WorldWideScience

Sample records for gavage safety assessment

  1. Evaluation of an acute oral gavage method for assessment of pesticide toxicity in terrestrial amphibians.

    Science.gov (United States)

    Fort, Douglas J; Mathis, Michael B; Kee, Faith; Whatling, Paul; Clerkin, David; Staveley, Jane; Habig, Clifford

    2018-02-01

    Development of an acute oral toxicity test with a terrestrial-phase amphibian was considered necessary to remove the uncertainty within the field of agrochemical risk assessments. The bullfrog (Lithobates catesbeianus) was selected for use as it is a representative of the family Ranidae and historically this species has been used as an amphibian test model species. Prior to definitive study, oral gavage methods were developed with fenthion and tetraethyl pyrophosphate. Dimethoate and malathion were subsequently tested with both male and female juvenile bullfrogs in comprehensive acute oral median lethal dose (LD50) studies. Juvenile bullfrogs were administered a single dose of the test article via oral gavage of a single gelatin capsule of dimethoate technical (dimethoate) or neat liquid Fyfanon ® Technical (synonym malathion), returned to their respective aquaria, and monitored for survival for 14 d. The primary endpoint was mortality, whereas behavioral responses, food consumption, body weight, and snout-vent length (SVL) were used to evaluate indications of sublethal toxicity (secondary endpoints). Acute oral LD50 values (95% fiducial interval) for dimethoate were 1459 (1176-1810, males) and 1528 (1275-1831, females), and for malathion they were 1829 (1480-2259, males) and 1672 (1280-2183, females) mg active substance/kg body weight, respectively. Based on the results of these studies, the methodology for the acute oral gavage administration of test items to terrestrial-phase amphibians was demonstrated as being a practical method of providing data for risk assessments. Environ Toxicol Chem 2018;37:436-450. © 2017 SETAC. © 2017 SETAC.

  2. A less stressful alternative to oral gavage for pharmacological and toxicological studies in mice

    International Nuclear Information System (INIS)

    Walker, Mary K.; Boberg, Jason R.; Walsh, Mary T.; Wolf, Valerie; Trujillo, Alisha; Duke, Melissa Skelton; Palme, Rupert; Felton, Linda A.

    2012-01-01

    Oral gavage dosing can induce stress and potentially confound experimental measurements, particularly when blood pressure and heart rate are endpoints of interest. Thus, we developed a pill formulation that mice would voluntarily consume and tested the hypothesis that pill dosing would be significantly less stressful than oral gavage. C57Bl/6 male mice were singly housed and on four consecutive days were exposed to an individual walking into the room (week 1, control), a pill being placed into the cage (week 2), and a dose of water via oral gavage (week 3). Blood pressure and heart rate were recorded by radiotelemetry continuously for 5 h after treatment, and feces collected 6–10 h after treatment for analysis of corticosterone metabolites. Both pill and gavage dosing significantly increased mean arterial pressure (MAP) during the first hour, compared to control. However, the increase in MAP was significantly greater after gavage and remained elevated up to 5 h, while MAP returned to normal within 2 h after a pill. Neither pill nor gavage dosing significantly increased heart rate during the first hour, compared to control; however, pill dosing significantly reduced heart rate while gavage significantly increased heart rate 2–5 h post dosing. MAP and heart rate did not differ 24 h after dosing. Lastly, only gavage dosing significantly increased fecal corticosterone metabolites, indicating a systemic stress response via activation of the hypothalamic–pituitary–adrenal axis. These data demonstrated that this pill dosing method of mice is significantly less stressful than oral gavage. -- Highlights: ► Developed a novel oral dosing method using a pill that mice will readily consume. ► Assessed stress by blood pressure, heart rate, and fecal corticosterone metabolites. ► Demonstrated that pill dosing is significantly less stressful than oral gavage.

  3. A less stressful alternative to oral gavage for pharmacological and toxicological studies in mice

    Energy Technology Data Exchange (ETDEWEB)

    Walker, Mary K., E-mail: mwalker@salud.unm.edu [Department of Pharmaceutical Sciences, College of Pharmacy, University of New Mexico Health Sciences Center, Albuquerque, NM (United States); Boberg, Jason R.; Walsh, Mary T.; Wolf, Valerie; Trujillo, Alisha; Duke, Melissa Skelton [Department of Pharmaceutical Sciences, College of Pharmacy, University of New Mexico Health Sciences Center, Albuquerque, NM (United States); Palme, Rupert [Department of Biomedical Sciences/Biochemistry, University of Veterinary Medicine, Vienna (Austria); Felton, Linda A. [Department of Pharmaceutical Sciences, College of Pharmacy, University of New Mexico Health Sciences Center, Albuquerque, NM (United States)

    2012-04-01

    Oral gavage dosing can induce stress and potentially confound experimental measurements, particularly when blood pressure and heart rate are endpoints of interest. Thus, we developed a pill formulation that mice would voluntarily consume and tested the hypothesis that pill dosing would be significantly less stressful than oral gavage. C57Bl/6 male mice were singly housed and on four consecutive days were exposed to an individual walking into the room (week 1, control), a pill being placed into the cage (week 2), and a dose of water via oral gavage (week 3). Blood pressure and heart rate were recorded by radiotelemetry continuously for 5 h after treatment, and feces collected 6–10 h after treatment for analysis of corticosterone metabolites. Both pill and gavage dosing significantly increased mean arterial pressure (MAP) during the first hour, compared to control. However, the increase in MAP was significantly greater after gavage and remained elevated up to 5 h, while MAP returned to normal within 2 h after a pill. Neither pill nor gavage dosing significantly increased heart rate during the first hour, compared to control; however, pill dosing significantly reduced heart rate while gavage significantly increased heart rate 2–5 h post dosing. MAP and heart rate did not differ 24 h after dosing. Lastly, only gavage dosing significantly increased fecal corticosterone metabolites, indicating a systemic stress response via activation of the hypothalamic–pituitary–adrenal axis. These data demonstrated that this pill dosing method of mice is significantly less stressful than oral gavage. -- Highlights: ► Developed a novel oral dosing method using a pill that mice will readily consume. ► Assessed stress by blood pressure, heart rate, and fecal corticosterone metabolites. ► Demonstrated that pill dosing is significantly less stressful than oral gavage.

  4. Acute Ethanol Gavage Attenuates Hemorrhage/Resuscitation-Induced Hepatic Oxidative Stress in Rats

    Directory of Open Access Journals (Sweden)

    B. Relja

    2012-01-01

    Full Text Available Acute ethanol intoxication increases the production of reactive oxygen species (ROS. Hemorrhagic shock with subsequent resuscitation (H/R also induces ROS resulting in cellular and hepatic damage in vivo. We examined the role of acute ethanol intoxication upon oxidative stress and subsequent hepatic cell death after H/R. 14 h before H/R, rats were gavaged with single dose of ethanol or saline (5 g/kg, EtOH and ctrl; H/R_EtOH or H/R_ctrl, resp.. Then, rats were hemorrhaged to a mean arterial blood pressure of 30±2 mmHg for 60 min and resuscitated. Two control groups underwent surgical procedures without H/R (sham_ctrl and sham_EtOH, resp.. Liver tissues were harvested at 2, 24, and 72 h after resuscitation. EtOH-gavage induced histological picture of acute fatty liver. Hepatic oxidative (4-hydroxynonenal, 4-HNE and nitrosative (3-nitrotyrosine, 3-NT stress were significantly reduced in EtOH-gavaged rats compared to controls after H/R. Proapoptotic caspase-8 and Bax expressions were markedly diminished in EtOH-gavaged animals compared with controls 2 h after resuscitation. EtOH-gavage increased antiapoptotic Bcl-2 gene expression compared with controls 2 h after resuscitation. iNOS protein expression increased following H/R but was attenuated in EtOH-gavaged animals after H/R. Taken together, the data suggest that acute EtOH-gavage may attenuate H/R-induced oxidative stress thereby reducing cellular injury in rat liver.

  5. 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

  6. Pharmacokinetics of Mequindox and Its Marker Residue 1,4-Bisdesoxymequindox in Swine Following Multiple Oral Gavage and Intramuscular Administration : An Experimental Study Coupled with Population Physiologically Based Pharmacokinetic Modeling

    NARCIS (Netherlands)

    Zeng, Dongping; Lin, Zhoumeng; Fang, Binghu; Li, Miao; Gehring, Ronette; Riviere, Jim E; Zeng, Zhenling

    2017-01-01

    Mequindox (MEQ) is a quinoxaline-N,N-dioxide antibiotic used in food-producing animals. MEQ residue in animal-derived foods is a food safety concern. The tissue distribution of MEQ and its marker residue 1,4-bisdesoxymequindox (M1) were determined in swine following oral gavage or intramuscular

  7. Impact of oral Lactobacillus acidophilus gavage on rooster seminal and cloacal Lactobacilli concentrations.

    Science.gov (United States)

    Kiess, A S; Hirai, J H; Triplett, M D; Parker, H M; McDaniel, C D

    2016-08-01

    The use of antibiotics in poultry is being heavily scrutinized, therefore alternatives such as probiotics are being investigated. Lactobacilli spp. are a commonly used bacteria in formulating probiotics, and the addition of Lactobacilli to broiler diets has demonstrated increased growth rates, stimulated immune systems, and reduced pathogen loads in the gastro-intestinal tract ( GI: ) tract. However, previous research has shown that when rooster semen is directly exposed to Lactobacillus acidophilus (L. acidophilus) sperm quality is reduced. Therefore, the objective of the current study was to determine if oral administration of L. acidophilus increases the concentration of Lactobacilli in semen as well as the cloaca. A total of 30 roosters were used: 15 roosters were gavaged with 1X PBS (Control) and 15 roosters were gavaged with 10(7) cfu/mL of L. acidophilus (Treated). All roosters were gavaged for 14 consecutive days. Semen was collected on a 3 d interval, and cloacal swabs were collected on a 2 d interval, beginning on the first day prior to oral administration. Semen and cloacal swabs were serial diluted, and 100 μL of each dilution was then plated on Man, Rogosa, Sharpe ( MRS: ) agar plates. All plates were incubated for 48 h at 37°C under anaerobic conditions and counted. All Lactobacilli counts were first log transformed, then log transformed (day 0) pre-counts were subtracted from the log transformed day counts providing log differences for the analysis. Seminal Lactobacilli counts were not altered by treatments. However, the main effect of treatment (P = 0.026) for cloacal counts indicated that roosters gavaged with Lactobacilli yielded higher counts than the controls. Additionally, cloaca samples also demonstrated a treatment by day interaction trend (P = 0.082), where Lactobacilli was higher in the L. acidophilus gavaged roosters than the controls only on days 3, 5, 13, and 15. In conclusion, the addition of L. acidophilus to the male breeder diet

  8. 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

  9. 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)

  10. 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

  11. 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

  12. Toxicological Assessment of β-(1à6-Glucan (Lasiodiplodan in Mice during a 28-Day Feeding Study by Gavage

    Directory of Open Access Journals (Sweden)

    Janaína A. Túrmina

    2012-12-01

    Full Text Available Studies evaluating the toxicity caused by fungal exopolysaccharides of the β-(1®6-D-glucan type are rare. In this study, the toxicological effects of sub-chronic treatments with lasiodiplodan (β-(1®6-D-glucan from Lasiodiplodia theobromae MMPI were evaluated in mice through the assessment of biochemical, hematological, and histopathological alterations. Thirty-two mice (16 male, 16 female were used in this study divided in two groups; one group received lasiodiplodan (50 mg/kg body weight daily for 28 days via gavage, and another (control group received saline during the same period. Blood samples were collected via cardiac puncture for hematological and biochemical analyses. Liver, heart, kidney, and spleen were collected for histopathological analysis. Statistical analysis was performed through one-way analysis of variance and only p < 0.05 F-values were presented. Significant reduction in blood glucose in the male group (35%; p < 0.01, transaminases activity in both sexes (AST and ALT; ~35%; p < 0.05, and urea (20%; p < 0.01 in the female group was observed with the lasiodiplodan treatment. The results showed that sub-chronic treatments with lasiodiplodan did not generate hematological and histopathological alterations leading to signs of toxicity in healthy mice, independent of gender.

  13. 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

  14. 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)

  15. 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

  16. 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)

  17. 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

  18. 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

  19. 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.

  20. 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)

  1. 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)

  2. 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)

  3. 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)

  4. 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)

  5. 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

  6. 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.

  7. 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

  8. 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

  9. 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

  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.? 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

  11. 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)

  12. 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)

  13. 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)

  14. 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)

  15. 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

  16. 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.

  17. 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

  18. 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)

  19. 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.

  20. 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

  1. 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

  2. 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.

  3. 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

  4. Effect of gavaged chemical form of 241Am on its retention in mice

    International Nuclear Information System (INIS)

    Hisamatsu, S.; Takizawa, Y.

    1987-01-01

    The retention of 241 Am in mice 48 h after administration by gavage is reported here. The 241 Am was given to mice in the form of either 241 Am nitrate or 241 Am citrate. The 241 Am was also injected into rats in the same form. The homogenized livers of those rats were subsequently administered by gavage to another group of mice. The retention of 241 Am citrate was 1.5 X 10(-2)% of the original dose and was the highest among the compounds examined. The retention of biologically incorporated 241 Am into the liver as 241 Am nitrate and as 241 Am citrate was 2.4 X 10(-3) and 2.6 X 10(-3)%, respectively, and was similar to the retention of 241 Am nitrate, which was 2.8 X 10(-3)%. The ratio of the retention in the carcass to that in the liver for the 241 Am citrate was lower than that of the 241 Am nitrate and the biologically incorporated 241 Am. This difference indicates that the distribution of 241 Am in the animal body depends on the chemical form administered. The retention of liver-incorporated 241 Am as citrate after autolysis of the liver is similar to that of fresh liver-incorporated 241 Am citrate

  5. 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)

  6. 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

  7. 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

  8. 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

  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). 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

  10. 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

  11. 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.

  12. 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)

  13. 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

  14. 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)

  15. 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

  16. 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)

  17. 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

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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)

  3. 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)

  4. 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

  5. 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

  6. 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

  7. 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)

  8. 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)

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. Stable engraftment of human microbiota into mice with a single oral gavage following antibiotic conditioning.

    Science.gov (United States)

    Staley, Christopher; Kaiser, Thomas; Beura, Lalit K; Hamilton, Matthew J; Weingarden, Alexa R; Bobr, Aleh; Kang, Johnthomas; Masopust, David; Sadowsky, Michael J; Khoruts, Alexander

    2017-08-01

    Human microbiota-associated (HMA) animal models relying on germ-free recipient mice are being used to study the relationship between intestinal microbiota and human disease. However, transfer of microbiota into germ-free animals also triggers global developmental changes in the recipient intestine, which can mask disease-specific attributes of the donor material. Therefore, a simple model of replacing microbiota into a developmentally mature intestinal environment remains highly desirable. Here we report on the development of a sequential, three-course antibiotic conditioning regimen that allows sustained engraftment of intestinal microorganisms following a single oral gavage with human donor microbiota. SourceTracker, a Bayesian, OTU-based algorithm, indicated that 59.3 ± 3.0% of the fecal bacterial communities in treated mice were attributable to the donor source. This overall degree of microbiota engraftment was similar in mice conditioned with antibiotics and germ-free mice. Limited surveys of systemic and mucosal immune sites did not show evidence of immune activation following introduction of human microbiota. The antibiotic treatment protocol described here followed by a single gavage of human microbiota may provide a useful, complimentary HMA model to that established in germ-free facilities. The model has the potential for further in-depth translational investigations of microbiota in a variety of human disease states.

  15. 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)

  16. 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)

  17. 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)

  18. 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

  19. 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

  20. 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)

  1. 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)

  2. 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

  3. 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.

  4. 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

  5. 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

  6. 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)

  7. 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

  8. 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

  9. Food safety assessment of Cry8Ka5 mutant protein using Cry1Ac as a control Bt protein.

    Science.gov (United States)

    Farias, Davi Felipe; Viana, Martônio Ponte; Oliveira, Gustavo Ramos; Santos, Vanessa Olinto; Pinto, Clidia Eduarda Moreira; Viana, Daniel Araújo; Vasconcelos, Ilka Maria; Grossi-de-Sa, Maria Fátima; Carvalho, Ana Fontenele Urano

    2015-07-01

    Cry8Ka5 is a mutant protein from Bacillus thuringiensis (Bt) that has been proposed for developing transgenic plants due to promising activity against coleopterans, like Anthonomus grandis (the major pest of Brazilian cotton culture). Thus, an early food safety assessment of Cry8Ka5 protein could provide valuable information to support its use as a harmless biotechnological tool. This study aimed to evaluate the food safety of Cry8Ka5 protein following the two-tiered approach, based on weights of evidence, proposed by ILSI. Cry1Ac protein was used as a control Bt protein. The history of safe use revealed no convincing hazard reports for Bt pesticides and three-domain Cry proteins. The bioinformatics analysis with the primary amino acids sequence of Cry8Ka5 showed no similarity to any known toxic, antinutritional or allergenic proteins. The mode of action of Cry proteins is well understood and their fine specificity is restricted to insects. Cry8Ka5 and Cry1Ac proteins were rapidly degraded in simulated gastric fluid, but were resistant to simulated intestinal fluid and heat treatment. The LD50 for Cry8Ka5 and Cry1Ac was >5000 mg/kg body weight when administered by gavage in mice. Thus, no expected relevant risks are associated with the consumption of Cry8Ka5 protein. Copyright © 2015 Elsevier Ltd. All rights reserved.

  10. Voluntary Ingestion of Antiparasitic Drugs Emulsified in Honey Represents an Alternative to Gavage in Mice

    OpenAIRE

    Küster, Tatiana; Zumkehr, Beatrice; Hermann, Corina; Theurillat, Regula; Thormann, Wolfgang; Gottstein, Bruno; Hemphill, Andrew

    2012-01-01

    The oral route is the most frequently used method of drug intake in humans. Oral administration of drugs to laboratory animals such as mice typically is achieved through gavage, in which a feeding needle is introduced into the esophagus and the drug is delivered directly into the stomach. This method requires technical skill, is stressful for animals, and introduces risk of injury, pain and morbidity. Here we investigated another method of drug administration. The benzimidazole derivative alb...

  11. 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

  12. 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

  13. 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.

  14. 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...

  15. 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

  16. 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.

  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. 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

  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. 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

  6. 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

  7. 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)

  8. 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

  9. 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

  10. 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.

  11. 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

  12. 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.

  13. 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

  14. 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.

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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.

  20. 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.

  1. 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

  2. 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)

  3. 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)

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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.

  9. 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

  10. 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.

  11. 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...

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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

  1. 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.)

  2. 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

  3. 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/

  4. 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/

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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...

  11. 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...

  12. 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.

  13. 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!

  14. 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

  15. 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

  16. 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

  17. 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.)

  18. 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

  19. 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)

  20. 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)

  1. 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)

  2. 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)

  3. 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)

  4. Pulmonary gene and microRNA expression changes in mice exposed to benzo(a)pyrene by oral gavage

    International Nuclear Information System (INIS)

    Halappanavar, Sabina; Wu, Dongmei; Williams, Andrew; Kuo, Byron; Godschalk, Roger W.; Van Schooten, Frederik J.; Yauk, Carole Lyn

    2011-01-01

    Highlights: → The study examines pulmonary response in mice exposed to BaP by oral gavage. → We examined pulmonary gene and miRNA expression changes and measured DNA adducts. → We compare the mechanisms of action that operate in lungs relative to the liver. → We show differences in biological pathways activated in lungs versus the liver. → We suggest that liver miRNAs are less sensitive to perturbations than lung miRNAs. -- Abstract: Exposure to the environmental mutagen benzo(a)pyrene (BaP) alters the expression of AHR-responsive genes as well as genes involved in other pathways. We recently reported that exposure of adult mice to BaP resulted in a robust transcriptome response in the liver, but this was accompanied by a complete lack of change in microRNA (miRNA) expression. Since BaP exposure does not result in hepatocarcinogenicity, but does cause lung cancer, in the present study we examine the pulmonary mRNA and miRNA responses to BaP in the same mice. Adult male B6C3F1 mice were exposed to 150 and 300 mg/kg BaP by oral gavage for three consecutive days and sacrificed 4 h after the last exposure. Serum clinical chemistry was performed for both the doses to assess the general toxicity of BaP; a modest decrease in serum inorganic phosphorous was observed at both the doses. A small decrease in serum glucose following 150 mg/kg and alkaline phosphatase following 300 mg/kg BaP was observed. BaP-DNA adduct levels in whole lung and liver tissues were assessed by 32 P postlabelling and similar dose dependent increases were observed for lung and liver. Using DNA microarrays, pulmonary mRNA and miRNA expressions were analysed. Over 1000 genes were statistically differentially expressed (p < 0.05). The perturbed pathways included oxidative stress, xenobiotic metabolism, cell proliferation, cell cycle, B and T-cell receptor signalling and primary immunodeficiency signalling pathways. Analysis of miRNA profiles revealed downregulation of miR-150, miR-142-5p, mi

  5. 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

  6. 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.

  7. 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

  8. 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.

  9. 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

  10. 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).

  11. 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

  12. Freezing and thawing effects on fat, protein, and lactose levels of human natural milk administered by gavage and continuous infusion

    Directory of Open Access Journals (Sweden)

    Andrea D. Abranches

    2014-07-01

    Full Text Available OBJECTIVES: to analyze the changes in human milk macronutrients: fat, protein, and lactose in natural human milk (raw, frozen and thawed, after administration simulation by gavage and continuous infusion. METHOD: an experimental study was performed with 34 human milk samples. The infrared spectrophotometry using the infrared analysis equipment MilkoScan Minor(r (Foss, Denmark equipment was used to analyze the macronutrients in human milk during the study phases. The analyses were performed in natural (raw samples and after freezing and fast thawing following two steps: gavage and continuous infusion. The non-parametric Wilcoxon test for paired samples was used for the statistical analysis. RESULTS: the fat content was significantly reduced after administration by continuous infusion (p < 0.001 during administration of both raw and thawed samples. No changes in protein and lactose content were observed between the two forms of infusion. However, the thawing process significantly increased the levels of lactose and milk protein. CONCLUSION: the route of administration by continuous infusion showed the greatest influence on fat loss among all the processes required for human milk administration.

  13. Freezing and thawing effects on fat, protein, and lactose levels of human natural milk administered by gavage and continuous infusion.

    Science.gov (United States)

    Abranches, Andrea D; Soares, Fernanda V M; Junior, Saint-Clair G; Moreira, Maria Elisabeth L

    2014-01-01

    to analyze the changes in human milk macronutrients: fat, protein, and lactose in natural human milk (raw), frozen and thawed, after administration simulation by gavage and continuous infusion. an experimental study was performed with 34 human milk samples. The infrared spectrophotometry using the infrared analysis equipment MilkoScan Minor® (Foss, Denmark) equipment was used to analyze the macronutrients in human milk during the study phases. The analyses were performed in natural (raw) samples and after freezing and fast thawing following two steps: gavage and continuous infusion. The non-parametric Wilcoxon test for paired samples was used for the statistical analysis. the fat content was significantly reduced after administration by continuous infusion (praw and thawed samples. No changes in protein and lactose content were observed between the two forms of infusion. However, the thawing process significantly increased the levels of lactose and milk protein. the route of administration by continuous infusion showed the greatest influence on fat loss among all the processes required for human milk administration. Copyright © 2014 Sociedade Brasileira de Pediatria. Published by Elsevier Editora Ltda. All rights reserved.

  14. 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

  15. 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

  16. 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

  17. 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.

  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. 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

  20. 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

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. 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)

  6. 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

  7. 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)

  8. 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)

  9. 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...

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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

  19. [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

  20. 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

  1. 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

  2. Intestinal lymphangiectasis and lipidosis in rats following subchronic exposure to indole-3-carbinol via oral gavage.

    Science.gov (United States)

    Boyle, Michael C; Crabbs, Torrie A; Wyde, Michael E; Painter, J Todd; Hill, Georgette D; Malarkey, David E; Lieuallen, Warren G; Nyska, Abraham

    2012-06-01

    To investigate the toxicity and carcinogenic potential of indole-3-carbinol (I3C), the National Toxicology Program has conducted 13-week subchronic studies in Fisher 344 rats and B6C3F1 mice, and chronic 2-year bioassays in Sprague-Dawley rats and B6C3F1 mice. While the chronic study results are not yet available, subchronic study results and short-term special evaluations of interim sacrifices in the 2-year rat bioassay are presented. F344 rats were orally gavaged ≤300 mg I3C/kg body weight 5 days a week for 13 weeks. Rats treated with ≥150 mg/kg demonstrated a dose-related dilation of lymphatics (lymphangiectasis) of the duodenum, jejunum, and mesenteric lymph nodes. Material within dilated lacteals stained positively for Oil Red O and Sudan Black, consistent with lipid. Electron microscopic evaluation confirmed extracellular lipid accumulation within the villar lamina propria, lacteals, and within villar macrophages. Analyses of hepatic and pulmonary CYP1A enzymes demonstrated dose-dependent I3C induction of CYP1A1 and 1A2. B6C3F1 mice orally gavaged ≤250 mg I3C/kg body weight did not demonstrate histopathological changes; however, hepatic CYP induction was similar to that in rats. The histopathologic changes of intestinal lymphangiectasis and lipidosis in this study share similarities with intestinal lymphangiectasia as observed in humans and dogs. However, the resultant clinical spectrum of protein-losing enteropathy was not present.

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. 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)

  9. 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...

  10. 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.)

  11. 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

  12. 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

  13. 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.

  14. 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)

  15. 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

  16. 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.

  17. 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.

  18. 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

  19. 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.

  20. 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

  1. 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.).

  2. 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.)

  3. 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 - ...

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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.

  9. 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

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. 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

  16. 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.

  17. 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

  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 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)

  20. 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...

  1. 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.

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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...

  9. 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/

  10. 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/

  11. 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)

  12. 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.

  13. 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.

  14. 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

  15. 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

  16. 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

  17. [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.

  18. 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

  19. 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.

  20. 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

  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. 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

  3. 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.

  4. 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

  5. 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...

  6. 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

  7. 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

  8. 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.

  9. 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

  10. 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.

  11. 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

  12. 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)

  13. 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)

  14. 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

  15. 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)

  16. 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.

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. 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

  2. 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)

  3. 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

  4. 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

  5. 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.)

  6. 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)

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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

  17. 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.

  18. 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

  19. 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)

  20. 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)

  1. 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.)

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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.

  7. 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)

  8. 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.

  9. 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)

  10. 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)

  11. 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.

  12. 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

  13. 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.

  14. 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...

  15. 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

  16. 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.

  17. 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)

  18. 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

  19. 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

  20. 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.

  1. 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.)

  2. 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

  3. 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.)

  4. 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

  5. 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)

  6. 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...

  7. 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.

  8. 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

  9. 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.)

  10. 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.

  11. 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.

  12. 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.)

  13. 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

  14. 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.

  15. 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

  16. 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

  17. 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)

  18. 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.

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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.

  7. 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.

  8. 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)

  9. 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)

  10. 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.

  11. 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

  12. 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

  13. 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.

  14. 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

  15. Safety assessment of a borehole type disposal facility using the ISAM methodology

    International Nuclear Information System (INIS)

    Blerk, J.J. van; Yucel, V.; Kozak, M.W.; Moore, B.A.

    2002-01-01

    As part of the IAEA's Co-ordinated Research Project (CRP) on Improving Long-term of Safety Assessment Methodologies for Near Surface Waste Disposal Facilities (ISAM), three example cases were developed. The aim was to test the ISAM safety assessment methodology using as realistic as possible data. One of the Test Cases, the Borehole Test Case (BTC), related to a proposed future disposal option for disused sealed radioactive sources. This paper uses the various steps of the ISAM safety assessment methodology to describe the work undertaken by ISAM participants in developing the BTC and provides some general conclusions that can be drawn from the findings of their work. (author)

  16. 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

  17. Safety assessments for centralized waste treatment and disposal facility in Puspokszilagy Hungary

    International Nuclear Information System (INIS)

    Berci, K.; Hauszmann, Z.; Ormai, P.

    2002-01-01

    The centralized waste treatment and disposal facility Puspokszilagy is a shallow land, near surface engineered type disposal unit. The site, together with its geographic, geological and hydrogeological characteristics, is described. Data are given on the radioactive inventory. The operational safety assessment and the post-closure safety assessment is outlined. (author)

  18. An Integrated Safety Assessment Methodology for Generation IV Nuclear Systems

    International Nuclear Information System (INIS)

    Leahy, Timothy J.

    2010-01-01

    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. Early work of the RSWG focused on defining a safety philosophy founded on lessons learned from current and prior generations of nuclear technologies, and on identifying technology characteristics that may help achieve Generation IV safety goals. More recent RSWG work has focused on the definition of an integrated safety assessment methodology for evaluating the safety of Generation IV systems. The methodology, tentatively called ISAM, is an integrated 'toolkit' consisting of analytical techniques that are available and matched to appropriate stages of Generation IV system concept development. 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.

  19. 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.

  20. ILK statement about the regulatory authorities' perception of operators' self-assessment of safety culture

    International Nuclear Information System (INIS)

    2005-01-01

    Over the past few years, German licensing and supervisory authorities have devoted increasing attention to safety management and safety culture issues. At present, German plant operators are introducing systems for self-assessment of the safety culture in their plants, such as the Safety Culture Assessment System developed by VGB Power Tech (VGB-SBS). In its statement, the International Committee on Nuclear Technology (ILK) addresses an effective approach of the authorities in evaluating the self-assessment of safety culture conducted by operators. ILK proposes a total of ten recommendations for evaluating the self-assessment system of the operators by the authority. The regulatory authorities should see to it that the operators establish a self-assessment system for aspects of organization and personnel, and use it continuously. The measures derived from this self-assessment by the operators, and the reasons underlying them, should be discussed with the authorities. In addition to the operators, also the regulatory authorities and the technical expert organizations commissioned by them should carry out self-assessments of their respective supervisory activities, taking into account also special events, such as changes in government, and develop appropriate programs of measures to be taken. In evaluating safety culture, the regulatory authorities should strive to support the activities of operators in improving their safety culture. A spirit of mutual confidence and cooperation should exist between operators and authorities. The recommendations expressed in the statement deliberately leave room for detailed implementation by the parties concerned. (orig.)

  1. Regulatory assessment of safety culture in nuclear organisations - current trends and challenges

    International Nuclear Information System (INIS)

    Tronea, M.

    2010-01-01

    The paper gives an overview of the current practices in the area of regulatory assessment of safety culture in nuclear organisations and of the associated challenges. While the assessment and inspection procedures currently in use by regulatory authorities worldwide are directed primarily at verifying compliance with the licensing basis, there is a recognised need for a more systematic approach to the identification, collection and review of data relevant to the safety culture in licensees' organisations. The paper presents a proposal for using the existing regulatory inspection practices for gathering information relevant to safety culture and for assessing it in an integrated manner. The proposal is based on the latest requirements and guidance issued by the International Atomic Energy Agency (IAEA) on management systems for nuclear facilities and activities, particularly as regards the attributes needed for a strong nuclear safety culture. (author)

  2. Risk-informed approaches to assess ecological safety of facilities with radioactive waste

    International Nuclear Information System (INIS)

    Vashchenko, V.N.; Zlochevskij, V.V.; Skalozubov, V.I.

    2011-01-01

    Ingenious risk-informed methods to assess ecological safety of facilities with radioactive waste are proposed in the paper. Probabilistic norms on lethal outcomes and reliability of safety barriers are used as safety criteria. Based on the probability measures, it is established that ecological safety conditions are met for the standard criterion of lethal outcomes

  3. Safety assessment for TA-48 radiochemical operations

    International Nuclear Information System (INIS)

    1994-08-01

    The purpose of this report is to document an assessment performed to evaluate the safety of the radiochemical operations conducted at the Los Alamos National Laboratory operations area designated as TA-48. This Safety Assessment for the TA-48 radiochemical operations was prepared to fulfill the requirements of US Department of Energy (DOE) Order 5481.1B, ''Safety Analysis and Review System.'' The area designated as TA-48 is operated by the Chemical Science and Technology (CST) Division and is involved with radiochemical operations associated with nuclear weapons testing, evaluation of samples collected from a variety of environmental sources, and nuclear medicine activities. This report documents a systematic evaluation of the hazards associated with the radiochemical operations that are conducted at TA-48. The accident analyses are limited to evaluation of the expected consequences associated with a few bounding accident scenarios that are selected as part of the hazard analysis. Section 2 of this report presents an executive summary and conclusions, Section 3 presents pertinent information concerning the TA-48 site and surrounding area, Section 4 presents a description of the TA-48 radiochemical operations, and Section 5 presents a description of the individual facilities. Section 6 of the report presents an evaluation of the hazards that are associated with the TA-48 operations and Section 7 presents a detailed analysis of selected accident scenarios

  4. Assessing the safety of the road network : a simple method.

    NARCIS (Netherlands)

    Janssen, S.T.M.C.

    1993-01-01

    Research by the SWOV Institute for Road Safety Research in The Netherlands assessing the safety of the Dutch road network is presented. In The Netherlands urban and rural roads and motorways are used roughly equally, however accident rates vary greatly. These are discussed in this paper. Tables show

  5. 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.

  6. SKI's and SSI's joint review of SKB's safety assessment report, SR 97. Summary

    International Nuclear Information System (INIS)

    2001-01-01

    The Swedish Nuclear Fuel and Waste Management Co (SKB) has a programme for the siting of a repository for spent nuclear fuel in Swedish bedrock. In 1996, the Swedish Government decided that SKB must perform an assessment of the repository's long-term safety before undertaking the next step of the programme which entails drilling in a minimum of two municipalities (site investigations). SKB has presented such a safety assessment in SR 97 Post-closure Safety (henceforth referred to as SR 97). SR 97 is one of the documents in the comprehensive reporting that SKB must provide when it proposes sites for investigation. The Swedish Nuclear Power Inspectorate (SKI) and the Swedish Radiation Protection Institute (SSI) have evaluated SR 97 in terms of its purposes which are to demonstrate a methodology for safety assessment, to show that Swedish bedrock can provide a safe repository using SKB's method, to provide a basis for specifying the factors that are important for site selection and to derive preliminary requirements on the function of the engineered barriers. The authorities have reached the following conclusions: SR 97 does not indicate any conditions that would mean that geological final disposal in accordance with SKB's method would have significant deficiencies in relation to the safety and radiation protection requirements of the authorities. SR 97 contains the elements required for a comprehensive assessment of safety and radiation protection. SKB's safety assessment methodology has improved within several important areas, such as the documentation of processes and properties that can affect repository performance and the development of models for safety assessment calculations. The methodology used in SR 97 has some deficiencies, for example, the specification of future events to be described in the safety assessment. SR 97 has not, to an adequate extent, dealt with unfavourable conditions that can affect the future safety of a repository. SKB states that the

  7. Safety assessment of multi-unit NPP sites subject to external events

    International Nuclear Information System (INIS)

    Samaddar, Sujit; Hibino, Kenta; Coman, Ovidiu

    2014-01-01

    This paper presents a framework for conducting a probabilistic safety assessment of multi-unit sites against external events. The treatment of multiple hazard on a unit, interaction between units, implementation of severe accident measures, human reliability, environmental conditions, metric of risk for both reactor and non-reactor sources, integration of risk and responses and many such important factors need to be addressed within the context of this framework. The framework facilitates the establishment of a comprehensive methodology that can be applied internationally to the peer review of safety assessment of multi-unit sites under the impact of multiple external hazards. In summary, it can be said that the site safety assessment for a multi-unit site will be quite complex and need to start with individual unit risk assessments, these need to be combined considering the interactions between units and their responses, and the fragilities of the installations established considering the combined demands from all interactions. Using newly established risk metric the risk can then be integrated for the overall site. Fig. 2 shows schematically such a proposal. Much work has to done and the IAEA has established a working group that is systematically establishing the structure and process to incorporate the many issues that are a part of a multi-unit site safety assessment. (authors)

  8. Climate considerations in long-term safety assessments for nuclear waste repositories.

    Science.gov (United States)

    Näslund, Jens-Ove; Brandefelt, Jenny; Liljedahl, Lillemor Claesson

    2013-05-01

    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.

  9. Assessment of Human Performance and Safety Culture at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Toth, Janos; Hadnagy, Lajos

    2002-01-01

    Evaluation of human performance and safety culture of the personnel at a Nuclear Power Plant is a very important element of the self assessment process. At the Paks NPP a systematic approach to this problem started in the early 90's. The first comprehensive analysis of the human performance of the personnel was performed by the Hungarian Research Institute for Electric Power (VEIKI). The analysis of human failures is also a part of the investigation and analysis of safety related reported events. This human performance analysis of events is carried out by the Laboratory of Psychology of the plant and a supporting organisation namely the Department of Ergonomics and Psychology of the Budapest University of Technical and Economical Sciences. The analysis of safety culture at the Paks NPP has been in the focus of attention since the implementation of the INSAG-4 document started world-wide. In 1993 an IAEA model project namely 'Strengthening Training for Operational Safety' was initiated with a sub-project called 'Enhancement of Safety Culture'. Within this project the first step was the initial assessment of the safety culture level at the Paks NPP. It was followed by some corrective actions and safety culture improvement programme. In 1999 the second assessment was performed in order to evaluate the progress as a result of the improvement programme. A few indicators reflecting the elements of safety culture were defined and compared. The assessment of the safety culture with a survey among the managers was performed in September 2000 and the results are being evaluated at the moment. The intention of the plant management is to repeat the assessment every 2-3 years and evaluate the trend of the indicator. (authors)

  10. Safety assessment principles for reactor protection systems in the United Kingdom

    International Nuclear Information System (INIS)

    Philp, W.

    1990-01-01

    The duty of Nuclear Installations Inspectorate (NII) is to see that the appropriate standards are developed, achieved and maintained by the plant operators, and to monitor and regulate the safety of the plant by means of its powers under the licence. It does not issue standards or codes of practice for NPPs, but it requires each plant operator to develop its own safety criteria and requirements. The following relevant issues are described: NII assessment principles and societal risks; principles and guidance for the assessment of rector protection systems; assessment of reactor shutdown systems

  11. Safety assessment principles for reactor protection systems in the United Kingdom

    Energy Technology Data Exchange (ETDEWEB)

    Philp, W

    1990-07-01

    The duty of Nuclear Installations Inspectorate (NII) is to see that the appropriate standards are developed, achieved and maintained by the plant operators, and to monitor and regulate the safety of the plant by means of its powers under the licence. It does not issue standards or codes of practice for NPPs, but it requires each plant operator to develop its own safety criteria and requirements. The following relevant issues are described: NII assessment principles and societal risks; principles and guidance for the assessment of rector protection systems; assessment of reactor shutdown systems.

  12. 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)

  13. 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.)

  14. Risk-informed decision making a keystone in advanced safety assessment

    International Nuclear Information System (INIS)

    Reinhart, M.

    2007-01-01

    Probabilistic Safety Assessment (PSA) has provided extremely valuable complementary insight, perspective, comprehension, and balance to deterministic nuclear reactor safety assessment. This integrated approach of risk-informed management and decision making has been called Risk-Informed Decision Making (RIDM). RIDM provides enhanced safety, reliability, operational flexibility, reduced radiological exposure, and improved fiscal economy. Applications of RIDM continuously increase. Current applications are in the areas of design, construction, licensing, operations, and security. Operational phase safety applications include the following: technical specifications improvement, risk-monitors and configuration control, maintenance planning, outage planning and management, in-service inspection, inservice testing, graded quality assurance, reactor oversight and inspection, inspection finding significance determination, operational events assessment, and rulemaking. Interestingly there is a significant spectrum of approaches, methods, programs, controls, data bases, and standards. The quest of many is to assimilate the full compliment of PSA and RIDM information and to achieve a balanced international harmony. The goal is to focus the best of the best, so to speak, for the benefit of all. Accordingly, this presentation will address the principles, benefits, and applications of RIDM. It will also address some of the challenges and areas to improve. Finally it will highlight efforts by the IAEA and others to capture the international thinking, experience, successes, challenges, and lessons in RIDM. (authors)

  15. Current status and new trends in the methodology of safety assessment for near surface disposal facilities

    International Nuclear Information System (INIS)

    Ilie, Petre; Didita, Liana; Danchiv, Alexandru

    2008-01-01

    The main goal of this paper is to present the status of the safety assessment methodology at the end of IAEA CRP 'Application of Safety Assessment Methodology for Near-Surface Radioactive Waste Disposal Facilities (ASAM)', and the new trends outlined at the launch of the follow-up project 'Practical Implementation of Safety Assessment Methodologies in a Context of Safety Case of Near-Surface Facilities (PRISM)'. Over the duration of the ASAM project, the ISAM methodology was confirmed as providing a good framework for conducting safety assessment calculations. In contrast, ASAM project identified the limitations of the ISAM methodology as currently formulated. The major limitations are situated in the area of the use of safety assessment for informing practical decisions about alternative waste and risk management strategies for real disposal sites. As a result of the limitation of the ISAM methodology, the PRISM project is established as an extension of the ISAM and ASAM projects. Based on the outcomes of the ASAM project, the main objective of the PRISM project are: 1 - to develop an overview of what constitutes an adequate safety case and safety assessment with a view to supporting decision making processes; 2 - to provide practical illustrations of how the safety assessment methodology could be used for addressing some specific issues arising from the ASAM project and national cases; 3 - to support harmonization with the IAEA's international safety standards. (authors)

  16. A Framework for Assessment of Aviation Safety Technology Portfolios

    Science.gov (United States)

    Jones, Sharon M.; Reveley, Mary S.

    2014-01-01

    The programs within NASA's Aeronautics Research Mission Directorate (ARMD) conduct research and development to improve the national air transportation system so that Americans can travel as safely as possible. NASA aviation safety systems analysis personnel support various levels of ARMD management in their fulfillment of system analysis and technology prioritization as defined in the agency's program and project requirements. This paper provides a framework for the assessment of aviation safety research and technology portfolios that includes metrics such as projected impact on current and future safety, technical development risk and implementation risk. The paper also contains methods for presenting portfolio analysis and aviation safety Bayesian Belief Network (BBN) output results to management using bubble charts and quantitative decision analysis techniques.

  17. 76 FR 45453 - New Car Assessment Program (NCAP); Safety Labeling

    Science.gov (United States)

    2011-07-29

    ... Testing F. Consumer Survey and Label Research G. Other Issues 1. Legend for Star Ratings 2. Overall... consumers to consider and assess the relative safety of vehicles before deciding which new vehicle they want..., and prominent fashion, and that the safety rating area of the Monroney label meet minimum size...

  18. Aging in probabilistic safety assessment

    International Nuclear Information System (INIS)

    Jordan Cizelj, R.; Kozuh, M.

    1995-01-01

    Aging is a phenomenon, which is influencing on unavailability of all components of the plant. The influence of aging on Probabilistic Safety Assessment calculations was estimated for Electrical Power Supply System. The average increase of system unavailability due to aging of system components was estimated and components were prioritized regarding their influence on change of system unavailability and relative increase of their unavailability due to aging. After the analysis of some numerical results, the recommendation for a detailed research of aging phenomena and its influence on system availability is given. (author)

  19. 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

  20. Probabilistic safety assessment - regulatory perspective

    International Nuclear Information System (INIS)

    Solanki, R.B.; Paul, U.K.; Hajra, P.; Agarwal, S.K.

    2002-01-01

    Full text: Nuclear power plants (NPPs) have been designed, constructed and operated mainly based on deterministic safety analysis philosophy. In this approach, a substantial amount of safety margin is incorporated in the design and operational requirements. Additional margin is incorporated by applying the highest quality engineering codes, standards and practices, and the concept of defence-in-depth in design and operating procedures, by including conservative assumptions and acceptance criteria in plant response analysis of postulated initiating events (PIEs). However, as the probabilistic approach has been improved and refined over the years, it is possible for the designer, operator and regulator to get a more detailed and realistic picture of the safety importance of plant design features, operating procedures and operational practices by using probabilistic safety assessment (PSA) along with the deterministic methodology. At present, many countries including USA, UK and France are using PSA insights in their decision making along with deterministic basis. India has also made substantial progress in the development of methods for carrying out PSA. However, consensus on the use of PSA in regulatory decision-making has not been achieved yet. This paper emphasises on the requirements (e.g.,level of details, key modelling assumptions, data, modelling aspects, success criteria, sensitivity and uncertainty analysis) for improving the quality and consistency in performance and use of PSA that can facilitate meaningful use of the PSA insights in the regulatory decision-making in India. This paper also provides relevant information on international scenario and various application areas of PSA along with progress made in India. The PSA perspective presented in this paper may help in achieving consensus on the use of PSA for regulatory / utility decision-making in design and operation of NPPs

  1. Safety assessment of a vault-based disposal facility using the ISAM methodology

    International Nuclear Information System (INIS)

    Kelly, E.; Kim, C.-L.; Lietava, P.; Little, R.; Simon, I.

    2002-01-01

    As part of the IAEA's Co-ordinated Research Project (CRP) on Improving Long-term of Safety Assessment Methodologies for Near Surface Waste Disposal Facilities (ISAM), three example cases were developed. The aim was to testing the ISAM safety assessment methodology using as realistic as possible data. One of the Test Cases, the Vault Test Case (VTC), related to the disposal of low level radioactive waste (LLW) to a hypothetical facility comprising a set of above surface vaults. This paper uses the various steps of the ISAM safety assessment methodology to describe the work undertaken by ISAM participants in developing the VTC and provides some general conclusions that can be drawn from the findings of their work. (author)

  2. Guidelines for the Review of Research Reactor Safety: Revised Edition. Reference Document for IAEA Integrated Safety Assessment of Research Reactors (INSARR)

    International Nuclear Information System (INIS)

    2013-01-01

    The Integrated Safety Assessment of Research Reactors (INSARR) is an IAEA safety review service available to Member States with the objective of supporting them in ensuring and enhancing the safety of their research reactors. This service consists of performing a comprehensive peer review and an assessment of the safety of the respective research reactor. The reviews are based on IAEA safety standards and on the provisions of the Code of Conduct on the Safety of Research Reactors. The INSARR can benefit both the operating organizations and the regulatory bodies of the requesting Member States, and can include new research reactors under design or operating research reactors, including those which are under a Project and Supply Agreement with the IAEA. The first IAEA safety evaluation of a research reactor operated by a Member State was completed in October 1959 and involved the Swiss 20 MW DIORIT research reactor. Since then, and in accordance with its programme on research reactor safety, the IAEA has conducted safety review missions in its Member States to enhance the safety of their research reactor facilities through the application of the Code of Conduct on the Safety of Research Reactors and the relevant IAEA safety standards. About 320 missions in 51 Member States were undertaken between 1972 and 2012. The INSARR missions and other limited scope safety review missions are conducted following the guidelines presented in this publication, which is a revision of Guidelines for the Review of Research Reactor Safety (IAEA Services Series No. 1), published in December 1997. This publication details those IAEA safety standards and guidance publications relevant to the safety of research reactors that have been revised or published since 1997. The purpose of this publication is to give guidance on the preparation, implementation, reporting and follow-up of safety review missions. It is also intended to be of assistance to operators and regulators in conducting

  3. Waste isolation safety assessment program

    International Nuclear Information System (INIS)

    Brandstetter, A.; Harwell, M.A.

    1979-05-01

    Associated with commercial nuclear power production in the United States is the generation of potentially hazardous radioactive wastes. The Department of Energy (DOE), through the National Waste Terminal Storage (NWTS) Program, is seeking to develop nuclear waste isolation systems in geologic formations that will preclude contact with the biosphere of waste radionuclides in concentrations which are sufficient to cause deleterious impact on humans or their environments. Comprehensive analyses of specific isolation systems are needed to assess the expectations of meeting that objective. The Waste Isolation Safety Assessment Program (WISAP) has been established at the Pacific Northwest Laboratory (operated by Battelle Memorial Institute) for developing the capability of making those analyses. Among the analyses required for isolation system evaluation is the detailed assessment of the post-closure performance of nuclear waste repositories in geologic formations. This assessment is essential, since it is concerned with aspects of the nuclear power program which previously have not been addressed. Specifically, the nature of the isolation systems (e.g., involving breach scenarios and transport through the geosphere), and the time-scales necessary for isolation, dictate the development, demonstration and application of novel assessment capabilities. The assessment methodology needs to be thorough, flexible, objective, and scientifically defensible. Further, the data utilized must be accurate, documented, reproducible, and based on sound scientific principles

  4. How to interpret safety critical failures in risk and reliability assessments

    International Nuclear Information System (INIS)

    Selvik, Jon Tømmerås; Signoret, Jean-Pierre

    2017-01-01

    Management of safety systems often receives high attention due to the potential for industrial accidents. In risk and reliability literature concerning such systems, and particularly concerning safety-instrumented systems, one frequently comes across the term ‘safety critical failure’. It is a term associated with the term ‘critical failure’, and it is often deduced that a safety critical failure refers to a failure occurring in a safety critical system. Although this is correct in some situations, it is not matching with for example the mathematical definition given in ISO/TR 12489:2013 on reliability modeling, where a clear distinction is made between ‘safe failures’ and ‘dangerous failures’. In this article, we show that different interpretations of the term ‘safety critical failure’ exist, and there is room for misinterpretations and misunderstandings regarding risk and reliability assessments where failure information linked to safety systems are used, and which could influence decision-making. The article gives some examples from the oil and gas industry, showing different possible interpretations of the term. In particular we discuss the link between criticality and failure. The article points in general to the importance of adequate risk communication when using the term, and gives some clarification on interpretation in risk and reliability assessments.

  5. Safety assessment methodologies for near surface disposal facilities. Results of a co-ordinated research project (ISAM). Volume 1: Review and enhancement of safety assessment approaches and tools. Volume 2: Test cases

    International Nuclear Information System (INIS)

    2004-07-01

    For several decades, countries have made use of near surface facilities for the disposal of low and intermediate level radioactive waste. In line with the internationally agreed principles of radioactive waste management, the safety of these facilities needs to be ensured during all stages of their lifetimes, including the post-closure period. By the mid 1990s, formal methodologies for evaluating the long term safety of such facilities had been developed, but intercomparison of these methodologies had revealed a number of discrepancies between them. Consequently, in 1997, the International Atomic Energy Agency launched a Co-ordinated Research Project (CRP) on Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities (ISAM). The particular objectives of the CRP were to provide a critical evaluation of the approaches and tools used in post-closure safety assessment for proposed and existing near-surface radioactive waste disposal facilities, enhance the approaches and tools used and build confidence in the approaches and tools used. The CRP ran until 2000 and resulted in the development of a harmonized assessment methodology (the ISAM project methodology), which was applied to a number of test cases. Over seventy participants from twenty-two Member States played an active role in the project and it attracted interest from around seven hundred persons involved with safety assessment in seventy-two Member States. The results of the CRP have contributed to the Action Plan on the Safety of Radioactive Waste Management which was approved by the Board of Governors and endorsed by the General Conference in September 2001. Specifically, they contribute to Action 5, which requests the IAEA Secretariat to 'develop a structured and systematic programme to ensure adequate application of the Agency's waste safety standards', by elaborating on the Safety Requirements on 'Near Surface Disposal of Radioactive Waste' (Safety Standards Series No. WS-R-1) and

  6. Assessment of safety engineering of circuits with dc micromotors

    Energy Technology Data Exchange (ETDEWEB)

    Pavlyuchenko, L.A.; Starchuk, S.E.

    1986-01-01

    Presents an assessment of safety engineering in d.c. micromotors operating as part of actuating devices in mining equipment. These micromotors should have RO (especially explosion proof) protection. The safety engineering should be assessed with an intermittent fault in the power line. Equations are given for calculation of the equivalent inductance of the micromotor circuit with an intermittent power line fault. If the circuit is not intrinsically safe, a diode in the forward direction is recommended for connection in series with the micromotor. If the power line is not intrinsically safe, a diode shunt is recommended. Comparative data for power sources (IBP) and micromotors (DPM, DPR, with permanent magnets) are given in tables. 4 refs.

  7. CLI technical commission. Additional safety assessments within the EDF nuclear stock

    International Nuclear Information System (INIS)

    2011-01-01

    This slides presentation addresses additional safety assessments within the EDF nuclear stock. It describes the context and challenges of these assessments (institutional framework, European coherence, major objectives for EDF). It describes how EDF is organised to perform these assessments: a global project after Fukushima, assessments and tests which are beyond the existing safety referential, the three defence lines and their tests. It addresses the content of assessment reports for each topic (earthquake, flooding, loss of water, loss of electric supply, accident management). It indicates some improvements proposed after the first assessments. It describes the improvements concerning the Gravelines power plant. It recalls the assessment agenda, proposes a brief overview of events, comments the results of the third decennial visit of production unit no. 1 which comprised some proof test on the primary circuit, on the reactor vessel and on the reactor building

  8. Automatic creation of Markov models for reliability assessment of safety instrumented systems

    International Nuclear Information System (INIS)

    Guo Haitao; Yang Xianhui

    2008-01-01

    After the release of new international functional safety standards like IEC 61508, people care more for the safety and availability of safety instrumented systems. Markov analysis is a powerful and flexible technique to assess the reliability measurements of safety instrumented systems, but it is fallible and time-consuming to create Markov models manually. This paper presents a new technique to automatically create Markov models for reliability assessment of safety instrumented systems. Many safety related factors, such as failure modes, self-diagnostic, restorations, common cause and voting, are included in Markov models. A framework is generated first based on voting, failure modes and self-diagnostic. Then, repairs and common-cause failures are incorporated into the framework to build a complete Markov model. Eventual simplification of Markov models can be done by state merging. Examples given in this paper show how explosively the size of Markov model increases as the system becomes a little more complicated as well as the advancement of automatic creation of Markov models

  9. Safety assessment for the 118-B-1 Burial Ground excavation treatability tests. Revision 2

    International Nuclear Information System (INIS)

    Zimmer, J.J.; Frain, J.M.

    1994-12-01

    This revision of the Safety Assessment provides an auditable safety analysis of the hazards for the proposed treatability test activities per DOE-EM-STD-5502-94, DOE Limited Standard, Hazard Baseline Documentation (DOE 1994). The proposed activities are classified as radiological activities and as such, no longer require Operational Safety Limits (OSLs). The OSLS, Prudent Actions, and Institutional and Organization Controls have been removed from this revision and replaced with ''Administrative Actions Important to Safety,'' as determined by the hazards analysis. Those Administrative Actions Important to Safety are summarized in Section 1.1, ''Assessment Summary.''

  10. Environment, Safety and Health progress assessment of the Idaho National Engineering Laboratory (INEL)

    International Nuclear Information System (INIS)

    1993-08-01

    The ES ampersand H Progress Assessments are part of the Department's continuous improvement process throughout DOE and its contractor organizations. The purpose of the INEL ES ampersand H Progress Assessment is to provide the Department with concise independent information on the following: (1) change in culture and attitude related to ES ampersand H activities; (2) progress and effectiveness of the ES ampersand H corrective actions resulting from previous Tiger Team Assessments; (3) adequacy and effectiveness of the ES ampersand H self-assessment programs of the DOE line organizations and the site management and operating contractor; and (4) effectiveness of DOE and contractor management structures, resources, and systems to effectively address ES ampersand H problems. It is not intended that this Progress Assessment be a comprehensive compliance assessments of ES ampersand H activities. The points of reference for assessing programs at the INEL were, for the most part, the 1991 INEL Tiger Team Assessment, the INEL Corrective Action Plan, and recent appraisals and self-assessments of INEL. Horizontal and vertical reviews of the following programmatic areas were conducted: Management: Corrective action program; self-assessment; oversight; directives, policies, and procedures; human resources management; and planning, budgeting, and resource allocation. Environment: Air quality management, surface water management, groundwater protection, and environmental radiation. Safety and Health: Construction safety, worker safety and OSHA, maintenance, packaging and transportation, site/facility safety review, and industrial hygiene

  11. Safety assessment for Dragon fuel element production

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1963-11-01

    This report shall be the Safety Assessment covering the manufacture of the First Charge of Fuel and Fuel Elements for the Dragon Reactor Experiment. It is issued in two parts, of which Part I is descriptive and Part II gives the Hazards Analysis, the Operating Limitations, the Standing Orders and the Emergency Drill. (author)

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

    Science.gov (United States)

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

    2017-06-01

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

  13. Improvement of operational safety: The self-assessment at the Russian NPPs

    International Nuclear Information System (INIS)

    Kolotov, Aleksander

    2002-01-01

    The operating organization has scheduled for the forthcoming year to elaborate the NPP safety self-assessment standards and to settle precise criterion for its performance. Toward this end it was decided to form a Working Team including of NPP and VNIIAES representatives to elaborate major tasks on the self-assessment improvement, development and implementation of new documentation and training the personnel in new methodology. Actions developed by R osenergoatom , one of its items is the participation of 'Rosenergoatom' and VNIIAES representatives in the workshop, contains the sequence of Russian NPP safety self-assessment improvement at the first stage

  14. Safety assessment of discharge chute isolation barrier preparation and installation activities. Revision 3

    International Nuclear Information System (INIS)

    Meichle, R.H.

    1994-01-01

    This revision adds a section addressing impacts of dropping surfacing tool and rack cutter on the basin floor, and corrects typographical errors. 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

  15. 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).

  16. Aligning the 3Rs with new paradigms in the safety assessment of chemicals.

    Science.gov (United States)

    Burden, Natalie; Mahony, Catherine; Müller, Boris P; Terry, Claire; Westmoreland, Carl; Kimber, Ian

    2015-04-01

    There are currently several factors driving a move away from the reliance on in vivo toxicity testing for the purposes of chemical safety assessment. Progress has started to be made in the development and validation of non-animal methods. However, recent advances in the biosciences provide exciting opportunities to accelerate this process and to ensure that the alternative paradigms for hazard identification and risk assessment deliver lasting 3Rs benefits, whilst improving the quality and relevance of safety assessment. The NC3Rs, a UK-based scientific organisation which supports the development and application of novel 3Rs techniques and approaches, held a workshop recently which brought together over 20 international experts in the field of chemical safety assessment. The aim of this workshop was to review the current scientific, technical and regulatory landscapes, and to identify key opportunities towards reaching these goals. Here, we consider areas where further strategic investment will need to be focused if significant impact on 3Rs is to be matched with improved safety science, and why the timing is right for the field to work together towards an environment where we no longer rely on whole animal data for the accurate safety assessment of chemicals.

  17. Review of probabilistic safety assessments: insights and recommendations regarding further developments

    International Nuclear Information System (INIS)

    Spitzer, C.

    1996-01-01

    Probabilistic Safety Assessments (PSAs) performed by utilities in the framework of Periodic Safety Reviews for German nuclear power plants are reviewed by TUeV Suedwest. Insights gained and recommendations concerning the necessity and focus of further developments and applications according to practical requests for the performance and assessment of PSAs within regulatory procedures are presented in this paper. Further on, recommendations are made in order to ensure the validity of the results of PSAs necessary in order to achieve the goals thereof. Beside some general points of view the emphasis of the paper is on methodological aspects with respect to evaluation methods and assessment of common cause failures as well as human reliability assessment

  18. Workshop on Regulatory Review and Safety Assessment Issues in Repository Licensing

    International Nuclear Information System (INIS)

    Wilmot, Roger D.

    2011-02-01

    The workshop described here was organised to address more general issues regarding regulatory review of SKB's safety assessment and overall review strategy. The objectives of the workshop were: - to learn from other programmes' experiences on planning and review of a license application for a nuclear waste repository, - to offer newly employed SSM staff an opportunity to learn more about selected safety assessment issues, and - to identify and document recommendations and ideas for SSM's further planning of the licensing review

  19. Concrete structures. Contribution to the safety assessment of existing structures

    Directory of Open Access Journals (Sweden)

    D. COUTO

    Full Text Available The safety evaluation of an existing concrete structure differs from the design of new structures. The partial safety factors for actions and resistances adopted in the design phase consider uncertainties and inaccuracies related to the building processes of structures, variability of materials strength and numerical approximations of the calculation and design processes. However, when analyzing a finished structure, a large number of unknown factors during the design stage are already defined and can be measured, which justifies a change in the increasing factors of the actions or reduction factors of resistances. Therefore, it is understood that safety assessment in existing structures is more complex than introducing security when designing a new structure, because it requires inspection, testing, analysis and careful diagnose. Strong knowledge and security concepts in structural engineering are needed, as well as knowledge about the materials of construction employed, in order to identify, control and properly consider the variability of actions and resistances in the structure. With the intention of discussing this topic considered complex and diffuse, this paper presents an introduction to the safety of concrete structures, a synthesis of the recommended procedures by Brazilian standards and another codes, associated with the topic, as well a realistic example of the safety assessment of an existing structure.

  20. Terrain Safety Assessment in Support of the Mars Science Laboratory Mission

    Science.gov (United States)

    Kipp, Devin

    2012-01-01

    In August 2012, the Mars Science Laboratory (MSL) mission will pioneer the next generation of robotic Entry, Descent, and Landing (EDL) systems by delivering the largest and most capable rover to date to the surface of Mars. The process to select the MSL landing site took over five years and began with over 50 initial candidate sites from which four finalist sites were chosen. The four finalist sites were examined in detail to assess overall science merit, EDL safety, and rover traversability on the surface. Ultimately, the engineering assessments demonstrated a high level of safety and robustness at all four finalist sites and differences in the assessment across those sites were small enough that neither EDL safety nor rover traversability considerations could significantly discriminate among the final four sites. Thus the MSL landing site at Gale Crater was selected from among the four finalists primarily on the basis of science considerations.

  1. Safety and performance indicators for the assessment of long-term safety of deep geological disposal of radioactive waste

    International Nuclear Information System (INIS)

    Hugi, M.; Schneider, J.W.; Dorp, F. van; Zuidema, P.

    2005-01-01

    The evaluation of the ability to isolate radioactive waste and the assessment of the long-term safety of a deep geological repository is usually done in terms of the calculated dose and/or risk for an average individual of the population which is potentially most affected by the potential impacts of the repository. At present, various countries and international organisations are developing so-called complementary indicators to supplement such calculations. These indicators are called ''safety indicators'' if they refer to the safety of the whole repository system; if they address the isolation capability of individual system components or the whole system from a more technical perspective, they are called ''performance indicators''. The need for complementary indicators follows from the long time frames which characterise the safety assessment of a geological repository, and the corresponding uncertainty of the calculated radiation dose. The main reason for these uncertainties is associated with the uncertain long-term prognosis of the surface environment and the related human behaviour. (orig.)

  2. Tank waste remediation system nuclear criticality safety inspection and assessment plan

    International Nuclear Information System (INIS)

    VAIL, T.S.

    1999-01-01

    This plan provides a management approved procedure for inspections and assessments of sufficient depth to validate that the Tank Waste Remediation System (TWRS) facility complies with the requirements of the Project Hanford criticality safety program, NHF-PRO-334, ''Criticality Safety General, Requirements''

  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. Assessment of oral toxicity and safety of pentamethylchromanol (PMCol), a potential chemopreventative agent, in rats and dogs

    International Nuclear Information System (INIS)

    Lindeblad, Matthew; Kapetanovic, Izet M.; Kabirov, Kasim K.; Detrisac, Carol J.; Dinger, Nancy; Mankovskaya, Irina; Zakharov, Alexander; Lyubimov, Alexander V.

    2010-01-01

    2,2,5,7,8-Pentamethyl-6-chromanol (PMCol) was administered by gavage in rats for 28 days at dose levels of 0, 100, 500, and 2000 mg/kg/day. PMCol administration induced decreases in body weight gains and food consumption, hepatotoxicity (increased TBILI, ALB, ALT, TP; increased relative liver weights; increased T4 and TSH), nephrotoxicity (increased BUN and BUN/CREAT, histopathology lesions), effect on lipid metabolism (increased CHOL), anemia, increase in WBC counts (total and differential), coagulation (FBGN↑and PT↓) and hyperkeratosis of the nonglandular stomach in the 2000 mg/kg/day dose group (in one or both sexes). In the 500 mg/kg/day dose group, toxicity was seen to a lesser extent. In the 100 mg/kg/day dose group, only increased CHOL (females) was observed. To assess the toxicity of PMCol in male dogs it was administered orally by capsule administration for 28 days at dose levels of 0, 50, 200 and 800 mg/kg/day (four male dogs/dose group). PMCol treatment at 800 mg/kg/day resulted in pronounced toxicity to the male dogs. Target organs of toxicity were liver and thymus. Treatment at 200 mg/kg/day resulted in toxicity consistent with slight adverse effect on the liver only. The results of the safety pharmacology study indicate that doses of 0, 50, 200 and 800 mg/kg administered orally did not have an effect on the QT interval, blood pressures and body temperatures following dosing over a 24-h recording period. Under the conditions of this study, the no-observed-adverse effect level (NOAEL) for daily oral administration of PMCol by gavage for 28 days to male rats was 100 mg/kg/day and 50 mg/kg in male dogs. In female rats, the NOAEL was not established due to statistically significant and biologically meaningful increases in CHOL level seen in the 100 mg/kg/day dose group. The results of these studies indicated that administration of PMCol at higher dose levels resulted in severe toxicity in dogs and moderate toxicity in rats, however, administration at

  5. Evaluation of changes in serum chemistry in association with feed withdrawal or high dose oral gavage with Dextran Sodium Sulfate (DSS) induced gut leakage in broiler chickens

    Science.gov (United States)

    Dextran sodium sulfate (DSS) has been shown to be effective at inducing enteric inflammation in broiler chickens, resulting in increased leakage of orally administered fluorescein isothiocyanate dextran to circulation. In a previous study, two doses of DSS (0.45g/dose) administered as oral gavage re...

  6. Early Safety Assessment of Automotive Systems Using Sabotage Simulation-Based Fault Injection Framework

    OpenAIRE

    Juez, Garazi; Amparan, Estíbaliz; Lattarulo, Ray; Ruíz, Alejandra; Perez, Joshue; Espinoza, Huascar

    2017-01-01

    As road vehicles increase their autonomy and the driver reduces his role in the control loop, novel challenges on dependability assessment arise. Model-based design combined with a simulation-based fault injection technique and a virtual vehicle poses as a promising solution for an early safety assessment of automotive systems. To start with, the design, where no safety was considered, is stimulated with a set of fault injection simulations (fault forecasting). By doing so, safety strategies ...

  7. A comparison of integrated safety analysis and probabilistic risk assessment

    International Nuclear Information System (INIS)

    Damon, Dennis R.; Mattern, Kevin S.

    2013-01-01

    The U.S. Nuclear Regulatory Commission conducted a comparison of two standard tools for risk informing the regulatory process, namely, the Probabilistic Risk Assessment (PRA) and the Integrated Safety Analysis (ISA). PRA is a calculation of risk metrics, such as Large Early Release Frequency (LERF), and has been used to assess the safety of all commercial power reactors. ISA is an analysis required for fuel cycle facilities (FCFs) licensed to possess potentially critical quantities of special nuclear material. A PRA is usually more detailed and uses more refined models and data than an ISA, in order to obtain reasonable quantitative estimates of risk. PRA is considered fully quantitative, while most ISAs are typically only partially quantitative. The extension of PRA methodology to augment or supplant ISAs in FCFs has long been considered. However, fuel cycle facilities have a wide variety of possible accident consequences, rather than a few surrogates like LERF or core damage as used for reactors. It has been noted that a fuel cycle PRA could be used to better focus attention on the most risk-significant structures, systems, components, and operator actions. ISA and PRA both identify accident sequences; however, their treatment is quite different. ISA's identify accidents that lead to high or intermediate consequences, as defined in 10 Code of Federal Regulations (CFR) 70, and develop a set of Items Relied on For Safety (IROFS) to assure adherence to performance criteria. PRAs identify potential accident scenarios and estimate their frequency and consequences to obtain risk metrics. It is acceptable for ISAs to provide bounding evaluations of accident consequences and likelihoods in order to establish acceptable safety; but PRA applications usually require a reasonable quantitative estimate, and often obtain metrics of uncertainty. This paper provides the background, features, and methodology associated with the PRA and ISA. The differences between the

  8. German data for risk based fire safety assessment

    International Nuclear Information System (INIS)

    Roewekamp, M.; Berg, H.P.

    1998-01-01

    Different types of data are necessary to perform risk based fire safety assessments and, in particular, to quantify the fire event tree considering the plant specific conditions. Data on fire barriers, fire detection and extinguishing, including also data on secondary effects of a fire, have to be used for quantifying the potential hazard and damage states. The existing German database on fires in nuclear power plants (NPPs) is very small. Therefore, in general generic data, mainly from US databases, are used for risk based safety assessments. Due to several differences in the plant design and conditions generic data can only be used as conservative assumptions. World-wide existing generic data on personnel failures in case of fire fighting have only to be adapted to the plant specific conditions inside the NPP to be investigated. In contrary, unavailabilities of fire barrier elements may differ strongly depending on different standards, testing requirements, etc. In addition, the operational behaviour of active fire protection equipment may vary depending on type and manufacturer. The necessity for more detailed and for additional plant specific data was the main reason for generating updated German data on the operational behaviour of active fire protection equipment/features in NPPs to support risk based fire safety analyses being recommended to be carried out as an additional tool to deterministic fire hazard analyses in the frame of safety reviews. The results of these investigations revealed a broader and more realistic database for technical reliability of active fire protection means, but improvements as well as collection of further data are still necessary. (author)

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

    Science.gov (United States)

    Zelkin, Natalie; Henriksen, Stephen

    2011-01-01

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

  10. Safety Assessment of Multi Purpose Small Payload Rack(MSPR)

    Science.gov (United States)

    Mizutani, Yoshinobu; Takada, Satomi; Murata, Kosei; Ozawa, Daisaku; Kobayashi, Ryoji; Nakamura, Yasuhiro

    2010-09-01

    We are reporting summary of preliminary safety assessment for Multi Purpose Small Payload Rack(MSPR), which is one of the micro gravity experiment facilities that are being developed for the 2nd phase JEM utilization(JEM: Japanese Experiment Module) that will be launched on H-II Transfer Vehicle(HTV) 2nd flight in 2011. MSPR is used for multi-purpose micro-g experiment providing experimental spaces and work stations. MSPR has three experimental spaces; first, there is a space called Work Volume(WV) with capacity volume of approximately 350 litters, in which multiple resources including electricity, communication, and moving image functions can be used. Within this space, installation of devices can be done by simple, prompt attachment by Velcro and pins with high degree of flexibility. Second, there is Small Experiment Area(SEA), with capacity volume of approximately 70 litters, in which electricity, communication, and moving image functions can also be used in the same way as WV. These spaces protect experiment devices and specimens from contingent loads by the crewmembers. Third, there is Work Bench with area of 0.5 square meters, on which can be used for maintenance, inspection and data operations of installed devices, etc. This bench can be stored in the rack during contingency. Chamber for Combustion Experiment(CCE) that is planned to be installed in WV is a pressure-resistant experimental container that can be used to seal hazardous materials from combustion experiments. This CCE has double sealing design in chamber itself, which resist gas leakage under normal the temperature and pressure. Electricity, communication, moving image function can be used in the same way as WV. JAXA Phase 2 Safety Review Panel(SRP) has been held in April, 2010. For safety analysis of MSPR, hazards were identified based on Fault Tree Analysis methodology and then these hazards were classified into either eight ISS standard-type hazards or eight unique-type hazards that requires

  11. 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.

  12. Fuel and canister process report for the safety assessment SR-Can

    International Nuclear Information System (INIS)

    Werme, Lars

    2006-10-01

    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-Can. The detailed assessment methodology, including the role of the process report in the assessment, is described in the SR-Can Main report. The report is written by, and for, experts in the relevant scientific fields. It should though be possible for a generalist in the area of long-term safety assessments of geologic nuclear waste repositories to comprehend the contents of the report. The report is an important part of the documentation of the SR-Can project and an essential reference within the project, providing a scientifically motivated plan for the handling of geosphere processes. It is, furthermore, foreseen that the report will be essential for reviewers scrutinising the handling of geosphere issues in the SR-Can assessment. Several types of fuel will be emplaced in the repository. For the reference case with 40 years of reactor operation, the fuel quantity from boiling water reactors, BWR fuel, is estimated at 7,000 tonnes, while the quantity from pressurized water reactors, PWR fuel, is estimated at about 2,300 tonnes. In addition, 23 tonnes of mixed-oxide fuel (MOX) fuel of German origin from BWR and PWR reactors and 20 tonnes of fuel from the decommissioned heavy water reactor in Aagesta will be disposed of. To allow for future changes in the Swedish nuclear programme, the safety assessment assumes a total of 6,000 canister corresponding to 12,000 tonnes of fuel

  13. Fuel and canister process report for the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Werme, Lars (ed.)

    2006-10-15

    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-Can. The detailed assessment methodology, including the role of the process report in the assessment, is described in the SR-Can Main report. The report is written by, and for, experts in the relevant scientific fields. It should though be possible for a generalist in the area of long-term safety assessments of geologic nuclear waste repositories to comprehend the contents of the report. The report is an important part of the documentation of the SR-Can project and an essential reference within the project, providing a scientifically motivated plan for the handling of geosphere processes. It is, furthermore, foreseen that the report will be essential for reviewers scrutinising the handling of geosphere issues in the SR-Can assessment. Several types of fuel will be emplaced in the repository. For the reference case with 40 years of reactor operation, the fuel quantity from boiling water reactors, BWR fuel, is estimated at 7,000 tonnes, while the quantity from pressurized water reactors, PWR fuel, is estimated at about 2,300 tonnes. In addition, 23 tonnes of mixed-oxide fuel (MOX) fuel of German origin from BWR and PWR reactors and 20 tonnes of fuel from the decommissioned heavy water reactor in Aagesta will be disposed of. To allow for future changes in the Swedish nuclear programme, the safety assessment assumes a total of 6,000 canister corresponding to 12,000 tonnes of fuel.

  14. 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...

  15. Food Safety Practices Assessment Tool: An Innovative Way to Test Food Safety Skills among Individuals with Special Needs

    Science.gov (United States)

    Carbone, Elena T.; Scarpati, Stanley E.; Pivarnik, Lori F.

    2013-01-01

    This article describes an innovative assessment tool designed to evaluate the effectiveness of a food safety skills curriculum for learners receiving special education services. As schools respond to the increased demand for training students with special needs about food safety, the need for effective curricula and tools is also increasing. A…

  16. Ensuring a proactive, evidence-based, patient safety approach to patient assessment.

    Science.gov (United States)

    Considine, Julie; Currey, Judy

    2015-01-01

    To argue that if all nurses were to adopt the primary survey approach (assessment of airway, breathing, circulation and disability) as the first element of patient assessment, they would be more focused on active detection of clinical deterioration rather than passive collection of patient data. Nurses are the professional group that carry the highest level of responsibility for patient assessment, accurate data collection and interpretation. The timely recognition of, and response to deteriorating patients, is dependent on the measurement and interpretation of pertinent physiological data by nurses. Discursive paper. Traditionally taught and commonly used approaches to patient assessment such as 'vital signs' and 'body systems' are not evidence-based nor framed in patient safety. The primary survey approach as the first element in patient assessment has three major advantages: (1) data are collected according to clinical importance; (2) data are collected using the same framework as most organisation's rapid response system activation criteria; and (3) the primary survey acts as a patient safety checklist, thereby decreasing the risk of failure to recognise, and therefore respond to, deteriorating patients. The vital signs and body systems approaches to patient assessment have significant limitations in identifying clinical deterioration. The primary survey approach provides nurses with a consistent, evidence-based and sequenced approach to patient assessment in every clinical setting. All nurses should use a primary survey approach as the first element of patient assessment in every patient encounter as a patient safety strategy. © 2014 John Wiley & Sons Ltd.

  17. 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

  18. Diversity for security: case assessment for FPGA-based safety-critical systems

    Directory of Open Access Journals (Sweden)

    Kharchenko Vyacheslav

    2016-01-01

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

  19. Planning report for the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-06-01

    This document is a planning report for SKB's next assessment of long-term safety for a KBS 3 repository. The assessment, SR-Can, is to be finished by the end of 2005 and will be used for SKB's application to build an Encapsulation plant for spent nuclear fuel. Apart from outlining the methodology, the report discusses the handling in SR-Can of a number of important issues regarding the near field, the geosphere, the biosphere, the climatic evolution etc. The Swedish nuclear safety and radiation protection authorities have recently issued regulations concerning the final disposal of nuclear waste. The principal compliance criterion states that the annual risk of harmful effects must not exceed 10{sup -6} for a representative individual in the group exposed to the greatest risk. There are also a number of requirements on methodological aspects of the safety assessment as well as on the contents of a safety report. The regulations are reproduced in an Appendix to this report. The primary safety function of the KBS 3 system is to completely isolate the spent nuclear fuel within copper canisters over the entire assessment period, which will be one million years in SR-Can. Should a canister be damaged, the secondary safety function is to retard any releases from the canisters. The main steps of the assessment are the following: 1. Qualitative system description, FEP processing: This step consists of defining a system boundary and of describing the system on a format suitable for the safety assessment. Databases of relevant features, events and processes influencing long-term safety are structured and used as one starting point for the assessment. 2. Initial state descriptions. 3. Process descriptions: In this step all identified processes within the system boundary involved in the long-term evolution of the system are described in detail. 4. Description of boundary conditions: This step is a broad description of the evolution of the boundaries of the system

  20. SCART guidelines. Reference report for IAEA Safety Culture Assessment Review Team (SCART)

    International Nuclear Information System (INIS)

    2008-01-01

    The IAEA Director General stressed the role of safety culture in his concluding remarks at the Meeting of the Contracting Parties to the Convention on Nuclear Safety in 2002: 'As we have learned in other areas, it is not enough simply to have a structure; it is not enough to say that we have the necessary laws and the appropriate regulatory bodies. All these are important, but equally important is that we have in place a safety culture that gives effect to the structure that we have developed. To me, effectiveness and transparency are keys. So, it is an issue which I am pleased to see, you are giving the attention it deserves and we will continue to work with you in clarifying, developing and applying safety culture through our programmes and through our technical cooperation activities.' The concept of safety culture was initially developed by the International Nuclear Safety Advisory Group (INSAG) after the Chernobyl accident in 1986. Since then the IAEA's perspective of safety culture has expanded with time as its recognition of the complexities of the concept developed. Safety culture is considered to be specific organizational culture in all types of organizations with activities that give rise to radiation risks. The aim is to make safety culture strong and sustainable, so that safety becomes a primary focus for all activities in such organizations, even for those, which might not look safety-related at first. SCART (Safety Culture Assessment Review Team) is a safety review service, which reflects the expressed interest of Members States for methods and tools for safety culture assessment. It is a replacement for the earlier service ASCOT (Assessment of Safety Culture in Organizations Team). The IAEA Safety Fundamentals, Requirements and Guides (Safety Standards) are the basis for the SCART Safety Review Service. The reports of INSAG, identifying important current nuclear safety issues, serve also as references during a SCART mission. SCART missions are based

  1. Liquefied Gaseous Fuels Safety and Environmental Control Assessment Program: second status report

    Energy Technology Data Exchange (ETDEWEB)

    1980-10-01

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

  2. Regulatory review and confidence building in post-closure safety assessments and safety cases for near surface disposal facilities-IAEA ASAM coordinated research programme

    International Nuclear Information System (INIS)

    Gonzales, A.; Simeonov, G.; Bennett, D.G.; Nys, V.; Ben Belfadhel, M.

    2005-01-01

    Some years ago, the IAEA successfully concluded a Coordinated Research Program (CRP) called Islam, which focussed on the development of an Improved Safety Assessment Methodology for near-surface radioactive waste disposal facilities. In November 2002, and as an extension of ISAM, the IAEA launched a new CRP called ASAM, designed to test the Application of the Safety Assessment Methodology by considering a range of near-surface disposal facilities. The ASAM work programme is being implemented by three application working groups and two cross-cutting working groups. The application working groups are testing the applicability of the ISAM methodology by assessing an existing disposal facility in Hungary, a copper mine in South Africa, and a hypothetical facility containing heterogenous wastes, such as disused sealed sources. The first cross-cutting working group is addressing a number of technical issues that are common to all near-surface disposal facilities, while the second group, the Regulatory Review Working Group (RRWG) is developing guidance on how to gain confidence in safety assessments and safety cases, and on how to conduct regulatory reviews of safety assessments. This paper provides a brief overview of the work being conducted by the Regulatory Review Working Group. (author)

  3. Assessment of occupational exposure due to intakes of radionuclides. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

  4. Assessment of occupational exposure due to intakes of radionuclides. Safety guide

    International Nuclear Information System (INIS)

    2000-01-01

    Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

  5. Assessment of occupational exposure due to intakes of radionuclides. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

  6. Assessment of occupational exposure due to intakes of radionuclides. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

  7. Assessment of occupational exposure due to intakes of radionuclides. Safety guide

    International Nuclear Information System (INIS)

    1999-01-01

    Occupational exposure due to radioactive materials can occur as a result of various human activities. These include work associated with the different stages of the nuclear fuel cycle, the use of radioactive sources in medicine, scientific research, agriculture and industry, and occupations which involve the handling of materials containing enhanced concentrations of naturally occurring radionuclides. In order to control this exposure, it is necessary to be able to assess the magnitude of the doses involved. Three interrelated Safety Guides, prepared jointly by the IAEA and the International Labour Office (ILO), provide guidance on the application of the requirements of the Basic Safety Standards with respect to occupational exposure. Reference [3] gives general advice on the exposure conditions for which monitoring programmes should be set up to assess radiation doses arising from external radiation and from intakes of radionuclides by workers. More specific guidance on the assessment of doses from external sources of radiation can be found in Ref. [4] and the present Safety Guide deals with intakes of radioactive materials. Recommendations related to occupational radiation protection have also been developed by the International Commission on Radiological Protection (ICRP) [5]. These and other current recommendations of the ICRP [6] have been taken into account in preparing this Safety Guide. The purpose of this Safety Guide is to provide guidance for regulatory authorities on conducting assessments of intakes of radioactive material arising from occupational exposure. This Guide will also be useful to those concerned with the planning, management and operation of occupational monitoring programmes, and to those involved in the design of equipment for use in internal dosimetry and workplace monitoring

  8. Effects of miglyol 812 on rats after 4 weeks of gavage as compared with methylcellulose/tween 80.

    Science.gov (United States)

    Sellers, Rani S; Antman, M; Phillips, J; Khan, K N; Furst, S M

    2005-01-01

    Miglyol 812 is a medium-chain triglyceride used in toxicology studies as an excipient to improve test compound solubility/absorption. As part of a larger toxicology study, 15 Wistar Han IGS rats/sex/group were dosed by oral gavage for 4 weeks with 10 mL kg(-1) day(-1) of 100% Miglyol 812 or 0.5% methylcellulose/0.1% Tween 80 in water (MC-T) followed by 4 weeks without treatment to evaluate the potential effects of this excipient in long-term toxicology studies relative to a traditional excipient such as MC-T. Clinical signs evident during the dosing phase included soft and/or mucoid stool in 12/15 males and 11/15 females treated with Miglyol 812 but in no animals treated with MC-T. Animals treated with Miglyol 812 had a 6-7% statistically significant reduction in body weight gain as compared to MC-T-treated animals. Statistically significant changes in clinical chemistry parameters as compared to MC-T included decreased blood urea nitrogen (50% and 29% in males and females, respectively), increased cholesterol (1.6-fold and 1.5-fold in males and females, respectively), decreased total protein (6% and 8% in males and females, respectively), decreased globulins (15% and 11% in males and females, respectively), and increased triglycerides (2.8-fold and 1.7-fold in males and females, respectively). Absolute and relative thymic weights decreased 28% and 24%, respectively, in males, and 18% and 17%, respectively, in females without histological alterations. Histopathology revealed increased alveolar histiocytosis with focal interstitial inflammation in lungs in 5/10 males and 7/10 females treated with Miglyol 812 compared to only 1/10 males and 1/10 females treated with MC-T. All effects were reversible during the recovery period. Results of this study indicate that 100% miglyol 812 produces reversible gastrointestinal effects and decreases in body weight gains along with changes in several serum chemistry parameters. Therefore, it should not be considered innocuous when

  9. 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.

  10. Flightdeck Automation Problems (FLAP) Model for Safety Technology Portfolio Assessment

    Science.gov (United States)

    Ancel, Ersin; Shih, Ann T.

    2014-01-01

    NASA's Aviation Safety Program (AvSP) develops and advances methodologies and technologies to improve air transportation safety. The Safety Analysis and Integration Team (SAIT) conducts a safety technology portfolio assessment (PA) to analyze the program content, to examine the benefits and risks of products with respect to program goals, and to support programmatic decision making. The PA process includes systematic identification of current and future safety risks as well as tracking several quantitative and qualitative metrics to ensure the program goals are addressing prominent safety risks accurately and effectively. One of the metrics within the PA process involves using quantitative aviation safety models to gauge the impact of the safety products. This paper demonstrates the role of aviation safety modeling by providing model outputs and evaluating a sample of portfolio elements using the Flightdeck Automation Problems (FLAP) model. The model enables not only ranking of the quantitative relative risk reduction impact of all portfolio elements, but also highlighting the areas with high potential impact via sensitivity and gap analyses in support of the program office. Although the model outputs are preliminary and products are notional, the process shown in this paper is essential to a comprehensive PA of NASA's safety products in the current program and future programs/projects.

  11. 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)

  12. Toxicokinetics of α-thujone following intravenous and gavage administration of α-thujone or α- and β-thujone mixture in male and female F344/N rats and B6C3F1 mice

    International Nuclear Information System (INIS)

    Waidyanatha, Suramya; Johnson, Jerry D.; Hong, S. Peter; Robinson, Veronica Godfrey; Gibbs, Seth; Graves, Steven W.; Hooth, Michelle J.; Smith, Cynthia S.

    2013-01-01

    Plants containing thujone have widespread use and hence have significant human exposure. α-Thujone caused seizures in rodents following gavage administration. We investigated the toxicokinetics of α-thujone in male and female F344/N rats and B6C3F1 mice following intravenous and gavage administration of α-thujone or a mixture of α- and β-thujone (which will be referred to as α,β-thujone). Absorption of α-thujone following gavage administration was rapid without any dose-, species-, sex- or test article-related effect. Absolute bioavailability of α-thujone following administration of α-thujone or α,β-thujone was generally higher in rats than in mice. In rats, females had higher bioavailability than males following administration of either test article although a sex difference was not observed in mice. C max and AUC ∞ increased greater than proportional to the dose in female rats following administration of α-thujone and in male and female mice following administration of α,β-thujone suggesting possible saturation of elimination kinetics with increasing dose. Dose-adjusted AUC ∞ for male and female rats was 5- to 15-fold and 3- to 24-fold higher than mice counterparts following administration of α-thujone and α,β-thujone, respectively (p-value < 0.0001 for all comparisons). Following both intravenous and gavage administration, α-thujone was distributed to the brains of rats and mice with females, in general, having higher brain:plasma ratios than males. These data are in support of the observed toxicity of α-thujone and α,β-thujone where females were more sensitive than males of both species to α-thujone-induced neurotoxicity. In general there was no difference in toxicokinetics between test articles when normalized to α-thujone concentration. - Highlights: • Absorption of α-thujone following gavage administration was rapid in rats and mice. • Rats undergo higher exposure to α-thujone than mice. • α-Thujone brain:plasma ratios

  13. Toxicokinetics of α-thujone following intravenous and gavage administration of α-thujone or α- and β-thujone mixture in male and female F344/N rats and B6C3F1 mice

    Energy Technology Data Exchange (ETDEWEB)

    Waidyanatha, Suramya, E-mail: waidyanathas@niehs.nih.gov [Division of National Toxicology Program, National Institute of Environmental Health Sciences, Research Triangle Park, NC 27709 (United States); Johnson, Jerry D.; Hong, S. Peter [Battelle Memorial Institute, Columbus, OH 43201 (United States); Robinson, Veronica Godfrey [Division of National Toxicology Program, National Institute of Environmental Health Sciences, Research Triangle Park, NC 27709 (United States); Gibbs, Seth; Graves, Steven W. [Battelle Memorial Institute, Columbus, OH 43201 (United States); Hooth, Michelle J.; Smith, Cynthia S. [Division of National Toxicology Program, National Institute of Environmental Health Sciences, Research Triangle Park, NC 27709 (United States)

    2013-09-01

    Plants containing thujone have widespread use and hence have significant human exposure. α-Thujone caused seizures in rodents following gavage administration. We investigated the toxicokinetics of α-thujone in male and female F344/N rats and B6C3F1 mice following intravenous and gavage administration of α-thujone or a mixture of α- and β-thujone (which will be referred to as α,β-thujone). Absorption of α-thujone following gavage administration was rapid without any dose-, species-, sex- or test article-related effect. Absolute bioavailability of α-thujone following administration of α-thujone or α,β-thujone was generally higher in rats than in mice. In rats, females had higher bioavailability than males following administration of either test article although a sex difference was not observed in mice. C{sub max} and AUC{sub ∞} increased greater than proportional to the dose in female rats following administration of α-thujone and in male and female mice following administration of α,β-thujone suggesting possible saturation of elimination kinetics with increasing dose. Dose-adjusted AUC{sub ∞} for male and female rats was 5- to 15-fold and 3- to 24-fold higher than mice counterparts following administration of α-thujone and α,β-thujone, respectively (p-value < 0.0001 for all comparisons). Following both intravenous and gavage administration, α-thujone was distributed to the brains of rats and mice with females, in general, having higher brain:plasma ratios than males. These data are in support of the observed toxicity of α-thujone and α,β-thujone where females were more sensitive than males of both species to α-thujone-induced neurotoxicity. In general there was no difference in toxicokinetics between test articles when normalized to α-thujone concentration. - Highlights: • Absorption of α-thujone following gavage administration was rapid in rats and mice. • Rats undergo higher exposure to α-thujone than mice. • α-Thujone brain

  14. Workshop on Regulatory Review and Safety Assessment Issues in Repository Licensing

    Energy Technology Data Exchange (ETDEWEB)

    Wilmot, Roger D. (Galson Sciences Limited (United Kingdom))

    2011-02-15

    The workshop described here was organised to address more general issues regarding regulatory review of SKB's safety assessment and overall review strategy. The objectives of the workshop were: - to learn from other programmes' experiences on planning and review of a license application for a nuclear waste repository, - to offer newly employed SSM staff an opportunity to learn more about selected safety assessment issues, and - to identify and document recommendations and ideas for SSM's further planning of the licensing review

  15. Safety Assessment Context for Croatian Low and Intermediate Level Radioactive Waste Repository

    International Nuclear Information System (INIS)

    Levanat, I.; Lokner, V.

    1998-01-01

    Safety assessments in a small country are usually performed to support the national waste management strategy, demonstrating compliance with national regulation for a particular facility. However, this assessment should - quite generally - provide reasonable assurance both to the public and to decision makers than the Croatian share of LILW from NPP Krsko can be safely disposed in Croatia. More specifically, assessment should clearly present all realistic options and compare the associated long term repository performances, demonstrating that desirable safety goals can be archived by an appropriate choice of (a) location, (b) facility design, (c) institutional control period and (d) waste acceptance criteria. As relevant national legislation is presently under review, generally recognized international safety standards, criteria and recommendations (e.g. as presented in the recent IAEA publications) should provide guidance for the assessment evaluation, since it is expected that they will be incorporated in the new national regulations. Finally, since Croatian radioactive waste management strategy is yet to be developed, such an assessment may contribute to its formulation and facilitate some specific decisions. (author)

  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. Complementary safety assessment assessment of nuclear facilities - FBFC Romans plant - AREVA

    International Nuclear Information System (INIS)

    2011-01-01

    This complementary safety assessment analyses the robustness of the FBFC Romans plant to extreme situations such as those that led to the Fukushima accident. This plant is dedicated to the fabrication of nuclear fuels for experimental reactors. 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 4 main measures to be taken to limit the risks linked to natural disasters: -) the implementation of a seismic detection and cutting system; -) the seismic reinforcement of the recycling workshop (R1 building); -) the suppression of the use of recycled water in the AP2 building; -) the determination of the critical water levels admitted in the buildings in case of strong rain periods. (A.C.)

  18. 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

  19. Study of the teratogenic potential of FD & C Yellow No. 5 when given by gavage to rats.

    Science.gov (United States)

    Collins, T F; Black, T N; Brown, L H; Bulhack, P

    1990-12-01

    FD & C Yellow No. 5 (tartrazine) was given to Osborne-Mendel rats by gavage at dose levels of 0, 60, 100, 200, 400, 600 or 1000 mg/kg body weight/day on days 0-19 of gestation. No maternal or developmental toxicity was observed when the rats were killed on day 20. The mean daily food consumption for the entire period of gestation was significantly greater in the females given 1000 mg/kg body weight/day than in the controls, but maternal body-weight gain was not affected. No dose-related effects were observed in implantations, foetal viability or external foetal development. Foetal skeletal and visceral development was similar among foetuses from all groups. At the doses given, FD & C Yellow No. 5 was neither toxic nor teratogenic.

  20. Environment, Safety and Health Progress Assessment of the Hanford Site

    International Nuclear Information System (INIS)

    1992-05-01

    This report documents the result of the US Department of Energy (DOE) Environment, Safety and Health (ES ampersand H) Progress Assessment of the Hanford Site, in Richland, Washington. The assessment, which was conducted from May 11 through May 22, 1992, included a selective-review of the ES ampersand H management systems and programs of the responsible DOE Headquarters Program Offices the DOE Richland Field Office, and the site contractors. The ES ampersand H Progress Assessments are part of the Secretary of Energy's continuing effort to institutionalize line management accountability and the self-assessment process throughout DOE and its contractor organizations. The purpose of the Hanford Site ES ampersand H Progress Assessment is to provide the Secretary with an independent assessment of the adequacy and effectiveness of the DOE and contractor management structures, resources, and systems to address ES ampersand H problems and requirements. They are not intended to be comprehensive compliance assessments of ES ampersand H activities. The point of reference for assessing programs at the Hanford Site was, for the most part, the Tiger Team Assessment of the Hanford Site, which was conducted from May 21 through July 18, 1990. A summary of issues and progress in the areas of environment, safety and health, and management is included

  1. Plasma, a plant safety monitoring and assessment system for VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hornaes, A.; Hulsund, J. E. [Institutt for energiteknikk (IFE), OECD Halden Reactor Project, Halden (Norway); Lipcsei, S.; Major, Cs.; Racz, A.; Vegh, J. [KFKI, Atomic Energy Research Institute, Budapest (Hungary); Eiler, J. [Paks, Nuclear Power Plant Ltd, Paks (Hungary)

    1999-05-15

    The objective with the Plant Safety Monitoring and Assessment System (PLASMA) is to develop an operator support system to support the execution of new symptom-based Emergency Operating Procedures for application in VVER reactors, with the Paks NPP in Hungary as the target plant. Many of the VVER reactors are rewriting their EOPs to comply more with Western standards of symptom-based EOPs. In this connection it is desirable to improve the data validation, information integration and presentation for operators when executing the EOPs. The entry-point to a symptom-oriented procedure is defined by the occurrence of a well-defined reactor operation status, with all its symptoms. However, the application of the EOF benefits from an operator support system, which performs plant status and symptom identification reliably and accurately. The development of the PLASMA system is a joint venture between Institutt for energiteknikk (IFE) and KFKI with the NPP Paks as the target plant. The project has been initiated and partly funded by the Science and Technology Agency (STA), Japan through the OECD NEA assistance program. In Hungary, considerable effort has concentrated on the safety reassessment of the Paks NPP and new EOPs are being written, but no comprehensive Operator Support System (OSS) for plant safety assessment is installed. Some safety parameter display functions are incorporated into diverse operator support systems, but an online 'plant safety monitoring and assessment system' is still missing. The present project comprises designing, constructing, testing and installing such an OSS, which to a great extent could support plant operators in their safety assessment work (author) (ml)

  2. The in-depth safety assessment (ISA) pilot projects in Ukraine

    International Nuclear Information System (INIS)

    Kot, C. A.

    1998-01-01

    Ukraine operates pressurized water reactors of the Soviet-designed type, VVER. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs). After approval of the SARS by the Ukrainian Nuclear Regulatory Authority, the plants will be granted longer-term operating licenses. In September 1995, the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine issued a new contents requirement for the safety analysis reports of VVERs in Ukraine. It contains requirements in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The DBA requirements are an expanded version of the older SAR requirements. The last two requirements, on PRA and BDBA, are new. The US Department of Energy (USDOE), through the International Nuclear Safety Program (INSP), has initiated an assistance and technology transfer program to Ukraine to assist their nuclear power stations in developing a Western-type technical basis for the new SARS. USDOE sponsored In-Depth Safety Assessments (ISAs) have been initiated at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1. USDOE/INSP have structured the ISA program in such a way as to provide maximum assistance and technology transfer to Ukraine while encouraging and supporting the Ukrainian plants to take the responsibility and initiative and to perform the required assessments

  3. Planning the Unplanned Experiment: Assessing the Efficacy of Standards for Safety Critical Software

    Science.gov (United States)

    Graydon, Patrick J.; Holloway, C. Michael

    2015-01-01

    We need well-founded means of determining whether software is t for use in safety-critical applications. While software in industries such as aviation has an excellent safety record, the fact that software aws have contributed to deaths illustrates the need for justi ably high con dence in software. It is often argued that software is t for safety-critical use because it conforms to a standard for software in safety-critical systems. But little is known about whether such standards `work.' Reliance upon a standard without knowing whether it works is an experiment; without collecting data to assess the standard, this experiment is unplanned. This paper reports on a workshop intended to explore how standards could practicably be assessed. Planning the Unplanned Experiment: Assessing the Ecacy of Standards for Safety Critical Software (AESSCS) was held on 13 May 2014 in conjunction with the European Dependable Computing Conference (EDCC). We summarize and elaborate on the workshop's discussion of the topic, including both the presented positions and the dialogue that ensued.

  4. Assessment of occupational exposure due to external sources of radiation. Safety guide

    International Nuclear Information System (INIS)

    2000-01-01

    Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy. The three Safety Guides on occupational radiation protection are jointly sponsored by the IAEA and the International Labour Office. The Agency gratefully acknowledges the contribution of the European Commission to the development of the present Safety Guide. The present Safety Guide addresses the assessment of exposure due to external sources of radiation in the workplace. Such exposure can result from a number of sources within a workplace, and the monitoring of workers and the workplace in such situations is an integral part of any occupational radiation protection programme. The assessment of exposure due to external radiation sources depends critically upon knowledge of the radiation type and energy and the conditions of exposure. The present Safety Guide reflects the major changes over the past decade in international practice in external dose assessment

  5. Assessment of occupational exposure due to external sources of radiation. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy. The three Safety Guides on occupational radiation protection are jointly sponsored by the IAEA and the International Labour Office. The Agency gratefully acknowledges the contribution of the European Commission to the development of the present Safety Guide. The present Safety Guide addresses the assessment of exposure due to external sources of radiation in the workplace. Such exposure can result from a number of sources within a workplace, and the monitoring of workers and the workplace in such situations is an integral part of any occupational radiation protection programme. The assessment of exposure due to external radiation sources depends critically upon knowledge of the radiation type and energy and the conditions of exposure. The present Safety Guide reflects the major changes over the past decade in international practice in external dose assessment

  6. Assessment of occupational exposure due to external sources of radiation. Safety guide

    International Nuclear Information System (INIS)

    1999-01-01

    Occupational exposure to ionizing radiation can occur in a range of industries, medical institutions, educational and research establishments and nuclear fuel cycle facilities. Adequate radiation protection of workers is essential for the safe and acceptable use of radiation, radioactive materials and nuclear energy. The three Safety Guides on occupational radiation protection are jointly sponsored by the IAEA and the International Labour Office. The Agency gratefully acknowledges the contribution of the European Commission to the development of the present Safety Guide. The present Safety Guide addresses the assessment of exposure due to external sources of radiation in the workplace. Such exposure can result from a number of sources within a workplace, and the monitoring of workers and the workplace in such situations is an integral part of any occupational radiation protection programme. The assessment of exposure due to external radiation sources depends critically upon knowledge of the radiation type and energy and the conditions of exposure. The present Safety Guide reflects the major changes over the past decade in international practice in external dose assessment

  7. 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)

  8. Ranking the types of intersections for assessing the safety of pedestrians using TOPSIS method

    Directory of Open Access Journals (Sweden)

    Călin ŞERBU

    2014-11-01

    Full Text Available Every year, more than 1500 accidents with pedestrian occur in the intersections in Romania. The number of accidents involving pedestrians in roundabouts intersections type increased approximately three times in 2013 compared to 2009 in Romania. This alarming increase led to the need of assessing the safety of pedestrians in intersections with or without safety systems. The safety systems for pedestrians and drivers include: the road marking, the pedestrian crossings marking, signal intersections with road signs, traffic lights or pedestrian safety barriers. We propose to assess the types of intersections with TOPSIS method.

  9. Assessment of the long-term safety for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Greis Dahlberg, Christina; Vahlund, Frederik [Svensk Kaernbraenslehantering AB, Stockholm (Sweden)

    2015-07-01

    During operation and decommissioning of the Swedish nuclear facilities, radioactive waste is generated that must be disposed of. Besides waste from the nuclear facilities, some waste derives from other activities such as industry, research, medical care, etc. Short-lived low- and intermediate-level waste from these activities is disposed of in the final repository for short-lived radioactive waste, SFR, in Forsmark. The facility, which has been in operation since 1988, is owned and operated by Svensk Karnbranslehantering AB, SKB. The existing facility has neither sufficient space nor a license to receive decommissioning waste. SFR must therefore be extended so that shortlived low- and intermediate-level decommissioning waste from the nuclear facilities can also be received. The need for additional capacity has been accentuated by the closure of two reactors in Barseback. These reactors cannot be dismantled until the SFR facility has been extended. The existing repository is built to receive, and after closure serve as a passive repository for, low- and intermediate-level radioactive waste. The disposal rooms are situated in the bedrock beneath the sea floor, covered by about 60 metres of rock. The repository has been designed so that it can be abandoned after closure without requiring further measures to maintain its function. The extension of SFR, is done at the -120 m level immediately adjacent to, and within the same depth range as, the existing facility. The basic function of the existing SFR and of the extended one will be the same. However, a clear difference is the design of the tunnel and the rock vault that are required to permit transport and storage of whole reactor pressure vessels. The application for a license to build this extension includes an assessment of the long-term safety (post-closure safety) of the facility. The safety assessment also contains an updated assessment of the long-term safety of the existing facility. The safety assessment for

  10. Outage Risk Assessment and Management (ORAM) technology to improve outage safety and economics

    International Nuclear Information System (INIS)

    Kalra, S.P.

    2004-01-01

    The Electric Power Research Institute (EPRI) has undertaken an aggressive program, called ORAM (Outage Risk Assessment and Management), to provide utilities with tools and technology to assist in managing risk during the planning and conduct of outages. The ORAM program consists of the following 6 steps: i) Perform utility surveys and visits on shutdown risk management needs, ii) Perform probabilistic shutdown safety assessments (PSSAs) to identify generic insights that can be incorporated into risk management guidelines and identify selected areas for the development of contingency actions, iii) Develop risk management guidelines (RMG's) that provide a systematic approach to the planning and conduct of outages from a safety perspective. Incorporate insights from the shutdown safety assessments and other operating experience into the RMG's. iv) Develop selected contingency actions including a thermalhydraulic tool kit to address higher risk time periods and activities identified in the shutdown safety assessments, v) Develop computer software that integrates all of the above capability into an easy to use tool for effective shutdown operation management for utilities, vi) Provide assistance in the transfer of this technology and the application of these tools. This paper briefly describes the technical approach and tools developed under EPRI's ORAM program and its applications for improving outage safety and economics. (author)

  11. 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.

  12. Adverse Outcome Pathways can drive non-animal approaches for safety assessment.

    Science.gov (United States)

    Burden, Natalie; Sewell, Fiona; Andersen, Melvin E; Boobis, Alan; Chipman, J Kevin; Cronin, Mark T D; Hutchinson, Thomas H; Kimber, Ian; Whelan, Maurice

    2015-09-01

    Adverse Outcome Pathways (AOPs) provide an opportunity to develop new and more accurate safety assessment processes for drugs and other chemicals, and may ultimately play an important role in regulatory decision making. Not only can the development and application of AOPs pave the way for the development of improved evidence-based approaches for hazard and risk assessment, there is also the promise of a significant impact on animal welfare, with a reduced reliance on animal-based methods. The establishment of a useable and coherent knowledge framework under which AOPs will be developed and applied has been a first critical step towards realizing this opportunity. This article explores how the development of AOPs under this framework, and their application in practice, could benefit the science and practice of safety assessment, while in parallel stimulating a move away from traditional methods towards an increased acceptance of non-animal approaches. We discuss here the key areas where current, and future initiatives should be focused to enable the translation of AOPs into routine chemical safety assessment, and lasting 3Rs benefits. © 2015 The Authors. Journal of Applied Toxicology published by John Wiley & Sons Ltd.

  13. Overview of waste isoltaion safety assessment program and description of source term characterization task at PNL

    International Nuclear Information System (INIS)

    Bradley, D.

    1977-01-01

    A project is being conducted to develop and illustrate the methods and obtain the data necessary to assess the safety of long-term disposal of high-level radioactive waste in geologic formations. The methods and data will initially focus on generic geologic isolation systems but will ultimately be applied to the long-term safety assessment of specific candidate sites that are selected in the NWTS Program. The activities of waste isolation safety assessment (WISAP) are divided into six tasks: (1) Safety Assessment Concepts and Methods, (2) Disruptive Event Analysis, (3) Source Characterization, (4) Transport Modeling, (5) Transport Data and (6) Societal Acceptance

  14. Assessment of the overall fire safety arrangements at nuclear power plants

    International Nuclear Information System (INIS)

    1996-01-01

    The present publication has been developed with the help of experts from regulatory, operating and engineering organizations, all with practical experience in the field of fire safety of nuclear power plants. The publication comprises a detailed checklist of the specific elements to be addressed when assessing the adequacy and effectiveness of the overall fire safety arrangements of operating nuclear power plants. The publication will be useful not only to regulators and safety assessors but also to operators and designers. The book addresses a specialized topic outlined in Safety Guide No. 50-SG-D2 (Rev.1), Fire Protection in Nuclear Power Plants, and it is recommended that it be used in conjunction with this Safety Series publication

  15. Novi Han Radioactive Waste Repository post-closure safety assessment, ver.2

    International Nuclear Information System (INIS)

    Mateeva, M.

    2003-01-01

    The methodology for the post-closure safety assessment is presented. The assessment context includes regulatory framework (protection principles); scope and time frame; radiological and technical requirements; modeling etc. The description of the Novi Han disposal system contains site location. meteorological, hydrological and seismological characteristics; waste and repository description and human activities characteristics. The next step in the methodology is scenario development and justification. The systematic generation os exposure scenarios is considered as central to the post-closure safety assessment. The most important requirements for the systematic scenario generation approach are: transparency, comprehensiveness (all possible FEPs influencing the the disposal system and the radionuclide release should be considered); relevant future evolutions; identification of critical issues and investigation of the robustness of the system. For the source-pathway-receptor analysis the Process System is divided into near-field, geosphere/atmosphere and biosphere, describing the key facets controlling the potential radionuclide migration to the environment. The schematic division of the Novi Han near-field Process System into lower-level conceptual features is presented and discussed. As a result of the examinations of the FEPs three classes of scenarios are identified for the Novi Han post-closure safety assessment: Environmental evolution scenarios (geological change and climate change); future human action scenarios (human intrusion and archaeological action); Scenarios with very low probability (terrorism, crashes, explosions). The safety assessment iteration leads to identification of a modern scenario generation approach, assessment of key radionuclide releases, geological and hydrological evaluation, identification of the key parameters from sensitivity analysis etc. Examples of conceptual models are given. For the mathematical modeling the AMBER code is used

  16. Risk assessment of computer-controlled safety systems for fusion reactors

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  17. 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

  18. The application of transcriptomics in the comparative safety assessment of (GMO-derived) plant products

    NARCIS (Netherlands)

    Kok, E.J.

    2008-01-01

    National and international organizations have discussed current approaches to the safety assessment of complex (plant) food products in general and the safety assessment of GMO-derived food products in particular. One of the recommendations of different expert meetings was that the new analytical

  19. Handling of future human actions in the safety assessment SR-Can

    International Nuclear Information System (INIS)

    Moren, Lena

    2006-10-01

    This report documents the future human actions (FHA) considered in the long-term safety analysis of a KBS-3 repository. The report is one of the supporting documents to the safety assessment SR-Can. The purpose of this report is to provide an account of: General considerations concerning FHA; The methodology applied in SR-Can to assess FHA; The aspects of FHA that need to be considered in the evaluation of their impact on a deep geological repository; and The selection of representative scenarios for illustrative consequence analysis

  20. Handling of future human actions in the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Moren, Lena

    2006-10-15

    This report documents the future human actions (FHA) considered in the long-term safety analysis of a KBS-3 repository. The report is one of the supporting documents to the safety assessment SR-Can. The purpose of this report is to provide an account of: General considerations concerning FHA; The methodology applied in SR-Can to assess FHA; The aspects of FHA that need to be considered in the evaluation of their impact on a deep geological repository; and The selection of representative scenarios for illustrative consequence analysis.

  1. 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

  2. Framework of nuclear safety and safety assessment

    International Nuclear Information System (INIS)

    Furuta, Kazuo

    2007-01-01

    Since enormous energy is released by nuclear chain reaction mainly as a form of radiation, a great potential risk accompanies utilization of nuclear energy. Safety has been continuously a critical issue therefore from the very beginning of its development. Though the framework of nuclear safety that has been established at an early developmental stage of nuclear engineering is still valid, more comprehensive approaches are required having experienced several events such as Three Mile Island, Chernobyl, and JCO. This article gives a brief view of the most basic principles how nuclear safety is achieved, which were introduced and sophisticated in nuclear engineering but applicable also to other engineering domains in general. (author)

  3. Extended biosphere dataset for safety assessment of radioactive waste geological disposal

    International Nuclear Information System (INIS)

    Kato, Tomoko; Suzuki, Yuji

    2007-01-01

    JAEA has an on-going programme of research and development relating to the safety assessment of the deep geological disposal systems of high-level radioactive waste (HLW) and transuranic waste (TRU). In the safety assessment of HLW and TRU disposal systems, biosphere assessment is necessary to estimate future radiological impacts on human beings (e.g. radiation dose). In order to estimate radiation dose, consideration needs to be given to the biosphere into which future releases of radionuclides might occur and to the associated future human behaviour. The data of some biosphere parameters needed to be updated by appropriate data sources for generic and site-specific biosphere assessment to improve reliability for the biosphere assessment, because some data published in the 1980's or the early 90's were found to be inappropriate for the recent biosphere assessment. Therefore, data of the significant parameters (especially for element-dependent) were set up on the basis of recent information, to update the generic biosphere dataset. (author)

  4. Project SAFE. Update of the SFR-1 safety assessment. Phase 1. Appendix A5: Radionuclide transport

    International Nuclear Information System (INIS)

    Moreno, L.

    1998-01-01

    A critical revision of the previous safety assessments made by SKB on the Final Repository for Radioactive Operational Waste, SFR is presented. The review of the Deepened Safety Assessment is also discussed. Based on this critical revision improvements are suggested. Hydrology, formation of complexes, and long-term behaviour of the barriers are some of the aspects where the safety assessment could be improved

  5. Handbook of safety assessment of nanomaterials from toxicological testing to personalized medicine

    CERN Document Server

    Fadeel, Bengt

    2014-01-01

    "The Handbook of Safety Assessment of Nanomaterials: From Toxicological Testing to Personalized Medicine provides a comprehensive overview of the state of the art of nanotoxicology and is a unique resource that fills up many knowledge gaps in the toxicity issue of nanomaterials in medical applications. The book is distinguished by up-to-date insights into creating a science-based framework for safety assessment of nanomedicines." -Prof. Yuliang Zhao, National Center for Nanosciences and Technology, China.

  6. Aspects of safety assessments for package with additional equipment components

    International Nuclear Information System (INIS)

    Reiche, I.; Boerst, F.M.; Krietsch, T.

    2004-01-01

    Many paragraphs in TS-R-1 contain the terms ''package'' or ''packaging''. These terms are defined in TS-R-1 paras 230 and 231 and explained in TS-G-1.1 paras 230.1 - 230.6. The importance of a consistent understanding of these definitions has been shown by recent discussions during the assessment of applications for package design approval. There was disagreement, if equipment components attached to the container body during transport, e.g. a transport frame, should be considered part of the package and taken into account in the safety assessment for the package. Discussions were also caused by the way inner design components are treated in the safety assessment of the package. This paper summarises the regulatory requirements to such additional equipment components and presents the way of their inclusion into the package design approval process in Germany

  7. 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

  8. Advanced Photon Source experimental beamline Safety Assessment Document: Addendum to the Advanced Photon Source Accelerator Systems Safety Assessment Document (APS-3.2.2.1.0)

    International Nuclear Information System (INIS)

    1995-01-01

    This Safety Assessment Document (SAD) addresses commissioning and operation of the experimental beamlines at the Advanced Photon Source (APS). Purpose of this document is to identify and describe the hazards associated with commissioning and operation of these beamlines and to document the measures taken to minimize these hazards and mitigate the hazard consequences. The potential hazards associated with the commissioning and operation of the APS facility have been identified and analyzed. Physical and administrative controls mitigate identified hazards. No hazard exists in this facility that has not been previously encountered and successfully mitigated in other accelerator and synchrotron radiation research facilities. This document is an updated version of the APS Preliminary Safety Analysis Report (PSAR). During the review of the PSAR in February 1990, the APS was determined to be a Low Hazard Facility. On June 14, 1993, the Acting Director of the Office of Energy Research endorsed the designation of the APS as a Low Hazard Facility, and this Safety Assessment Document supports that designation

  9. 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

  10. 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...

  11. 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.))

  12. Safety risk assessment using analytic hierarchy process (AHP) during planning and budgeting of construction projects.

    Science.gov (United States)

    Aminbakhsh, Saman; Gunduz, Murat; Sonmez, Rifat

    2013-09-01

    The inherent and unique risks on construction projects quite often present key challenges to contractors. Health and safety risks are among the most significant risks in construction projects since the construction industry is characterized by a relatively high injury and death rate compared to other industries. In construction project management, safety risk assessment is an important step toward identifying potential hazards and evaluating the risks associated with the hazards. Adequate prioritization of safety risks during risk assessment is crucial for planning, budgeting, and management of safety related risks. In this paper, a safety risk assessment framework is presented based on the theory of cost of safety (COS) model and the analytic hierarchy process (AHP). The main contribution of the proposed framework is that it presents a robust method for prioritization of safety risks in construction projects to create a rational budget and to set realistic goals without compromising safety. The framework provides a decision tool for the decision makers to determine the adequate accident/injury prevention investments while considering the funding limits. The proposed safety risk framework is illustrated using a real-life construction project and the advantages and limitations of the framework are discussed. Copyright © 2013 National Safety Council and Elsevier Ltd. All rights reserved.

  13. Geosphere process report for the safety assessment SR-Can

    International Nuclear Information System (INIS)

    Skagius, Kristina

    2006-09-01

    This report documents geosphere processes identified as relevant to the long-term safety of a KBS- repository, and forms an important part of the reporting of the safety assessment SR-Can. The detailed assessment methodology, including the role of the process report in the assessment, is described in the SR-Can Main report. The following excerpts describe the methodology, and clarify the role of this process report in the assessment. 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, will evolve over time. Future states of the system will depend on the initial state of the system, a number of radiation related, thermal, hydraulic, mechanical, chemical and biological processes acting within the repository system over time, and external influences acting on the system. A methodology in ten steps has been developed for SR-Can described below. Identification of factors to consider (FEP processing): This step consists of identifying all the factors that need to be included in the analysis. Experience from earlier safety assessments and KBS-specific and international databases of relevant features, events and processes influencing long-term safety are utilised. Based on the results of the FEP processing, an SR-Can FEP catalogue, containing FEPs to be handled in SR-Can, has been established. The initial state of the system is described based on the design specifications of the KBS repository, a descriptive model of the repository site and a site-specific layout of the repository. The initial state of the fuel and the engineered components is that immediately after deposition, as described in the SR-Can Initial state report. The initial state of the geosphere and the biosphere is that of the natural system prior to excavation, as described in the site descriptive models. The repository layouts adapted to the sites are provided in underground

  14. Geosphere process report for the safety assessment SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Skagius, Kristina [Kemakta Konsult AB, Stockholm (SE)] (ed.)

    2006-09-15

    This report documents geosphere processes identified as relevant to the long-term safety of a KBS- repository, and forms an important part of the reporting of the safety assessment SR-Can. The detailed assessment methodology, including the role of the process report in the assessment, is described in the SR-Can Main report. The following excerpts describe the methodology, and clarify the role of this process report in the assessment. 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, will evolve over time. Future states of the system will depend on the initial state of the system, a number of radiation related, thermal, hydraulic, mechanical, chemical and biological processes acting within the repository system over time, and external influences acting on the system. A methodology in ten steps has been developed for SR-Can described below. Identification of factors to consider (FEP processing): This step consists of identifying all the factors that need to be included in the analysis. Experience from earlier safety assessments and KBS-specific and international databases of relevant features, events and processes influencing long-term safety are utilised. Based on the results of the FEP processing, an SR-Can FEP catalogue, containing FEPs to be handled in SR-Can, has been established. The initial state of the system is described based on the design specifications of the KBS repository, a descriptive model of the repository site and a site-specific layout of the repository. The initial state of the fuel and the engineered components is that immediately after deposition, as described in the SR-Can Initial state report. The initial state of the geosphere and the biosphere is that of the natural system prior to excavation, as described in the site descriptive models. The repository layouts adapted to the sites are provided in underground

  15. RADON-type disposal facility safety case 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)

    Guskov, A.; Batanjieva, B.; Kozak, M.W.; Torres-Vidal, C.

    2002-01-01

    The ISAM safety assessment methodology was applied to RADON-type facilities. The assessments conducted through the ISAM project were among the first conducted for these kinds of facilities. These assessments are anticipated to lead to significantly improved levels of safety in countries with such facilities. Experience gained though this RADON-type Safety Case was already used in Russia while developing national regulatory documents. (author)