kataoka safety-first model and a Poisson probability representing the uncertainty effect or jump processes associated with food safety. Analysis indicates that incorporating jump processes in food safety risk assessment provides more efficient cost/risk tradeoffs. Nevertheless, increased margin of safety...

  13. Immunogenicity of biologically-derived therapeutics: assessment and interpretation of nonclinical safety studies.

    Science.gov (United States)

    Ponce, Rafael; Abad, Leslie; Amaravadi, Lakshmi; Gelzleichter, Thomas; Gore, Elizabeth; Green, James; Gupta, Shalini; Herzyk, Danuta; Hurst, Christopher; Ivens, Inge A; Kawabata, Thomas; Maier, Curtis; Mounho, Barbara; Rup, Bonita; Shankar, Gopi; Smith, Holly; Thomas, Peter; Wierda, Dan

    2009-07-01

    An evaluation of potential antibody formation to biologic therapeutics during the course of nonclinical safety studies and its impact on the toxicity profile is expected under current regulatory guidance and is accepted standard practice. However, approaches for incorporating this information in the interpretation of nonclinical safety studies are not clearly established. Described here are the immunological basis of anti-drug antibody formation to biopharmaceuticals (immunogenicity) in laboratory animals, and approaches for generating and interpreting immunogenicity data from nonclinical safety studies of biotechnology-derived therapeutics to support their progression to clinical evaluation. We subscribe that immunogenicity testing strategies should be adapted to the specific needs of each therapeutic development program, and data generated from such analyses should be integrated with available clinical and anatomic pathology, pharmacokinetic, and pharmacodynamic data to properly interpret nonclinical studies.

  14. Review of light--water reactor safety studies. Volume 3 of health and safety impacts of nuclear, geothermal, and fossil-fuel electric generation in California

    International Nuclear Information System (INIS)

    Nero, A.V.; Farnaam, M.R.K.

    1977-01-01

    This report summarizes and compares important studies of light-water nuclear reactor safety, emphasizing the Nuclear Regulatory Commission's Reactor Safety Study, work on risk assessment funded by the Electric Power Research Institute, and the Report of the American Physical Society study group on light-water reactor safety. These reports treat risk assessment for nuclear power plants and provide an introduction to the basic issues in reactor safety and the needs of the reactor safety research program. Earlier studies are treated more briefly. The report includes comments on the Reactor Safety Study. The manner in which these studies may be used and alterations which would increase their utility are discussed

  15. Value of preapproval safety data in predicting postapproval hepatic safety and assessing the legitimacy of class warning.

    Science.gov (United States)

    Lin, Yeong-Liang; Wu, Ya-Chi; Gau, Churn-Shiouh; Lin, Min-Shung

    2012-02-01

    The objective of this study was to systematically evaluate whether preapproval safety data for nonhepatotoxic drugs and hepatotoxic drugs can be compared to improve preapproval prediction of postapproval hepatic safety and to assess the legitimacy of applying class warnings. Drugs within a therapeutic class that included at least one drug that had been withdrawn from the market because of liver toxicity or had a warning of potential liver toxicity issued by major regulatory agencies, and at least one drug free from such regulatory action, were identified and divided into two groups: drugs with and drugs without regulatory action. Preapproval clinical data [including the elevation rates of alanine aminotransferse (ALT) and withdrawal due to liver toxicity, the number of patients with combined elevation of ALT and bilirubin, and liver failure] and nonclinical data (including chemical structures, metabolic pathways, and other significant findings in animal studies) were compared between the two groups. Six drug classes were assessed in this study: thiazolidinediones, cyclooxygenase-2 inhibitors, fluoroquinolones, catechol-O-methyltransferase (COMT) inhibitors, leukotriene receptor inhibitors, and endothelin receptor antagonists. In two classes (COMT inhibitors and endothelin receptor antagonists), drugs with regulatory action had significantly higher rates of ALT elevation of more than threefold and greater numbers of patients with combined elevation of ALT and bilirubin than drugs without regulatory action. Drugs with regulatory action also had chemical structures or metabolic pathways associated with the toxicity. The legitimacy of class warnings was refuted in all six classes of drugs. Preapproval safety data may help predict postapproval hepatic safety and can be used to assess the legitimacy of applying class warnings.

  16. 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

  17. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Executive summary: main report

    International Nuclear Information System (INIS)

    1975-10-01

    Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks

  18. Use of the Home Safety Self-Assessment Tool (HSSAT) within Community Health Education to Improve Home Safety.

    Science.gov (United States)

    Horowitz, Beverly P; Almonte, Tiffany; Vasil, Andrea

    2016-10-01

    This exploratory research examined the benefits of a health education program utilizing the Home Safety Self-Assessment Tool (HSSAT) to increase perceived knowledge of home safety, recognition of unsafe activities, ability to safely perform activities, and develop home safety plans of 47 older adults. Focus groups in two senior centers explored social workers' perspectives on use of the HSSAT in community practice. Results for the health education program found significant differences between reported knowledge of home safety (p = .02), ability to recognize unsafe activities (p = .01), safely perform activities (p = .04), and develop a safety plan (p = .002). Social workers identified home safety as a major concern and the HSSAT a promising assessment tool. Research has implications for reducing environmental fall risks.

  19. Safety culture assessment in petrochemical industry: a comparative study of two algerian plants.

    Science.gov (United States)

    Boughaba, Assia; Hassane, Chabane; Roukia, Ouddai

    2014-06-01

    To elucidate the relationship between safety culture maturity and safety performance of a particular company. To identify the factors that contribute to a safety culture, a survey questionnaire was created based mainly on the studies of Fernández-Muñiz et al. The survey was randomly distributed to 1000 employees of two oil companies and realized a rate of valid answer of 51%. Minitab 16 software was used and diverse tests, including the descriptive statistical analysis, factor analysis, reliability analysis, mean analysis, and correlation, were used for the analysis of data. Ten factors were extracted using the analysis of factor to represent safety culture and safety performance. The results of this study showed that the managers' commitment, training, incentives, communication, and employee involvement are the priority domains on which it is necessary to stress the effort of improvement, where they had all the descriptive average values lower than 3.0 at the level of Company B. Furthermore, the results also showed that the safety culture influences the safety performance of the company. Therefore, Company A with a good safety culture (the descriptive average values more than 4.0), is more successful than Company B in terms of accident rates. The comparison between the two petrochemical plants of the group Sonatrach confirms these results in which Company A, the managers of which are English and Norwegian, distinguishes itself by the maturity of their safety culture has significantly higher evaluations than the company B, who is constituted of Algerian staff, in terms of safety management practices and safety performance.

  20. 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.

  1. 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

  2. Integrated safety assessment report: Integrated Safety Assessment Program: Millstone Nuclear Power Station, Unit 1 (Docket No. 50-245): Draft report

    International Nuclear Information System (INIS)

    1987-04-01

    The Integrated Safety Assessment Program (ISAP) was initiated in November 1984, by the US Nuclear Regulatory Commission to conduct integrated assessments for operating nuclear power reactors. The integrated assessment is conducted in a plant-specific basis to evaluate all licensing actions, licensee initiated plant improvements and selected unresolved generic/safety issues to establish implementation schedules for each item. In addition, procedures will be established to allow for a periodic updating of the schedules to account for licensing issues that arise in the future. This report documents the review of Millstone Nuclear Power Station, Unit No. 1, operated by Northeast Nuclear Energy Company (located in Waterford, Connecticut). Millstone Nuclear Power Station, Unit No. 1, is one of two plants being reviewed under the pilot program for ISAP. This report indicates how 85 topics selected for review were addressed. This report presents the staff's recommendations regarding the corrective actions to resolve the 85 topics and other actions to enhance plant safety. The report is being issued in draft form to obtain comments from the licensee, nuclear safety experts, and the Advisory Committee for Reactor Safeguards (ACRS). Once those comments have been resolved, the staff will present its positions, along with a long-term implementation schedule from the licensee, in the final version of this report

  3. Assessing progress in the development of safety culture

    International Nuclear Information System (INIS)

    Rotaru, Ioan; Ghita, Sorin

    1999-01-01

    The concept of safety culture was introduced by the International Nuclear Safety Advisory Group (INSAG) in the Summary Report on the Post-Accident Meeting on the Chernobyl Accident in 1986. The concept was further expanded in the 1988 INSAG-3 report, Basic Safety Principles for Nuclear Power Plants, and again in 1991 in the INSAG-4 report. Recognizing the increasing role that safety culture is expected to play in nuclear installations worldwide, the Convention on Nuclear Safety states the Contracting Parties' desire 'to promote an effective nuclear safety culture'. The concept of safety culture is defined in INSAG-4 as follows: Safety culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance. Safety culture is also an amalgamation of values, standards, morals and norms of acceptable behaviour. These are aimed at maintaining a self disciplined approach to the enhancement of safety beyond legislative and regulatory requirements. Therefore, the safety culture has to be inherent in the thoughts and actions of all the individuals at every level in an organization. The leadership provided by top management is crucial. Safety culture applies to conventional and personal safety as well as nuclear safety. All safety consideration are affected by common points of beliefs, attitudes, behaviour, and cultural differences, closely linked to a shared system of values and standards. The paper poses questions and tries to find answers relative to issues like: - how to assess progress; - specific organizational indicators of a progressive safety culture; - detection of incipient weaknesses in safety culture (organizational issues, employee issues, technology issues); - revitalizing a weakened safety culture; - overall assesment of safety culture; - general evaluation model. In conclusion, there is no consistent and

  4. Safety analysis, risk assessment, and risk acceptance criteria

    International Nuclear Information System (INIS)

    Jamali, K.

    1997-01-01

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

  5. A Microbial Assessment Scheme to measure microbial performance of Food Safety Management Systems.

    Science.gov (United States)

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

    2009-08-31

    A Food Safety Management System (FSMS) implemented in a food processing industry is based on Good Hygienic Practices (GHP), Hazard Analysis Critical Control Point (HACCP) principles and should address both food safety control and assurance activities in order to guarantee food safety. One of the most emerging challenges is to assess the performance of a present FSMS. The objective of this work is to explain the development of a Microbial Assessment Scheme (MAS) as a tool for a systematic analysis of microbial counts in order to assess the current microbial performance of an implemented FSMS. It is assumed that low numbers of microorganisms and small variations in microbial counts indicate an effective FSMS. The MAS is a procedure that defines the identification of critical sampling locations, the selection of microbiological parameters, the assessment of sampling frequency, the selection of sampling method and method of analysis, and finally data processing and interpretation. Based on the MAS assessment, microbial safety level profiles can be derived, indicating which microorganisms and to what extent they contribute to food safety for a specific food processing company. The MAS concept is illustrated with a case study in the pork processing industry, where ready-to-eat meat products are produced (cured, cooked ham and cured, dried bacon).

  6. Development of safety function assessment trees for pressurized heavy water reactor LP/SD operations

    International Nuclear Information System (INIS)

    Yang, Hui Chang; Chung, Chang Hyun; Kim, Ki Yong; Jee, Moon Hak; Sung, Chang Kyoung

    2003-01-01

    The objective of Configuration Risk Management Program(CRMP) is to maintain the safety level by assuring the defense-in-depth of nuclear power plant while the configurations are changed during plant operations, especially for the LP/SD. Such a safety purpose can be achieved by establishing the risk monitoring programs with both quantitative and qualitative features. Generally, the quantitative risk evaluation models, i.e., PRA models are used for the risk evaluation during full power operation, and the qualitative risk evaluation models such as safety function assessment trees are used. Through this study, safety function assessment trees were developed

  7. Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities

    International Nuclear Information System (INIS)

    Batandjieva, B.; Torres-Vidal, C.

    2002-01-01

    The International Atomic Energy Agency (IAEA) Coordinated research program ''Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities'' (ISAM) has developed improved safety assessment methodology for near surface disposal facilities. The program has been underway for three years and has included around 75 active participants from 40 countries. It has also provided examples for application to three safety cases--vault, Radon type and borehole radioactive waste disposal facilities. The program has served as an excellent forum for exchange of information and good practices on safety assessment approaches and methodologies used worldwide. It also provided an opportunity for reaching broad consensus on the safety assessment methodologies to be applied to near surface low and intermediate level waste repositories. The methodology has found widespread acceptance and the need for its application on real waste disposal facilities has been clearly identified. The ISAM was finalized by the end of 2000, working material documents are available and an IAEA report will be published in 2002 summarizing the work performed during the three years of the program. The outcome of the ISAM program provides a sound basis for moving forward to a new IAEA program, which will focus on practical application of the safety assessment methodologies to different purposes, such as licensing radioactive waste repositories, development of design concepts, upgrading existing facilities, reassessment of operating repositories, etc. The new program will also provide an opportunity for development of guidance on application of the methodology that will be of assistance to both safety assessors and regulators

  8. [Agricultural biotechnology safety assessment].

    Science.gov (United States)

    McClain, Scott; Jones, Wendelyn; He, Xiaoyun; Ladics, Gregory; Bartholomaeus, Andrew; Raybould, Alan; Lutter, Petra; Xu, Haibin; Wang, Xue

    2015-01-01

    Genetically modified (GM) crops were first introduced to farmers in 1995 with the intent to provide better crop yield and meet the increasing demand for food and feed. GM crops have evolved to include a thorough safety evaluation for their use in human food and animal feed. Safety considerations begin at the level of DNA whereby the inserted GM DNA is evaluated for its content, position and stability once placed into the crop genome. The safety of the proteins coded by the inserted DNA and potential effects on the crop are considered, and the purpose is to ensure that the transgenic novel proteins are safe from a toxicity, allergy, and environmental perspective. In addition, the grain that provides the processed food or animal feed is also tested to evaluate its nutritional content and identify unintended effects to the plant composition when warranted. To provide a platform for the safety assessment, the GM crop is compared to non-GM comparators in what is typically referred to as composition equivalence testing. New technologies, such as mass spectrometry and well-designed antibody-based methods, allow better analytical measurements of crop composition, including endogenous allergens. Many of the analytical methods and their intended uses are based on regulatory guidance documents, some of which are outlined in globally recognized documents such as Codex Alimentarius. In certain cases, animal models are recommended by some regulatory agencies in specific countries, but there is typically no hypothesis or justification of their use in testing the safety of GM crops. The quality and standardization of testing methods can be supported, in some cases, by employing good laboratory practices (GLP) and is recognized in China as important to ensure quality data. Although the number of recommended, in some cases, required methods for safety testing are increasing in some regulatory agencies, it should be noted that GM crops registered to date have been shown to be

  9. 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

  10. Assessment of reliability and validity of a new safety culture questionnaire

    Directory of Open Access Journals (Sweden)

    A.A. Farshad

    2010-04-01

    Full Text Available Background and aims   As a Development of Industrial process, human, environment, equipment, material and validity of system has been exposed to hazardous conditions. Regards of 32.3 percent of occupations in industries, this study focused on risk assessment of foundry unit by energy trace and barrier analysis (ETBA method and presented approaches to control of accident.     Methods   the recent study is as a case study one to risk assessment in a foundry unit in Qazvin industrial city in1387. In this study risks were founded by ETBA method and evaluated by MILSTD- 882B. Data were collected by direct observations, interview with workers and supervisor and engineers, walking-talking through method, documents investigation of operational processors, preventive maintenances, equipment technical properties, accidental and medical documents. Finally ETBA worksheets completed.     Results   totally 154 risks has been found. 40 from total are been unacceptable risk, 68 unfavorable and also 46 acceptable but with remediation action. Casting workshop had risks more than other workshops (with 74 identified risks.Potential and heat energies were founded as most   hazardous energies, with respectively 51 and 38 risk cases.     Conclusion   This study recommended to be done actions for identification and control risk, such as: safety training, occupation training, preventive maintenance, contract safety, safety  communication and safety audit group.  

  11. Assessment of the factors with significant influence on safety culture

    International Nuclear Information System (INIS)

    Farcasiu, M.; Nitoi, M.

    2013-01-01

    In this paper, a qualitative and a quantitative evaluation of the factors with significant impact on safety culture were performed. These techniques were established and applied in accordance with IAEA standards. In order to show the applicability and opportunity of the methodology a specific case study was prepared: safety culture evaluation for INR Pitesti. The qualitative evaluation was performed using specific developed questionnaires. Through analysis of the completed questionnaires was established the development stage of safety culture at INR. The quantitative evaluation was performed using a guide to rate the influence factors. For each factor was identified the influence (negative or positive) and ranking score was estimated using scoring criteria. The results have emphasized safety culture stages. The paper demonstrates the fact that using both quantitative and qualitative assessment techniques, a practical value of the safety culture concept is given. (authors)

  12. Application of the Integrated Safety Assessment methodology to safety margins. Dynamic Event Trees, Damage Domains and Risk Assessment

    International Nuclear Information System (INIS)

    Ibánez, L.; Hortal, J.; Queral, C.; Gómez-Magán, J.; Sánchez-Perea, M.; Fernández, I.; Meléndez, E.; Expósito, A.; Izquierdo, J.M.; Gil, J.; Marrao, H.; Villalba-Jabonero, E.

    2016-01-01

    The Integrated Safety Assessment (ISA) methodology, developed by the Consejo de Seguridad Nuclear, has been applied to an analysis of Zion NPP for sequences with Loss of the Component Cooling Water System (CCWS). The ISA methodology proposal starts from the unfolding of the Dynamic Event Tree (DET). Results from this first step allow assessing the sequence delineation of standard Probabilistic Safety Analysis results. For some sequences of interest of the outlined DET, ISA then identifies the Damage Domain (DD). This is the region of uncertain times and/or parameters where a safety limit is exceeded, which indicates the occurrence of certain damage situation. This paper illustrates application of this concept obtained simulating sequences with MAAP and with TRACE. From information of simulation results of sequence transients belonging to the DD and the time-density probability distributions of the manual actions and of occurrence of stochastic phenomena, ISA integrates the dynamic reliability equations proposed to obtain the sequence contribution to the global Damage Exceedance Frequency (DEF). Reported results show a slight increase in the DEF for sequences investigated following a power uprate from 100% to 110%. This demonstrates the potential use of the method to help in the assessment of design modifications. - Highlights: • This paper illustrates an application of the ISA methodology to safety margins. • Dynamic Event Trees are useful tool for verifying the standard PSA Event Trees. • The ISA methodology takes into account the uncertainties in human action times. • The ISA methodology shows the Damage Exceedance Frequency increase in power uprates.

  13. Guidance on the safety assessment methodology for storage of radioactive waste

    International Nuclear Information System (INIS)

    Kinyanjui, M.N.

    2014-04-01

    This project on safety assessment on storage was carried out with the main objective of ensuring safety of human life and our environment. This is the fundamental principle of radiation protection. Safety assessment has been evaluated as a tool in the safety case in the pre-construction, operational and the post closure phase of storage. In particular the iterative process of evaluating and predicting safety scenarios at each stage of the process has proved to be prudent. It is important that this concept be adopted for this type of facility to ensure safety of mankind and the environment now and in the future.

  14. MAPLE-X10 reactor safety assessment

    International Nuclear Information System (INIS)

    Cotnam, K.D.; Lounsbury, R.I.; Gillespie, G.E.

    1990-01-01

    This paper reports on the safety assessment of the 10 MW MAPLE-X10 reactor which has involved a substantial component of PSA analysis to supplement deterministic analysis. Initiating events are identified through the use of a master logic diagram. The events are then examined through event sequence diagrams, at the concept design stage, followed by a set of reliability analyses that are coordinated with the event sequence diagrams. Improvements identified through the reliability analyses are incorporated into the design to ensure that safety objectives are attained

  15. Real-time safety risk assessment based on a real-time location system for hydropower construction sites.

    Science.gov (United States)

    Jiang, Hanchen; Lin, Peng; Fan, Qixiang; Qiang, Maoshan

    2014-01-01

    The concern for workers' safety in construction industry is reflected in many studies focusing on static safety risk identification and assessment. However, studies on real-time safety risk assessment aimed at reducing uncertainty and supporting quick response are rare. A method for real-time safety risk assessment (RTSRA) to implement a dynamic evaluation of worker safety states on construction site has been proposed in this paper. The method provides construction managers who are in charge of safety with more abundant information to reduce the uncertainty of the site. A quantitative calculation formula, integrating the influence of static and dynamic hazards and that of safety supervisors, is established to link the safety risk of workers with the locations of on-site assets. By employing the hidden Markov model (HMM), the RTSRA provides a mechanism for processing location data provided by the real-time location system (RTLS) and analyzing the probability distributions of different states in terms of false positives and negatives. Simulation analysis demonstrated the logic of the proposed method and how it works. Application case shows that the proposed RTSRA is both feasible and effective in managing construction project safety concerns.

  16. Intrusion resistant underground structure (IRUS) - safety assessment and licensing

    International Nuclear Information System (INIS)

    Lange, B. A.

    1997-01-01

    This paper describes the safety goals, human exposure scenarios and critical groups, the syvac-nsure performance assessment code, groundwater pathway safety results, and inadvertent human intrusion of the IRUS. 2 tabs

  17. Planning the Unplanned Experiment: Towards Assessing the Efficacy of Standards for Safety-Critical Software

    Science.gov (United States)

    Graydon, Patrick J.; Holloway, C. M.

    2015-01-01

    Safe use of software in safety-critical applications requires well-founded means of determining whether software is fit for such use. While software in industries such as aviation has a good safety record, little is known about whether standards for software in safety-critical applications 'work' (or even what that means). It is often (implicitly) argued that software is fit for safety-critical use because it conforms to an appropriate standard. Without knowing whether a standard works, such reliance is an experiment; without carefully collecting assessment data, that experiment is unplanned. To help plan the experiment, we organized a workshop to develop practical ideas for assessing software safety standards. In this paper, we relate and elaborate on the workshop discussion, which revealed subtle but important study design considerations and practical barriers to collecting appropriate historical data and recruiting appropriate experimental subjects. We discuss assessing standards as written and as applied, several candidate definitions for what it means for a standard to 'work,' and key assessment strategies and study techniques and the pros and cons of each. Finally, we conclude with thoughts about the kinds of research that will be required and how academia, industry, and regulators might collaborate to overcome the noted barriers.

  18. Safety Culture Perceptions in a Collegiate Aviation Program: A Systematic Assessment

    OpenAIRE

    Adjekum, Daniel Kwasi

    2014-01-01

    An assessment of the perceptions of respondents on the safety culture at an accredited Part 141 four year collegiate aviation program was conducted as part of the implementation of a safety management system (SMS). The Collegiate Aviation Program Safety Culture Assessment Survey (CAPSCAS), which was modified and revalidated from the existing Commercial Aviation Safety Survey (CASS), was used. Participants were drawn from flight students and certified flight instructors in the program. The sur...

  19. Specifications of the International Atomic Energy Agency's international project on safety assessment driven radioactive waste management solutions

    International Nuclear Information System (INIS)

    Ghannadi, M.; Asgharizadeh, F.; Assadi, M. R.

    2008-01-01

    Radioactive waste is produced in the generation of nuclear power and the production and use of radioactive materials in the industry, research, and medicine. The nuclear waste management facilities need to perform a safety assessment in order to ensure the safety of a facility. Nuclear safety assessment is a structured and systematic way of examining a proposed facility, process, operation and activity. In nuclear waste management point of view, safety assessment is a process which is used to evaluate the safety of radioactive waste management and disposal facilities. In this regard the International Atomic Energy Agency is planed to implement an international project with cooperation of some member states. The Safety Assessment Driving Radioactive Waste Management Solutions Project is an international programme of work to examine international approaches to safety assessment in aspects of p redisposal r adioactive waste management, including waste conditioning and storage. This study is described the rationale, common aspects, scope, objectives, work plan and anticipated outcomes of the project with refer to International Atomic Energy Agency's documents, such as International Atomic Energy Agency's Safety Standards, as well as the Safety Assessment Driving Radioactive Waste Management Solutions project reports

  20. Role and meaning of safety assessment from the point of view of IAEA

    International Nuclear Information System (INIS)

    Lyubarskiy, A.

    2012-01-01

    In 2006, the IAEA published its revised Safety Fundamentals. This states that the ''fundamental safety objective is to protect people and the environment from harmful effects of ionizing radiation''. This objective has to be achieved for all facilities and activities and for all stages over the lifetime of a facility by adherence to ten fundamental principles. This leads, inter alia, to the requirement for a safety assessment to be carried out. In particular, the text accompanying Principle 3 on leadership and management for safety states that: ''3.15. Safety has to be assessed for all facilities and activities, consistent with a graded approach. Safety assessment involves the systematic analysis of normal operation and its effects, of the ways in which failures might occur and of the consequences of such failures. Safety assessments cover the safety measures necessary to control the hazard, and the design and engineered safety features are assessed to demonstrate that they fulfill the safety functions required of them. Where control measures or operator actions are called on to maintain safety, an initial safety assessment has to be carried out to demonstrate that the arrangements made are robust and that they can be relied on. A facility may only be constructed and commissioned or an activity may only be commenced once it has been demonstrated to the satisfaction of the regulatory body that the proposed safety measures are adequate.'' Principle 3 further states that the process of safety assessment for facilities and activities is repeated in the conduct of operations in order to take into account changed circumstances (such as the application of new standards or scientific and technological developments), the feedback of operating experience, modifications and the effects of ageing. Continuation of operations over long periods of time requires reassessments demonstrating that the safety measures remain adequate. (orig.)

  1. Case studies in the application of probabilistic safety assessment techniques to radiation sources. Final report of a coordinated research project 2001-2003

    International Nuclear Information System (INIS)

    2006-04-01

    Radiation sources are used worldwide in many industrial and medical applications. In general, the safety record associated with their use has been very good. However, accidents involving these sources have occasionally resulted in unplanned exposures to individuals. When assessed prospectively, this type of exposure is termed a 'potential exposure'. The International Commission on Radiological Protection (ICRP) has recommended the assessment of potential exposures that may result from radiation sources and has suggested that probabilistic safety assessment (PSA) techniques may be used in this process. Also, Paragraph 2.13 of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources (BSS) requires that the authorization process for radiation sources include an assessment of all exposures, including potential exposures, which may result from the use of a radiation source. In light of the ICRP's work described above, and the possibility that PSA techniques could be used in exposure assessments that are required by the BSS, the IAEA initiated a coordinated research project (CRP) to study the benefits and limitations of the application of PSA techniques to radiation sources. The results of this CRP are presented in this publication. It should be noted that these results are based solely on the work performed, and the conclusions drawn, by the research teams involved in this CRP. It is intended that international organizations involved in radiation protection will review the information in this report and will take account of it during the development of guidance and requirements related to the assessment of potential exposures from radiation sources. Also, it is anticipated that the risk insights obtained through the studies will be considered by medical practitioners, facility staff and management, equipment designers, and regulators in their safety management and risk evaluation activities. A draft

  2. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-04-15

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10{sup -6} per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective

  3. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    International Nuclear Information System (INIS)

    1990-04-01

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10 -6 per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective of

  4. Methodology for Safety Assessment Applied to Predisposal Waste Management. Report of the Results of the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) 2004–2010)

    International Nuclear Information System (INIS)

    2015-12-01

    Report of the Results of the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) (2004–2010) The IAEA’s progamme on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) focused on approaches and mechanisms for application of safety assessment methodologies for the predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts, which have since been incorporated into IAEA Safety Standards Series No. GSG-3, Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste. In 2005, an initial specification was developed for the Safety Assessment Framework (SAFRAN) software tool to apply the SADRWMS flowcharts. In 2008, an in-depth application of the SAFRAN tool and the SADRWMS methodology was carried out on the predisposal management facilities of the Thailand Institute of Nuclear Technology Radioactive Waste Management Centre (TINT Facility). This publication summarizes the content and outcomes of the SADRWMS programme. The Chairman’s Report of the SADRWMS Project and the Report of the TINT test case are provided on the CD-ROM which accompanies this report

  5. Pilot Study on Harmonisation of Reactor Safety in WENRA Countries

    International Nuclear Information System (INIS)

    2003-03-01

    Most of the objectives, set for the Pilot Study, were met. It can be concluded that the methodology was adequate for its purpose. National requirements on selected safety issues have been systematically compared and the major gaps and differences have been identified. Convenient overviews have been provided of differences and similarities between the countries. Furthermore, the conclusions are based on a safety justification and are detailed enough to provide input to a further more detailed analysis on the national level. It was not possible, however, to provide fully verified conclusions about the implementation of the reference levels in the different countries. This has to do with the following constraints on the study: In line with the Terms of Reference, the comparison of formal requirements did not address the more detailed use of criteria and methods to verify compliance. The same requirement could be enforced differently in different regulatory systems, and hence lead to different implementation. The Pilot Study also assessed the implementation, but it was not possible to do this in sufficient detail to identify such differences. The implementation was assessed on the basis of current knowledge of the respective regulatory body, but it was not possible to provide the panels with evidence of the implementation. For these reasons, conclusions about implemented safety provisions in the different countries should be drawn with precaution. The introduction of the panel assessments greatly improved the quality and consistency of the comparison assessments. Uncertainties in the assessments are mainly connected with lack of time to make a detailed analysis in some cases. The reliability of the assessments seems to be sufficient for the objectives of the Pilot Study. The introduction of the IAEA safety standards in the study proved to be helpful and provided confidence in the scope and strictness of the reference levels. This Pilot Study has contributed to

  6. Safety assessment of human and organizational factors in French fuel cycle facilities

    International Nuclear Information System (INIS)

    Menuet, Lise; Beauquier, Sophie

    2013-01-01

    According to the French law, each nuclear facility has to provide a safety demonstration every ten years. The assessment of this demonstration supports the decision of the French Safety Authority regarding the authorisation of operating for the ten years to come. In addition, transversal topics, which are linked with safety performance, such as safety management, management of competencies, maintenance's policy are periodically evaluated. One aspect of these assessments relates to Human and Organizational Factors (HOF) and their contribution to safety. Our communication will describe the assessment of the HOF-related part, performed by the Institute for Radioprotection and Nuclear Safety Institute (IRSN) the Technical Support Organisation of the French Safety Authority). It will focus on the methodological framework, the tools which are developed and used for assessing the integration of HOF in safety demonstration, and the main difficulties of this kind of assessment. Each situation will be illustrated by concrete examples coming from safety assessments concerning fuel cycle's plants: Areva's plants dedicated to uranium conversion, uranium enrichment, fuel manufacturing, spent fuel reprocessing, treatment facilities and CEA's laboratories dedicated to research and development and to interim spent fuel storage. The methodological framework for assessing HOF currently implements three main steps which will be precisely described: - checking that the nuclear plant has made an exhaustive analysis of the risks linked with HOF. Regarding to HOF, the Licensee safety demonstration is based on the description of the main human activities which are considered as hazardous regarding safety. These activities are accomplished with a human contribution and they require a safe realisation. - assessing the human, organisational and technical barriers that the nuclear plant have planed in order to make the operations safe, to avoid, prevent or detect an

  7. Assessment of safety culture in the Iranian nuclear installations

    International Nuclear Information System (INIS)

    Farahani, H.F.; Davilu, H.; Sepanloo, K.

    2005-01-01

    The deficient safety culture (S.C) is the center of safety issues of nuclear industry. To benefit from the advantages of nuclear technology and considering the fact of potential hazards of accidents in nuclear installations it is essential to view safety as the highest priority. S.C is an amalgamation of values, standards, morals and norms of acceptable behavior. Organizations having effective S.C show constant commitment to safety as a top level priority. Furthermore, the personnel of a nuclear facility shall recognize the safety significance of their tasks. Many people even those who work in the field of safety do not have a correct understanding of what S.C looks like in practical sense. In this study, by conducting a survey according to IAEA-TECDOC-1329 in some nuclear facilities, the S.C within the Iranian nuclear facilities is assessed. The human and organizational factors in Tehran Research Reactor are evaluated using a questionnaire method with active participation of the reactor operators. The results sho w that the operators are pretty aware of the subject. Also it has been identified some areas of improvement. (authors)

  8. A Study on Safety and Risk Assessment of Dangerous Cargo Operations in Oil/Chemical Tankers

    Directory of Open Access Journals (Sweden)

    Cenk ŞAKAR

    2017-12-01

    Full Text Available The safety and risk assessment of dangerous cargo operations in oil and chemical tankers is a necessary process to prevent possible accidents during these operations. Fire and explosion are the major accidents encountered in tanker operations. In this study, a model was constructed through the Fuzzy Bayes Network Method for the probabilistic relationships between the causes of fire and explosion accidents that could occur during the tank cleaning process. The study is composed of two stages. Firstly, the variables that are the subject of the problem and that constitute the graphical structure of the Bayes Networks are identified. Then, expert opinion was sought as the statistical data on accident reports were not recorded properly while identifying the conditional probability of the relationships between the variables. Linguistic variables whose fuzzy membership functions were identified were used in detecting the probabilities. The findings of the sensitivity test revealed that the major reasons that could lead to fire and explosion during the tank cleaning process are ignition sources, reaction and safety culture.

  9. Effects of a team-based assessment and intervention on patient safety culture in general practice

    DEFF Research Database (Denmark)

    Hoffmann, B; Müller, V; Rochon, J

    2014-01-01

    Background: The measurement of safety culture in healthcare is generally regarded as a first step towards improvement. Based on a self-assessment of safety culture, the Frankfurt Patient Safety Matrix (FraTrix) aims to enable healthcare teams to improve safety culture in their organisations....... In this study we assessed the effects of FraTrix on safety culture in general practice. Methods: We conducted an open randomised controlled trial in 60 general practices. FraTrix was applied over a period of 9 months during three facilitated team sessions in intervention practices. At baseline and after 12...... months, scores were allocated for safety culture as expressed in practice structure and processes (indicators), in safety climate and in patient safety incident reporting. The primary outcome was the indicator error management. Results: During the team sessions, practice teams reflected on their safety...

  10. Quantitative assessment of safety barrier performance in the prevention of domino scenarios triggered by fire

    International Nuclear Information System (INIS)

    Landucci, Gabriele; Argenti, Francesca; Tugnoli, Alessandro; Cozzani, Valerio

    2015-01-01

    The evolution of domino scenarios triggered by fire critically depends on the presence and the performance of safety barriers that may have the potential to prevent escalation, delaying or avoiding the heat-up of secondary targets. The aim of the present study is the quantitative assessment of safety barrier performance in preventing the escalation of fired domino scenarios. A LOPA (layer of protection analysis) based methodology, aimed at the definition and quantification of safety barrier performance in the prevention of escalation was developed. Data on the more common types of safety barriers were obtained in order to characterize the effectiveness and probability of failure on demand of relevant safety barriers. The methodology was exemplified with a case study. The results obtained define a procedure for the estimation of safety barrier performance in the prevention of fire escalation in domino scenarios. - Highlights: • We developed a methodology for the quantitative assessment of safety barriers. • We focused on safety barriers aimed at preventing domino effect triggered by fire. • We obtained data on effectiveness and availability of the safety barriers. • The methodology was exemplified with a case study of industrial interest. • The results showed the role of safety barriers in preventing fired domino escalation

  11. Contribution of operating feedback to probabilistic safety studies

    International Nuclear Information System (INIS)

    Guio, J.M. de; Lannoy, A.

    1992-03-01

    This paper presents the method used for PWR unit operation feedback analysis and its contribution to probabilistic safety studies. The targets were as follows: - use of failure data banks to assess reliability parameters, - use of event data banks to identify and quantify main system initiating events, - determination of a standard operating profile. These studies, performed in the context of nuclear power plant safety programs, prove useful not only to safety engineers but also to equipment experts, designers, operators and maintenance specialists. They constitute basic data for studies in all these areas or the departure point for new investigations. (authors). 3 figs., 3 tabs., 3 refs

  12. The role of probabilistic safety assessment in the design

    International Nuclear Information System (INIS)

    Green, A.; Ingham, E.L.

    1989-01-01

    The use of probabilistic safety assessment (PSA) for Heysham 2 and Torness marked a major change in the design approach to nuclear safety within the U.K. Design Safety Guidelines incorporating probabilistic safety targets required that design justification would necessitate explicit consideration of the consequence of accidents in relation to their frequency. The paper discusses these safety targets and their implications, the integration of PSA into the design process and an outline of the methodology. The influence of PSA on the design is discussed together with its role in the overall demonstration of reactor safety. (author)

  13. Procedures for conducting probabilistic safety assessment for non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    2002-01-01

    A well performed and adequately documented safety assessment of a nuclear facility will serve as a basis to determine whether the facility complies with the safety objectives, principles and criteria as stipulated by the national regulatory body of the country where the facility is in operation. International experience shows that the practices and methodologies used to perform safety assessments and periodic safety re-assessment for non-reactor nuclear facilities differ significantly from county to country. Most developing countries do not have methods and guidance for safety assessment that are prescribed by the regulatory body. Typically the safety evaluation for the facility is based on a case by case assessment. Whilst conservative deterministic analyses are predominantly used as a licensing basis in many countries, recently probabilistic safety assessment (PSA) techniques have been applied as a useful complementary tool to support safety decision making. The main benefit of PSA is to provide insights into the safety aspects of facility design and operation. PSA points up the potential environmental impacts of postulated accidents, including the dominant risk contributors, and enables safety analysts to compare options for reducing risk. In order to advise on how to apply PSA methodology for the safety assessment of non-reactor nuclear facilities, the IAEA organized several consultants meetings, which led to the preparation of this TECDOC. This document is intended as guidance for the conduct of PSA in non-nuclear facilities. The main emphasis here is on the general procedural steps of a PSA that is specific for a non-reactor nuclear facility, rather than the details of the specific methods. The report is directed at technical staff managing or performing such probabilistic assessments and to promote a standardized framework, terminology and form of documentation for these PSAs. It is understood that the level of detail implied in the tasks presented in this

  14. Selected component failure rate values from fusion safety assessment tasks

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, L.C.

    1998-09-01

    This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers.

  15. Selected Component Failure Rate Values from Fusion Safety Assessment Tasks

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, Lee Charles

    1998-09-01

    This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers.

  16. Selected component failure rate values from fusion safety assessment tasks

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    1998-01-01

    This report is a compilation of component failure rate and repair rate values that can be used in magnetic fusion safety assessment tasks. Several safety systems are examined, such as gas cleanup systems and plasma shutdown systems. Vacuum system component reliability values, including large vacuum chambers, have been reviewed. Values for water cooling system components have also been reported here. The report concludes with the examination of some equipment important to personnel safety, atmospheres, combustible gases, and airborne releases of radioactivity. These data should be useful to system designers to calculate scoping values for the availability and repair intervals for their systems, and for probabilistic safety or risk analysts to assess fusion systems for safety of the public and the workers

  17. Safety assessment and quality control of medical x-ray facilities in some hospitals in Ghana

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  18. Environment, safety and health progress assessment manual

    International Nuclear Information System (INIS)

    1992-12-01

    On June 27, 1989, the Secretary of Energy announced a 10-Point Initiative to strengthen environment, safety, and health (ES ampersand H) programs, and waste management activities at DOE production, research, and testing facilities. One of the points involved conducting dent Tiger Team Assessments of DOE operating facilities. The Office of Special independent Projects (OSP), EH-5, in the Office of the Assistant Secretary for Environment, Safety and Health, EH-1, was assigned the responsibility to conduct the Tiger Team Assessments. Through June 1992, a total of 35 Tiger Team Assessments were completed. The Secretary directed that Corrective Action Plans be developed and implemented to address the concerns identified by the Tiger Teams. In March 1991, the Secretary approved a plan for assessments that are ''more focused, concentrating on ES ampersand H management, ES ampersand H corrective actions, self-assessment programs, and root-cause related issues.'' In July 1991, the Secretary approved the initiation of ES ampersand H Progress Assessments, as a followup to the Tiger Team Assessments, and in the continuing effort to institutionalize the self-assessment process and line management accountability in the ES ampersand H areas. This manual documents the processes to be used to perform the ES ampersand H Progress Assessments. It was developed based upon the lessons learned from Tiger Team Assessments, the two pilot Progress Assessments, and Progress Assessments that have been completed. The manual will be updated periodically to reflect lessons learned or changes in policy

  19. Assessment of Contributions to Patient Safety Knowledge by the Agency for Healthcare Research and Quality-Funded Patient Safety Projects

    Science.gov (United States)

    Sorbero, Melony E S; Ricci, Karen A; Lovejoy, Susan; Haviland, Amelia M; Smith, Linda; Bradley, Lily A; Hiatt, Liisa; Farley, Donna O

    2009-01-01

    Objective To characterize the activities of projects funded in Agency for Healthcare Research and Quality (AHRQ)' patient safety portfolio and assess their aggregate potential to contribute to knowledge development. Data Sources Information abstracted from proposals for projects funded in AHRQ' patient safety portfolio, information on safety practices from the AHRQ Evidence Report on Patient Safety Practices, and products produced by the projects. Study Design This represented one part of the process evaluation conducted as part of a longitudinal evaluation based on the Context–Input–Process–Product model. Principal Findings The 234 projects funded through AHRQ' patient safety portfolio examined a wide variety of patient safety issues and extended their work beyond the hospital setting to less studied parts of the health care system. Many of the projects implemented and tested practices for which the patient safety evidence report identified a need for additional evidence. The funded projects also generated a substantial body of new patient safety knowledge through a growing number of journal articles and other products. Conclusions The projects funded in AHRQ' patient safety portfolio have the potential to make substantial contributions to the knowledge base on patient safety. The full value of this new knowledge remains to be confirmed through the synthesis of results. PMID:21456108

  20. Study on improvement in reliability of inventory assessment in vitrified waste for long-term safety of geological disposal

    International Nuclear Information System (INIS)

    Ishikawa, Masumi; Kaneko, Satoru; Kitayama, Kazumi; Ishiguro, Katsuhiko; Ueda, Hiroyoshi; Wakasugi, Keiichiro; Shinohara, Nobuo; Okumura, Keisuke; Chino, Masamichi; Moriya, Noriyasu

    2009-01-01

    Since quality control issues for vitrified waste are defined mainly with the focus on the transport and interim storage of the waste rather than the long-term safety of geological disposal, they do not cover inventories of long-lived nuclides that are of most interest in the safety assessment of geological disposal. Therefore, we suggest a flow chart for the assessment of inventories of long-lived nuclides in the vitrified waste focusing on the measured values. This includes an indirect assessment with indicative nuclides that have been already measured in the returned vitrified wastes from abroad. In order to apply this flow chart for commercial operation, its applicability should be examined for cases with a variation in burn-up history and with an uncertainty associated with carry-over fraction at reprocessing. We started an R and D program to examine the applicability as well as to improve the reliability of the nuclide generation/decay code and the nuclear data library using liquid waste from spent fuel with a clear irradiation history. To solve the issue of quality control for vitrified waste, a comprehensive study is needed in aspects not only of geological disposal field but also of operation of a nuclear power plant, reprocessing of spent fuel and vitrification of liquid waste. This study is a pioneering study conducted to integrate them. (author)

  1. Safety Assessment of Food and Feed from GM Crops in Europe: Evaluating EFSA's Alternative Framework for the Rat 90-day Feeding Study.

    Science.gov (United States)

    Hong, Bonnie; Du, Yingzhou; Mukerji, Pushkor; Roper, Jason M; Appenzeller, Laura M

    2017-07-12

    Regulatory-compliant rodent subchronic feeding studies are compulsory regardless of a hypothesis to test, according to recent EU legislation for the safety assessment of whole food/feed produced from genetically modified (GM) crops containing a single genetic transformation event (European Union Commission Implementing Regulation No. 503/2013). The Implementing Regulation refers to guidelines set forth by the European Food Safety Authority (EFSA) for the design, conduct, and analysis of rodent subchronic feeding studies. The set of EFSA recommendations was rigorously applied to a 90-day feeding study in Sprague-Dawley rats. After study completion, the appropriateness and applicability of these recommendations were assessed using a battery of statistical analysis approaches including both retrospective and prospective statistical power analyses as well as variance-covariance decomposition. In the interest of animal welfare considerations, alternative experimental designs were investigated and evaluated in the context of informing the health risk assessment of food/feed from GM crops.

  2. A feasibility study for Arizona's roadway safety management process using the Highway Safety Manual and SafetyAnalyst : final report.

    Science.gov (United States)

    2016-07-01

    To enable implementation of the American Association of State Highway Transportation (AASHTO) Highway Safety Manual using : SaftetyAnalyst (an AASHTOWare software product), the Arizona Department of Transportation (ADOT) studied the data assessment :...

  3. Plasma-safety assessment model and safety analyses of ITER

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Bartels, H.-H.; Uckan, N.A.; Sugihara, M.; Seki, Y.

    2001-01-01

    A plasma-safety assessment model has been provided on the basis of the plasma physics database of the International Thermonuclear Experimental Reactor (ITER) to analyze events including plasma behavior. The model was implemented in a safety analysis code (SAFALY), which consists of a 0-D dynamic plasma model and a 1-D thermal behavior model of the in-vessel components. Unusual plasma events of ITER, e.g., overfueling, were calculated using the code and plasma burning is found to be self-bounded by operation limits or passively shut down due to impurity ingress from overheated divertor targets. Sudden transition of divertor plasma might lead to failure of the divertor target because of a sharp increase of the heat flux. However, the effects of the aggravating failure can be safely handled by the confinement boundaries. (author)

  4. Assessing medical students' perceptions of patient safety: the medical student safety attitudes and professionalism survey.

    Science.gov (United States)

    Liao, Joshua M; Etchegaray, Jason M; Williams, S Tyler; Berger, David H; Bell, Sigall K; Thomas, Eric J

    2014-02-01

    To develop and test the psychometric properties of a survey to measure students' perceptions about patient safety as observed on clinical rotations. In 2012, the authors surveyed 367 graduating fourth-year medical students at three U.S. MD-granting medical schools. They assessed the survey's reliability and construct and concurrent validity. They examined correlations between students' perceptions of organizational cultural factors, organizational patient safety measures, and students' intended safety behaviors. They also calculated percent positive scores for cultural factors. Two hundred twenty-eight students (62%) responded. Analyses identified five cultural factors (teamwork culture, safety culture, error disclosure culture, experiences with professionalism, and comfort expressing professional concerns) that had construct validity, concurrent validity, and good reliability (Cronbach alphas > 0.70). Across schools, percent positive scores for safety culture ranged from 28% (95% confidence interval [CI], 13%-43%) to 64% (30%-98%), while those for teamwork culture ranged from 47% (32%-62%) to 74% (66%-81%). They were low for error disclosure culture (range: 10% [0%-20%] to 27% [20%-35%]), experiences with professionalism (range: 7% [0%-15%] to 23% [16%-30%]), and comfort expressing professional concerns (range: 17% [5%-29%] to 38% [8%-69%]). Each cultural factor correlated positively with perceptions of overall patient safety as observed in clinical rotations (r = 0.37-0.69, P safety behavioral intent item. This study provided initial evidence for the survey's reliability and validity and illustrated its applicability for determining whether students' clinical experiences exemplify positive patient safety environments.

  5. Safety assessment methodology for waste repositories in deep geological formations

    International Nuclear Information System (INIS)

    Chapuis, A.M.; Lewi, J.; Pradel, J.; Queniart, D.; Raimbault, P.; Assouline, M.

    1986-06-01

    The long term safety of a nuclear waste repository relies on the evaluation of the doses which could be transferred to man in the future. This implies a detailed knowledge of the medium where the waste will be confined, the identification of the basic phenomena which govern the migration of the radionuclides and the investigation of all possible scenarios that may affect the integrity of the barriers between the waste and the biosphere. Inside the Institute of protection and nuclear safety of the French Atomic Energy Commission (CEA/IPSN), the Department of the Safety Analysis (DAS) is currently developing a methodology for assessing the safety of future geological waste repositories, and is in charge of the modelling development, while the Department of Technical Protection (DPT) is in charge of the geological experimental studies. Both aspects of this program are presented. The methodology for risk assessment stresses the needs for coordination between data acquisition and model development which should result in the obtention of an efficient tool for safety evaluation. Progress needs to be made in source and geosphere modelling. Much more sophisticated models could be used than the ones which is described; however sensitivity analysis will determine the level of sophistication which is necessary to implement. Participation to international validation programs are also very important for gaining confidence in the approaches which have been chosen

  6. Safety assessment for the above ground storage of Cadmium Safety and Control Rods at the Solid Waste Management Facility

    International Nuclear Information System (INIS)

    Shaw, K.W.

    1993-11-01

    The mission of the Savannah River Site is changing from radioisotope production to waste management and environmental restoration. As such, Reactor Engineering has recently developed a plan to transfer the safety and control rods from the C, K, L, and P reactor disassembly basin areas to the Transuranic (TRU) Waste Storage Pads for long-term, retrievable storage. The TRU pads are located within the Solid Waste Management Facilities at the Savannah River Site. An Unreviewed Safety Question (USQ) Safety Evaluation has been performed for the proposed disassembly basin operations phase of the Cadmium Safety and Control Rod Project. The USQ screening identified a required change to the authorization basis; however, the Proposed Activity does not involve a positive USQ Safety Evaluation. A Hazard Assessment for the Cadmium Safety and Control Rod Project determined that the above-ground storage of the cadmium rods results in no change in hazard level at the TRU pads. A Safety Assessment that specifically addresses the storage (at the TRU pads) phase of the Cadmium Safety and Control Rod Project has been performed. Results of the Safety Assessment support the conclusion that a positive USQ is not involved as a result of the Proposed Activity

  7. Review and assessment of nuclear facilities by the regulatory body. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    The purpose of this Safety Guide is to provide recommendations for regulatory bodies on reviewing and assessing the various safety related submissions made by the operator of a nuclear facility at different stages (siting, design, construction, commissioning, operation and decommissioning or closure) in the facility's lifetime to determine whether the facility complies with the applicable safety objectives and requirements. This Safety Guide covers the review and assessment of submissions in relation to the safety of nuclear facilities such as: enrichment and fuel manufacturing plants. Nuclear power plants. Other reactors such as research reactors and critical assemblies. Spent fuel reprocessing plants. And facilities for radioactive waste management, such as treatment, storage and disposal facilities. This Safety Guide also covers issues relating to the decommissioning of nuclear facilities, the closure of waste disposal facilities and site rehabilitation. Objectives, management, planning and organizational matters relating to the review and assessment process are presented in Section 2. Section 3 deals with the bases for decision making and conduct of the review and assessment process. Section 4 covers aspects relating to the assessment of this process. The Appendix provides a generic list of topics to be covered in the review and assessment process

  8. Review and assessment of nuclear facilities by the regulatory body. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    The purpose of this Safety Guide is to provide recommendations for regulatory bodies on reviewing and assessing the various safety related submissions made by the operator of a nuclear facility at different stages (siting, design, construction, commissioning, operation and decommissioning or closure) in the facility's lifetime to determine whether the facility complies with the applicable safety objectives and requirements. This Safety Guide covers the review and assessment of submissions in relation to the safety of nuclear facilities such as: enrichment and fuel manufacturing plants. Nuclear power plants. Other reactors such as research reactors and critical assemblies. Spent fuel reprocessing plants. And facilities for radioactive waste management, such as treatment, storage and disposal facilities. This Safety Guide also covers issues relating to the decommissioning of nuclear facilities, the closure of waste disposal facilities and site rehabilitation. Objectives, management, planning and organizational matters relating to the review and assessment process are presented in Section 2. Section 3 deals with the bases for decision making and conduct of the review and assessment process. Section 4 covers aspects relating to the assessment of this process. The Appendix provides a generic list of topics to be covered in the review and assessment process

  9. Safety assessment guidance in the International Atomic Energy Agency RADWASS Program

    Energy Technology Data Exchange (ETDEWEB)

    Vovk, I.F.; Seitz, R.R.

    1995-12-31

    The IAEA RADWASS programme is aimed at establishing a coherent and comprehensive set of principles and standards for the safe management of waste and formulating the guidelines necessary for their application. A large portion of this programme has been devoted to safety assessments for various waste management activities. Five Safety Guides are planned to be developed to provide general guidance to enable operators and regulators to develop necessary framework for safety assessment process in accordance with international recommendations. They cover predisposal, near surface disposal, geological disposal, uranium/thorium mining and milling waste, and decommissioning and environmental restoration. The Guide on safety assessment for near surface disposal is at the most advanced stage of preparation. This draft Safety Guide contains guidance on description of the disposal system, development of a conceptual model, identification and description of relevant scenarios and pathways, consequence analysis, presentation of results and confidence building. The set of RADWASS publications is currently undergoing in-depth review to ensure a harmonized approach throughout the Safety Series.

  10. Uncertainty in safety : new techniques for the assessment and optimisation of safety in process industry

    NARCIS (Netherlands)

    Rouvroye, J.L.; Nieuwenhuizen, J.K.; Brombacher, A.C.; Stavrianidis, P.; Spiker, R.Th.E.; Pyatt, D.W.

    1995-01-01

    At this moment there is no standardised method for the assessment for safety in the process industry. Many companies and institutes use qualitative techniques for safety analysis while other companies and institutes use quantitative techniques. The authors of this paper will compare different

  11. A quantitative assessment of organizational factors affecting safety using a system dynamics model

    International Nuclear Information System (INIS)

    Yoo, J. K.; Yoon, T. S.

    2003-01-01

    The purpose of this study is to develop a system dynamics model for the assessment of organizational and human factors in the nuclear power plant safety. Previous studies are classified into two major approaches. One is the engineering approach such as ergonomics and Probabilistic Safety Assessment (PSA). The other is socio-psychology one. Both have contributed to find organizational and human factors and increased nuclear safety However, since these approaches assume that the relationship among factors is independent they do not explain the interactions between factors or variables in NPP's. To overcome these restrictions, a system dynamics model, which can show causal relations between factors and quantify organizational and human factors, has been developed. Operating variables such as degree of leadership, adjustment of number of employee, and workload in each department, users can simulate various situations in nuclear power plants in the organization side. Through simulation, user can get an insight to improve safety in plants and to find managerial tools in the organization and human side

  12. Nuclear utility self-assessment as viewed by the corporate nuclear safety committee

    International Nuclear Information System (INIS)

    Corcoran, W.R.

    1992-01-01

    This paper discusses how corporate nuclear safety committees use the principles of self-assessment to enhance nuclear power plant safety performance. Corporate nuclear safety committees function to advise the senior nuclear power executive on matters affecting nuclear safety. These committees are required by the administrative controls section of the plant technical specifications which are part of the final safety analysis report and the operating license. Committee membership includes senior utility executives, executives from sister utilities, utility senior technical experts, and outside consultants. Current corporate nuclear safety committees often have a finely tuned intuitive feel for self-assessment that they use to probe the underlying opportunities for quality and safety enhancements. The questions prompted by the self-assessment orientation enable the utility line organization members to gain better perspectives on the characteristics of the organizational systems that they manage and work in

  13. Quantitative reliability assessment for safety critical system software

    International Nuclear Information System (INIS)

    Chung, Dae Won; Kwon, Soon Man

    2005-01-01

    An essential issue in the replacement of the old analogue I and C to computer-based digital systems in nuclear power plants is the quantitative software reliability assessment. Software reliability models have been successfully applied to many industrial applications, but have the unfortunate drawback of requiring data from which one can formulate a model. Software which is developed for safety critical applications is frequently unable to produce such data for at least two reasons. First, the software is frequently one-of-a-kind, and second, it rarely fails. Safety critical software is normally expected to pass every unit test producing precious little failure data. The basic premise of the rare events approach is that well-tested software does not fail under normal routine and input signals, which means that failures must be triggered by unusual input data and computer states. The failure data found under the reasonable testing cases and testing time for these conditions should be considered for the quantitative reliability assessment. We will present the quantitative reliability assessment methodology of safety critical software for rare failure cases in this paper

  14. Safety assessment of botanicals and botanical preparations used as ingredients in food supplements: testing an European Food Safety Authority-tiered approach.

    Science.gov (United States)

    Speijers, Gerrit; Bottex, Bernard; Dusemund, Birgit; Lugasi, Andrea; Tóth, Jaroslav; Amberg-Müller, Judith; Galli, Corrado L; Silano, Vittorio; Rietjens, Ivonne M C M

    2010-02-01

    This article describes results obtained by testing the European Food Safety Authority-tiered guidance approach for safety assessment of botanicals and botanical preparations intended for use in food supplements. Main conclusions emerging are as follows. (i) Botanical ingredients must be identified by their scientific (binomial) name, in most cases down to the subspecies level or lower. (ii) Adequate characterization and description of the botanical parts and preparation methodology used is needed. Safety of a botanical ingredient cannot be assumed only relying on the long-term safe use of other preparations of the same botanical. (iii) Because of possible adulterations, misclassifications, replacements or falsifications, and restorations, establishment of adequate quality control is necessary. (iv) The strength of the evidence underlying concerns over a botanical ingredient should be included in the safety assessment. (v) The matrix effect should be taken into account in the safety assessment on a case-by-case basis. (vi) Adequate data and methods for appropriate exposure assessment are often missing. (vii) Safety regulations concerning toxic contaminants have to be complied with. The application of the guidance approach can result in the conclusion that safety can be presumed, that the botanical ingredient is of safety concern, or that further data are needed to assess safety.

  15. Measuring Best Practices for Workplace Safety, Health, and Well-Being: The Workplace Integrated Safety and Health Assessment.

    Science.gov (United States)

    Sorensen, Glorian; Sparer, Emily; Williams, Jessica A R; Gundersen, Daniel; Boden, Leslie I; Dennerlein, Jack T; Hashimoto, Dean; Katz, Jeffrey N; McLellan, Deborah L; Okechukwu, Cassandra A; Pronk, Nicolaas P; Revette, Anna; Wagner, Gregory R

    2018-05-01

    To present a measure of effective workplace organizational policies, programs, and practices that focuses on working conditions and organizational facilitators of worker safety, health and well-being: the workplace integrated safety and health (WISH) assessment. Development of this assessment used an iterative process involving a modified Delphi method, extensive literature reviews, and systematic cognitive testing. The assessment measures six core constructs identified as central to best practices for protecting and promoting worker safety, health and well-being: leadership commitment; participation; policies, programs, and practices that foster supportive working conditions; comprehensive and collaborative strategies; adherence to federal and state regulations and ethical norms; and data-driven change. The WISH Assessment holds promise as a tool that may inform organizational priority setting and guide research around causal pathways influencing implementation and outcomes related to these approaches.

  16. Development of Safety Culture Assessment Strategy for Korean NPP

    International Nuclear Information System (INIS)

    Park, Jung Hwan; Kim, Jong Hyun

    2014-01-01

    This paper aims at developing the requirements for a method to evaluate the operational safety culture, evaluating currently available methods based on the requirements, and suggesting a method to evaluate and improve the operational safety culture for Korean nuclear power plants. This paper reviews the widely-used methods to assess safety culture for NPPs and their basis. Then, this paper develops the requirements for the method to evaluate operational safety culture for Korean NPPs. Based on these requirements, Korean Safety Culture Indicators (KSCI) and evaluation measures are also suggested. Finally this paper proposes the guidelines to develop improvements to safety culture from the evaluation results

  17. Development of Safety Culture Assessment Strategy for Korean NPP

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jung Hwan; Kim, Jong Hyun [KEPCO, Ulsan (Korea, Republic of)

    2014-08-15

    This paper aims at developing the requirements for a method to evaluate the operational safety culture, evaluating currently available methods based on the requirements, and suggesting a method to evaluate and improve the operational safety culture for Korean nuclear power plants. This paper reviews the widely-used methods to assess safety culture for NPPs and their basis. Then, this paper develops the requirements for the method to evaluate operational safety culture for Korean NPPs. Based on these requirements, Korean Safety Culture Indicators (KSCI) and evaluation measures are also suggested. Finally this paper proposes the guidelines to develop improvements to safety culture from the evaluation results.

  18. Comparative assessment of safety indicators for vehicle trajectories on the highway

    NARCIS (Netherlands)

    Mullakkal Babu, F.A.; Wang, M.; Farah, H.; van Arem, B.; Happee, R.

    2017-01-01

    Safety measurement and analysis have been a challenging and well-researched topic in transportation. Conventionally, surrogate safety measures have been used as safety indicators in simulation models for safety assessment, in control formulations for driver assistance systems, and in data analysis

  19. Safety Culture Assessment Tools in Nuclear and Non-Nuclear Domains

    International Nuclear Information System (INIS)

    Mkrtchyan, L.; Turcanu, C.

    2012-01-01

    Over the last decades, in many domains especially in high risk industries, the authorities paid increasing attention to safety management systems and, in particular, to safety culture. Consequently, in the applied and academic literature a huge amount of studies explored the main challenges, issues and obstacles related with safety culture. We undertake a survey of safety culture experiences in the main safety-critical industries such as nuclear, railways, offshore, aviation, airlines, health care, etc. We review both academic and applied literature up to the year 2011. Our results help to establish a comprehensive view on the subject, its main terminologies, existing tools, and main difficulties. The purpose of this report is to raise awareness about the current tools of safety culture assessment, both in the nuclear as well as in the non-nuclear domain. The report provides also practical recommendations about the possible use of each tool given different circumstances and different factors. We do not aim to rank the tools pointing the best one, but we highlight instead the unique features of these tools, pointing their strong and weak sides

  20. Safety Culture Assessment Tools in Nuclear and Non-Nuclear Domains

    Energy Technology Data Exchange (ETDEWEB)

    Mkrtchyan, L; Turcanu, C

    2012-03-15

    Over the last decades, in many domains especially in high risk industries, the authorities paid increasing attention to safety management systems and, in particular, to safety culture. Consequently, in the applied and academic literature a huge amount of studies explored the main challenges, issues and obstacles related with safety culture. We undertake a survey of safety culture experiences in the main safety-critical industries such as nuclear, railways, offshore, aviation, airlines, health care, etc. We review both academic and applied literature up to the year 2011. Our results help to establish a comprehensive view on the subject, its main terminologies, existing tools, and main difficulties. The purpose of this report is to raise awareness about the current tools of safety culture assessment, both in the nuclear as well as in the non-nuclear domain. The report provides also practical recommendations about the possible use of each tool given different circumstances and different factors. We do not aim to rank the tools pointing the best one, but we highlight instead the unique features of these tools, pointing their strong and weak sides.

  1. Safety assessment of Novi Han radioactive waste repository - features, problems, results and perspectives

    International Nuclear Information System (INIS)

    Mateeva, M.

    2000-01-01

    This paper summarizes the work done and the achievements reached in the Novi Han radioactive waste repository safety assessment within the IAEA Model Project 'Increasing the safety of Novi Han radioactive waste repository BUL 4/005'. The overall safety assessment has a wide context, but the work reported here relates only to some details and results concerning the development and implementation of the appropriate methodology approach, model and computer code used for the calculations. Different steps and procedures are included for a better practical understanding of the obtained results during the safety assessment performance. The methodology approach is widely based on an international experience in safety analysis and implemented for evaluation computer code AMBER, which is one of the recommended from the safety assessments experts. (author)

  2. 76 FR 74723 - New Car Assessment Program (NCAP); Safety Labeling

    Science.gov (United States)

    2011-12-01

    ... [Docket No. NHTSA 2010-0025] RIN 2127-AK51 New Car Assessment Program (NCAP); Safety Labeling AGENCY... NHTSA's regulation on vehicle labeling of safety rating information to reflect the enhanced NCAP ratings... Traffic Safety Administration under the enhanced NCAP testing and rating program. * * * * * (e) * * * (4...

  3. Abstract of results of safety study. Nuclear fuel cycle field in fiscal 2003

    International Nuclear Information System (INIS)

    2004-11-01

    This report descried the results of studies of nuclear fuel cycle field (nuclear fuel facilities, seismic design, all subjects of environmental radiation and waste disposal, and subjects on nuclear fuel cycle in probabilistic safety assessment) in fiscal 2003 on the basis of the principle project of safety study (from fiscal 2001 to 2005). It consists of four chapters; the first chapter is outline of the principle of project, the second is objects and subjects of safety study in the nuclear fuel cycle field, the third list of questionnaire of results of safety study and the forth investigation of results of safety study in fiscal 2003. There are 49 lists, which include 22 reports on the nuclear fuel facility, one on the seismic design, 4 on the probabilistic safety assessment, 7 on the environmental radiation and 15 on the waste disposal. (S.Y.)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-11-15

    . The standards are also applied by regulatory bodies and operators around the world to enhance safety in nuclear power generation and in nuclear applications in medicine, industry, agriculture and research. Safety is not an end in itself but a prerequisite for the purpose of the protection of people in all States and of the environment - now and in the future. The risks associated with ionizing radiation must be assessed and controlled without unduly limiting the contribution of nuclear energy to equitable and sustainable development. Governments, regulatory bodies and operators everywhere must ensure that nuclear material and radiation sources are used beneficially, safely and ethically. The IAEA safety standards are designed to facilitate this, and I encourage all Member States to make use of them.

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

    International Nuclear Information System (INIS)

    2011-01-01

    . The standards are also applied by regulatory bodies and operators around the world to enhance safety in nuclear power generation and in nuclear applications in medicine, industry, agriculture and research. Safety is not an end in itself but a prerequisite for the purpose of the protection of people in all States and of the environment - now and in the future. The risks associated with ionizing radiation must be assessed and controlled without unduly limiting the contribution of nuclear energy to equitable and sustainable development. Governments, regulatory bodies and operators everywhere must ensure that nuclear material and radiation sources are used beneficially, safely and ethically. The IAEA safety standards are designed to facilitate this, and I encourage all Member States to make use of them.

  6. Guidelines for the review research reactor safety. Reference document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

    International Nuclear Information System (INIS)

    1997-01-01

    In 1992, the IAEA published new safety standards for research reactors as part of the set of publications considered by its Research Reactor Safety Programme (RRSP). This set also includes publications giving guidance for all safety aspects related to the lifetime of a research reactor. In addition, the IAEA has also revised the Safety Standards for radiation protection. Consequently, it was considered advisable to revise the Integrated Safety Assessment of Research Reactors (INSARR) procedures to incorporate the new requirements and guidance as well as to extend the scope of the safety reviews to currently operating research reactors. The present report is the result of this revision. The purpose of this report is to give guidance on the preparation, execution, reporting and follow-up of safety review mission to research reactors as conducted by the IAEA under its INSARR missions safety service. However, it will also be of assistance to operators and regulators in conducting: (a) ad hoc safety assessments of research reactors to address individual issues such as ageing or safety culture; and (b) other types of safety reviews such as internal and peer reviews and regulatory inspections

  7. The safety assessment of radioactive material transpotation at sea

    International Nuclear Information System (INIS)

    Satoh, K.; Ozaki, S.; Watabe, N.; Fukuda, S.; Iida, T.; Miyao, S.; Noguchi, K.; Nakajima, K.

    1989-01-01

    Large quantities of low level wastes are prepared for transportation by special use vessels from each power plant to the storage facility at Rokkasho-mura in Aomori Prefecture. Large quantities of reprocessed wastes are also planned for return by similar vessels to the same place from France and the UK. In this paper the authors describe the safety assessment in hypothetical accident conditions during such mass transportation at sea. Although the possibilities of the sinking of the special use vessels as shown in figure 1 are considered to be very low on account of their double-hull structure, it is necessary to estimate the radiological risks of the transportation in order to obtain public acceptance. In this study, the following procedure is taken: (i) assumption of accident; (ii) establishment of safety assessment procedure; (iii) determination of source terms; (iv) diffusion calculation of radionuclide; (v) estimation of radiation exposure of the public

  8. The DYLAM approach to systems safety and reliability assessment

    International Nuclear Information System (INIS)

    Amendola, A.

    1988-01-01

    A survey of the principal features and applications of DYLAM (Dynamic Logical Analytical Methodology) is presented, whose basic principles can be summarized as follows: after a particular modelling of the component states, computerized heuristical procedures generate stochastic configurations of the system, whereas the resulting physical processes are simultaneously simulated to give account of the possible interactions between physics and states and, on the other hand, to search for system dangerous configurations and related probabilities. The association of probabilistic techniques for describing the states with physical equations for describing the process results in a very powerful tool for safety and reliability assessment of systems potentially subjected to dangerous incidental transients. A comprehensive picture of DYLAM capability for manifold applications can be obtained by the review of the study cases analyzed (LMFBR core accident, systems reliability assessment, accident simulation, man-machine interaction analysis, chemical reactors safety, etc.)

  9. Do whole-food animal feeding studies have any value in the safety assessment of GM crops?

    Science.gov (United States)

    Herman, Rod A; Ekmay, Ricardo

    2014-02-01

    The use of whole-food (grain meal contained in feed) animal-feeding studies to support the safety assessment of genetically modified crops has been contentious. This may be, in part, a consequence of poorly agreed upon study objectives. Whole-food animal-feeding studies have been postulated to be useful in detecting both expected and unexpected effects on the composition of genetically modified crops. While the justification of animal feeding studies to detect unexpected effects may be inadequately supported, there may be better justification to conduct such studies in specific cases to investigate the consequences of expected compositional effects including expression of transgenic proteins. Such studies may be justified when (1) safety cannot reasonably be predicted from other evidence, (2) reasonable hypothesis for adverse effects are postulated, (3) the compositional component in question cannot be isolated or enriched in an active form for inclusion in animal feeding studies, and (4) reasonable multiples of exposure can be accomplished relative to human diets. The study design for whole-food animal-feeding studies should be hypotheses-driven, and the types of data collected should be consistent with adverse effects that are known to occur from dietary components of biological origin. Copyright © 2013 The Authors. Published by Elsevier Inc. All rights reserved.

  10. 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

  11. Safety assessment of starch-based personal care products: Nanocapsules and pickering emulsions.

    Science.gov (United States)

    Marto, J; Pinto, P; Fitas, M; Gonçalves, L M; Almeida, A J; Ribeiro, H M

    2018-03-01

    The safety profile of the ingredients used in topical dosage forms and its evaluation is an issue of utmost importance. A suitable equilibrium between safety and efficacy is crucial before promoting a dermatological product. The aim of this work was to assess the safety and biological effects of starch-based vehicles (St-BV) used in such products. The hazard, exposure and dose-response assessment were used to characterize the risk of each ingredient. The EpiSkin™ assay and human repeat insult patch tests were performed to compare the theoretical safety assessment to in vitro and in vivo data. The efficacy of the St-BV was studied using biophysical measurements in human volunteers during 28 days, showing that all ingredients and their combinations were safe for the consumer. Tissue viability determined using the EpiSkin™ testing reached values between 84.0 ± 5.0% and 98.0 ± 8.6% after application of St-BV, which were considered as non-irritant to the skin. These observations were confirmed by the in vivo studies where the St-BV did not induce any sensitization on the volunteers, being safe for human use. Moreover, St-BV increased skin hydration and microcirculation, emerging as an attractive alternative to chemical raw materials. Copyright © 2018 Elsevier Inc. All rights reserved.

  12. Data report for the safety assessment SR-Site

    International Nuclear Information System (INIS)

    2010-12-01

    This report compiles, documents, and qualifies input data identified as essential for the long-term safety assessment of a KBS-3 repository, and forms an important part of the reporting of the safety assessment project SR-Site. The input data concern the repository system, broadly defined as the deposited spent nuclear fuel, the engineered barriers surrounding it, the host rock, and the biosphere in the proximity of the repository. The input data also concern external influences acting on the system, in terms of climate related data. Data are provided for a selection of relevant conditions and are qualified through traceable standardised procedures

  13. Data report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    2010-12-15

    This report compiles, documents, and qualifies input data identified as essential for the long-term safety assessment of a KBS-3 repository, and forms an important part of the reporting of the safety assessment project SR-Site. The input data concern the repository system, broadly defined as the deposited spent nuclear fuel, the engineered barriers surrounding it, the host rock, and the biosphere in the proximity of the repository. The input data also concern external influences acting on the system, in terms of climate related data. Data are provided for a selection of relevant conditions and are qualified through traceable standardised procedures

  14. Psychometric model for safety culture assessment in nuclear research facilities

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, C.S. do, E-mail: claudio.souza@ctmsp.mar.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), Av. Professor Lineu Prestes 2468, 05508-000 São Paulo, SP (Brazil); Andrade, D.A., E-mail: delvonei@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN – SP), Av. Professor Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil); Mesquita, R.N. de, E-mail: rnavarro@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN – SP), Av. Professor Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil)

    2017-04-01

    Highlights: • A psychometric model to evaluate ‘safety climate’ at nuclear research facilities. • The model presented evidences of good psychometric qualities. • The model was applied to nuclear research facilities in Brazil. • Some ‘safety culture’ weaknesses were detected in the assessed organization. • A potential tool to develop safety management programs in nuclear facilities. - Abstract: A safe and reliable operation of nuclear power plants depends not only on technical performance, but also on the people and on the organization. Organizational factors have been recognized as the main causal mechanisms of accidents by research organizations through USA, Europe and Japan. Deficiencies related with these factors reveal weaknesses in the organization’s safety culture. A significant number of instruments to assess the safety culture based on psychometric models that evaluate safety climate through questionnaires, and which are based on reliability and validity evidences, have been published in health and ‘safety at work’ areas. However, there are few safety culture assessment instruments with these characteristics (reliability and validity) available on nuclear literature. Therefore, this work proposes an instrument to evaluate, with valid and reliable measures, the safety climate of nuclear research facilities. The instrument was developed based on methodological principles applied to research modeling and its psychometric properties were evaluated by a reliability analysis and validation of content, face and construct. The instrument was applied to an important nuclear research organization in Brazil. This organization comprises 4 research reactors and many nuclear laboratories. The survey results made possible a demographic characterization and the identification of some possible safety culture weaknesses and pointing out potential areas to be improved in the assessed organization. Good evidence of reliability with Cronbach's alpha

  15. Psychometric model for safety culture assessment in nuclear research facilities

    International Nuclear Information System (INIS)

    Nascimento, C.S. do; Andrade, D.A.; Mesquita, R.N. de

    2017-01-01

    Highlights: • A psychometric model to evaluate ‘safety climate’ at nuclear research facilities. • The model presented evidences of good psychometric qualities. • The model was applied to nuclear research facilities in Brazil. • Some ‘safety culture’ weaknesses were detected in the assessed organization. • A potential tool to develop safety management programs in nuclear facilities. - Abstract: A safe and reliable operation of nuclear power plants depends not only on technical performance, but also on the people and on the organization. Organizational factors have been recognized as the main causal mechanisms of accidents by research organizations through USA, Europe and Japan. Deficiencies related with these factors reveal weaknesses in the organization’s safety culture. A significant number of instruments to assess the safety culture based on psychometric models that evaluate safety climate through questionnaires, and which are based on reliability and validity evidences, have been published in health and ‘safety at work’ areas. However, there are few safety culture assessment instruments with these characteristics (reliability and validity) available on nuclear literature. Therefore, this work proposes an instrument to evaluate, with valid and reliable measures, the safety climate of nuclear research facilities. The instrument was developed based on methodological principles applied to research modeling and its psychometric properties were evaluated by a reliability analysis and validation of content, face and construct. The instrument was applied to an important nuclear research organization in Brazil. This organization comprises 4 research reactors and many nuclear laboratories. The survey results made possible a demographic characterization and the identification of some possible safety culture weaknesses and pointing out potential areas to be improved in the assessed organization. Good evidence of reliability with Cronbach's alpha

  16. 78 FR 14912 - International Aviation Safety Assessment (IASA) Program Change

    Science.gov (United States)

    2013-03-08

    ... Aviation Safety Assessment (IASA) Program Change AGENCY: Federal Aviation Administration (FAA), DOT. ACTION..., into the U.S., or codeshare with a U.S. air carrier, complies with international aviation safety... subject to that country's aviation safety oversight can serve the United States using its own aircraft or...

  17. Fuel and canister process report for the safety assessment SR-Site

    International Nuclear Information System (INIS)

    Werme, Lars; Lilja, Christina

    2010-12-01

    This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

  18. Fuel and canister process report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    Werme, Lars; Lilja, Christina (eds.)

    2010-12-15

    This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Site. The detailed assessment methodology, including the role of the process reports in the assessment, is described in the SR-Site Main report /SKB 2011/

  19. Probabilistic safety assessment in radioactive waste disposal

    International Nuclear Information System (INIS)

    Robinson, P.C.

    1987-07-01

    Probabilistic safety assessment codes are now widely used in radioactive waste disposal assessments. This report gives an overview of the current state of the field. The relationship between the codes and the regulations covering radioactive waste disposal is discussed and the characteristics of current codes is described. The problems of verification and validation are considered. (author)

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

    Directory of Open Access Journals (Sweden)

    Ernest S. Fung

    2018-06-01

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

  1. Edible safety requirements and assessment standards for agricultural genetically modified organisms.

    Science.gov (United States)

    Deng, Pingjian; Zhou, Xiangyang; Zhou, Peng; Du, Zhong; Hou, Hongli; Yang, Dongyan; Tan, Jianjun; Wu, Xiaojin; Zhang, Jinzhou; Yang, Yongcun; Liu, Jin; Liu, Guihua; Li, Yonghong; Liu, Jianjun; Yu, Lei; Fang, Shisong; Yang, Xiaoke

    2008-05-01

    This paper describes the background, principles, concepts and methods of framing the technical regulation for edible safety requirement and assessment of agricultural genetically modified organisms (agri-GMOs) for Shenzhen Special Economic Zone in the People's Republic of China. It provides a set of systematic criteria for edible safety requirements and the assessment process for agri-GMOs. First, focusing on the degree of risk and impact of different agri-GMOs, we developed hazard grades for toxicity, allergenicity, anti-nutrition effects, and unintended effects and standards for the impact type of genetic manipulation. Second, for assessing edible safety, we developed indexes and standards for different hazard grades of recipient organisms, for the influence of types of genetic manipulation and hazard grades of agri-GMOs. To evaluate the applicability of these criteria and their congruency with other safety assessment systems for GMOs applied by related organizations all over the world, we selected some agri-GMOs (soybean, maize, potato, capsicum and yeast) as cases to put through our new assessment system, and compared our results with the previous assessments. It turned out that the result of each of the cases was congruent with the original assessment.

  2. Safety assessment of emergency power systems for nuclear power plants

    International Nuclear Information System (INIS)

    1992-01-01

    This publication is intended to assist the safety assessor within a regulatory body, or one working as a consultant, in assessing the safety of a given design of the emergency power systems (EPS) for a nuclear power plant. The present publication refers closely to the NUSS Safety Guide 50-SG-D7 (Rev. 1), Emergency Power Systems at Nuclear Power Plants. It covers therefore exactly the same technical subject as that Safety Guide. In view of its objective, however, it attempts to help in the evaluation of possible technical solutions which are intended to fulfill the safety requirements. Section 2 clarifies the scope further by giving an outline of the assessment steps in the licensing process. After a general outline of the assessment process in relation to the licensing of a nuclear power plant, the publication is divided into two parts. First, all safety issues are presented in the form of questions that have to be answered in order for the assessor to be confident of a safe design. The second part presents the same topics in tabulated form, listing the required documentation which the assessor has to consult and those international and national technical standards pertinent to the topics. An extensive reference list provides information on standards. 1 tab

  3. Use of operational experience in fire safety assessment of nuclear power plants

    International Nuclear Information System (INIS)

    2000-01-01

    Fire hazard has been identified as a major contributor to a plant's operational risk and the international nuclear power industry has been studying and developing tools for defending against this hazard. Considerable progress in design and regulatory requirements for fire safety, in fire protection technology and in related analytical techniques has been made in the past two decades. Substantial efforts have been undertaken worldwide to implement these advances in the interest of improving fire safety both at new and existing nuclear power plants. To assist in these efforts, the IAEA initiated a programme on fire safety that was intended to provide assistance to Member States in improving fire safety in nuclear power plants. In order to achieve this general objective, the IAEA programme aimed at the development of guidelines and good practices, the promotion of advanced fire safety assessment techniques, the exchange of state of the art information between practitioners and the provision of engineering safety advisory services and training in the implementation of internationally accepted practices. During the period 1993-1994, the IAEA activities related to fire safety concentrated on the development of guidelines and good practice documents related to fire safety and fire protection of operating plants. One of the first tasks was the development of a Safety Guide that formulates specific requirements with regard to the fire safety of operating nuclear power plants. Several documents, which provide advice on fire safety inspection, were developed to assist in its implementation. In the period 1995-1996, the programme focused on the preparation of guidelines for the systematic analysis of fire safety at nuclear power plants (NPPs). The IAEA programme on fire safety for 1997-1998 includes tasks aimed at promoting systematic assessment of fire safety related occurrences and dissemination of essential insights from this assessment. One of the topics addressed is the

  4. Reactor Safety Assessment System--A situation assessment aid for USNRC emergency response

    International Nuclear Information System (INIS)

    Bray, M.A.; Sebo, D.E.; Dixon, B.W.

    1985-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is intended for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system uses plant parameter data and status information from the power plant. It has a rule base that uses the parametric values, the known operator actions, and the time sequence information in the data to generate situation assessment conclusions for use by the NRC Reactor Safety Team. RSAS rules currently cover one specific reactor type and use setpoints specific to one power plant

  5. Reactor Safety Assessment System: a situation assessment aid for USNRC emergency response

    International Nuclear Information System (INIS)

    Bray, M.A.; Sebo, D.E.; Dixon, B.W.

    1985-04-01

    The Reactor Safety Assessment System is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is intended for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system uses plant parameter data and status information from the power plant. It has a rule base which uses the parametric values, the known operator actions and the time sequence information in the data to generate situation assessment conclusions for use by the NRC Reactor Safety Team. RSAS rules currently cover one specific reactor type and use setpoints specific to one power plant. 5 figs

  6. [Safety assessment of foods derived from genetically modified plants].

    Science.gov (United States)

    Pöting, A; Schauzu, M

    2010-06-01

    The placing of genetically modified plants and derived food on the market falls under Regulation (EC) No. 1829/2003. According to this regulation, applicants need to perform a safety assessment according to the Guidance Document of the Scientific Panel on Genetically Modified Organisms of the European Food Safety Authority (EFSA), which is based on internationally agreed recommendations. This article gives an overview of the underlying legislation as well as the strategy and scientific criteria for the safety assessment, which should generally be based on the concept of substantial equivalence and carried out in relation to an unmodified conventional counterpart. Besides the intended genetic modification, potential unintended changes also have to be assessed with regard to potential adverse effects for the consumer. All genetically modified plants and derived food products, which have been evaluated by EFSA so far, were considered to be as safe as products derived from the respective conventional plants.

  7. The Fukushima Daiichi Accident. Technical Volume 2/5. Safety Assessment

    International Nuclear Information System (INIS)

    2015-08-01

    Technical Volume 1 of this report has described what happened during the accident at the Fukushima Daiichi nuclear power plant (NPP). This volume begins (Section 2.1) with a review of how the design basis of the site for external events was assessed initially and then reassessed over the life of the NPP. The section also describes the physical changes that were made to the units as a result. The remainder of the volume describes the treatment of beyond design basis events in the safety assessment of the site, the accident management provisions, the effectiveness of regulatory programmes, human and organizational factors and the safety culture, and the role of operating experience. Further background information is contained in three annexes included on the CD-ROM of this Technical Volume which describe analytical investigations of the accident along with information on topics such as system performance, defence in depth and severe accident phenomena. Section 2.2 provides an assessment of the systems that failed, resulting in a failure to maintain the fundamental safety functions in Units 1–3, which were in operation at the time of the tsunami and in which the reactor pressure vessels (RPV) and containment vessels failed. The section also describes Units 4-6, which were shut down at the time of the tsunami, and the site’s central spent fuel storage facility. Section 2.3 discusses the probabilistic and deterministic safety assessments of beyond design basis accidents (BDBAs) that had been performed for the plant and the insights from these assessments that had led to changes in the plant’s design. The section pays particular attention to the assessment of extreme natural hazards, such as the one which led to the total loss of AC power supply on the site. The additional loss of DC power supply in Units 1 and 2 played a key role in the progression of the accident because it impeded the diagnosis of plant conditions and made the operators unaware of the status of

  8. Safety assessment of Vitis vinifera (grape)-derived ingredients as used in cosmetics.

    Science.gov (United States)

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

    2014-01-01

    The Cosmetic Ingredient Review Expert Panel (Panel) assessed the safety of 24 Vitis vinifera (grape)-derived ingredients and found them safe in the present practices of use and concentration in cosmetics. These ingredients function in cosmetics mostly as skin-conditioning agents, but some function as antioxidants, flavoring agents, and/or colorants. The Panel reviewed the available animal and clinical data to determine the safety of these ingredients. Additionally, some constituents of grapes have been assessed previously for safety as cosmetic ingredients by the Panel, and others are compounds that have been discussed in previous Panel safety assessments. © The Author(s) 2014.

  9. Self-assessment of safety culture in nuclear installations. Highlights and good practices

    International Nuclear Information System (INIS)

    2002-11-01

    This report summarizes the findings of two IAEA Technical Committee Meetings on Safety Culture Self-Assessment Highlights and Good Practices. The meetings took place on 3-5 June 1998 and 23-25 October 2000 in Vienna, and involved an international cross-section of representatives who participated both in plenary discussions and working groups. The purpose of the meetings was to discuss the practical implications of evolutionary changes in the development of safety culture, and to share international experience, particularly on the methods used for the assessment of safety culture and good practices for its enhancement in an organization. The working groups were allocated specific topics for discussion, which included the following: organizational factors influencing the implementation of actions to improve safety culture; how to measure, effectively, progress in implementing solutions to safety culture problems; the symptoms of a weakening safety culture; the suitability of different methods for assessing safety culture; the achievement of sustainable improvements in safety culture using the results of assessment; the potential threats to the continuation of a strong safety culture in an organization from the many challenges facing the nuclear industry. The working groups, when appropriate, considered issues from both the utility's and the regulator's perspectives. This report will be of interest to all organizations who wish to assess and achieve a strong and sustainable safety culture. This includes not only nuclear power plants, but also other sectors of the nuclear industry such as uranium mines and mills, nuclear fuel fabrication facilities, nuclear waste repositories, research reactors, accelerators, radiography facilities, etc. The report specifically supplements other IAEA publications on this subject

  10. Method for Pedestrian Crossing Risk Assessment and Safety Level Determination: the Case Study of Tallinn

    Energy Technology Data Exchange (ETDEWEB)

    Pashkevich, M.; Krasilnikova, A.; Antov, D.

    2016-07-01

    Pedestrians are a part of vulnerable road users which safety requires a special attention. Official statistics in Estonia from the last decade returns the following numbers: around 30 % of all road traffic accidents in the country were accidents with pedestrians, 32 % of all traffic fatalities were finished with pedestrian death. Pedestrian crossing has the biggest risk level between all kinds of pedestrian facilities, because it includes a direct conflict point between vehicle and pedestrian traffics. The article presents a method to assess risk of pedestrian crossing users and to determine safety level of this road infrastructure element. This approach is based on observation and collection of infrastructural as well as traffic data, which includes: (1) information about each pedestrian crossing facility, its location and state, (2) data about accidents with pedestrians and their features, (3) data from road traffic measurements. The main advantages of the described method are universality and comprehensiveness. The case study was done in Kristiine district of the city Tallinn, which was chosen as the most typical average district of Estonian capital. Results of this study are also presented in the article. (Author)

  11. Report on probabilistic safety assessment (PSA) quality assurance in utilization of risk information

    International Nuclear Information System (INIS)

    2006-12-01

    Recently in Japan, introduction of nuclear safety regulations using risk information such as probabilistic safety assessment (PSA) has been considered and utilization of risk information in the rational and practical measures on safety assurance has made a progress to start with the operation or inspection area. The report compiled results of investigation and studies of PSA quality assurance in risk-informed activities in the USA. Relevant regulatory guide and standard review plan as well as issues and recommendations were reviewed for technical adequacy and advancement of probabilistic risk assessment technology in risk-informed decision making. Useful and important information to be referred as issues in PSA quality assurance was identified. (T. Tanaka)

  12. South African safety assessment framework for the pebble bed modular reactor - HTR2008-58192

    International Nuclear Information System (INIS)

    Joubert, J.; Kohtz, N.; Coe, I.

    2008-01-01

    It is planned to construct a first of a kind Pebble Bed Modular Reactor (PBMR) in South Africa. A need has been recognized to accompany the licensing process for the PBMR with independent safety assessments to ensure that the safety case submitted by the applicant complies with the licensing requirements of the NNR. At the HTR 2006 Conference, the framework and major challenges on safety assessment that the South African National Nuclear Regulator (NNR) faces in developing and applying appropriate strategies and tools were presented. This paper discusses the current status of the various NNR assessment activities and describes how this will be considered in the NNR Final Report on the PBMR Safety Case. The traditional safety assessment process has been adapted to take into account the developmental nature of the project. By performing safety assessments, the designer and applicant must ensure that the design as proposed for construction and as-built meets the safety requirements defined by the regulatory framework. The regulator performs independent safety assessments, including independent analyses in areas deemed safety significant and potentially safety significant. The developmental nature of the project also led to the identification of a series of regulatory assessment activities preceding the formal assessment of the safety case. Besides an assessment of the resolution of Key Licensing Issues which have been defined in an early stage of the project and are discussed in /l/, these activities comprise the participation in an SAR Early Intervention Process, the execution of a regulatory HAZOP and the development of a regulatory assessment specification for the formal assessment of the safety case. This paper briefly describes these activities and their current status. During the last two years, significant progress was made with the development or adjustment of tools for the independent analysis by the regulator of the steady state core design, of the transient

  13. Recent Trends In The Methods Of Safety Assessment Of Rad Waste Treatment And Disposal

    International Nuclear Information System (INIS)

    Mahmoud, N.S.

    2012-01-01

    Radioactive waste management system involves a huge variety of processes and activities. This includes; collection and segregation, pretreatment, treatment, conditioning, storage and finally disposal. To assure the safety of the different facility of each step in the waste management system, the operator should prepare a safety analysis report to be assessed by the national regulatory body. The content of the safety analysis report must include all data about the site, facility design, operational phase, waste materials, and safety assessment methodologies. Safety assessment methodologies are iterative processes involving site-specific, prospective modeling evaluations of the pre-operational, operational, and post-closure time in case of disposal facilities. The safety assessment focuses primarily on a decision about compliance with performance objectives, rather than the much more difficult problem of predicting actual radiological impacts on the public at far future times. The recent organization processes of the safety assessment are improved by the ISAM working group from IAEA for waste disposal site. These safety assessment methodologies have been modified within SADRWMS IAEA project for the establishment of safety methodologies for the pre-disposal facilities (treatment and storage facilities) and the disposal site.

  14. Safety assessment of genetically modified foods

    NARCIS (Netherlands)

    Kleter, G.A.; Noordam, M.Y.

    2016-01-01

    The cultivation of genetically modified (GM) crops has steadily increased since their introduction to the market in the mid-1990s. Before these crops can be grown and sold they have to obtain regulatory approval in many countries, the process of which includes a pre-market safety assessment. The

  15. Qualified Presumption of Safety (QPS) is a generic risk assessment approach applied by the European Food Safety Authority (EFSA)

    DEFF Research Database (Denmark)

    Leuschner, R. G. K.; Robinson, T. P.; Hugas, M.

    2010-01-01

    Qualified Presumption of Safety (QPS) is a generic risk assessment approach applied by the European Food Safety Authority (EFSA) to notified biological agents aiming at simplifying risk assessments across different scientific Panels and Units. The aim of this review is to outline the implementation...... and value of the QPS assessment for EFSA and to explain its principles such as the unambiguous identity of a taxonomic unit, the body of knowledge including potential safety concerns and how these considerations lead to a list of biological agents recommended for QPS which EFSA keeps updated through...

  16. Safety assessment of discharge chute isolation barrier preparation and installation

    International Nuclear Information System (INIS)

    Meichle, R.H.

    1994-01-01

    This analysis examines activities associated with the installation of isolation barriers in the K Basins at the Hanford Reservation. This revision adds evaluation of barrier drops on stored fuel and basin floor, identifies fuel which will be moved and addresses criticality issues with sludge. The safety assessment is made for the activities for the preparation and installation of the discharge chute isolation barriers. The safety assessment includes a hazard assessment and comparisons of potential accidents/events to those addressed by the current safety basis documentation. No significant hazards were identified. An evaluation against the USQ evaluation questions was made and the determination made that the activities do not represent a USQ. Hazard categorization techniques were used to provide a basis for readiness review classifications

  17. 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.

  18. Experience in the implementation of quality assurance program and safety culture assessment of research reactor operation and maintenance

    International Nuclear Information System (INIS)

    Syarip; Suryopratomo, K.

    2001-01-01

    The implementation of quality assurance program and safety culture for research reactor operation are of importance to assure its safety status. It comprises an assessment of the quality of both technical and organizational aspects involved in safety. The method for the assessment is based on judging the quality of fulfillment of a number of essential issues for safety i.e. through audit, interview and/or discussions with personnel and management in plant. However, special consideration should be given to the data processing regarding the fuzzy nature of the data i.e. in answering the questionnaire. To accommodate this situation, the SCAP, a computer program based on fuzzy logic for assessing plant safety status, has been developed. As a case study, the experience in the assessment of Kartini research reactor safety status shows that it is strongly related to the implementation of quality assurance program in reactor operation and awareness of reactor operation staffs to safety culture practice. It is also shown that the application of the fuzzy rule in assessing reactor safety status gives a more realistic result than the traditional approach. (author)

  19. Independent Safety Assessment of the TOPAZ-II space nuclear reactor power system (Revised)

    International Nuclear Information System (INIS)

    1993-09-01

    The Independent Safety Assessment described in this study report was performed to assess the safety of the design and launch plans anticipated by the U.S. Department of Defense (DOD) in 1993 for a Russian-built, U.S.-modified, TOPAZ-II space nuclear reactor power system. Its conclusions, and the bases for them, were intended to provide guidance for the U.S. Department of Energy (DOE) management in the event that the DOD requested authorization under section 91b. of the Atomic Energy Act of 1954, as amended, for possession and use (including ground testing and launch) of a nuclear-fueled, modified TOPAZ-II. The scientists and engineers who were engaged to perform this assessment are nationally-known nuclear safety experts in various disciplines. They met with participants in the TOPAZ-II program during the spring and summer of 1993 and produced a report based on their analysis of the proposed TOPAZ-II mission. Their conclusions were confined to the potential impact on public safety and did not include budgetary, reliability, or risk-benefit analyses

  20. Evaluation of safety assessment methodologies in Rocky Flats Risk Assessment Guide (1985) and Building 707 Final Safety Analysis Report (1987)

    International Nuclear Information System (INIS)

    Walsh, B.; Fisher, C.; Zigler, G.; Clark, R.A.

    1990-01-01

    FSARs. Rockwell International, as operating contractor at the Rocky Flats plant, conducted a safety analysis program during the 1980s. That effort resulted in Final Safety Analysis Reports (FSARs) for several buildings, one of them being the Building 707 Final Safety Analysis Report, June 87 (707FSAR) and a Plant Safety Analysis Report. Rocky Flats Risk Assessment Guide, March 1985 (RFRAG85) documents the methodologies that were used for those FSARs. Resources available for preparation of those Rocky Flats FSARs were very limited. After addressing the more pressing safety issues, some of which are described below, the present contractor (EG ampersand G) intends to conduct a program of upgrading the FSARs. This report presents the results of a review of the methodologies described in RFRAG85 and 707FSAR and contains suggestions that might be incorporated into the methodology for the FSAR upgrade effort

  1. Validity of instruments to assess students' travel and pedestrian safety

    Directory of Open Access Journals (Sweden)

    Baranowski Tom

    2010-05-01

    Full Text Available Abstract Background Safe Routes to School (SRTS programs are designed to make walking and bicycling to school safe and accessible for children. Despite their growing popularity, few validated measures exist for assessing important outcomes such as type of student transport or pedestrian safety behaviors. This research validated the SRTS school travel survey and a pedestrian safety behavior checklist. Methods Fourth grade students completed a brief written survey on how they got to school that day with set responses. Test-retest reliability was obtained 3-4 hours apart. Convergent validity of the SRTS travel survey was assessed by comparison to parents' report. For the measure of pedestrian safety behavior, 10 research assistants observed 29 students at a school intersection for completion of 8 selected pedestrian safety behaviors. Reliability was determined in two ways: correlations between the research assistants' ratings to that of the Principal Investigator (PI and intraclass correlations (ICC across research assistant ratings. Results The SRTS travel survey had high test-retest reliability (κ = 0.97, n = 96, p Conclusions These validated instruments can be used to assess SRTS programs. The pedestrian safety behavior checklist may benefit from further formative work.

  2. Safety assessment of automated vehicle functions by simulation-based fault injection

    OpenAIRE

    Juez, Garazi; Amparan, Estibaliz; Lattarulo, Ray; Rastelli, Joshue Perez; Ruiz, Alejandra; Espinoza, Huascar

    2017-01-01

    As automated driving vehicles become more sophisticated and pervasive, it is increasingly important to assure its safety even in the presence of faults. This paper presents a simulation-based fault injection approach (Sabotage) aimed at assessing the safety of automated vehicle functions. In particular, we focus on a case study to forecast fault effects during the model-based design of a lateral control function. The goal is to determine the acceptable fault detection interval for pe...

  3. Assessment of safety of the nuclear installations of the world

    International Nuclear Information System (INIS)

    Thomas, B.A.; Pozniakov, N.; Banga, U.

    1992-01-01

    Incidents and accidents periodically remind us that preventive measures at nuclear installations are not fully reliable. Although sound design is widely recognized to be prerequisite for safe operation, it is not sufficient. An active management that compensates for the weak aspects of the installations design by redundant operational provisions, is the key factor to ensure safe operation. Safety of nuclear installations cannot be assessed on an emotional basis. Since 1986, accurate safety assessment techniques based on an integrated approach to operational safety have been made available by the ASSET services and are applicable to any industrial process dealing with nuclear materials. The ASSET methodology enables to eliminate in advance the Root Causes of the future accidents by introducing practical safety culture principles in the current managerial practices

  4. Safety assessment driving radioactive waste management solutions (SADRWMS Methodology) implemented in a software tool (SAFRAN)

    Energy Technology Data Exchange (ETDEWEB)

    Kinker, M., E-mail: M.Kinker@iaea.org [International Atomic Energy Agency (IAEA), Vienna (Austria); Avila, R.; Hofman, D., E-mail: rodolfo@facilia.se [FACILIA AB, Stockholm (Sweden); Jova Sed, L., E-mail: jovaluis@gmail.com [Centro Nacional de Seguridad Nuclear (CNSN), La Habana (Cuba); Ledroit, F., E-mail: frederic.ledroit@irsn.fr [IRSN PSN-EXP/SSRD/BTE, (France)

    2013-07-01

    In 2004, the International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts which could be used to improve the mechanisms for applying safety assessment methodologies to predisposal management of radioactive waste. These flowcharts have since been incorporated into DS284 (General Safety Guide on the Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste), and were also considered during the early development stages of the Safety Assessment Framework (SAFRAN) Tool. In 2009 the IAEA presented DS284 to the IAEA Waste Safety Standards Committee, during which it was proposed that the graded approach to safety case and safety assessment be illustrated through the development of Safety Reports for representative predisposal radioactive waste management facilities and activities. To oversee the development of these reports, it was agreed to establish the International Project on Complementary Safety Reports: Development and Application to Waste Management Facilities (CRAFT). The goal of the CRAFT project is to develop complementary reports by 2014, which the IAEA could then publish as IAEA Safety Reports. The present work describes how the DS284 methodology and SAFRAN Tool can be applied in the development and review of the safety case and safety assessment to a range of predisposal waste management facilities or activities within the Region. (author)

  5. Safety assessment driving radioactive waste management solutions (SADRWMS Methodology) implemented in a software tool (SAFRAN)

    International Nuclear Information System (INIS)

    Kinker, M.; Avila, R.; Hofman, D.; Jova Sed, L.; Ledroit, F.

    2013-01-01

    In 2004, the International Atomic Energy Agency (IAEA) organized the International Project on Safety Assessment Driving Radioactive Waste Management Solutions (SADRWMS) to examine international approaches to safety assessment for predisposal management of radioactive waste. The initial outcome of the SADRWMS Project was achieved through the development of flowcharts which could be used to improve the mechanisms for applying safety assessment methodologies to predisposal management of radioactive waste. These flowcharts have since been incorporated into DS284 (General Safety Guide on the Safety Case and Safety Assessment for Predisposal Management of Radioactive Waste), and were also considered during the early development stages of the Safety Assessment Framework (SAFRAN) Tool. In 2009 the IAEA presented DS284 to the IAEA Waste Safety Standards Committee, during which it was proposed that the graded approach to safety case and safety assessment be illustrated through the development of Safety Reports for representative predisposal radioactive waste management facilities and activities. To oversee the development of these reports, it was agreed to establish the International Project on Complementary Safety Reports: Development and Application to Waste Management Facilities (CRAFT). The goal of the CRAFT project is to develop complementary reports by 2014, which the IAEA could then publish as IAEA Safety Reports. The present work describes how the DS284 methodology and SAFRAN Tool can be applied in the development and review of the safety case and safety assessment to a range of predisposal waste management facilities or activities within the Region. (author)

  6. Safety assessment of genetically modified crops

    International Nuclear Information System (INIS)

    Atherton, Keith T.

    2002-01-01

    The development of genetically modified (GM) crops has prompted widespread debate regarding both human safety and environmental issues. Food crops produced by modern biotechnology using recombinant techniques usually differ from their conventional counterparts only in respect of one or a few desirable genes, as opposed to the use of traditional breeding methods which mix thousands of genes and require considerable efforts to select acceptable and robust hybrid offspring. The difficulties of applying traditional toxicological testing and risk assessment procedures to whole foods are discussed along with the evaluation strategies that are used for these new food products to ensure the safety of these products for the consumer

  7. Assessing progress in the development of safety culture

    International Nuclear Information System (INIS)

    Rotaru, I.; Ghita, S.; Biro, L.

    2002-01-01

    This paper is focussed on the organizational culture and learning processes required for the implementation of all aspects of safety culture. There is no prescriptive formula for improving safety culture. However, some common characteristics and practices are emerging that can be adopted by organizations in order to make progress. The paper refers to some approaches that have been successful in a number of countries. The experience of the international nuclear industry in the development and improvement of safety culture could be extended and found useful in other nuclear activities, irrespective of scale. The examples given of specific practice cover a wide range of activities including analysis of events, the regulatory approach on safety culture, employee participation and safety performance measures. Many of these practices may be relevant to smaller organizations and could contribute to improving safety culture, whatever the size of the organization. The most effective approach is to pursue a range of practices that can be mutually supportive in the development of a progressive safety culture, supported by professional standards, organizational and management commitment. Some guidance is also given on the assessment of safety culture and on the detection of a weakening safety culture. Few suggestions for accelerating the safety culture development and improvement process are also provided. (author)

  8. Crane Safety Assessment Method Based on Entropy and Cumulative Prospect Theory

    Directory of Open Access Journals (Sweden)

    Aihua Li

    2017-01-01

    Full Text Available Assessing the safety status of cranes is an important problem. To overcome the inaccuracies and misjudgments in such assessments, this work describes a safety assessment method for cranes that combines entropy and cumulative prospect theory. Firstly, the proposed method transforms the set of evaluation indices into an evaluation vector. Secondly, a decision matrix is then constructed from the evaluation vectors and evaluation standards, and an entropy-based technique is applied to calculate the index weights. Thirdly, positive and negative prospect value matrices are established from reference points based on the positive and negative ideal solutions. Thus, this enables the crane safety grade to be determined according to the ranked comprehensive prospect values. Finally, the safety status of four general overhead traveling crane samples is evaluated to verify the rationality and feasibility of the proposed method. The results demonstrate that the method described in this paper can precisely and reasonably reflect the safety status of a crane.

  9. An approach for risk informed safety culture assessment for Canadian nuclear power stations

    International Nuclear Information System (INIS)

    Nelson, W.R.

    2010-01-01

    One of the most important components of effective safety and risk management for nuclear power stations is a healthy safety culture. DNV has developed an approach for risk informed safety culture assessment that combines two complementary paradigms for safety and risk management: loss prevention - for preventing and intervening in accidents; and critical function management - for achieving safety and performance goals. Combining these two paradigms makes it possible to provide more robust systems for safety management and to support a healthy safety culture. This approach is being applied to safety culture assessment in partnership with a Canadian nuclear utility. (author)

  10. Safety assessment of existing highway bridges and viaducts

    NARCIS (Netherlands)

    Maljaars, J.; Steenbergen, R.; Abspoel, L.; Kolstein, M.H.

    2012-01-01

    The assessment of the structural safety of existing br idges and viaducts becomes increasingly important in many countries owing to an increase in traffic loads. Most existing standards, however, are developed for the design of new structures. For this reason, an assessment method for determining

  11. Food Safety Management in a Global Environment: The Role of Risk Assessment Models

    OpenAIRE

    Fuentes-Pila, Joaquin; Jimeno, Vicente; Manzano, Amparo; Rodriguez Monroy, Carlos; Mar Fernandez, Maria Del

    2006-01-01

    Quantitative risk assessment models are playing a minor role in the development of the new EU legal framework for food safety. There is a tendency of the EU institutions to apply the precautionary principle versus the predisposition of the USA institutions to rely on risk analysis. This paper provides a comparison of the role played by quantitative risk assessment models in the development of new policies on food safety in the EU and in the USA, focusing on a study case: the supply chain of s...

  12. Overview of the ISAM safety assessment methodology

    International Nuclear Information System (INIS)

    Simeonov, G.

    2003-01-01

    The ISAM safety assessment methodology consists of the following key components: specification of the assessment context description of the disposal system development and justification of scenarios formulation and implementation of models running of computer codes and analysis and presentation of results. Common issues run through two or more of these assessment components, including: use of methodological and computer tools, collation and use of data, need to address various sources of uncertainty, building of confidence in the individual components, as well as the overall assessment. The importance of the iterative nature of the assessment should be recognised

  13. Analysis of third-party certification approaches using an occupational health and safety conformity-assessment model.

    Science.gov (United States)

    Redinger, C F; Levine, S P

    1998-11-01

    The occupational health and safety conformity-assessment model presented in this article was developed (1) to analyze 22 public and private programs to determine the extent to which these programs use third parties in conformity-assessment determinations, and (2) to establish a framework to guide future policy developments related to the use of third parties in occupational health and safety conformity-assessment activities. The units of analysis for this study included select Occupational Safety and Health Administration programs and standards, International Organization for Standardization-based standards and guidelines, and standards and guidelines developed by nongovernmental bodies. The model is based on a 15-cell matrix that categorizes first-, second-, and third-party activities in terms of assessment, accreditation, and accreditation-recognition activities. The third-party component of the model has three categories: industrial hygiene/safety testing and sampling; product, equipment, and laboratory certification; and, occupational health and safety management system registration/certification. Using the model, 16 of the 22 programs were found to have a third-party component in their conformity-assessment structure. The analysis revealed that (1) the model provides a useful means to describe and analyze various third-party approaches, (2) the model needs modification to capture aspects of traditional governmental conformity-assessment/enforcement activities, and (3) several existing third-party conformity-assessment systems offer robust models that can guide future third-party policy formulation and implementation activities.

  14. Building quality into performance and safety assessment software

    International Nuclear Information System (INIS)

    Wojciechowski, L.C.

    2011-01-01

    Quality assurance is integrated throughout the development lifecycle for performance and safety assessment software. The software used in the performance and safety assessment of a Canadian deep geological repository (DGR) follows the CSA quality assurance standard CSA-N286.7 [1], Quality Assurance of Analytical, Scientific and Design Computer Programs for Nuclear Power Plants. Quality assurance activities in this standard include tasks such as verification and inspection; however, much more is involved in producing a quality software computer program. The types of errors found with different verification methods are described. The integrated quality process ensures that defects are found and corrected as early as possible. (author)

  15. Safety assessment input for site selection - the Swedish example - 59031

    International Nuclear Information System (INIS)

    Andersson, Johan

    2012-01-01

    Svensk Kaernbraenslehantering AB (SKB) has performed comprehensive investigations of two candidate sites for a final repository for Sweden's spent nuclear fuel. In March 2011 SKB decided to submit licence applications for a final repository at Forsmark. Before selection, SKB stated that the site that offers the best prospects for achieving long-term safety in practice would be selected. Based on experiences previous safety assessments, a number of issues related to long-term safety need to be considered in the context of site comparison. The factors include sensitivity to climate change such as periods of permafrost and glaciations, rock mechanics evolution including the potential for thermally induced spalling and sensitivity to potential future earthquakes, current and future groundwater flow, evolution of groundwater composition and proximity to mineral resources. Each of these factors related to long-term safety for the two candidate sites is assessed in a comparative analysis of site characteristics. The assessment also considers differences in biosphere conditions and in the confidence of the site descriptions. The comparison is concluded by an assessment on how the identified differences would affect the estimated radiological risk from a repository located at either of the sites. The assessment concludes that there are a number of safety related site characteristics for which the analyses do not show any decisive differences in terms of implications on safety, between the sites Forsmark and Laxemar. However, the frequency of water conducting fractures at repository depth is much smaller at Forsmark than at Laxemar. This difference, in turn, affects the future stability of the current favourable groundwater composition, which combined with the much higher flows at Laxemar would, for the current repository design, lead to a breach in the safety functions for the buffer and the canister for many more deposition positions at Laxemar than at Forsmark. Thereby

  16. [Safety Walkround as a risk assessment tool: the first Italian experience].

    Science.gov (United States)

    Levati, A; Amato, S; Adrario, E; De Flaviis, C; Delia, C; Milesi, S; Petrini, F; Bevilacqua, L

    2009-01-01

    In 2007 the Study Group "Clinical Risk Management" of the Italian Society of Anaesthesia and Intensive Care Unit (SIAARTI) performed a multicentric study in Intensive Care Unit (ICU) to assess the feasibility and efficacy of the Safety WalkRound (SWR) as a tool for the risk assessment. As the environment and organization of ICU are more complex than anaesthesia ones, mainly due to the severity of patients, high number of involved healthcare givers and different kinds of procedures, the Study Group decided that a check list is not fit for ICU and , after a careful review of the literature, chose to test the Safety WalkRound. in four Italian General ICUs. The SWR was born in 2003 when Frankel plans a structured interview of 15 questions (about 50% open) to collect operators' opinion about rate and type of errors, near misses, communication, problems regarding the report of adverse events and suggestions to increase patient safety. Consequently SWR is a tool of risk assessment alternative to the Incident Reporting which is marked by a diffuse underreporting of operators. Although the SWR is a new tool not validated in Italian language neither published in Italy on PubMed journals , the Study Group has decided that it might be fit for the organization of Italian Healthcare System. A back translation of the validated model of Joint Commission was provided and the translated version has been lightly changed to be employed in hospitals with and without Incident Reporting . The questions have been changed or introduced on the basis of the organization vulnerabilities detected with observational techniques or Focus Group. The interview performed in Italy contains 16 questions classified into five groups: a) error, b) error prevention, c) communication, teamwork and leadership, d) error discussion and e) relationship with patients and their families. The answers collected have been analyzed to detect the vulnerabilities in the organizations and specify the improvements to

  17. Survey of bayesian belif nets for quantitative reliability assessment of safety critical software used in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Eom, H.S.; Sung, T.Y.; Jeong, H.S.; Park, J.H.; Kang, H.G.; Lee, K

    2001-03-01

    As part of the Probabilistic Safety Assessment of safety grade digital systems used in Nuclear Power plants research, measures and methodologies applicable to quantitative reliability assessment of safety critical software were surveyed. Among the techniques proposed in the literature we selected those which are in use widely and investigated their limitations in quantitative software reliability assessment. One promising methodology from the survey is Bayesian Belief Nets (BBN) which has a formalism and can combine various disparate evidences relevant to reliability into final decision under uncertainty. Thus we analyzed BBN and its application cases in digital systems assessment area and finally studied the possibility of its application to the quantitative reliability assessment of safety critical software.

  18. Survey of bayesian belif nets for quantitative reliability assessment of safety critical software used in nuclear power plants

    International Nuclear Information System (INIS)

    Eom, H. S.; Sung, T. Y.; Jeong, H. S.; Park, J. H.; Kang, H. G.; Lee, K.

    2001-03-01

    As part of the Probabilistic Safety Assessment of safety grade digital systems used in Nuclear Power plants research, measures and methodologies applicable to quantitative reliability assessment of safety critical software were surveyed. Among the techniques proposed in the literature we selected those which are in use widely and investigated their limitations in quantitative software reliability assessment. One promising methodology from the survey is Bayesian Belief Nets (BBN) which has a formalism and can combine various disparate evidences relevant to reliability into final decision under uncertainty. Thus we analyzed BBN and its application cases in digital systems assessment area and finally studied the possibility of its application to the quantitative reliability assessment of safety critical software

  19. A quantitative assessment of organizational factors affecting safety using a system dynamics model

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, J. K. [Systemix Company, Seoul (Korea, Republic of); Yoon, T. S. [Korea Electric Power Research Institute (Korea, Republic of)

    2003-07-01

    The purpose of this study is to develop a system dynamics model for the assessment of organizational and human factors in the nuclear power plant safety. Previous studies are classified into two major approaches. One is the engineering approach such as ergonomics and Probabilistic Safety Assessment (PSA). The other is socio-psychology one. Both have contributed to find organizational and human factors and increased nuclear safety However, since these approaches assume that the relationship among factors is independent they do not explain the interactions between factors or variables in NPP's. To overcome these restrictions, a system dynamics model, which can show causal relations between factors and quantify organizational and human factors, has been developed. Operating variables such as degree of leadership, adjustment of number of employee, and workload in each department, users can simulate various situations in nuclear power plants in the organization side. Through simulation, user can get an insight to improve safety in plants and to find managerial tools in the organization and human side.

  20. Initialization of Safety Assessment Process for the Croatian Radioactive Waste repository on Trgovska gora

    International Nuclear Information System (INIS)

    Lokner, V.; Levanat, I.; Subasic, D.

    2000-01-01

    An iterative process of safety assessment, presently focusing on the site-specific evaluation of the post-closure phase for the prospective LILW repository on Trgovska gora in Croatia, has recently been initiated. The primary aim of the first assessment iterations is to provide the experts involved, the regulators and the general public with a reasonable assurance that the applicable long term performance and safety objectives can be met. Another goal is to develop a sufficient understanding of the system behavior to support decisions about the site investigation, the facility design, the waste acceptance criteria and the closure conditions. In this initial phase, the safety assessment is structured in a manner following closely methodology of the ISAM. The International Programme for Improving Long Term Safety Assessment Methodologies for Near Surface Radioactive Waste Disposal Facilities the IAEA coordinated research program started in 1997. Results of the safety assessment first iteration will be organized and presented in the form of a preliminary safety analysis report (PSAR), expected to be completed in the second part of the year 2000. As the first report on the initiated safety assessment activities, the PSAR will describe the concept and aims of the assessment process. Particular emphasis will be placed on description of the key elements of a safety assessment approach by: a) defining the assessment context; b) providing description of the disposal system; c) developing and justifying assessment scenarios; d) formulating and implementing models; and e) interpreting the scoping calculations. (author)

  1. Assessment of passive safety system of a Small Modular Reactor (SMR)

    International Nuclear Information System (INIS)

    Butt, Hassan Nawaz; Ilyas, Muhammad; Ahmad, Masroor; Aydogan, Fatih

    2016-01-01

    Highlights: • The MASLWR test facility has been modeled in RELAP5-SCDAP. The model is validated by comparing the simulation results with the experimental data. • Results obtained from various transients show that high pressure vent and sump recirculation lines provide natural circulation flow path for long term cooling of core. • New scenarios are considered in which the effect of vent and sump recirculation valves failure has been investigated. • It is found from the results that continuous loss of inventory occurs due to lack of recirculation. • It is concluded that the high pressure vent valves in the MASLWR safety system require more redundancy. - Abstract: Innovative SMRs are designed with enhanced safety features based on lessons learnt from past experience of plant operation. Reliance on natural circulation and addition of passive safety systems made them inherently safe and simple in design. It is required to study reliability assessment of passive safety systems during postulated transients prior to their deployment on commercial scale. Test facilities and best estimate system codes are playing significant role in assessment of passive safety systems as well as in design, certification and evaluation of these innovative types of reactors. RELAP5 code is widely used for thermal-hydraulic analysis of nuclear reactors. In this work, the passive safety systems of Multi-Application Small Light Water (MASLWR) have been assessed. The complete loop of the MASLWR test facility has been modeled in RELAP5-SCDAP Mod 4.0. The RELAP5 model is validated by comparing the simulation results with the experimental data. Results obtained for various transients show that high pressure vent and sump recirculation lines provide natural circulation flow path for long term cooling of core to avoid core heat up. Some of the components of passive safety system of MASLWR still rely on active power. Therefore, it was necessary to investigate their performance under failure

  2. The Nirex safety assessment research programme

    International Nuclear Information System (INIS)

    Cooper, M.J.

    1988-07-01

    This report describes progress on the Nirex Safety Assessment Research Programme in 1987/88. The programme is concerned with research into the disposal of low-level waste (LLW) and intermediate-level waste (ILW) into underground repositories. At the beginning of 1987/88 a range of techniques for measuring and modelling far-field phenomena were being applied to near-surface disposal of low-level waste in clay. However, during the year the far-field studies were redirected to consider generic geological materials of interest for deep disposal of low and intermediate-level waste, which is now the preferred option in the UK. A substantial part of the programme is concerned with the effectiveness of near-field barriers to water-borne leakage of radionuclides from cementitious repositories. Considerable progress has been made in quantifying this and laying the foundations for robust and reliable radiological assessments to be made with appropriate models. New projects have also been initiated to study the evolution and migration of gases from an underground repository and to consider the contribution of the biosphere to the retardation of radionuclides. (author)

  3. Quantification of human reliability in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Hirschberg, S.; Dankg, Vinh N.

    1996-01-01

    Human performance may substantially influence the reliability and safety of complex technical systems. For this reason, Human Reliability Analysis (HRA) constitutes an important part of Probabilistic Safety Assessment (PSAs) or Quantitative Risk Analyses (QRAs). The results of these studies as well as analyses of past accidents and incidents clearly demonstrate the importance of human interactions. The contribution of human errors to the core damage frequency (CDF), as estimated in the Swedish nuclear PSAs, are between 15 and 88%. A survey of the FRAs in the Swiss PSAs shows that also for the Swiss nuclear power plants the estimated HE contributions are substantial (49% of the CDF due to internal events in the case of Beznau and 70% in the case of Muehleberg; for the total CDF, including external events, 25% respectively 20%). Similar results can be extracted from the PSAs carried out for French, German, and US plants. In PSAs or QRAs, the adequate treatment of the human interactions with the system is a key to the understanding of accident sequences and their relative importance to overall risk. The main objectives of HRA are: first, to ensure that the key human interactions are systematically identified and incorporated into the safety analysis in a traceable manner, and second, to quantify the probabilities of their success and failure. Adopting a structured and systematic approach to the assessment of human performance makes it possible to provide greater confidence that the safety and availability of human-machine systems is not unduly jeopardized by human performance problems. Section 2 discusses the different types of human interactions analysed in PSAs. More generally, the section presents how HRA fits in the overall safety analysis, that is, how the human interactions to be quantified are identified. Section 3 addresses the methods for quantification. Section 4 concludes the paper by presenting some recommendations and pointing out the limitations of the

  4. Development and Application of Level 2 Probabilistic Safety Assessment for Nuclear Power Plants. Specific Safety Guide

    International Nuclear Information System (INIS)

    2010-01-01

    The objective of this Safety Guide is to provide recommendations for meeting the IAEA safety requirements in performing or managing a level 2 probabilistic safety assessment (PSA) project for a nuclear power plant; thus it complements the Safety Guide on level 1 PSA. One of the aims of this Safety Guide is to promote a standard framework, standard terms and a standard set of documents for level 2 PSAs to facilitate regulatory and external peer review of their results. It describes all elements of the level 2 PSA that need to be carried out if the starting point is a fully comprehensive level 1 PSA. Contents: 1. Introduction; 2. PSA project management and organization; 3. Identification of design aspects important to severe accidents and acquisition of information; 4. Interface with level 1 PSA: Grouping of sequences; 5. Accident progression and containment analysis; 6. Source terms for severe accidents; 7. Documentation of the analysis: Presentation and interpretation of results; 8. Use and applications of the PSA; Annex I: Example of a typical schedule for a level 2 PSA; Annex II: Computer codes for simulation of severe accidents; Annex III: Sample outline of documentation for a level 2 PSA study.

  5. Development of the KINS Safety Culture Maturity Model for Self and Independent Assessment

    International Nuclear Information System (INIS)

    Sheen, C.; Choi, Y.S.

    2016-01-01

    Safety culture of an organization is cultivated and affected not only by societal and regulatory environment of the organization, but by its philosophies, policies, events and activities experienced in the process of accomplishing its mission. The safety culture would be continuously changed by the interactions between its members along with time as an organic entity. In order to perform a systematic self- or independent assessment of safety culture, a safety culture assessment model (SCAM) properly reflecting cultural characteristics should be necessary. In addition, a SCAM should be helpful not only to establish correct directions, goals, and strategies for safety culture development, but should anticipating obstacles against safety culture development in the implementation process derived from the assessment. In practical terms, a SCAM should be useful for deriving effective guidelines and implementing of corrective action programs for the evaluated organization. Korea Institute of Nuclear Safety (KINS) performed a research project for six years to develop a SCAM satisfying the above prerequisites for self- and independent assessment. The KINS SCAM was developed based on the five stage safety culture maturity model proposed by Professor Patrick Hudson and was modified into four stages to reflect existing safety culture assessment experiences at Korean nuclear power plants. In order to define the change mechanism of safety culture for development and reversion, the change model proposed by Prochaska and DiClemente was introduced into KINS SCAM and developed into the Spiral Change Model.

  6. Assessing the Safety Attitudes Questionnaire (SAQ), German language version in Swiss university hospitals--a validation study.

    Science.gov (United States)

    Zimmermann, Natalie; Küng, Kaspar; Sereika, Susan M; Engberg, Sandra; Sexton, Bryan; Schwendimann, René

    2013-09-10

    Improving patient safety has become a major focus of clinical care and research over the past two decades. An institution's patient safety climate represents an essential component of ensuring a safe environment and thereby can be vital to the prevention of adverse events. Covering six patient safety related factors, the Safety Attitudes Questionnaire (SAQ) is a validated and widely used instrument to measure the patient safety climate in clinical areas. The objective of this study was to assess the psychometric properties of the German language version of the SAQ. A survey was carried out in two University Hospitals in Switzerland in autumn 2009 where the SAQ was distributed to a sample of 406 nurses and physicians in medical and surgical wards. Following the American Educational Research Association guidelines, we tested the questionnaire validity by levels of evidence: content validity, internal structure and relations to other variables. Confirmatory factor analysis was used to examine factor structure. Cronbach's alphas and inter-item correlations were calculated to examine internal consistency reliability. A total of 319 questionnaires were completed representing an overall response rate of 78.6%. For three items, the item content validity index was <0.75. Confirmatory factor analysis showed acceptable model fit (RMSEA = 0.045; CFI = 0.944) for the six-factor model. Additional exploratory factor analysis could not identify a better factor model. SAQ factor scores showed positive correlations with the Safety Organizing Scale (r = .56-.72). The SAQ German version showed moderate to strong internal consistency reliability indices (Cronbach alpha = .65-.83). The German language version of the SAQ demonstrated acceptable to good psychometric properties and therefore shows promise to be a sound instrument to measure patient safety climate in Swiss hospital wards. However, the low item content validity and large number of missing responses for several items suggest

  7. 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

  8. 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

  9. Integrated safety assessment report, Haddam Neck Plant (Docket No. 50-213): Integrated Safety Assessment Program: Draft report

    International Nuclear Information System (INIS)

    1987-07-01

    The integrated assessment is conducted on a plant-specific basis to evaluate all licensing actions, licensee initiated plant improvements and selected unresolved generic/safety issues to establish implementation schedules for each item. Procedures allow for a periodic updating of the schedules to account for licensing issues that arise in the future. The Haddam Neck Plant is one of two plants being reviewed under the pilot program. This report indicates how 82 topics selected for review were addressed, and presents the staff's recommendations regarding the corrective actions to resolve the 82 topics and other actions to enhance plant safety. 135 refs., 4 figs., 5 tabs

  10. United States program for the safety assessment of geologic disposal of commercial radioactive wastes

    International Nuclear Information System (INIS)

    Claiborne, H.C.

    1977-01-01

    The safe disposal of commercial radioactive wastes in deep geologic formations is the goal of the National Waste Terminal Storage (NWTS) Program. A comprehensive safety assessment program has been established which will proceed on a schedule consistent with the start-up of two waste repositories in late 1985. Safety assessment begins with selection of a disposal site; that is, all geologic and hydrologic factors must indicate long-term stability of the formation and prospective isolation of wastes from circulating around waters for hundreds of thousands of years. The long-term stability of each site must be demonstrated by sophisticated rock mechanics analyses. To help provide answers on the mechanism and consequences of an unlikely breach in the integrity of the repository, a Waste Isolation Safety Assessment Program (WISAP) is being sponsored at the Battelle Pacific Northwest Laboratories. Methods and data necessary to characterize the safety of generic geological waste disposal concepts, which are to be applied in the assessment of specific sties, will be developed. Other long-term safety-related studies that complement WISAP are in progress, for example, borehole plugging, salt dissolutioning, and salt transport in vertical boreholes. Requirements for licensing are in the process of being formulated by the NRC

  11. Radionuclide transport report for the safety assessment SR-Site

    Energy Technology Data Exchange (ETDEWEB)

    2010-12-15

    This document compiles radionuclide transport calculations of a KBS-3 repository for the safety assessment SR-Site. The SR-Site assessment supports the licence application for a final repository at Forsmark, Sweden

  12. IAEA Issues Report on Mission to Review Japan's Nuclear Power Plant Safety Assessment Process

    International Nuclear Information System (INIS)

    2012-01-01

    Full text: A team of international nuclear safety experts has delivered its report on a mission it conducted from 21-31 January 2012 to review Japan's process for assessing nuclear safety at the nation's nuclear power plants. International Atomic Energy Agency (IAEA) officials delivered the IAEA Mission Report to Japanese officials yesterday and made it publicly available today. Following the 11 March 2011 accident at TEPCO's Fukushima Daiichi Nuclear Power Station, Japan's Nuclear and Industrial Safety Agency (NISA) announced the development of a revised safety assessment process for the nation's nuclear power reactors. At the request of the Government of Japan, the IAEA organized a team of five IAEA and three international nuclear safety experts and visited Japan to review NISA's approach to the Comprehensive Assessments for the Safety of Existing Power Reactor Facilities and how NISA examines the results submitted by nuclear operators. A Preliminary Summary Report was issued on 31 January. 'The mission report provides additional information regarding the team's recommendations and overall finding that NISA's instructions to power plants and its review process for the Comprehensive Safety Assessments are generally consistent with IAEA Safety Standards', said team leader James Lyons, Director of the IAEA's Nuclear Installation Safety Division. National safety assessments and their peer review by the IAEA are a key component of the IAEA Action Plan on Nuclear Safety, which was approved by the Agency's Member States following last year's nuclear accident at Fukushima Daiichi Nuclear Power Station. The IAEA safety review mission held meetings in Tokyo with officials from NISA, the Japanese Nuclear Energy Safety Organization (JNES), and the Kansai Electric Power Company (KEPCO), and the team visited the Ohi Nuclear Power Station to see an example of how Japan's Comprehensive Safety Assessment is being implemented by nuclear operators. In its report delivered today

  13. Safety assessment of geologic repositories for nuclear waste

    International Nuclear Information System (INIS)

    Bartlett, J.W.; Burkholder, H.C.; Winegardner, W.K.

    1977-01-01

    Consideration of geologic isolation for final disposition of radioactive wastes has led to the need for evaluation of the safety of the concept. Such evaluations require consideration of factors not encountered in conventional risk analysis: consequences at times and places far removed from the repository site; indirect, complex, and alternative pathways between the waste and the point of potential consequences; a highly limited data base; and limited opportunity for experimental verification of results. R and D programs to provide technical safety evaluations are under way. Three methods are being considered for the probabilistic aspects of the evaluations: fault tree analysis, repository simulation analysis, and system stability analysis. Nuclide transport models, currently in a relatively advanced state of development, are used to evaluate consequences of postulated loss of geologic isolation. This paper outlines the safety assessment methods, unique features of the assessment problem that affect selection of methods and reliability of results, and available results. It also discusses potential directions for future work

  14. Safety Assessment in the AREVA Group: Operating Experience from a Self-Assessment Tool

    International Nuclear Information System (INIS)

    Coye de Brunélis, T.; Mignot, E.; Sidaner, J.-F.

    2016-01-01

    The expression “safety culture” first appeared following analysis of the Chernobyl accident in 1986. It was first defined in INSAG-4 (International Nuclear Safety Advisory Group safety series) in 1991. Other events have occurred in nuclear facilities and during transportation since Chernobyl: Tokai Mura in 1999, Roissy Transport in 2002, Davis Besse in 2002, Thorp in 2005. These events show that the initial approach was too simplistic. Based on this observation, the definition of safety culture was supplemented by including concepts of cultural value (associated with the country and the company) and human and organizational factors, and was integrated in that form with the emergence and implementation of integrated management systems (IMS). Today, the concept of nuclear safety culture covers a wide set of factors such as safety, quality, corporate culture, defined processes and policies, organizations and related resources. Any assessment of people’s safety culture, particularly people directly involved in facility operations, is thus part of a comprehensive policy and contributes to a de facto demonstration of the priority which management assigns to safety.

  15. Environment, Safety, and Health Self-Assessment Report, Fiscal Year 2008

    Energy Technology Data Exchange (ETDEWEB)

    Chernowski, John

    2009-02-27

    Lawrence Berkeley National Laboratory's Environment, Safety, and Health (ES&H) Self-Assessment Program ensures that Integrated Safety Management (ISM) is implemented institutionally and by all divisions. The Self-Assessment Program, managed by the Office of Contract Assurance (OCA), provides for an internal evaluation of all ES&H programs and systems at LBNL. The functions of the program are to ensure that work is conducted safely, and with minimal negative impact to workers, the public, and the environment. The Self-Assessment Program is also the mechanism used to institute continuous improvements to the Laboratory's ES&H programs. The program is described in LBNL/PUB 5344, Environment, Safety, and Health Self-Assessment Program and is composed of four distinct assessments: the Division Self-Assessment, the Management of Environment, Safety, and Health (MESH) review, ES&H Technical Assurance, and the Appendix B Self-Assessment. The Division Self-Assessment uses the five core functions and seven guiding principles of ISM as the basis of evaluation. Metrics are created to measure performance in fulfilling ISM core functions and guiding principles, as well as promoting compliance with applicable regulations. The five core functions of ISM are as follows: (1) Define the Scope of Work; (2) Identify and Analyze Hazards; (3) Control the Hazards; (4) Perform the Work; and (5) Feedback and Improvement. The seven guiding principles of ISM are as follows: (1) Line Management Responsibility for ES&H; (2) Clear Roles and Responsibilities; (3) Competence Commensurate with Responsibilities; (4) Balanced Priorities; (5) Identification of ES&H Standards and Requirements; (6) Hazard Controls Tailored to the Work Performed; and (7) Operations Authorization. Performance indicators are developed by consensus with OCA, representatives from each division, and Environment, Health, and Safety (EH&S) Division program managers. Line management of each division performs the

  16. Overview of fourth generation reactors. Assessment in terms of safety and radiation protection

    International Nuclear Information System (INIS)

    Couturier, J.; Baudrand, O.; Blanc, D.; Bourgois, T.; Hache, G.; Ivanov, E.; Bonneville, H.; Meignen, R.; Nicaise, G.; Bruna, G.; Clement, B.; Kissane, M.; Monhardt, B.

    2012-01-01

    Based on a systematic analysis of the different concepts of fourth generation nuclear reactors, this report gives an overview of specific aspects regarding safety and radiation protection for six concepts: sodium fast reactors (SFR), gas fast reactors (GFR), lead fast reactors (LFR), molten salt reactors (MSR), very high or high temperature reactors (V/HTR) and supercritical water reactors (SCWR). This assessment is based on different studies and researches performed by the IRSN at an international level. For each reactor concept, the report proposes a presentation of the current status of development and its perspectives, describes the safety aspects which are specific to this concept, identifies and discusses elements for safety analysis, and assesses the concept with respect to the Fukushima accident and IAEA recommendations and predefined themes

  17. HERMES probabilistic risk assessment. Pilot study

    International Nuclear Information System (INIS)

    Parisot, F.; Munoz, J.

    1993-01-01

    The study was performed in 1989 of the contribution of probabilistic analysis for the optimal construction of system safety status in aeronautical and European nuclear industries, shows the growing trends towards incorporation of quantitative safety assessment and lead to an agreement to undertake a prototype proof study on Hermes. The main steps of the study and results are presented in the paper

  18. Safety assessment of envisaged systems for automotive hydrogen supply and utilization

    Energy Technology Data Exchange (ETDEWEB)

    Landucci, Gabriele [Dipartimento di Ingegneria Chimica, Chimica Industriale e Scienza dei Materiali, Universita di Pisa, via Diotisalvi n.2, 56126 Pisa (Italy); Tugnoli, Alessandro; Cozzani, Valerio [Dipartimento di Ingegneria Chimica, Mineraria e delle Tecnologie Ambientali, Alma Mater Studiorum - Universita di Bologna, via Terracini n.28, 40131 Bologna (Italy)

    2010-02-15

    A novel consequence-based approach was applied to the inherent safety assessment of the envisaged hydrogen production, distribution and utilization systems, in the perspective of the widespread hydrogen utilization as a vehicle fuel. Alternative scenarios were assessed for the hydrogen system chain from large scale production to final utilization. Hydrogen transportation and delivery was included in the analysis. The inherent safety fingerprint of each system was quantified by a set of Key Performance Indicators (KPIs). Rules for KPIs aggregation were considered for the overall assessment of the system chains. The final utilization stage resulted by large the more important for the overall expected safety performance of the system. Thus, comparison was carried out with technologies proposed for the use of other low emission fuels, as LPG and natural gas. The hazards of compressed hydrogen-fueled vehicles resulted comparable, while reference innovative hydrogen technologies evidenced a potentially higher safety performance. Thus, switching to the inherently safer technologies currently under development may play an important role in the safety enhancement of hydrogen vehicles, resulting in a relevant improvement of the overall safety performance of the entire hydrogen system. (author)

  19. A study on optimization of the nuclear safety system

    International Nuclear Information System (INIS)

    Lee, Sang Hoon; Koh, Byung Joon; Kim, Jin Soo; Kim, Byoung Do; Cho, Seong Won; Kwon, Seog Kwon; Choi, Kwang Sik

    1986-12-01

    The number of nuclear facilities (nuclear power plants, research reactors, nuclear fuel facilities) under construction or in operation in Korea continues to increase and this has brought about increased importance and concerns toward nuclear safety in Korea. Also, domestic nuclear related organizations are increasingly carrying out the design/construction of nuclear power plants and the development /supply of nuclear fuels. In order to flexibly respond to these changes and to suggest direction to take, it is necessary to re-examine the current nuclear safety regulation system. This study is carried out in two stages and this report describes the results of the analysis and the assessment of the nuclear licencing system of such foreign countries as sweden and German, as the first of the two. In this regard, this study includes the analysis on the backgrounds on the choice of nuclear licensing system, the analysis on the licensing procedures, the analysis on the safety inspection system and the enforcement laws, the analysis on the structure and function of the regulatory, business and research organizations as well as the analysis on the relationship between the safety research and the regulatory duties. In this study, the German safety inspection system and the enforcement procedures and the Swedish nuclear licensing system are analyzed in detail. By comparing and assessing the finding with the current Korea Nuclear Licensing System, this study points out some reform measures of the Korean system that needs to improved. With the changing situations in mind, this study aims to develop the nuclear safety regulation system optimized for Korean situation by re-examining the current regulation system. (Author)

  20. Deep repository for long-lived low- and intermediate-level waste. Preliminary safety assessment

    International Nuclear Information System (INIS)

    1999-11-01

    A preliminary safety assessment has been performed of a deep repository for long-lived low- and intermediate-level waste, SFL 3-5. The purpose of the study is to investigate the capacity of the facility to act as a barrier to the release of radionuclides and toxic pollutants, and to shed light on the importance of the location of the repository site. A safety assessment (SR 97) of a deep repository for spent fuel has been carried out at the same time. In SR 97, three hypothetical repository sites have been selected for study. These sites exhibit fairly different conditions in terms of hydrogeology, hydrochemistry and ecosystems. To make use of information and data from the SR 97 study, we have assumed that SFL 3-5 is co-sited with the deep repository for spent fuel. A conceivable alternative is to site SFL 3-5 as a completely separate repository. The focus of the SFL 3-5 study is a quantitative analysis of the environmental impact for a reference scenario, while other scenarios are discussed and analyzed in more general terms. Migration in the repository's near- and far-field has been taken into account in the reference scenario. Environmental impact on the three sites has also been calculated. The calculations are based on an updated forecast of the waste to be disposed of in SFL 3-5. The forecast includes radionuclide content, toxic metals and other substances that have a bearing on a safety assessment. The safety assessment shows how important the site is for safety. Two factors stand out as being particularly important: the water flow at the depth in the rock where the repository is built, and the ecosystem in the areas on the ground surface where releases may take place in the future. Another conclusion is that radionuclides that are highly mobile and long-lived, such as 36 Cl and 93 Mo , are important to take into consideration. Their being long-lived means that barriers and the ecosystems must be regarded with a very long time horizon

  1. Safety cases for the co-ordinated research project on improvement of safety assessment methodologies for near surface radioactive waste disposal facilities (ISAM)

    International Nuclear Information System (INIS)

    Kozak, M.W.; Torres-Vidal, C.; Kelly, E.; Guskov, A.; Blerk, J. van

    2002-01-01

    A Co-ordinated Research Project (CRP) has recently been completed on the Improvement of Safety Assessment Methodologies for Near-Surface Radioactive Waste Disposal Facilities (ISAM). A major aspect of the project was the use of safety cases for the practical application of safety assessment. An overview of the ISAM safety cases is given in this paper. (author)

  2. Confidence building in safety assessment

    International Nuclear Information System (INIS)

    Osthols, E.

    1999-01-01

    Engineered disposal systems are necessary to isolate radioactive waste from humans and the environment. It is essential to have access to basic thermochemical data relevant to varying geological environments for the radioactive elements involved. The OECD/NEA Thermochemical Data Base project (TDB) aims to make widely available basic thermochemical data of the type needed for safety assessment of nuclear storage facilities. The history and the present status of the project are presented. (K.A.)

  3. Framework for continuous assessment and improvement of occupational health and safety issues in construction companies.

    Science.gov (United States)

    Mahmoudi, Shahram; Ghasemi, Fakhradin; Mohammadfam, Iraj; Soleimani, Esmaeil

    2014-09-01

    Construction industry is among the most hazardous industries, and needs a comprehensive and simple-to-administer tool to continuously assess and promote its health and safety performance. Through the study of various standard systems (mainly Health, Safety, and Environment Management System; Occupational Health and Safety Assessment Series 180001; and British Standard, occupational health and safety management systems-Guide 8800), seven main elements were determined for the desired framework, and then, by reviewing literature, factors affecting these main elements were determined. The relative importance of each element and its related factors was calculated at organizational and project levels. The provided framework was then implemented in three construction companies, and results were compared together. THE RESULTS OF THE STUDY SHOW THAT THE RELATIVE IMPORTANCE OF THE MAIN ELEMENTS AND THEIR RELATED FACTORS DIFFER BETWEEN ORGANIZATIONAL AND PROJECT LEVELS: leadership and commitment are the most important elements at the organization level, whereas risk assessment and management are most important at the project level. The present study demonstrated that the framework is easy to administer, and by interpreting the results, the main factors leading to the present condition of companies can be determined.

  4. Neuropsychological assessment of driving safety risk in older adults with and without neurologic disease.

    Science.gov (United States)

    Anderson, Steven W; Aksan, Nazan; Dawson, Jeffrey D; Uc, Ergun Y; Johnson, Amy M; Rizzo, Matthew

    2012-01-01

    Decline in cognitive abilities can be an important contributor to the driving problems encountered by older adults, and neuropsychological assessment may provide a practical approach to evaluating this aspect of driving safety risk. The purpose of the present study was to evaluate several commonly used neuropsychological tests in the assessment of driving safety risk in older adults with and without neurological disease. A further goal of this study was to identify brief combinations of neuropsychological tests that sample performances in key functional domains and thus could be used to efficiently assess driving safety risk. A total of 345 legally licensed and active drivers over the age of 50, with no neurologic disease (N = 185), probable Alzheimer's disease (N = 40), Parkinson's disease (N = 91), or stroke (N = 29), completed vision testing, a battery of 10 neuropsychological tests, and an 18-mile drive on urban and rural roads in an instrumented vehicle. Performances on all neuropsychological tests were significantly correlated with driving safety errors. Confirmatory factor analysis was used to identify 3 key cognitive domains assessed by the tests (speed of processing, visuospatial abilities, and memory), and several brief batteries consisting of one test from each domain showed moderate corrected correlations with driving performance. These findings are consistent with the notion that driving places demands on multiple cognitive abilities that can be affected by aging and age-related neurological disease, and that neuropsychological assessment may provide a practical off-road window into the functional status of these cognitive systems.

  5. Interim main report of the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Hedin, Allan [and others

    2004-08-01

    This document is an interim report on the safety assessment SR-Can (SR in the acronym stands for Safety Report and Can is short for canister). The final SR-Can report will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of the present interim report is to demonstrate the methodology for safety assessment so that it can be reviewed before it is used in a license application. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. Preliminary data from the Forsmark site, presently being investigated by SKB as one of the candidate for a KBS-3 repository are used to some extent as examples. However, the collected data are yet too sparse to allow an evaluation of safety for this site. An important aim of this report is to demonstrate the proper handling of requirements on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Authority are duplicated in an Appendix. The principal acceptance criterion requires that 'the annual risk of harmful effects after closure does not exceed 10{sup -6} for a representative individual in the group exposed to the greatest risk'. 'Harmful effects' refer to cancer and hereditary effects. Following the introductory chapter 1, this report outlines the methodology for the SR-Can assessment in chapter 2, and presents in chapters 3, 4 and 5 the initial state of the system and the plans and methods for handling external influences and internal processes, respectively. Function indicators are introduced in chapter 6 and a preliminary evaluation of these is given in chapter 7. The material presented in the first seven chapters is utilised in the scenario selection in chapter 8

  6. Interim main report of the safety assessment SR-Can

    International Nuclear Information System (INIS)

    Hedin, Allan

    2004-08-01

    This document is an interim report on the safety assessment SR-Can (SR in the acronym stands for Safety Report and Can is short for canister). The final SR-Can report will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of the present interim report is to demonstrate the methodology for safety assessment so that it can be reviewed before it is used in a license application. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. Preliminary data from the Forsmark site, presently being investigated by SKB as one of the candidate for a KBS-3 repository are used to some extent as examples. However, the collected data are yet too sparse to allow an evaluation of safety for this site. An important aim of this report is to demonstrate the proper handling of requirements on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Authority are duplicated in an Appendix. The principal acceptance criterion requires that 'the annual risk of harmful effects after closure does not exceed 10 -6 for a representative individual in the group exposed to the greatest risk'. 'Harmful effects' refer to cancer and hereditary effects. Following the introductory chapter 1, this report outlines the methodology for the SR-Can assessment in chapter 2, and presents in chapters 3, 4 and 5 the initial state of the system and the plans and methods for handling external influences and internal processes, respectively. Function indicators are introduced in chapter 6 and a preliminary evaluation of these is given in chapter 7. The material presented in the first seven chapters is utilised in the scenario selection in chapter 8. Hydrogeological

  7. Performance assessment studies for the long-term safety evaluation of radioactive waste disposal facility

    International Nuclear Information System (INIS)

    Bujoreanu, D.; Olteanu, M.; Bujoreanu, L.

    2008-01-01

    Especially during the last ten years, a part of Romanian research program 'Management of Radioactive Waste and Spent Fuel' was focused mainly on applicative research for the design of near-surface disposal facility, which intends to accommodate the low and intermediate radioactive waste generated from Cernavoda NPP. In this frame, our contribution was at the acquisition of technical data for the characterization of the future disposal facility. In the present, the project of the disposal facility, located on the Saligny site, near Cernavoda NPP, must be licensed. As regards to the safe disposal, the location of final disposal, the Saligny site, has been characterized through the five geological formations which contain potential routes for transport of radionuclide released from disposal facility, in the receiving zones(potential receiving zones), into liquid and gaseous phases. The technical characteristics of the disposal facility were adapted at the Romanian disposal concept using the reference data from IAEA technical report (IAEA,1999). Input parameters which characterized from physical and chemical point of view the disposal system, were partially taken from literature. The performance assessment studies, which follows the preliminary design development phases and the selection, describes how the source term is affected by the infiltration of water through the disposal facility, degradation process of engineering barriers (reflected in the distribution coefficient values) and solubility limit. The studies regard the evaluation of the source term, sensitivity and uncertainty analysis provide the information on 'how' and 'why' were evaluated, following: (i) radiological safety assessment of near-surface disposal facility on Saligny site; (ii) complexity standard assessment of the Engineering Barriers Systems (EBS); (iii) identification of the elements which must be elaborated for the increase of the disposal safety and the necessity for new technical data for

  8. NANoREG framework for the safety assessment of nanomaterials

    OpenAIRE

    Gottardo, Stefania; Alessandrelli, Maria; Amenta, Valeria; Atluri, Rambabu; Barberio, Grazia; Bekker, Cindy; Bergonzo, Philippe; Bleeker, Eric; Booth, Andy; Borges, Teresa; Buttol, Patrizia; Carlander, David; Castelli, Stefano; Chevillard, Sylvie

    2017-01-01

    The NANoREG framework addresses the need to ease the nanomaterials safety assessment in the REACH Regulation context. It offers forward-looking strategies: Safe-by-Design, a Nanospecific Prioritisation and Risk Assessment, and Life Cycle Assessment. It is intended for scientific experts, regulatory authorities and industry.

  9. COLD-SAT feasibility study safety analysis

    Science.gov (United States)

    Mchenry, Steven T.; Yost, James M.

    1991-01-01

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

  10. Work-related stress risk assessment in Italy: the validation study of health safety and executive indicator tool.

    Science.gov (United States)

    Rondinone, Bruna Maria; Persechino, Benedetta; Castaldi, Tiziana; Valenti, Antonio; Ferrante, Pierpaolo; Ronchetti, Matteo; Iavicoli, Sergio

    2012-01-01

    In compliance with the Italian occupational health and safety regulatory framework, as provided by the Lgs. Decree 81/2008, the "work-related stress" risk assessment should follow the same principles as other risk assessments, in accordance with the European Agreement of 8 October 2004; therefore, validated and scientifically proven methodological tools are needed to conduct an adequate work-related stress risk assessment. The UK's Health Safety and Executive (HSE) Indicator Tool (IT) is used for assessing the risk of work-related stress. The aim of this study is to test the factor structure of IT as a measure of work-related stress in a sample of Italian workers. Data collected from 65 Italian organizations (6378 workers) was used for a Confirmatory Factor Analysis (CFA) on the 35-item seven-factor model. The results showed acceptable fit to the data (CFI .90; TLI .89, RMSEA .045). A second CFA was done to test a 35-item six-factor model (CFI .89, TLI .87, RMSEA .047). Both models were tested after removing six items (factor loadings less than .50.), resulting in a 29-item model. Here again, there was an acceptable fit to the data (29-item seven-factor model: CFI .93, TLI .91, RMSEA .044; 29-item six-factor model: CFI .92, TLI .90, RMSEA .046). These findings show that the HSE model satisfactorily adapts to use in a sample of Italian workers. One of the most important innovations introduced in the assessment of work-related stress with the HSE IT is the global approach for identifying work-related stress risk factors, aimed at establishing the best strategy from the viewpoints of prevention officers and also of workers.

  11. Management of safety, safety culture and self assessment

    International Nuclear Information System (INIS)

    Carnino, A.

    2000-01-01

    Safety management is the term used for the measures required to ensure that an acceptable level of safety is maintained throughout the life of an installation, including decommissioning. The safety culture concept and its implementation are described in part one of the paper. The principles of safety are now quite well known and are implemented worldwide. It leads to a situation where harmonization is being achieved as indicated by the entry into force of the Convention on Nuclear Safety. To go beyond the present nuclear safety levels, management of safety and safety culture will be the means for achieving progress. Recent events which took place in major nuclear power countries have shown the importance of the management and the consequences on safety. At the same time, electricity deregulation is coming and will impact on safety through reductions in staffing and in operation and maintenance cost at nuclear installations. Management of safety as well as its control and monitoring by the safety authorities become a key to the future of nuclear energy.(author)

  12. Ex-ante assessment of the safety effects of intelligent transport systems.

    Science.gov (United States)

    Kulmala, Risto

    2010-07-01

    There is a need to develop a comprehensive framework for the safety assessment of Intelligent Transport Systems (ITS). This framework should: (1) cover all three dimensions of road safety-exposure, crash risk and consequence, (2) cover, in addition to the engineering effect, also the effects due to behavioural adaptation and (3) be compatible with the other aspects of state of the art road safety theories. A framework based on nine ITS safety mechanisms is proposed and discussed with regard to the requirements set to the framework. In order to illustrate the application of the framework in practice, the paper presents a method based on the framework and the results from applying that method for twelve intelligent vehicle systems in Europe. The framework is also compared to two recent frameworks applied in the safety assessment of intelligent vehicle safety systems. Copyright 2010 Elsevier Ltd. All rights reserved.

  13. Re-assessment of seismic loads in conjunction with periodic safety review

    International Nuclear Information System (INIS)

    Jonczyk, Josef

    2002-01-01

    The objective of this paper is the fundamental consideration of a safeguard-aim-oriented approach for use in the re-assessment of seismic events with regard to the periodic safety review (PSR) of nuclear power plants (NPP). The re-assessment aspects of site-specific design earthquakes (DEQ), specially the procedure for seismic hazard analysis, will not, however, be considered in detail here. The proposed assessment concept clearly presents a general approach for safety assessments. The approach is based on a successive screening review of components that are considered sufficiently earthquake-resistant. In this respect, the principle of maximum practical application of the design documentation has been considered in the re-assessment process. On the other hand, the safeguard-aim-oriented evaluation will also be applied with regard to whether the requirements of the safety regulations are fulfilled with respect to the safety goals. The review in conjunction with PSR does not, however, attempt to perform this under all technical aspects. Moreover, it is possible to make extensive use of experimental knowledge and engineering judgement with regard to the structural capacity behaviour in case of a seismic event. Compared with design procedures, however, this proposed approach differs from the one applied in licensing procedures, in which such assessment freedom will not usually be exhausted. (author)

  14. Use of cut-off values as meaningfulness limits in probabilistic studies and its effect on NPPs risk assessment and safety improvement

    International Nuclear Information System (INIS)

    Petrangeli, G.; Valeri, A.; Zaffiro, C.

    1991-01-01

    This paper discusses the use of cut-off values in probabilistic risk assessment/probabilistic safety assessment (PRA/PSA) of nuclear power plants (NPPs), in order to explore under which conditions this practice may help improve the meaningfulness of the results of the analyses and safety of plants, and how it may affect the assessment of risk. Reference is made, in particular, to some past practical applications, also taken from the experience of the authors within the frame of the Italian licensing process. The paper describes the Italian probabilistic criteria which use probabilistic targets and cut-off values to assess safety and identify plant safety improvements. The rationale of the approach is also discussed in the paper and results of sample applications are illustrated. The paper concludes that the use of cut-off values, if properly implemented, could be productive to improve the plant safety as it helps the analyst to focus on a restricted field of analysis, ignoring lower probability and less known events. It also points out that cut-off values should be considered as living numbers to be lowered and even eliminated as soon as significant advancements are made, through research and operational experience, in the knowledge of the pertinent events

  15. Overview of the reactor safety study consequence model

    International Nuclear Information System (INIS)

    Wall, I.B.; Yaniv, S.S.; Blond, R.M.; McGrath, P.E.; Church, H.W.; Wayland, J.R.

    1977-01-01

    The Reactor Safety Study (WASH-1400) is a comprehensive assessment of the potential risk to the public from accidents in light water power reactors. The engineering analysis of the plants is described in detail in the Reactor Safety Study: it provides an estimate of the probability versus magnitude of the release of radioactive material. The consequence model, which is the subject of this paper, describes the progression of the postulated accident after the release of the radioactive material from the containment. A brief discussion of the manner in which the consequence calculations are performed is presented. The emphasis in the description is on the models and data that differ significantly from those previously used for these types of assessments. The results of the risk calculations for 100 light water power reactors are summarized

  16. Application of REPAS Methodology to Assess the Reliability of Passive Safety Systems

    Directory of Open Access Journals (Sweden)

    Franco Pierro

    2009-01-01

    Full Text Available The paper deals with the presentation of the Reliability Evaluation of Passive Safety System (REPAS methodology developed by University of Pisa. The general objective of the REPAS is to characterize in an analytical way the performance of a passive system in order to increase the confidence toward its operation and to compare the performances of active and passive systems and the performances of different passive systems. The REPAS can be used in the design of the passive safety systems to assess their goodness and to optimize their costs. It may also provide numerical values that can be used in more complex safety assessment studies and it can be seen as a support to Probabilistic Safety Analysis studies. With regard to this, some examples in the application of the methodology are reported in the paper. A best-estimate thermal-hydraulic code, RELAP5, has been used to support the analyses and to model the selected systems. Probability distributions have been assigned to the uncertain input parameters through engineering judgment. Monte Carlo method has been used to propagate uncertainties and Wilks' formula has been taken into account to select sample size. Failure criterions are defined in terms of nonfulfillment of the defined design targets.

  17. Patient safety culture assessment in oman.

    Science.gov (United States)

    Al-Mandhari, Ahmed; Al-Zakwani, Ibrahim; Al-Kindi, Moosa; Tawilah, Jihane; Dorvlo, Atsu S S; Al-Adawi, Samir

    2014-07-01

    To illustrate the patient safety culture in Oman as gleaned via 12 indices of patient safety culture derived from the Hospital Survey on Patient Safety Culture (HSPSC) and to compare the average positive response rates in patient safety culture between Oman and the USA, Taiwan, and Lebanon. This was a cross-sectional research study employed to gauge the performance of HSPSC safety indices among health workers representing five secondary and tertiary care hospitals in the northern region of Oman. The participants (n=398) represented different professional designations of hospital staff. Analyses were performed using univariate statistics. The overall average positive response rate for the 12 patient safety culture dimensions of the HSPSC survey in Oman was 58%. The indices from HSPSC that were endorsed the highest included 'organizational learning and continuous improvement' while conversely, 'non-punitive response to errors' was ranked the least. There were no significant differences in average positive response rates between Oman and the United States (58% vs. 61%; p=0.666), Taiwan (58% vs. 64%; p=0.386), and Lebanon (58% vs. 61%; p=0.666). This study provides the first empirical study on patient safety culture in Oman which is similar to those rates reported elsewhere. It highlights the specific strengths and weaknesses which may stem from the specific milieu prevailing in Oman.

  18. Patient Safety Culture Assessment in Oman

    Science.gov (United States)

    Al-Mandhari, Ahmed; Al-Zakwani, Ibrahim; Al-Kindi, Moosa; Tawilah, Jihane; Dorvlo, Atsu S.S.; Al-Adawi, Samir

    2014-01-01

    Objective To illustrate the patient safety culture in Oman as gleaned via 12 indices of patient safety culture derived from the Hospital Survey on Patient Safety Culture (HSPSC) and to compare the average positive response rates in patient safety culture between Oman and the USA, Taiwan, and Lebanon. Methods This was a cross-sectional research study employed to gauge the performance of HSPSC safety indices among health workers representing five secondary and tertiary care hospitals in the northern region of Oman. The participants (n=398) represented different professional designations of hospital staff. Analyses were performed using univariate statistics. Results The overall average positive response rate for the 12 patient safety culture dimensions of the HSPSC survey in Oman was 58%. The indices from HSPSC that were endorsed the highest included ‘organizational learning and continuous improvement’ while conversely, ‘non-punitive response to errors’ was ranked the least. There were no significant differences in average positive response rates between Oman and the United States (58% vs. 61%; p=0.666), Taiwan (58% vs. 64%; p=0.386), and Lebanon (58% vs. 61%; p=0.666). Conclusion This study provides the first empirical study on patient safety culture in Oman which is similar to those rates reported elsewhere. It highlights the specific strengths and weaknesses which may stem from the specific milieu prevailing in Oman. PMID:25170407

  19. Innovative Modelling Approach of Safety Culture Assessment in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ahn, N.

    2016-01-01

    A culture is commonly defined as the shared set of norms and values that govern appropriate individual behavior. Safety culture is the subset of organizational culture that reflects the general attitude and approaches to safety and risk management. While safety is sometimes narrowly defined in terms of human death and injury, we use a more inclusive definition that also considers mission loss as a safety problem and is thus applicable to nuclear power plants and missions. The recent accident reports and investigations of the nuclear power plant mission failures (i.e., TMI, Chernobyl, and Fukushima) point to safety cultural problems in nuclear power plants. Many assessment approaches have been developed by organizations such as IAEA and INPO based on the assessment of parameters at separate levels — individuals, groups, and organizations.

  20. Self assessment of safety culture in HANARO using the code of conduct on the safety of research reactor by IAEA

    International Nuclear Information System (INIS)

    Lim, I.C.; Hwang, S.Y.; Woo, J.S.; Lee, M.; Jun, B.J.

    2003-01-01

    Full text: The safety culture in HANARO was self-assessed in accordance with the Code of Conduct on the Safety of Research Reactor drafted by IAEA. From 2002, IAEA has worked on the development of the Code of Conduct to achieve and maintain high level of nuclear safety in research reactors worldwide through the enhancement of national measures and international co-operation including, where appropriate, safety related technical cooperation. It defines the role of the state, the role of the regulatory body, the role of the operating organization and the role of the IAEA. As for the role of operating organization, the code specifies general requirements in assessment and verification of safety, financial and human resources, quality assurance, human factors, radiation protection and emergency preparedness. It also defines the role of operating organization for safety of research reactor in siting, design, operation, maintenance, modification and utilization as well. All of these items are the subjects for safety culture implementation, which means the Code could be a guideline for an operating organization to assess its safety culture. The self-assessment of safety culture in HANARO was made by using the sections of the Code describing the role of the operating organization for safety of research reactor. The major assessment items and the practices in HANARO for each items are as follow: The SAR of HANARO was reviewed by the regulatory body before the construction and the fuel loading of HANARO. Major design modifications and new installation of utilization facility needs the approval from regulatory body and safety assessment is a requirement for the approval. The Tech. Spec. for HANARO Operation specifies the analysis, surveillance, testing and inspection for HANARO operation. The reactor operation is mainly supported by the government and partly by nuclear R and D fund. The education and training of operation staff are one of major tasks of operating organization

  1. Assessment of the safety of foods derived from genetically modified (GM) crops.

    Science.gov (United States)

    König, A; Cockburn, A; Crevel, R W R; Debruyne, E; Grafstroem, R; Hammerling, U; Kimber, I; Knudsen, I; Kuiper, H A; Peijnenburg, A A C M; Penninks, A H; Poulsen, M; Schauzu, M; Wal, J M

    2004-07-01

    This paper provides guidance on how to assess the safety of foods derived from genetically modified crops (GM crops); it summarises conclusions and recommendations of Working Group 1 of the ENTRANSFOOD project. The paper provides an approach for adapting the test strategy to the characteristics of the modified crop and the introduced trait, and assessing potential unintended effects from the genetic modification. The proposed approach to safety assessment starts with the comparison of the new GM crop with a traditional counterpart that is generally accepted as safe based on a history of human food use (the concept of substantial equivalence). This case-focused approach ensures that foods derived from GM crops that have passed this extensive test-regime are as safe and nutritious as currently consumed plant-derived foods. The approach is suitable for current and future GM crops with more complex modifications. First, the paper reviews test methods developed for the risk assessment of chemicals, including food additives and pesticides, discussing which of these methods are suitable for the assessment of recombinant proteins and whole foods. Second, the paper presents a systematic approach to combine test methods for the safety assessment of foods derived from a specific GM crop. Third, the paper provides an overview on developments in this area that may prove of use in the safety assessment of GM crops, and recommendations for research priorities. It is concluded that the combination of existing test methods provides a sound test-regime to assess the safety of GM crops. Advances in our understanding of molecular biology, biochemistry, and nutrition may in future allow further improvement of test methods that will over time render the safety assessment of foods even more effective and informative. Copryright 2004 Elsevier Ltd.

  2. Assessment of elementary school safety restraint programs.

    Science.gov (United States)

    1985-06-01

    The purpose of this research was to identify elementary school (K-6) safety belt : education programs in use in the United States, to review their development, and : to make administrative and impact assessments of their use in selected States. : Six...

  3. Safety-licensing assessment of NASAP reactor concepts and fuel cycle facilities

    International Nuclear Information System (INIS)

    Lipinski, W.C.; Prohammer, F.G.; van Erp, J.B.; Seefeldt, W.B.

    1978-06-01

    Assessments are presented of the safety/licensability of reactor concepts based on information supplied by the Nonproliferation Alternative Systems Assessment Program (NASAP) characterization contractors in their updated responses to the data package for NASAP Rolling Report II. The assessment of the LMFBR includes information from a characterization contractor on alternate fuel cycles but does not include information provided by a characterization contractor on plant-related safety issues. The information provided by the characterization contractors was supplemented by assessments provided by the U. S. Nuclear Regulatory Commission

  4. Ultraviolet safety assessments of insect light traps.

    Science.gov (United States)

    Sliney, David H; Gilbert, David W; Lyon, Terry

    2016-01-01

    Near-ultraviolet (UV-A: 315-400 nm), "black-light," electric lamps were invented in 1935 and ultraviolet insect light traps (ILTs) were introduced for use in agriculture around that time. Today ILTs are used indoors in several industries and in food-service as well as in outdoor settings. With recent interest in photobiological lamp safety, safety standards are being developed to test for potentially hazardous ultraviolet emissions. A variety of UV "Black-light" ILTs were measured at a range of distances to assess potential exposures. Realistic time-weighted human exposures are shown to be well below current guidelines for human exposure to ultraviolet radiation. These UV-A exposures would be far less than the typical UV-A exposure in the outdoor environment. Proposals are made for realistic ultraviolet safety standards for ILT products.

  5. Activities on safety culture study. Study status in public and private sectors

    International Nuclear Information System (INIS)

    Makino, Maomi; Takano, Kenichi

    2004-01-01

    Around after entering in the 21st century, organizational accidents had occurred in Japan at various industries including nuclear industry, which were caused directly by unsafe action, human error and illegal conduct of personnel but there were problems in safety culture of organization such as slow retreat of safety system stimulated by management, schedule control and procedure management becoming a dead letter, lack of safety education, and workplace climate of schedule priority. This article referred to organizational factors common to many severe accidents and introduced safety culture study in public and private sectors to overcome those factors. Safety Culture Evaluation Support Tool (SCEST) was developed for self-evaluation of safety culture of organization as well as Organizational Reliability model (OR model) for analysis of initiation and propagation process of risk event. Safety diagnosis system and feedback type risk assessment system for promoting safe organizational climate and culture were also developed. (T. Tanaka)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  7. Biosphere models for safety assessment of radioactive waste disposal

    International Nuclear Information System (INIS)

    Proehl, G.; Olyslaegers, G.; Zeevaert, T.; Kanyar, B.; Bergstroem, U.; Hallberg, B.; Mobbs, S.; Chen, Q.; Kowe, R.

    2004-01-01

    The aim of the BioMoSA project has been to contribute in the confidence building of biosphere models, for application in performance assessments of radioactive waste disposal. The detailed objectives of this project are: development and test of practical biosphere models for application in long-term safety studies of radioactive waste disposal to different European locations, identification of features, events and processes that need to be modelled on a site-specific rather than on a generic base, comparison of the results and quantification of the variability of site-specific models developed according to the reference biosphere methodology, development of a generic biosphere tool for application in long term safety studies, comparison of results from site-specific models to those from generic one, Identification of possibilities and limitations for the application of the generic biosphere model. (orig.)

  8. Probabilistic safety assessment for food irradiation facility

    International Nuclear Information System (INIS)

    Solanki, R.B.; Prasad, M.; Sonawane, A.U.; Gupta, S.K.

    2012-01-01

    Highlights: ► Different considerations are required in PSA for Non-Reactor Nuclear Facilities. ► We carried out PSA for food irradiation facility as a part of safety evaluation. ► The results indicate that the fatal exposure risk is below the ‘acceptable risk’. ► Adequate operator training and observing good safety culture would reduce the risk. - Abstract: Probabilistic safety assessment (PSA) is widely used for safety evaluation of Nuclear Power Plants (NPPs) worldwide. The approaches and methodologies are matured and general consensus exists on using these approaches in PSA applications. However, PSA applications for safety evaluation for non-reactor facilities are limited. Due to differences in the processes in nuclear reactor facilities and non-reactor facilities, the considerations are different in application of PSA to these facilities. The food irradiation facilities utilize gamma irradiation sources, X-ray machines and electron accelerators for the purpose of radiation processing of variety of food items. This is categorized as Non-Reactor Nuclear Facility. In this paper, the application of PSA to safety evaluation of food irradiation facility is presented considering the ‘fatality due to radiation overexposure’ as a risk measure. The results indicate that the frequency of the fatal exposure is below the numerical acceptance guidance for the risk to the individual. Further, it is found that the overall risk to the over exposure can be reduced by providing the adequate operator training and observing good safety culture.

  9. Modeling and Analysis on Radiological Safety Assessment of Low- and Intermediate Level Radioactive Waste Repository

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youn Myoung; Jung, Jong Tae; Kang, Chul Hyung (and others)

    2008-04-15

    Modeling study and analysis for technical support for the safety and performance assessment of the low- and intermediate level (LILW) repository partially needed for radiological environmental impact reporting which is essential for the licenses for construction and operation of LILW has been fulfilled. Throughout this study such essential area for technical support for safety and performance assessment of the LILW repository and its licensing as gas generation and migration in and around the repository, risk analysis and environmental impact during transportation of LILW, biosphere modeling and assessment for the flux-to-dose conversion factors for human exposure as well as regional and global groundwater modeling and analysis has been carried out.

  10. Modeling and Analysis on Radiological Safety Assessment of Low- and Intermediate Level Radioactive Waste Repository

    International Nuclear Information System (INIS)

    Lee, Youn Myoung; Jung, Jong Tae; Kang, Chul Hyung

    2008-04-01

    Modeling study and analysis for technical support for the safety and performance assessment of the low- and intermediate level (LILW) repository partially needed for radiological environmental impact reporting which is essential for the licenses for construction and operation of LILW has been fulfilled. Throughout this study such essential area for technical support for safety and performance assessment of the LILW repository and its licensing as gas generation and migration in and around the repository, risk analysis and environmental impact during transportation of LILW, biosphere modeling and assessment for the flux-to-dose conversion factors for human exposure as well as regional and global groundwater modeling and analysis has been carried out

  11. Study on real working performance and overload safety factor of high arch dam

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    Considering the fact that high arch dams have problems such as complicated stress,high cost,and hazards after being damaged,this paper intends to study the effects of load,material strength,and safety analysis method on dam safety and working performance of arch dams.In this article,the effects of temperature,self weight exaction way and water loading on structure response are first discussed,and a more reasonable way of considering is then put forward.By taking into consideration the mechanical property of materials and comparing the effects of different yield criteria on overloading safety of high arch dams,this paper concludes that brittle characteristics of concrete should be fully considered when conducting safety assessment for high arch dams to avoid overestimating the bearing capacity of the dams.By comparing several typical projects,this paper works out a safety assessment system of multiple safety and relevant engineering analogical analysis methods,which is closer to the actual situation,and thus is able to assess the response of high arch dam structure in a more comprehensive way,elicit the safety coefficients in different situations,and provide a new way of considering the safety assessment of high arch dams.

  12. Assessment of Quality and Safety of Street-vended Fried West ...

    African Journals Online (AJOL)

    This study was carried out to assess the quality and safety of street-vended fried West African ilisha (Ilisha africana) from major markets in Ibadan, Oyo State, Nigeria. A total number of 100 fried fish samples obtained from ten major markets in Ibadan, Oyo State, Nigeria were analysed in the laboratory for: proximate ...

  13. The safety assessment of OPR-1000 nuclear power plant for station blackout accident applying the combined deterministic and probabilistic procedure

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dong Gu, E-mail: littlewing@kins.re.kr [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2014-08-15

    Highlights: • The combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of the BDBAs. • The safety assessment of OPR-1000 nuclear power plant for SBO accident is performed by applying the CDPP. • By estimating the offsite power restoration time appropriately, the SBO risk is reevaluated. • It is concluded that the CDPP is applicable to safety assessment of BDBAs without significant erosion of the safety margin. - Abstract: Station blackout (SBO) is a typical beyond design basis accident (BDBA) and significant contributor to overall plant risk. The risk analysis of SBO could be important basis of rulemaking, accident mitigation strategy, etc. Recently, studies on the integrated approach of deterministic and probabilistic method for nuclear safety in nuclear power plants have been done, and among them, the combined deterministic and probabilistic procedure (CDPP) was proposed for safety assessment of the BDBAs. In the CDPP, the conditional exceedance probability obtained by the best estimate plus uncertainty method acts as go-between deterministic and probabilistic safety assessments, resulting in more reliable values of core damage frequency and conditional core damage probability. In this study, the safety assessment of OPR-1000 nuclear power plant for SBO accident was performed by applying the CDPP. It was confirmed that the SBO risk should be reevaluated by eliminating excessive conservatism in existing probabilistic safety assessment to meet the targeted core damage frequency and conditional core damage probability. By estimating the offsite power restoration time appropriately, the SBO risk was reevaluated, and it was finally confirmed that current OPR-1000 system lies in the acceptable risk against the SBO. In addition, it is concluded that the CDPP is applicable to safety assessment of BDBAs in nuclear power plants without significant erosion of the safety margin.

  14. Assessment of efficacy and safety of praziquantel in the treatment of ...

    African Journals Online (AJOL)

    This study assessed the efficacy and safety of PZQ in school-aged children in four ... All pupils were also treated with a single dose of PZQ at 40 mg/kg. ... The presence of S. haematobium was neither age nor sex dependent (p>0.05).

  15. Cyber Security Risk Assessment for the KNICS Safety Systems

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  16. Relationship between Risk Assessment and Compliance to Health and Safety in Ugandan Secondary Schools

    Science.gov (United States)

    Sekiwu, Denis; Kabanda, Milly; Naluwemba, Esther Frances; Kaggwa, Victoria Tamale

    2015-01-01

    Health hazards are part and parcel of human life necessitating the provision of safety in every organizational environment (WHO regional Office for Africa, 2004). Likewise, the area of safety and accident prevention is of great concern to school improvement. The study sought to investigate the relationship between Risk Assessment and Compliancy to…

  17. Indicators for traffic safety assessment and prediction and their application in micro-simulation modelling : a study of urban and suburban intersections

    OpenAIRE

    Archer, Jeffery

    2005-01-01

    In order to achieve sustainable long-term transport infrastructure development, there is a growing need for fast, reliable and effective methods to evaluate and predict the impact of traffic safety measures. Recognising this need, and the need for an active traffic safety approach, this thesis focuses on traffic safety assessment and prediction based on the use of safety indicators that measure the spatial and/or temporal proximity of safety critical events. The main advantage of such measure...

  18. Assessment of Primary Production of Horticultural Safety Management Systems of Mushroom Farms in South Africa.

    Science.gov (United States)

    Dzingirayi, Garikayi; Korsten, Lise

    2016-07-01

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

  19. Assessing Risk-Based Performance Indicators in Safety-Critical Systems for Nuclear Power Plants

    OpenAIRE

    TONT Gabriela

    2011-01-01

    The paper proposes framework for a multidisciplinary nuclear risk and safety assessment by modeling uncertainty and combining diverse evidence provided in such a way that it could be used to represent an entire argument about a system's dependability. The identified safety issues are being treated by means of probabilistic safety assessment (PSA). The behavior simulation of power plant in thepresence of risk factors is analyzed from the vulnerability, risk and functional safety viewpoints, hi...

  20. 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

  1. Safety study application guide

    International Nuclear Information System (INIS)

    1993-07-01

    Martin Marietta Energy Systems, Inc., (Energy Systems) is committed to performing and documenting safety analyses for facilities it manages for the Department of Energy (DOE). Included are analyses of existing facilities done under the aegis of the Safety Analysis Report Upgrade Program, and analyses of new and modified facilities. A graded approach is used wherein the level of analysis and documentation for each facility is commensurate with the magnitude of the hazard(s), the complexity of the facility and the stage of the facility life cycle. Safety analysis reports (SARs) for hazard Category 1 and 2 facilities are usually detailed and extensive because these categories are associated with public health and safety risk. SARs for Category 3 are normally much less extensive because the risk to public health and safety is slight. At Energy Systems, safety studies are the name given to SARs for Category 3 (formerly open-quotes lowclose quotes) facilities. Safety studies are the appropriate instrument when on-site risks are limited to irreversible consequences to a few people, and off-site consequences are limited to reversible consequences to a few people. This application guide provides detailed instructions for performing safety studies that meet the requirements of DOE Orders 5480.22, open-quotes Technical Safety Requirements,close quotes and 5480.23, open-quotes Nuclear Safety Analysis Reports.close quotes A seven-chapter format has been adopted for safety studies. This format allows for discussion of all the items required by DOE Order 5480.23 and for the discussions to be readily traceable to the listing in the order. The chapter titles are: (1) Introduction and Summary, (2) Site, (3) Facility Description, (4) Safety Basis, (5) Hazardous Material Management, (6) Management, Organization, and Institutional Safety Provisions, and (7) Accident Analysis

  2. Establishment of safety goal and its quantification based on risk assessment

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Muramatsu, Ken

    2017-01-01

    We must clarify the safety objectives sought by society in securing the safety of nuclear reactors and nuclear power plants. For that purpose, it is useful to utilize risk assessment. Quantitative methods including probabilistic risk assessment (PRA) are superior in terms of scientific rationality and quantitative performance compared with conventional deterministic methods, and able to indicate an objective numerical value of safety level. Consequently, quantitative methods can enhance the transparency, consistency, compliance, predictability, and explanatory power of regulatory decisions toward business operators and citizens. Business operators can explain the validity of their own safety assurance activities to regulators and citizens. The goal to be secured becomes clear by incorporating the safety goal into the specific performance goal required for the nuclear power plant from the viewpoint of deep safeguard, and it becomes easy to evaluate the effectiveness of the safety measures. It helps us greatly in judging and selecting the appropriateness of safety measures. It should be noted: the fact that the result of implementing the PRA satisfies the safety goal is not a sufficient condition in the sense of guaranteeing complete safety but a necessary condition. The nuclear power field is a region with large uncertainty, and research/efforts for accuracy improvement and evaluation validity will be required continuously. (A.O.)

  3. Regulatory status on the safety assessment of a HLW repository in other countries

    International Nuclear Information System (INIS)

    Lee, Sung Ho; Hwang, Yong Soo

    2008-12-01

    To construct a HLW repository, it is essential to meet the requirements on the regulation for a deep geological disposal. Even if the construction of a HLW repository is determined positively, technical standards which assert the performance of a repository will be needed. Among various technical standards, safety assessment based on the repository evolution in the future will play an important role in the licensing process. The foreign countries' technical standards on the safety assessment of a HLW repository may be an indicator to carry out the R and D activities on geological disposal effectively. In this report, assessment period, limit of radiation dose and uncertainty related to the safety assessment are investigated and analyzed in detail. Especially, the technical reviews of USA regulation bodies seems to be reasonable in the point of the intrinsic attribute of safety assessment

  4. Non-technical issues in safety assessments for nuclear disposal facilities

    International Nuclear Information System (INIS)

    Kallenbach-Herbert, Beate; Brohmann, Bettina

    2010-09-01

    The paper highlights that a comprehensive approach to safety affords the consideration of technology, organisation, personnel and social environment. In several safety relevant contexts of nuclear waste disposal these fields are closely interrelated. The approach for the consideration of socio-scientific aspects which is sketched in this paper supports the systematic treatment of safety relevant non-technical issues in the safety case or in safety assessments for a disposal project. Furthermore it may foster the dialogue among specialists from the technical, the natural- and the socio-scientific field on questions of disposal safety. In this way it may contribute to a better understanding among the affected scientific disciplines in nuclear waste disposal.

  5. Assessment of freeway work zone safety with improved cellular automata model

    Directory of Open Access Journals (Sweden)

    Guohua Liang

    2014-08-01

    Full Text Available To accurately assess the safety of freeway work zones, this paper investigates the safety of vehicle lane change maneuvers with improved cellular automata model. Taking the traffic conflict and standard deviation of operating speed as the evaluation indexes, the study evaluates the freeway work zone safety. With improved deceleration probability in car-following raies and the addition of lanechanging rules under critical state, the lane-changing behavior under critical state is defined as a conflict count. Through 72 schemes of simulation runs, the possible states of the traffic flow are carefully studied. The results show that under the condition of constant saturation traffic conflict count and vehicle speed standard deviation reach their maximums when the mixed rate of heave vehicles is 40%. Meanwhile, in the case of constant heavy vehicles mix, traffic conflict count and vehicle speed standard deviation reach maximum values when saturation rate is 0. 75. Integrating ail simulation results, it is known the traffic safety in freeway work zones is classified into four levels : safe, relatively safe, relatively dangerous, and dangerous.

  6. Assessing the safety attitudes questionnaire (SAQ), German language version in Swiss university hospitals - a validation study

    Science.gov (United States)

    2013-01-01

    Background Improving patient safety has become a major focus of clinical care and research over the past two decades. An institution’s patient safety climate represents an essential component of ensuring a safe environment and thereby can be vital to the prevention of adverse events. Covering six patient safety related factors, the Safety Attitudes Questionnaire (SAQ) is a validated and widely used instrument to measure the patient safety climate in clinical areas. The objective of this study was to assess the psychometric properties of the German language version of the SAQ. Methods A survey was carried out in two University Hospitals in Switzerland in autumn 2009 where the SAQ was distributed to a sample of 406 nurses and physicians in medical and surgical wards. Following the American Educational Research Association guidelines, we tested the questionnaire validity by levels of evidence: content validity, internal structure and relations to other variables. Confirmatory factor analysis was used to examine factor structure. Cronbach’s alphas and inter-item correlations were calculated to examine internal consistency reliability. Results A total of 319 questionnaires were completed representing an overall response rate of 78.6%. For three items, the item content validity index was <0.75. Confirmatory factor analysis showed acceptable model fit (RMSEA = 0.045; CFI = 0.944) for the six-factor model. Additional exploratory factor analysis could not identify a better factor model. SAQ factor scores showed positive correlations with the Safety Organizing Scale (r = .56 - .72). The SAQ German version showed moderate to strong internal consistency reliability indices (Cronbach alpha = .65 - .83). Conclusions The German language version of the SAQ demonstrated acceptable to good psychometric properties and therefore shows promise to be a sound instrument to measure patient safety climate in Swiss hospital wards. However, the low item content validity and large number

  7. Safety assessment of Olkiluoto NPP units 1 and 2. Decision of the Radiation and Nuclear Safety Authority regarding the periodic safety review of the Olkiluoto NPP

    International Nuclear Information System (INIS)

    2010-02-01

    In this safety assessment the Radiation and Nuclear Safety Authority (STUK) has evaluated the safety of the Olkiluoto Nuclear Power Plant units 1 and 2 in connection with the periodic safety review. This safety assessment provides a summary of the reviews, inspections and continuous oversight carried out by STUK. The issues addressed in the assessment and the related evaluation criteria are set forth in the nuclear energy and radiation safety legislation and the regulations issued thereunder. The provisions of the Nuclear Energy Act concerning the safe use of nuclear energy, security and emergency preparedness arrangements, and waste management are specified in more detail in the Government Decrees and Regulatory Guides issued by STUK. Based on the assessment, STUK consideres that the Olkiluoto Nuclear Power Plant units 1 and 2 meet the set safety requirements for operational nuclear power plants, the emergency preparedness arrangements are sufficient and the necessary control to prevent the proliferation of nuclear weapons has been appropriately arranged. The physical protection of the Olkiluoto nuclear power plant is not yet completely in compliance with the requirements of Government Decree 734/2008, which came into force in December 2008. Further requirements concerning this issue based also on the principle of continuous improvement were included in the decision relating to the periodic safety review. The safety of the Olkiluoto nuclear power plant was assessed in compliance with the Government Decree on the Safety of Nuclear Power Plants (733/2008), which came into force in 2008. The decree notes that existing nuclear power plants need not meet all the requirements set out for new plants. Most of the design bases pertaining to the Olkiluoto 1 and 2 nuclear power plant units were set in the 1970s. Substantial modernisations have been carried out at the Olkiluoto 1 and 2 nuclear power plant units since their commissioning to improve safety. This is in line with

  8. Interim main report of the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Hedin, Allan (ed.) [and others

    2004-08-01

    This document is an interim report on the safety assessment SR-Can (SR in the acronym stands for Safety Report and Can is short for canister). The final SR-Can report will support SKB's application to build an Encapsulation plant for spent nuclear fuel and is to be produced in 2006. The purpose of the present interim report is to demonstrate the methodology for safety assessment so that it can be reviewed before it is used in a license application. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. Preliminary data from the Forsmark site, presently being investigated by SKB as one of the candidate for a KBS-3 repository are used to some extent as examples. However, the collected data are yet too sparse to allow an evaluation of safety for this site. An important aim of this report is to demonstrate the proper handling of requirements on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Authority are duplicated in an Appendix. The principal acceptance criterion requires that 'the annual risk of harmful effects after closure does not exceed 10{sup -6} for a representative individual in the group exposed to the greatest risk'. 'Harmful effects' refer to cancer and hereditary effects. Following the introductory chapter 1, this report outlines the methodology for the SR-Can assessment in chapter 2, and presents in chapters 3, 4 and 5 the initial state of the system and the plans and methods for handling external influences and internal processes, respectively. Function indicators are introduced in chapter 6 and a preliminary evaluation of these is given in chapter 7. The material presented in the first seven chapters is utilised in the scenario selection

  9. Nirex Safety Assessment Research Programme bibliography, 1988

    International Nuclear Information System (INIS)

    Cooper, M.J.

    1988-05-01

    This bibliography lists reports and papers written as part of the Nirex Safety Assessment Research Programme, which is concerned with disposal of low-level and intermediate-level waste (LLW and ILW) and associated radiological assessments. All work referred to has been funded, or partly funded, by UK Nirex Limited, previously known as the Nuclear Industry Radioactive Waste Executive (NIREX). The bibliography has been divided into two sections, a List of Publications in roughly chronological order and an Author Index. The topics involved include near-field and far-field studies. The near-field includes the waste, its immobilising medium, its container, the engineered structure in which the containers are emplaced, and the immediately adjacent geological formation disturbed by the construction of the repository. The far-field is the undisturbed geological formation between the near-field and the biosphere. (author)

  10. A measurement tool to assess culture change regarding patient safety in hospital obstetrical units.

    Science.gov (United States)

    Kenneth Milne, J; Bendaly, Nicole; Bendaly, Leslie; Worsley, Jill; FitzGerald, John; Nisker, Jeff

    2010-06-01

    Clinical error in acute care hospitals can only be addressed by developing a culture of safety. We sought to develop a cultural assessment survey (CAS) to assess patient safety culture change in obstetrical units. Interview prompts and a preliminary questionnaire were developed through a literature review of patient safety and "high reliability organizations," followed by interviews with members of the Managing Obstetrical Risk Efficiently (MOREOB) Program of the Society of Obstetricians and Gynaecologists of Canada. Three hundred preliminary questionnaires were mailed, and 21 interviews and 9 focus groups were conducted with the staff of 11 hospital sites participating in the program. To pilot test the CAS, 350 surveys were mailed to staff in participating hospitals, and interviews were conducted with seven nurses and five physicians who had completed the survey. Reliability analysis was conducted on four units that completed the CAS prior to and following the implementation of the first MOREOB module. Nineteen values and 105 behaviours, practices, and perceptions relating to patient safety were identified and included in the preliminary questionnaire, of which 143 of 300 (47.4%) were returned. Among the 220 cultural assessment surveys returned (62.9%), six cultural scales emerged: (1) patient safety as everyone's priority; (2) teamwork; (3) valuing individuals; (4) open communication; (5) learning; and (6) empowering individuals. The reliability analysis found all six scales to have internal reliability (Cronbach alpha), ranging from 0.72 (open communication) to 0.84 (valuing individuals). The CAS developed for this study may enable obstetrical units to assess change in patient safety culture.

  11. Seismic safety margin assessment program (Annual safety research report, JFY 2010)

    International Nuclear Information System (INIS)

    Suzuki, Kenichi; Iijima, Toru; Inagaki, Masakatsu; Taoka, Hideto; Hidaka, Shinjiro

    2011-01-01

    Seismic capacity test data, analysis method and evaluation code provided by Seismic Safety Margin Assessment Program have been utilized for the support of seismic back-check evaluation of existing plants. The summary of the program in 2010 is as follows. 1. Component seismic capacity test and quantitative seismic capacity evaluation. Many seismic capacity tests of various snubbers were conducted and quantitative seismic capacities were evaluated. One of the emergency diesel generator partial-model seismic capacity tests was conducted and quantitative seismic capacity was evaluated. Some of the analytical evaluations of piping-system seismic capacities were conducted. 2. Analysis method for minute evaluation of component seismic response. The difference of seismic response of large components such as primary containment vessel and reactor pressure vessel when they were coupled with 3-dimensional FEM building model or 1-dimensional lumped mass building model, was quantitatively evaluated. 3. Evaluation code for quantitative evaluation of seismic safety margin of systems, structures and components. As the example, quantitative evaluation of seismic safety margin of systems, structures and components were conducted for the reference plant. (author)

  12. Validity of instruments to assess students' travel and pedestrian safety.

    Science.gov (United States)

    Mendoza, Jason A; Watson, Kathy; Baranowski, Tom; Nicklas, Theresa A; Uscanga, Doris K; Hanfling, Marcus J

    2010-05-18

    Safe Routes to School (SRTS) programs are designed to make walking and bicycling to school safe and accessible for children. Despite their growing popularity, few validated measures exist for assessing important outcomes such as type of student transport or pedestrian safety behaviors. This research validated the SRTS school travel survey and a pedestrian safety behavior checklist. Fourth grade students completed a brief written survey on how they got to school that day with set responses. Test-retest reliability was obtained 3-4 hours apart. Convergent validity of the SRTS travel survey was assessed by comparison to parents' report. For the measure of pedestrian safety behavior, 10 research assistants observed 29 students at a school intersection for completion of 8 selected pedestrian safety behaviors. Reliability was determined in two ways: correlations between the research assistants' ratings to that of the Principal Investigator (PI) and intraclass correlations (ICC) across research assistant ratings. The SRTS travel survey had high test-retest reliability (kappa = 0.97, n = 96, p < 0.001) and convergent validity (kappa = 0.87, n = 81, p < 0.001). The pedestrian safety behavior checklist had moderate reliability across research assistants' ratings (ICC = 0.48) and moderate correlation with the PI (r = 0.55, p = < 0.01). When two raters simultaneously used the instrument, the ICC increased to 0.65. Overall percent agreement (91%), sensitivity (85%) and specificity (83%) were acceptable. These validated instruments can be used to assess SRTS programs. The pedestrian safety behavior checklist may benefit from further formative work.

  13. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks.

  14. Safety culture perceptions of pharmacists in Malaysian hospitals and health clinics: a multicentre assessment using the Safety Attitudes Questionnaire

    Science.gov (United States)

    Samsuri, Srima Elina; Pei Lin, Lua; Fahrni, Mathumalar Loganathan

    2015-01-01

    Objective To assess the safety attitudes of pharmacists, provide a profile of their domains of safety attitude and correlate their attitudes with self-reported rates of medication errors. Design A cross-sectional study utilising the Safety Attitudes Questionnaire (SAQ). Setting 3 public hospitals and 27 health clinics. Participants 117 pharmacists. Main outcome measure(s) Safety culture mean scores, variation in scores across working units and between hospitals versus health clinics, predictors of safety culture, and medication errors and their correlation. Results Response rate was 83.6% (117 valid questionnaires returned). Stress recognition (73.0±20.4) and working condition (54.8±17.4) received the highest and lowest mean scores, respectively. Pharmacists exhibited positive attitudes towards: stress recognition (58.1%), job satisfaction (46.2%), teamwork climate (38.5%), safety climate (33.3%), perception of management (29.9%) and working condition (15.4%). With the exception of stress recognition, those who worked in health clinics scored higher than those in hospitals (psafety culture. As perceptions improved, the number of medication errors reported decreased. Group-specific interventions that target specific domains are necessary to improve the safety culture. PMID:26610761

  15. A methodology for a quantitative assessment of safety culture in NPPs based on Bayesian networks

    International Nuclear Information System (INIS)

    Kim, Young Gab; Lee, Seung Min; Seong, Poong Hyun

    2017-01-01

    Highlights: • A safety culture framework and a quantitative methodology to assess safety culture were proposed. • The relation among Norm system, Safety Management System and worker's awareness was established. • Safety culture probability at NPPs was updated by collecting actual organizational data. • Vulnerable areas and the relationship between safety culture and human error were confirmed. - Abstract: For a long time, safety has been recognized as a top priority in high-reliability industries such as aviation and nuclear power plants (NPPs). Establishing a safety culture requires a number of actions to enhance safety, one of which is changing the safety culture awareness of workers. The concept of safety culture in the nuclear power domain was established in the International Atomic Energy Agency (IAEA) safety series, wherein the importance of employee attitudes for maintaining organizational safety was emphasized. Safety culture assessment is a critical step in the process of enhancing safety culture. In this respect, assessment is focused on measuring the level of safety culture in an organization, and improving any weakness in the organization. However, many continue to think that the concept of safety culture is abstract and unclear. In addition, the results of safety culture assessments are mostly subjective and qualitative. Given the current situation, this paper suggests a quantitative methodology for safety culture assessments based on a Bayesian network. A proposed safety culture framework for NPPs would include the following: (1) a norm system, (2) a safety management system, (3) safety culture awareness of worker, and (4) Worker behavior. The level of safety culture awareness of workers at NPPs was reasoned through the proposed methodology. Then, areas of the organization that were vulnerable in terms of safety culture were derived by analyzing observational evidence. We also confirmed that the frequency of events involving human error

  16. Determination of the number of software tests using probabilistic safety assessment

    International Nuclear Information System (INIS)

    Kang, H. K.; Seong, T. Y.; Lee, K. Y.

    2000-01-01

    The broader usage of digital equipment in nuclear power plants gives rise to the safety problems of software. The field test should be performed before the software is used in critical applications because it is well known that software shows non-linear response when it is applied to different target systems in different environment. In the case of safety-critical applications, the result of tests contains usually zero failure case and the satisfiable number of tests is hard to be determined. In this paper, we suggests the method to determine the number of software tests without failure using the probabilistic safety assessment. From the result of the probabilistic safety assessment on total system, the desirable unavailability of software is calculated and the number of tests is determined

  17. Pushing and pulling: an assessment tool for occupational health and safety practitioners.

    Science.gov (United States)

    Lind, Carl Mikael

    2018-03-01

    A tool has been developed for supporting practitioners when assessing manual pushing and pulling operations based on an initiative by two global companies in the manufacturing industry. The aim of the tool is to support occupational health and safety practitioners in risk assessment and risk management of pushing and pulling operations in the manufacturing and logistics industries. The tool is based on a nine-multiplier equation that includes a wide range of factors affecting an operator's health risk and capacity in pushing and pulling. These multipliers are based on psychophysical, physiological and biomechanical studies in combination with judgments from an expert group consisting of senior researchers and ergonomists. In order to consider usability, more than 50 occupational health and safety practitioners (e.g., ergonomists, managers, safety representatives and production personnel) participated in the development of the tool. An evaluation by 22 ergonomists supports that the push/pull tool is user friendly in general.

  18. Probabilistic safety assessment of Narora Atomic Power Project

    International Nuclear Information System (INIS)

    Babar, A.K.; Saraf, R.K.; Kakodkar, A.; Sanyasi Rao, V.V.S.

    1989-01-01

    Various safety studies on Pressurised Water and Boiling Water reactors have been conducted. However, a detailed report on probabilistic safety assessment (PSA) of PHWRs is not available. PSA level I results of the standardised 235 MWe PHWR under construction at Narora are presented herein. Fault Tree analysis of various initiating events (IEs), safety systems has been completed. Event Tree analysis has been performed for all the dominating IEs to identify the accident sequences and a list of the dominating accident sequences is included. Analysis has been carried out using Monte Carlo simulation to propagate the uncertanities in failure rate data. Further uncertainty analysis is extended to obtain distributions for the accident sequences and core damage frequency. Some noteworthy results of the study apart from the various design modifications incorporated during the design phase are: (i) The accident sequences resulting from station blackout are dominant contributors to the core damage frequency. (ii) Class-IV transients, small break LOCA are significant IEs. Main steam line break is likely to induce steam generator tube ruptures. (iii) Moderator circulation, fire fighting system, secondary steam relief are relatively important in core damage frequency reductions. (iv) Under accidental situations human errors are likely to be asociated with valving in shutdown cooling and fire fighting systems. (author). 14 tabs., 14 figs., 15 refs

  19. An Assessment of SKB's Performance Assessment Calculations in the Interim Main Report for the Safety Assessment SR-Can

    International Nuclear Information System (INIS)

    Maul, Philip; Robinson, Peter

    2005-03-01

    SKB have published their Interim Main Report of the safety assessment SR-Can, which is intended to establish the framework for what will be submitted in 2006 in support of a licence application for construction of the spent fuel encapsulation plant. This follows on from the SR-Can Planning Document published in 2003. The purpose of the Interim Report is stated to be to demonstrate the methodology that will be used for safety assessment. The present report evaluates the information provided in the Interim SR-Can Report that is relevant to the Performance Assessment (PA) calculations that SKB intend to undertake, using independent calculations to facilitate this process. SKB consider that the primary safety function is to isolate completely the fuel within the canisters over the entire assessment period. Should a canister be damaged, the secondary safety function is to ensure that any release is retarded and dispersed sufficiently to ensure that concentrations levels in the accessible environment cannot cause unacceptable consequences. In this report PA calculations are considered to include both a high-level representation of the evolution of the system (relevant to the primary safety function), and any subsequent radionuclide transport (relevant to the secondary safety function). The main conclusions drawn are: 1. The effects of climate evolution on engineered barriers have not been analysed in detail in the Interim Report, and this limits the usefulness of the preliminary calculations that have been undertaken. 2. A key aspect of SKB's approach is the use of an integrated near-field evolution model. The information provided on this model demonstrates its capability efficiently to reproduce calculations from individual process models, but insufficient information is given at the present time to justify statements about interactions between processes. In particular it is assumed that relatively short term thermal and resaturation processes do not affect the

  20. Safety assessment of plant food supplements (PFS)

    NARCIS (Netherlands)

    Berg, van den S.J.P.L.; Serra-Majem, L.; Coppens, P.; Rietjens, I.

    2011-01-01

    Botanicals and botanical preparations, including plant food supplements (PFS), are widely used in Western diets. The growing use of PFS is accompanied by an increasing concern because the safety of these PFS is not generally assessed before they enter the market. Regulatory bodies have become more

  1. Comparative safety assessment of plant-derived foods

    NARCIS (Netherlands)

    Kok, E.J.; Keijer, J.; Kleter, G.A.; Kuiper, H.A.

    2008-01-01

    The second generation of genetically modified (GM) plants that are moving towards the market are characterized by modifications that may be more complex and traits that more often are to the benefit of the consumer. These developments will have implications for the safety assessment of the resulting

  2. Safety assessment for the passive system of the nuclear power plants (NPPs) using safety margin estimation

    International Nuclear Information System (INIS)

    Woo, Tae-Ho; Lee, Un-Chul

    2010-01-01

    The probabilistic safety assessment (PSA) for gas-cooled nuclear power plants has been investigated where the operational data are deficient, because there is not any commercial gas-cooled nuclear power plant. Therefore, it is necessary to use the statistical data for the basic event constructions. Several estimations for the safety margin are introduced for the quantification of the failure frequency in the basic event, which is made by the concept of the impact and affordability. Trend of probability of failure (TPF) and fuzzy converter (FC) are introduced using the safety margin, which shows the simplified and easy configurations for the event characteristics. The mass flow rate in the natural circulation is studied for the modeling. The potential energy in the gravity, the temperature and pressure in the heat conduction, and the heat transfer rate in the internal stored energy are also investigated. The values in the probability set are compared with those of the fuzzy set modeling. Non-linearity of the safety margin is expressed by the fuzziness of the membership function. This artificial intelligence analysis of the fuzzy set could enhance the reliability of the system comparing to the probabilistic analysis.

  3. International cooperation in the safety and environmental assessment for the ITER engineering design activities

    International Nuclear Information System (INIS)

    Gordon, C.; Baker, D.J.; Bartels, H-W.

    1998-01-01

    The ITER Project includes design and assessment activities to ensure the safety and environmental attractiveness of ITER and demonstrate that it can be sited in any of the sponsoring Parties with a minimum of site-specific redesign. This paper highlights some of the efforts to develop an international consensus approach for ITER safety design and assessment, including: development of general safety and environmental design criteria; development of quantitative dose-release assessment criteria; development of a radiation protection program; waste characterization; and development of safety analysis guidelines. The high level of interaction, cooperation and collaboration between the Joint Central Team and the Home Teams, and between the safety team and designers, and the spirit of consensus that has guided them have resulted in a safe design for ITER and a safety design and assessment that can meet the needs of the potential host countries. (author)

  4. Climate Considerations in Long-Term Safety Assessments for Nuclear Waste Repositories

    Energy Technology Data Exchange (ETDEWEB)

    Naeslund, Jens-Ove; Brandefelt, Jenny; Claesson Liljedahl, Lillemor [Svensk Kaernbraenslehantering AB, Stockholm (Sweden)], E-mail: jens-ove.naslund@skb.se

    2013-05-15

    For a deep geological repository for spent nuclear fuel planned in Sweden, the safety assessment covers up to 1 million years. Climate scenarios range from high-end global warming for the coming 100 000 years, through deep permafrost, to large ice sheets during glacial conditions. In contrast, in an existing repository for short-lived waste the activity decays to low levels within a few tens of thousands of years. The shorter assessment period, 100 000 years, requires more focus on climate development over the coming tens of thousands of years, including the earliest possibility for permafrost growth and freezing of the engineered system. The handling of climate and climate change in safety assessments must be tailor-made for each repository concept and waste type. However, due to the uncertain future climate development on these vast time scales, all safety assessments for nuclear waste repositories require a range of possible climate scenarios.

  5. Human reliability analysis in probabilistic safety assessment for nuclear power plants. A Safety Practice. A publication within the NUSS programme

    International Nuclear Information System (INIS)

    1995-01-01

    Probabilistic safety assessment (PSA) is playing an increasingly important role in the safe operation of nuclear power plants throughout the world. In order to establish a consistent framework for conducting PSA studies, for promoting technology transfer of the state of the art, and for encouraging uniformity in the way PSA is carried out, the IAEA is preparing a set of publications which gives guidance on various aspects of PSA. This document presents a practical approach for incorporating human reliability analysis (HRA) into PSA. It describes the steps needed and the documentation that should be provided both to support the PSA itself and to ensure effective communication of important information arising from the studies. It also describes a framework for analysing those human actions which could affect safety and for relating such human influences to specific parts of a PSA. This Safety Practice also addresses the limitations of PSA in taking account of human factors in relation to safety and risk. Refs, figs and tabs

  6. Safety assessment of biotechnology-derived pharmaceuticals: ICH and beyond.

    Science.gov (United States)

    Serabian, M A; Pilaro, A M

    1999-01-01

    Many scientific discussions, especially in the past 8 yr, have focused on definition of criteria for the optimal assessment of the preclinical toxicity of pharmaceuticals. With the current overlap of responsibility among centers within the Food and Drug Administration (FDA), uniformity of testing standards, when appropriate, would be desirable. These discussions have extended beyond the boundaries of the FDA and have culminated in the acceptance of formalized, internationally recognized guidances. The work of the International Committee on Harmonisation (ICH) and the initiatives developed by the FDA are important because they (a) represent a consensus scientific opinion, (b) promote consistency, (c) improve the quality of the studies performed, (d) assist the public sector in determining what may be generally acceptable to prepare product development plans, and (e) provide guidance for the sponsors in the design of preclinical toxicity studies. Disadvantages associated with such initiatives include (a) the establishment of a historical database that is difficult to relinquish, (b) the promotion of a check-the-box approach, i.e., a tendancy to perform only the minimum evaluation required by the guidelines, (c) the creation of a disincentive for industry to develop and validate new models, and (d) the creation of state-of-the-art guidances that may not allow for appropriate evaluation of novel therapies. The introduction of biotechnology-derived pharmaceuticals for clinical use has often required the application of unique approaches to assessing their safety in preclinical studies. There is much diversity among these products, which include the gene and cellular therapies, monoclonal antibodies, human-derived recombinant regulatory proteins, blood products, and vaccines. For many of the biological therapies, there will be unique product issues that may require specific modifications to protocol design and may raise additional safety concerns (e.g., immunogenicity

  7. Deep repository for long-lived low- and intermediate-level waste. Preliminary safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-11-01

    A preliminary safety assessment has been performed of a deep repository for long-lived low- and intermediate-level waste, SFL 3-5. The purpose of the study is to investigate the capacity of the facility to act as a barrier to the release of radionuclides and toxic pollutants, and to shed light on the importance of the location of the repository site. A safety assessment (SR 97) of a deep repository for spent fuel has been carried out at the same time. In SR 97, three hypothetical repository sites have been selected for study. These sites exhibit fairly different conditions in terms of hydrogeology, hydrochemistry and ecosystems. To make use of information and data from the SR 97 study, we have assumed that SFL 3-5 is co-sited with the deep repository for spent fuel. A conceivable alternative is to site SFL 3-5 as a completely separate repository. The focus of the SFL 3-5 study is a quantitative analysis of the environmental impact for a reference scenario, while other scenarios are discussed and analyzed in more general terms. Migration in the repository's near- and far-field has been taken into account in the reference scenario. Environmental impact on the three sites has also been calculated. The calculations are based on an updated forecast of the waste to be disposed of in SFL 3-5. The forecast includes radionuclide content, toxic metals and other substances that have a bearing on a safety assessment. The safety assessment shows how important the site is for safety. Two factors stand out as being particularly important: the water flow at the depth in the rock where the repository is built, and the ecosystem in the areas on the ground surface where releases may take place in the future. Another conclusion is that radionuclides that are highly mobile and long-lived, such as {sup 36}Cl and {sup 93}Mo , are important to take into consideration. Their being long-lived means that barriers and the ecosystems must be regarded with a very long time horizon.

  8. Studying the Safety Impact of Autonomous Vehicles Using Simulation-Based Surrogate Safety Measures

    Directory of Open Access Journals (Sweden)

    Mark Mario Morando

    2018-01-01

    Full Text Available Autonomous vehicle (AV technology has advanced rapidly in recent years with some automated features already available in vehicles on the market. AVs are expected to reduce traffic crashes as the majority of crashes are related to driver errors, fatigue, alcohol, or drugs. However, very little research has been conducted to estimate the safety impact of AVs. This paper aims to investigate the safety impacts of AVs using a simulation-based surrogate safety measure approach. To this end, safety impacts are explored through the number of conflicts extracted from the VISSIM traffic microsimulator using the Surrogate Safety Assessment Model (SSAM. Behaviours of human-driven vehicles (HVs and AVs (level 4 automation are modelled within the VISSIM’s car-following model. The safety investigation is conducted for two case studies, that is, a signalised intersection and a roundabout, under various AV penetration rates. Results suggest that AVs improve safety significantly with high penetration rates, even when they travel with shorter headways to improve road capacity and reduce delay. For the signalised intersection, AVs reduce the number of conflicts by 20% to 65% with the AV penetration rates of between 50% and 100% (statistically significant at p<0.05. For the roundabout, the number of conflicts is reduced by 29% to 64% with the 100% AV penetration rate (statistically significant at p<0.05.

  9. Current issues and perspectives in food safety and risk assessment.

    Science.gov (United States)

    Eisenbrand, G

    2015-12-01

    In this review, current issues and opportunities in food safety assessment are discussed. Food safety is considered an essential element inherent in global food security. Hazard characterization is pivotal within the continuum of risk assessment, but it may be conceived only within a very limited frame as a true alternative to risk assessment. Elucidation of the mode of action underlying a given hazard is vital to create a plausible basis for human toxicology evaluation. Risk assessment, to convey meaningful risk communication, must be based on appropriate and reliable consideration of both exposure and mode of action. New perspectives, provided by monitoring human exogenous and endogenous exposure biomarkers, are considered of great promise to support classical risk extrapolation from animal toxicology. © The Author(s) 2015.

  10. Canadian Consumer Food Safety Practices and Knowledge: Foodbook Study.

    Science.gov (United States)

    Murray, Regan; Glass-Kaastra, Shiona; Gardhouse, Christine; Marshall, Barbara; Ciampa, Nadia; Franklin, Kristyn; Hurst, Matt; Thomas, M Kate; Nesbitt, Andrea

    2017-10-01

    Understanding consumers' food safety practices and knowledge supports food safety education for the prevention of foodborne illness. The objective of this study was to describe Canadian consumer food safety practices and knowledge. This study identifies demographic groups for targeted food safety education messaging and establishes a baseline measurement to assess the effectiveness of food safety interventions over time. Questions regarding consumer food safety practices and knowledge were included in a population-based telephone survey, Foodbook, conducted from November 2014 to March 2015. The results were analyzed nationally by age group and by gender. The results showed that approximately 90% of Canadians reported taking the recommended cleaning and separating precautions when handling raw meat to prevent foodborne illness. Only 29% of respondents reported using a food thermometer when cooking any meat, and even fewer (12%) reported using a food thermometer for small cuts of meat such as chicken pieces. The majority (>80%) of Canadians were aware of the foodborne illness risks related to chicken and hamburger, but fewer (poultry.

  11. Improvement of safety approach for accident during operation of LILW disposal facility: Application for operational safety assessment of the near-surface LILW disposal facility in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Joo; Kim, Min Seong; Park, Jin Beak [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

    2017-06-15

    To evaluate radiological impact from the operation of a low- and intermediate-level radioactive waste disposal facility, a logical presentation and explanation of expected accidental scenarios is essential to the stakeholders of the disposal facility. The logical assessment platform and procedure, including analysis of the safety function of disposal components, operational hazard analysis, operational risk analysis, and preparedness of remedial measures for operational safety, are improved in this study. In the operational risk analysis, both design measures and management measures are suggested to make it possible to connect among design, operation, and safety assessment within the same assessment platform. For the preparedness of logical assessment procedure, classifcation logic of an operational accident is suggested based on the probability of occurrence and consequences of assessment results. The improved assessment platform and procedure are applied to an operational accident analysis of the Korean low- and intermediate-level radioactive waste disposal facility and partly presented in this paper.

  12. Improvement of safety approach for accident during operation of LILW disposal facility: Application for operational safety assessment of the near-surface LILW disposal facility in Korea

    International Nuclear Information System (INIS)

    Kim, Hyun Joo; Kim, Min Seong; Park, Jin Beak

    2017-01-01

    To evaluate radiological impact from the operation of a low- and intermediate-level radioactive waste disposal facility, a logical presentation and explanation of expected accidental scenarios is essential to the stakeholders of the disposal facility. The logical assessment platform and procedure, including analysis of the safety function of disposal components, operational hazard analysis, operational risk analysis, and preparedness of remedial measures for operational safety, are improved in this study. In the operational risk analysis, both design measures and management measures are suggested to make it possible to connect among design, operation, and safety assessment within the same assessment platform. For the preparedness of logical assessment procedure, classifcation logic of an operational accident is suggested based on the probability of occurrence and consequences of assessment results. The improved assessment platform and procedure are applied to an operational accident analysis of the Korean low- and intermediate-level radioactive waste disposal facility and partly presented in this paper

  13. Validation of innovative technologies and strategies for regulatory safety assessment methods: challenges and opportunities.

    Science.gov (United States)

    Stokes, William S; Wind, Marilyn

    2010-01-01

    Advances in science and innovative technologies are providing new opportunities to develop test methods and strategies that may improve safety assessments and reduce animal use for safety testing. These include high throughput screening and other approaches that can rapidly measure or predict various molecular, genetic, and cellular perturbations caused by test substances. Integrated testing and decision strategies that consider multiple types of information and data are also being developed. Prior to their use for regulatory decision-making, new methods and strategies must undergo appropriate validation studies to determine the extent that their use can provide equivalent or improved protection compared to existing methods and to determine the extent that reproducible results can be obtained in different laboratories. Comprehensive and optimal validation study designs are expected to expedite the validation and regulatory acceptance of new test methods and strategies that will support improved safety assessments and reduced animal use for regulatory testing.

  14. Probabilistic assessment of nuclear safety and safeguards

    International Nuclear Information System (INIS)

    Higson, D.J.

    1987-01-01

    Nuclear reactor accidents and diversions of materials from the nuclear fuel cycle are perceived by many people as particularly serious threats to society. Probabilistic assessment is a rational approach to the evaluation of both threats, and may provide a basis for decisions on appropriate actions to control them. Probabilistic method have become standard tools used in the analysis of safety, but there are disagreements on the criteria to be applied when assessing the results of analysis. Probabilistic analysis and assessment of the effectiveness of nuclear material safeguards are still at an early stage of development. (author)

  15. Statistical analysis applied to safety culture self-assessment

    International Nuclear Information System (INIS)

    Macedo Soares, P.P.

    2002-01-01

    Interviews and opinion surveys are instruments used to assess the safety culture in an organization as part of the Safety Culture Enhancement Programme. Specific statistical tools are used to analyse the survey results. This paper presents an example of an opinion survey with the corresponding application of the statistical analysis and the conclusions obtained. Survey validation, Frequency statistics, Kolmogorov-Smirnov non-parametric test, Student (T-test) and ANOVA means comparison tests and LSD post-hoc multiple comparison test, are discussed. (author)

  16. Results of the CANDU 3 probabilistic safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Jaitly, R K [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    The purpose of the Conceptual Probabilistic Safety Assessment (PSA) of the CANDU 3 reactor was to provide safety assistance in the early stages of design to ensure that the design included adequate redundancy and functional separation of the mitigating systems; the final design should therefore give better results, particularly after modifications involving control, electrical power, instrument air, and service water. The initial PSA gave a total CANDU 3 core damage frequency of 7.8 x 10{sup -6}/year. 4 refs., 1 fig.

  17. Results of the CANDU 3 probabilistic safety assessment

    International Nuclear Information System (INIS)

    Jaitly, R.K.

    1995-01-01

    The purpose of the Conceptual Probabilistic Safety Assessment (PSA) of the CANDU 3 reactor was to provide safety assistance in the early stages of design to ensure that the design included adequate redundancy and functional separation of the mitigating systems; the final design should therefore give better results, particularly after modifications involving control, electrical power, instrument air, and service water. The initial PSA gave a total CANDU 3 core damage frequency of 7.8 x 10 -6 /year. 4 refs., 1 fig

  18. HERMES docking/berthing system pilot study. Quantitative assessment

    International Nuclear Information System (INIS)

    Munoz Blasco, J.; Goicoechea Sanchez, F.J.

    1993-01-01

    This study falls within the framework of the incorporation of quantitative risk assessment to the activities planned for the ESA-HERMES project (ESA/ CNES). The main objective behind the study was the analysis and evaluation of the potential contribution of so-called probabilistic or quantitative safety analysis to the optimization of the safety development process for the systems carrying out the safety functions required by the new and complex HERMES Space Vehicle. For this purpose, a pilot study was considered a good start in quantitative safety assessments (QSA), as this approach has been frequently used in the past to establish a solid base in large-scale QSA application programs while avoiding considerable economic risks. It was finally decided to select the HERMES docking/berthing system with Man Tender Free Flyer as the case-study. This report describes the different steps followed in the study, along with the main insights obtained and the general conclusions drawn from the study results. (author)

  19. Analysis on evaluation ability of nonlinear safety assessment model of coal mines based on artificial neural network

    Institute of Scientific and Technical Information of China (English)

    SHI Shi-liang; LIU Hai-bo; LIU Ai-hua

    2004-01-01

    Based on the integration analysis of goods and shortcomings of various methods used in safety assessment of coal mines, combining nonlinear feature of mine safety sub-system, this paper establishes the neural network assessment model of mine safety, analyzes the ability of artificial neural network to evaluate mine safety state, and lays the theoretical foundation of artificial neural network using in the systematic optimization of mine safety assessment and getting reasonable accurate safety assessment result.

  20. Developing guidance in the nuclear criticality safety assessment for fuel cycle facilities

    International Nuclear Information System (INIS)

    Galet, C.; Evo, S.

    2012-01-01

    In this poster IRSN (Institute for radiation protection and nuclear safety) presents its safety guides whose purpose is to transmit the safety assessment know-how to any 'junior' staff or even to give a view of the safety approach on the overall risks to any staff member. IRSN has written a first version of such a safety guide for fuel cycle facilities and laboratories. It is organized into several chapters: some refer to types of assessments, others concern the types of risks. Currently, this guide contains 13 chapters and each chapter consists of three parts. In parallel to the development of criticality chapter of this guide, the IRSN criticality department has developed a nuclear criticality safety guide. It follows the structure of the three parts fore-mentioned, but it presents a more detailed first part and integrates, in the third part, the experience feedback collected on nuclear facilities. The nuclear criticality safety guide is online on the IRSN's web site